ML13254A171

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T. Smith Ltr Independent Confirmatory Survey Summary and Results for the Ford Nuclear Reactor, Revision 1, Ann Arbor, Michigan
ML13254A171
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Site: University of Illinois
Issue date: 08/01/2013
From: Altic N
Oak Ridge Associated Universities
To:
NRC/FSME
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DCN 5176-SR-01-1, RFTA 12-008
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INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE FORD NUCLEAR REACTOR, REVISION 1 ANN ARBOR, MICHIGAN Nick Altic Prepared for the U.S. Nuclear Regulatory Commission Approved for public release; further dissemination unlimited.

Oak Ridge Associated Universities manages the Oak Ridge Institute for Science and Education (ORISE) contract for the U.S. Department of Energy. ORISE focuses on scientific initiatives to research health risks from occupational hazards, assess environmental cleanup, respond to radiation medical emergencies, support national security and emergency preparedness, and educate the next generation of scientists.

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 FORD NUCLEAR REACTOR, REVISION 1 ANN ARBOR, MICHIGAN Prepared by Nick Altic 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 AUGUST 2013 Prepared by Oak Ridge Associated Universities under the Oak Ridge Institute of 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.

5176-SR-01-1

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

1. INTRODUCTION....................................................................................................................................... 1
2. SITE DESCRIPTION ................................................................................................................................. 1 2.1 CONFIRMATORY UNITS .................................................................................................................... 2
3. APPLICABLE SITE GUIDELINES ........................................................................................................ 4
4. OBJECTIVES................................................................................................................................................ 4
5. PROCEDURES ............................................................................................................................................ 5 5.1 DOCUMENT REVIEW ......................................................................................................................... 5 5.2 REFERENCE SYSTEM ......................................................................................................................... 5 5.3 SURFACE SCANS ................................................................................................................................. 5 5.4 SURFACE ACTIVITY MEASUREMENTS ............................................................................................. 6 5.5 SOIL AND MISCELLANEOUS MATERIAL SAMPLING ...................................................................... 6
6. SAMPLE ANALYSIS AND DATA INTERPRETATION ................................................................. 7
7. FINDINGS AND RESULTS ..................................................................................................................... 7 7.1 DOCUMENT REVIEW ......................................................................................................................... 7 7.2 SURFACE SCANS ................................................................................................................................. 8 7.3 SURFACE ACTIVITY MEASUREMENTS ............................................................................................. 9 7.4 SURFACE ACTIVITY DATA COMPARISON .....................................................................................10 7.5 RADIONUCLIDE CONCENTRATIONS IN SOIL SAMPLES ..............................................................11 7.5.1 Inter-Laboratory Comparison ...............................................................................................12
8.

SUMMARY

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

9. REFERENCES ...........................................................................................................................................14 APPENDIX A: FIGURES APPENDIX B: SCAN RESULTS APPENDIX C: TABLES APPENDIX D: MAJOR INSTRUMENTATION APPENDIX E: SURVEY AND ANALYTICAL PROCEDURES Ford Reactor Survey Report ii 5176-SR-01-1

TABLES Table 3.1. FNR Radiological Contaminants and Decommissioning Criteria ........................................... 4 Table 7.1. Retrospective Analysis of FSS Data Packages............................................................................. 8 Table 7.2. Confirmatory Surface Activity Comparison ..............................................................................10 Table 7.3. Side-by-Side Beta Measurements for Survey Unit 3-1 .............................................................11 Table 7.4. Radionuclide Concentrations in Soil (pCi/g) ............................................................................12 FIGURES Fig. A-1. Location of the University of Michigan .................................................................................... A-1 Fig. A-2. Ford Nuclear Reactor, Basement Confirmatory Units and Scan Coverage ........................ A-2 Fig. A-3. Ford Nuclear Reactor, First Floor Scan Coverage .................................................................. A-3 Fig. A-4. Ford Nuclear Reactor, Second Floor Scan Coverage ............................................................. A-4 Fig. A-5. Ford Nuclear Reactor, Third Floor Scan Coverage ................................................................ A-5 Fig. A-6. Ford Nuclear Reactor, Cooling Tower Scan Coverage .......................................................... A-6 Fig. A-7. Ford Nuclear Reactor, Confirmatory Unit 1 Direct Measurement and Soil Sample Locations ....................................................................................................................................... A-7 Fig. A-8. Ford Nuclear Reactor, Confirmatory Unit 2 Direct Measurement Locations .................... A-8 Fig. A-9. Ford Nuclear Reactor, Confirmatory Unit 3 Direct Measurement Locations .................... A-9 Ford Reactor Survey Report iii 5176-SR-01-1

ACRONYMS COC contaminant of concern cpm counts per minute CU confirmatory unit DCGLW derived concentration guideline level DER normalized absolute difference dpm disintegrations per minute FNR Ford Nuclear Reactor FSS final status surveys FSSP final status survey plan IA impacted areas IEAV Independent Environmental Assessment and Verification ISM integrated safety management LBGR lower bound of the gray region MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC minimum detectable concentration MeV megaelectron volts NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission NRIP NIST Radiochemistry Intercomparison Program ORAU Oak Ridge Associated Universities ORISE Oak Ridge Institute for Science and Education pCi/g picocuries per gram Q quantile SOF sum of fractions SU survey unit TAP total absorption peak UM University of Michigan Ford Reactor Survey Report iv 5176-SR-01-1

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE FORD NUCLEAR REACTOR, REVISION 1 ANN ARBOR, MICHIGAN

1. INTRODUCTION The Ford Nuclear Reactor (FNR) at the University of Michigan (UM) was a light-water cooled and moderated open-pool design reactor (UM 2012). It had a heterogeneous core composed of aluminum and enriched uranium-235 (UM 2006). The FNR went critical in 1957 and provided neutron and gamma irradiation services, neutron beam port experimental facilities, and training facilities for use by faculty, students, and researchers from UM, other universities, and industrial organizations. The FNR was operated by the Michigan Memorial Phoenix Project of UM under U.S. Nuclear Regulatory Commission (NRC) License R-28, Docket 50-2, until it was shut down in July 2003. The fuel was removed in December 2003.

CH2M HILL conducted the historical site assessment in 2003 to assess and detail the radiological status of the FNR. Results of radiological surveys showed that many potentially impacted areas were free of contamination, including a major portion of the FNR structure. However, non-routine occurrences, accidents, or spills between 1959 and 2001 contributed to the contamination of several systems and surfaces associated with the reactor (UM 2006).

At the NRCs request, Oak Ridge Associated Universities (ORAU) Independent Environmental Assessment and Verification (IEAV) Program conducted in-process confirmatory survey activities at the FNR.

2. SITE DESCRIPTION Located on the North Campus of UM at 2301 Bonisteel Boulevard, approximately 1.25 miles northeast of the central business district of Ann Arbor (Fig. A-1), the windowless FNR building is constructed of reinforced concrete with a brick veneer. The footprint of the FNR building is approximately 4,760 square feet (440 square meters) with a height near 69 ft. (21 m) and is conjoined with the Phoenix Memorial Laboratory. Though several services were interconnected, the two structures operated independently. The FNR facility was divided into five levels (UM 2006):

Ford Reactor Survey Report 1 5176-SR-01-1

Basement (liquid cooling and waste systems), 1st Floor (beamport experimental area), 2nd Floor (maintenance and other support facilities and systems), 3rd Floor (reactor access and control), and the 4th Floor (cooling tower) (Figs. A-2 to A-6).

Results from the characterization surveys in 2006 identified contamination in the following areas (UM 2006):

  • Basement o Floor drains o Sumps o Floor
  • 1st Floor o Floor drain o Floor trench around the west and north pool walls o Storage ports in the west wall
  • 3rd Floor o Floor drains o Floor near the pool o South wall above the pool o Room 3103 The extensive removal, disposition, and/or decontamination of these components, structures, and systemsas well as several others throughout the FNRwere performed during the remediation phase of the decommissioning process. Post-remediation sampling and routine surveys have been performed to confirm that volumetric and surface contamination are not present in/on the remaining FNR structures (UM 2012).

2.1 CONFIRMATORY UNITS The operational history for various areas of the facility resulted in different levels of potential exposure to residual radiological contamination. Therefore, different areas required different levels of survey coverage to determine if remaining residual radioactivity levels meet the NRC release criteria. UM has divided the FNR facility into multiple survey units (SUs) in accordance with the guidance in the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (NRC 2000).

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MARSSIM designates three category classifications for impacted areas (IAs), or areas that have some potential for containing contaminated material. Based on contamination potential, IAs are categorized as Class 1, 2, or 3. Descriptions for each classification for IAs are as follows:

Class 1: Buildings or land areas that have a significant potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiological surveys) that exceeds the expected derived concentration guideline level (DCGLW)

Class 2: Buildings or land areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGLW Class 3: Any impacted areas that are not expected to contain residual contamination, or are expected to contain levels of residual contamination at a small fraction of the DCGLW Non-impacted areas are areas that should not have the potential to contain contaminated materials.

For confirmatory purposes, ORAU grouped several of UMs finished final status surveys (FSS) SUs into three confirmatory units (CUs). All CUs were given a Class 1 designation. Two CUs were located in the basement and one was located on the third floor. UM had just begun FSS activities on the 1st or 2nd floors; therefore, ORAU could not perform confirmatory surveys of those areas.

Instead, ORAU performed in-process surveys consisting of high- to medium density scans in the several areas that had not received FSS.

Located in the basement, CU 1 consisted of SUs B-1, B-2, B-4, B-5, B-6, and a portion of SU B-3.

Also located in the basement, CU 2 consisted of a portion of SU B-3 and SU 1-13. CU 2 did not have a ceiling and was open to the first floor; this CU contained the bottom portion of the pool.

CU 3 was located on the third floor and consisted of SUs 3-3, 3-4, 3-5, 3-6, 3-7, and 3-8.

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3. APPLICABLE SITE GUIDELINES The primary contaminants of concern (COCs) for the FNR are beta-gamma emittersfission and activation productsresulting from reactor operation. Results from samples collected during remediation activities indicate that residual amounts of cobalt-60 (Co-60), silver-108m (Ag-108m),

silver-110m (Ag-110m), and cesium-137 (Cs-137) may be present in construction materials and exposed surface soils (UM 2012). In addition, europium-152 (Eu-152) and carbon-14 (C-14) were detected in subsurface soil at concentrations below their respective default screening values.

Table 3.1 provides further detail regarding the respective contaminants and associated DCGLs (UM 2012).

Table 3.1. FNR Radiological Contaminants and Decommissioning Criteria Structure Surface DCGL Surface Soil (dpm/100cm2)

Radionuclide DCGL (pCi/g)

Total Removable Co-60 3.8 7,050 705 Cs-137 11 28,000 2,800 Ag-108m 8.2 17,000 1,700 Ag-110m 4.92 10,200 1,020 C-14 N/A N/A N/A Eu-152 N/A N/A N/A Gross beta N/A 5125 512 aN/A = not applicable or not present as a contaminant bIncludes short-lived daughter products present due to assumed ingrowth period of 20 years

4. OBJECTIVES UM was still in the process of performing FSS at the time of the ORAU site visit. Therefore, ORAU performed in-process inspections in areas where UM was working during the visit and confirmatory surveys where FSS were completed. The objectives were to independently review contractor documentation and field data, evaluate UMs survey process, and generate independent radiological data to assist NRC in evaluating the adequacy and accuracy of UMs FSS results.

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5. PROCEDURES During the period December 4 through December 6, 2012, ORAU performed an in-process confirmatory survey of the FNR. The survey was in accordance with a plan dated November 30, 2012, and submitted to and approved by the NRC headquarters (ORAU 2012a). Survey activities were conducted in accordance with the ORAU/ORISE Survey Procedures and ORAU Quality Program Manuals (ORAU/ORISE 2013a and ORAU 2012b). This report summarizes the procedures and results of the survey.

5.1 DOCUMENT REVIEW Prior to on-site activities, ORAU was tasked with reviewing the final status survey plan (FSSP) for the FNR provided by UM. The FSSP was reviewed for adequacy and appropriateness while taking MARSSIM guidance into account (NRC 2000).ORAU also reviewed UMs FSS data packages of the SUs listed in Section 2.1 to ensure that survey objectives stated in the FSSP were met.

5.2 REFERENCE SYSTEM ORAU used specific X, Y coordinates from the southwest corner of the respective CU floor and lower left corner of walls.

5.3 SURFACE SCANS Gamma scans were performed using sodium iodide (NaI) scintillation detectors coupled to ratemeter-scalers with audible indicators. Beta surface scans were performed using both large (floor monitor) and hand-held gas proportional detectors coupled to ratemeter-scalers with audible indicators. Both NaI and gas proportional detector/instrument combinations were connected to hand-held electronic data collectors equipped with real-time data-logging software to record instrument response during scans.

High-density scans were performed in all three CUs. Judgmental confirmatory surveys were also performed within portions of SUs where FSS activities had not been completed. Medium- to high-density scans were performed for judgmental scan locations. Judgmental scan locations were as follows:

Ford Reactor Survey Report 5 5176-SR-01-1

  • 1st Floor - Lower walls and floor of Room 1101
  • 2nd Floor - Exhaust plenum and floor in Room 2111
  • 3rd Floor - Lower walls and floor of Rooms 3110, 3109, and 3108, and Corridor 3101
  • 4th Floor - Cooling tower floor, walls and support bracing 5.4 SURFACE ACTIVITY MEASUREMENTS Construction material-specific background measurements were collected for correcting gross activity measurements performed on structural SUs. Material-specific backgrounds were collected from the same area by both ORAU and the licensee. Direct measurements for total beta activity were performed at random locations in each CU (Figs. A-7 to A-9). The number of measurements performed was determined by the relative shift used by UM. No judgmental direct measurement locations were identified were collected. Smear samples to determine removable gross beta activity levels were collected from any direct measurement location that was above 10% of the DCGLW.

Additionally, ten measurement locations were selected to correspond to licensee locations (i.e., side-by-side measurements) for direct data comparison. The direct measurements were performed using hand-held gas proportional detectors. Detectors were coupled to portable ratemeter-scalers.

5.5 SOIL AND MISCELLANEOUS MATERIAL SAMPLING Two judgmental surface soil samples were collected from the basement area (Fig. A-7). Soil sample 5176S0001 was collected from the cold sump drain line trench along the south wall. Soil sample 5176S0002 was collected from the hot sump trench area where it meets the east wall. Selected sample locations were based on the results of gamma scans and previously identified contamination.

In addition to the soil samples collected by the survey team, UM submitted five samples for an inter-laboratory comparison at NRCs request (Samples 5176S0003 through 5176S0007). ORAU also received five split samples from UM (Samples 5176S0008 through 5176S0012), in addition to the samples submitted for the inter-laboratory comparison.

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6. SAMPLE ANALYSIS AND DATA INTERPRETATION Samples were returned to the ORAU/ORISE Radiological and Environmental Analysis Laboratory in Oak Ridge, Tennessee, for analysis and interpretation. Sample analyses were performed in accordance with the ORISE Laboratory Procedures Manual (ORAU/ORISE 2013b). Soil and miscellaneous roofing material samples were analyzed by solid-state gamma spectroscopy for gamma-emitting COCs. Analytical results were reported in units of picocuries per gram (pCi/g).

Direct measurement data were converted to units of disintegrations per minute per 100 square centimeters (dpm/100 cm2). The data generated were compared with the approved DCGLWs established for the FNR.

7. FINDINGS AND RESULTS The results for each of the verification activities are discussed below.

7.1 DOCUMENT REVIEW The ORAU review of UMs FSSP indicated that an incorrect surface efficiency was used for calculating the total efficiency. Per the FSSP, a surface efficiency of 0.37 would be applied. In the absence of site-specific data, MARSSIM prescribes that a surface efficiency of 0.25 should be applied for beta emitters with a maximum energy between 0.15 and 0.4 megaelectron volt (MeV),

and a surface efficiency of 0.5 should be applied for maximum beta energies above 0.4 MeV. UM opted to assign a surface efficiency of 0.25 to all beta emitters, which is a conservative approach.

Other observations made during the confirmatory survey site visit are described below.

Table 7.1 provides a retrospective review of UMs FSS data packages for the 13 SUs in which confirmatory survey activities were performed. The FSS surface activity data reviewed was calculated with a surface efficiency of 0.25. For FSS planning, UM chose to set the lower bound of the gray region (LBGR) and expected contaminant variation at 50% and 30%, respectively, of the gross beta DCGLW. The resulting relative shift (/) is equal to 1.67, which translates to 17 direct measurement locations required for each survey unit (NRC 2000). A retrospective analysis of the FSS data shows that the mean surface activity and contaminant variability were less than the LBGR Ford Reactor Survey Report 7 5176-SR-01-1

and expected variability used as planning inputs, meaning that more direct measurements were collected than required.

Table 7.1. Retrospective Analysis of FSS Data Packages ORAU Surface Activity n (dpm/100 cm2) Retrospective Confirmatory UM SU Unit Collected /

Mean CU 1 B-1 22 287 247 20 CU 1 B-2 19 164 284 17 CU 1/CU 2 B-3 19 181 363 14 CU 1 B-4 26 -379 367 15 CU 1 B-5 19 -142 353 15 CU 1 B-6 18 93 285 18 CU 2 1-13 23 -164 1059 5.0 CU 3 3-3 17 857 335 13 CU 3 3-4 17 497 556 8.3 CU 3 3-5 18 528 490 9.4 CU 3 3-6 25 324 327 15 CU 3 3-7 20 294 437 11 CU 3 3-8 22 524 537 8.6 aCalculated by setting the LBGR at the SU mean surface activity.

7.2 SURFACE SCANS The gross count rates for beta and gamma radiation surface scan data for each ORAU CU and the corresponding UM SUs were prepared for report presentation using quantile (Q) plots. The Q-plots are presented in Appendix B. They are a graphical technique for determining if there is a common distribution in data sets. The advantage of the Q-plot is that population distributional aspects can be evaluated simultaneously. The detectable aspects include:

  • Shifts in scale
  • Changes in symmetry (skewness of the data)
  • The presence of outliers Ford Reactor Survey Report 8 5176-SR-01-1

Q-plots were generated by uploading the scan data files into the U.S. Environmental Protection Agencys ProUCL software. In the Q-plots provided in Appendix B, the Y-axis represents observed count rates in counts per minute (cpm). The X-axis represents the data quantiles about the mean value. A normal distribution that is not skewed by outliers will appear as a straight line with the slope of the line subject to the degree of variability among the data population (i.e., a background radiation population). Values less than the mean are represented in the negative quantiles, and values greater than the mean are represented in the positive quantiles. The presence of more than one populatione.g., background radiation population and contaminationwould display on a Q-plot as a step function. Small areas of localized contamination will appear on the Q-plot as outlier points in the upper right quadrant.

Instrument response for beta scans ranged from 3 to 805 gross cpm for the walls and 167 to 2,196 cpm for the floor over all areas investigated during confirmatory surveys. Instrument responses are low because data capture was initiated before the instrument was turned on; therefore, the instruments increase to background was captured. Instrument response for gamma scans ranged from 1,054 to 28,393 gross cpm over all areas investigated during confirmatory surveys. Beta floor scans are reported separately from wall scans because different detectors were used. The detector for the beta floor scans had a much larger background reading than the detector used for beta wall scans. The ORAU survey team detected residual radioactivity in CU 3, outside of the doorway to Room 3103, while performing surface scans with hand-held gas proportional and NaI detectors.

Even though the elevated area was small and below the action level, UM still remediated the area to background levels. Elevated instrument response for gamma scans for the floor and walls on the first floor is apparent by looking at the Q-plot; however, this elevated response is due to source storage area outside of the study boundary. No other areas of elevated activity were detected from surface scans.

7.3 SURFACE ACTIVITY MEASUREMENTS Total surface activity levels for the three CUs are provided in Tables C-1 through C-3. The reported surface activities represent gross levels that have been corrected for background. Background measurements were collected from the lower pool monolith for concrete and the first floor mens restroom in the Phoenix Memorial Lab. These were the same area that UM collected their Ford Reactor Survey Report 9 5176-SR-01-1

background measurements. Table 7.2 provides a summary of the confirmatory measurement data for each CU relative to FSS data of UM.

Table 7.2. Confirmatory Surface Activity Comparison Surface Activity (dpm/100 cm2)

Confirmatory Unit Min Max Mean 95% Confidence Interval of Mean ORAU UM ORAU UM ORAU UM ORAU UM 1 -370 -1,084 270 821 10 -10 -300 to 321 -808 to 789 2 -610 -1,511 300 3,596 -160 -85 -552 to 232 -1,853 to 1,682 3 -140 -870 610 1,875 187 488 -157 to 531 -447 to 1,423 The variation in the surface activity levels reported by UM was much larger than those determined by ORAU for all CUs. This large variation is most likely because the FSS instrumentation used by UM had a small efficiency; thus, a small change in background resulted in large change in surface activity. The mean surface activity reported by UM was within the 95% confidence interval of the mean reported by ORAU for each of the three CUs.

7.4 SURFACE ACTIVITY DATA COMPARISON During the site visit ORAU collected side-by-side direct measurements with UM in SU 3-1, located in Corridor 3101, on the third floor. The instrument/detector combinations used were:

  • ORAULudlum Model 2221 ratemeter-scaler coupled to a Ludlum Model 43-68 gas proportional detector
  • UMModel 2360 data-logger coupled to a Ludlum Model 43-93 alpha/beta scintillation detector The total efficiencies for the instrument and detector combinations were 7.36% for UM and 14% for ORAU. Geometry correction factors for ORAU and UM detectors used were 1.26 and 1.00, respectively. The results of the side-by-side measurements are shown in Table 7.3. The surface activity for the individual measurements reported by UM are all above zero, indicating that Ford Reactor Survey Report 10 5176-SR-01-1

background may not be properly defined. However, it appears that UM under-estimated the background levels, which is conservative.

Table 7.3. Side-by-Side Beta Measurements for Survey Unit 3-1 Surface Activity Gross Counts (cpm)

UM Location Code (dpm/100 cm2)

ORAU UM ORAU UM FNR_3-1_C1_A_011 399 305 96 313 FNR_3-1_C1_C_012 384 399 11 1,698 FNR_3-1_C1_C_013 359 291 -130 231 FNR_3-1_C1_C_014 302 277 -450 41 FNR_3-1_W1_A_015 378 298 -23 217 FNR_3-1_C1_C_016 361 278 -120 326 FNR_3-1_F1_C_017 393 314 62 543 FNR_3-1_W1_A_018 412 297 170 204 FNR_3-1_C1_C_019 405 316 130 571 FNR_3-1_F1_C_022 415 293 190 258 Mean: -6 440 7.5 RADIONUCLIDE CONCENTRATIONS IN SOIL SAMPLES Individual sample results for the gamma-emitting fission/activation products that UM has identified as COCs are presented in Table 7.4. Samples 5176S0001 and 5176S0002 were collected from the drain line and hot sump soil trenches in the basement by the survey team during the December 2012 site visit. The remaining samples (5176S0008 through 5176S0012) presented in Table 7.4 were submitted as split samples to the ORAU/ORISE Radiological and Environmental Analysis Laboratory. These split samples were collected from the basement and first floor exposed soil areasincluding the cold sump, hot sump, and trench between the cold sump and drain line.

Sample results for 5176S0002 indicated the most notable detected concentrations of elevated COCs were for Ag-108m and Co-60; however, the levels were below the individual COC DCGLWs.

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Table 7.4. Radionuclide Concentrations in Soil (pCi/g)

ORAU Sample Ag-108M Ag-110M Cs-137 Co-60 SOFa ID 5176S0001b 0.53 +/- 0.04c -0.05 +/- 0.02 0.05 +/- 0.02 0.06 +/- 0.02 0.08 5176S0002d 1.64 +/- 0.12 -0.23 +/- 0.06 0.45 +/- 0.05 2.85 +/- 0.19 0.99 5176S0008 0.00e +/- 0.00 0.00 +/- 0.01 0.00 +/- 0.01 0.03 +/- 0.02 0.01 5176S0009 0.01 +/- 0.01 0.00 +/- 0.01 -0.01 +/- 0.02 0.17 +/- 0.04 0.05 5176S0010 0.01 +/- 0.01 -0.01 +/- 0.02 0.01 +/- 0.02 0.02 +/- 0.02 0.01 5176S0011 0.03 +/- 0.01 -0.02 +/- 0.02 0.00 +/- 0.02 0.04 +/- 0.02 0.01 5176S0012 0.22 +/- 0.02 0.00 +/- 0.01 0.21 +/- 0.02 0.28 +/- 0.03 0.12 aSOF = sum of fractions. Negative values were not included in the SOF calculations bCollected from drain line soil trench cErrors represent the total propagated uncertainties reported at the 95% confidence level dCollected from hot sump soil trench eZero values are due to rounding 7.5.1 Inter-Laboratory Comparison Five of UMs FSS soil samples (5176S0003 through -0007) were submitted to the ORAU/ORISE Radiological and Environmental Analysis Laboratory for an inter-lab comparison. Radionuclide concentrations determined by each laboratory are presented in Table C-4. The criterion used to evaluate the samples was the normalized absolute difference (DER). The DER is defined in the equation below (DOE 2012).

l l

= 3

( )2 + ( )2 Where:

S = sample concentration D = Duplicated sample concentration CSUS = 1 sigma uncertainty of the sample CSUD = 1 sigma uncertainty of the duplicate If the DER is less than 3, the results are in agreement at the 99% confidence level. Sample 5176S0005 was the only sample that had a DER value of greater than 3, which was for Co-60 only.

Results reported by both labs were below the Co-60 DCGLW.

Ford Reactor Survey Report 12 5176-SR-01-1

8.

SUMMARY

At the NRCs request, ORAU conducted confirmatory surveys of the FNR during the period of December 4 through 6, 2012. The survey activities included visual inspections and measurement and sampling activities. Confirmatory activities also included the review and assessment of UMs project documentation and methodologies.

Surface scans identified elevated activity in two areas. The first area was on a wall outside of Room 3103 and the second area was in the southwest section on the first floor. The first area was remediated to background levels. However, the second area was due to gamma shine from a neighboring source storage area.

A retrospective analysis of UMs FSS data shows that for the SUs investigated by the ORAU survey team, UM met the survey requirements set forth in the FSSP. The total mean surface activity values were directly compared with the mean total surface activity reported by UM. Mean surface activity values determined by UM were within two standard deviations of the mean determined by ORAU.

Additionally, all surface activity values were less than the corresponding gross beta DCGLW.

Laboratory analysis of the soil showed that COC concentrations were less than the respective DCGLW values. For the inter-lab comparison, the DER was above 3 for only one sample. However, since the sum of fractions (see Table C-5) for each of the soil samples was below 1, thus none of the samples would fail to meet release guidelines.

Based on the findings of the side-by-side direct measurements, and after discussion with the NRC and ORAU, UM decided to use a more appropriate source efficiency in their direct measurement calculations and changed their source efficiency from 0.37 to 0.25.

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9. REFERENCES DOE 2012. Quality Systems for Analytical Services. Revision 2.8. U.S. Department of Energy.

Washington, DC. January.

NRC 1998. Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions. U.S. Nuclear Regulatory Commission. 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.

ORAU 2012a. Project-Specific Plan for Independent Confirmatory Survey Activities Associated With the Ford Nuclear Reactor at the University of Michigan, Ann Arbor, Michigan. Prepared by Oak Ridge Associated Universities under the Oak Ridge Institute for Science and Education contract. Oak Ridge, Tennessee. November 30.

ORAU 2012b. Quality Program Manual for the Independent Environmental Assessment and Verification Program. Oak Ridge Associated Universities. Oak Ridge, Tennessee. November 29.

ORAU 2012c. ORAU/ORISE Health and Safety Manual. Oak Ridge Associated Universities.

Oak Ridge, Tennessee. May 18.

ORAU/ORISE 2011. ORAU/ORISE 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. Survey Procedures Manual for the Independent Environmental Assessment and Verification Program. Oak Ridge Institute for Science and Education, managed and operated by Oak Ridge Associated Universities. Oak Ridge, Tennessee. January 18.

ORAU/ORISE 2013b. Laboratory Procedures Manual for the Independent Environmental Assessment and Verification Program. Oak Ridge Institute for Science and Education, managed and operated by Oak Ridge Associated Universities. Oak Ridge, Tennessee. May 3.

UM 2006. Decommissioning Plan for the Ford Nuclear Reactor. Revision 1. University of Michigan.

Ann Arbor, Michigan. January 5.

UM 2012. Ford Nuclear Reactor - Technical Specification Amendment Request Decommissioning PlanRevised Section 4.5.10.6 (Final Status Survey). Docket 50-2/License R-28. University of Michigan. Ann Arbor, Michigan. November 2.

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APPENDIX A FIGURES Ford Reactor Survey Report 5176-SR-01-1

Fig. A-1. Location of the University of Michigan Ford Reactor Survey Report A-1 5176-SR-01-1

Figure provided by University of Michigan Scan Coverage Percentage Confirmatory Unit 1 (CU1) consists of Survey Units B-1, B-2, B-4, Ford Nuclear Reactor B-5 and B-6 Beta floor, lower wall and ceiling scan coverage ~ 80% University of Michigan Gamma floor scan coverage ~ 80% Ann Arbor, Michigan Gamma lower wall scan coverage ~ 50%

Confirmatory Unit 2 (CU2) consists of a portion of Survey Unit B-3 Basement Scans and Beta floor, lower wall and ceiling scan coverage ~ 80%

Confirmatory Units Gamma floor scan coverage ~ 80%

Gamma lower wall scan coverage ~ 50%

Fig. A-2. Ford Nuclear Reactor, Basement Confirmatory Units and Scan Coverage Ford Reactor Survey Report A-2 5176-SR-01-1

Figure provided by University of Michigan Scan Coverage Percentage Ford Nuclear Reactor Beta floor and lower wall scan coverage ~ 80% University of Michigan Gamma floor scan coverage ~ 80% Ann Arbor, Michigan Gamma lower wall scan coverage ~ 50%

1st Floor Scan Coverage Fig. A-3. Ford Nuclear Reactor, First Floor Scan Coverage Ford Reactor Survey Report A-3 5176-SR-01-1

Figure provided by University of Michigan Ford Nuclear Reactor Scan Coverage Percentage University of Michigan Beta floor and exhaust plenum scan coverage ~ 80%

Gamma floor scan coverage ~ 80% Ann Arbor, Michigan 2nd Floor Scan Coverage Fig. A-4. Ford Nuclear Reactor, Second Floor Scan Coverage Ford Reactor Survey Report A-4 5176-SR-01-1

Figure provided by University of Michigan Scan Coverage Percentage Confirmatory Unit 3 (CU3) consists of Survey Units 3-3, 3-4, 3-5.

Ford Nuclear Reactor 3-6, 3-7 and 3-8 (in Rooms 3103, 3104 and 3105J)

Beta floor and lower wall scan coverage ~ 80% University of Michigan Gamma floor scan coverage ~ 80% Ann Arbor, Michigan Gamma lower wall scan coverage ~ 50%

Remaining Areas Beta floor and lower wall scan coverage ~ 80% 3rd Floor Scan Coverage Gamma floor scan coverage ~ 80%

Gamma lower wall scan coverage ~ 50%

Fig. A-5. Ford Nuclear Reactor, Third Floor Scan Coverage Ford Reactor Survey Report A-5 5176-SR-01-1

Figure provided by University of Michigan Ford Nuclear Reactor Scan Coverage Percentage University of Michigan Cooling Tower Beta floor and tresses scan coverage ~ 80% Ann Arbor, Michigan Gamma floor scan coverage ~ 80%

Gamma tresses scan coverage ~ 50%

Cooling Tower Scan Coverage Fig. A-6. Ford Nuclear Reactor, Cooling Tower Scan Coverage Ford Reactor Survey Report A-6 5176-SR-01-1

Figure provided by University of Michigan Direct Measurement and Soil Sample Locations

  1. Direct Measurement/#Smear - Lower Walls and Floor Ford Nuclear Reactor
  1. Direct Measurement/#Smear - Upper Walls and Ceiling University of Michigan S000# Soil Sample Location Ann Arbor, Michigan CU 1 Direct Measurement and Soil Sample Locations Fig. A-7. Ford Nuclear Reactor, Confirmatory Unit 1 Direct Measurement and Soil Sample Locations Ford Reactor Survey Report A-7 5176-SR-01-1

Figure provided by University of Michigan Direct Measurement Locations

  1. Direct Measurement/#Smear - Lower Walls and Floor Ford Nuclear Reactor
  1. Direct Measurement/#Smear - Upper Walls and Ceiling University of Michigan Ann Arbor, Michigan CU 2 Direct Measurement Locations Fig. A-8. Ford Nuclear Reactor, Confirmatory Unit 2 Direct Measurement Locations Ford Reactor Survey Report A-8 5176-SR-01-1

Figure provided by University of Michigan Direct Measurement Locations

  1. Direct Measurement/#Smear - Lower Walls and Floor Ford Nuclear Reactor
  1. Direct Measurement/#Smear - Upper Walls and Ceiling University of Michigan Ann Arbor, Michigan CU 3 Direct Measurement Locations Fig. A-9. Ford Nuclear Reactor, Confirmatory Unit 3 Direct Measurement Locations Ford Reactor Survey Report A-9 5176-SR-01-1

APPENDIX B SCAN RESULTS Ford Reactor Survey Report 5176-SR-01-1

Confirmatory Unit 1 Beta Scan Summary Statistics Surface Min Max Mean Median SD Walls 57 654 328 325 78.86 Ceiling 138 568 326 321 79.48 Floor 732 1,975 1,343 1,351 179.7 Confirmatory Unit 1 Gamma Scan Summary Statistics Surface Min Max Mean Median SD Walls 6,725 11,647 9,393 9,360 700.1 Floor 7,545 11,269 9,497 9,467 675.2 Ford Reactor Survey Report B-1 5176-SR-01-1

Confirmatory Unit 2 Beta Scan Summary Statistics Surface Min Max Mean Median SD Walls 125 599 327.8 324.5 74.57 Floor 1,057 1,932 1,476 1,463 164.9 Confirmatory Unit 2 Gamma Scan Summary Statistics Surface Min Max Mean Median SD Walls 4,998 12,803 9,353 9,578 1,231 Floor 5,785 11,976 9,416 9,677 1,349 Ford Reactor Survey Report B-2 5176-SR-01-1

Confirmatory Unit 3 Beta Scan Summary Statistics Surface Min Max Mean Median SD Walls 3 768 422.4 397 105.9 Floor 167 2,133 1,592 1,590 186.3 Confirmatory Unit 3 Gamma Scan Summary Statistics Surface Min Max Mean Median SD Walls 8,604 16,462 11,739 11,029 1,851 Floor 9,019 15,255 11,169 10,909 1,203 Ford Reactor Survey Report B-3 5176-SR-01-1

First Floor Beta Scan Summary Statistics Surface Min Max Mean Median SD Walls 110 805 329.9 323 85.14 Floor 732 1,975 1,372 1,388 176.7 First Floor Gamma Scan Summary Statistics Surface Min Max Mean Median SD Walls 5,354 17,795 8,623 8,816 1,734 Floor 6,404 28,393 10,753 9,694 3,331 Ford Reactor Survey Report B-4 5176-SR-01-1

Rm 2111 Beta Scan Summary Statistics Surface Min Max Mean Median SD Plenum 92 619 288 282 77.5 Floor 19 1,751 1,327 1,322 146.6 Ford Reactor Survey Report B-5 5176-SR-01-1

Third Floor Beta Scan Summary Statistics Surface Min Max Mean Median SD Walls 252 586 347.1 344 45.63 Floor 616 2,196 1,517 1,531 158.5 Third Floor Gamma Scan Summary Statistics Surface Min Max Mean Median SD Walls 7,862 12,289 9,732 9,576 969.1 Floor 1,054 12,252 9,013 9,171 1,300 Ford Reactor Survey Report B-6 5176-SR-01-1

Cooling Tower Beta Scan Summary Statisticsa Surface Min Max Mean Median SD Walls and Floor 0 601 243.8 247 109.4 aTechnician also performed gamma scans of this area but due to an electronic issue the data was not recorded. No contamination was identified by the technician.

Scans could not be redone due to time constraints.

Ford Reactor Survey Report B-7 5176-SR-01-1

Pool Beta Scan Summary Statistics Surface Min Max Mean Median SD Pool 57 562 303.1 305 83.66 Pool Gamma Scan Summary Statistics Surface Min Max Mean Median SD Pool 5,262 10,337 8,142 8,353 1,200 Ford Reactor Survey Report B-8 5176-SR-01-1

APPENDIX C TABLES Ford Reactor Survey Report 5176-SR-01-1

Table C-1. Confirmatory Unit 1 Surface Activity Total Surface Location Surface Gross Activity (dpm/100 cm2) 1 Ceiling 321 -200 2 Floor 380 130 3 Floor 376 110 4 Ceiling 335 -120 5 Floor 405 270a 6 Ceiling 343 -79 7 Ceiling 357 0 8 Floor 385 160 9 Wall 3 375 100 10 Floor 374 96 11 Ceiling 341 -91 12 Ceiling 365 45 13 Ceiling 390 190 14 Wall 2 291 -370 15 Ceiling 338 -110 16 Floor 364 40 17 Wall 1 358 6 Mean 10 aRemovable surface activity was -2 dpm/100 cm2 Ford Reactor Survey Report C-1 5176-SR-01-1

Table C-2. Confirmatory Unit 2 Surface Activity Total Surface Location Surface Gross Activity (dpm/100 cm2) 1 Floor 435 300a 2 Floor 352 -170 3 Floor 360 -120 4 Floor 380 -11 5 Floor 400 100 6 Floor 327 -310 7 Floor 331 -290 8 Wall 4 341 -230 9 Wall 4 321 -350 10 Wall 4 333 -280 11 Wall 4 364 -100 12 Wall 3 342 -230 13 Wall 3 344 -220 14 Floor 368 -79 15 Wall 1 366 -91 16 Wall 3 376 -34 17 Floor 274 -610 Mean -160 aRemovable activity was -1 dpm/100 cm2 Ford Reactor Survey Report C-2 5176-SR-01-1

Table C-3. Confirmatory Unit 3 Surface Activity Surface Activity (dpm/100 cm2)

Location Surface Gross Total Removable 1 Ceiling 395 220 2 2 Ceiling 415 330 1 3 Wall 1 464 610 -3 4 Wall 3 373 91 a 5 Floor 393 200 6 Wall 3 376 110 7 Wall 1 372 85 8 Wall 2 381 140 9 Wall 4 386 160 10 Wall 1 409 290 -3 11 Floor 398 230 -4 12 Wall 3 333 -140 13 Floor 416 330 -1 14 Ceiling 412 310 2 15 Wall 4 406 280 -3 16 Wall 3 351 -34 17 Wall 2 680 -40 -1 Mean 187 aSmear sample for removable activity was not collected Ford Reactor Survey Report C-3 5176-SR-01-1

Table C-4. Inter-Laboratory Comparison of Soil Samples Sample ID Ag-108m (pCi/g) Ag-110m (pCi/g)

DERa DER ORAU UM ORAU UM ORAU UM 5176S0003 FNR-SOIL-2-007 0.02 +/- 0.01 b 0.0305 +/- 0.0146c 1.2 0.02 +/- 0.02 0.00928 +/- 0.0182 0.8 5176S0004 FNR-SOIL-2-008 0.92 +/- 0.07 0.889 +/- 0.0354 0.8 -0.02 +/- 0.03 0.00871 +/- 0.0289 1.4 5176S0005 FNR-SOIL-2-009 0.02 +/- 0.02 0.0169 +/- 0.00947 0.3 -0.02 +/- 0.03 0.00357 +/- 0.0111 1.4 5176S0006 FNR-SOIL-2-013 0.02 +/- 0.02 -0.00280 +/- 0.0166 1.7 -0.02 +/- 0.03 -0.00303 +/- 0.0196 0.9 5176S0007 FNR-SOIL-2-014 0.00 +/- 0.02 0.0182 +/- 0.0196 1.3 0.00 +/- 0.02 0.00580 +/- 0.0262 0.3 Sample ID Co-60 (pCi/g) Cs-137 (pCi/g)

DER DER ORAU UM ORAU UM ORAU UM 5176S0003 FNR-SOIL-2-007 0.03 +/- 0.02 0.0383 +/- 0.0252 0.5 0.00 +/- 0.02 -0.00607 +/- 0.0202 0.4 5176S0004 FNR-SOIL-2-008 0.37 +/- 0.04 0.313 +/- 0.0403 2.0 0.05 +/- 0.02 0.0123 +/- 0.0341 1.9 5176S0005 FNR-SOIL-2-009 1.79 +/- 0.12 1.54 +/- 0.0602 3.6 -0.01 +/- 0.03 0.000233 +/- 0.0125 0.6 5176S0006 FNR-SOIL-2-013 0.30 +/- 0.04 0.282 +/- 0.0383 0.6 0.00 +/- 0.02 -0.0146 +/- 0.0227 0.9 5176S0007 FNR-SOIL-2-014 0.45 +/- 0.05 0.374 +/- 0.0488 2.1 0.01 +/- 0.02 0.000816 +/- 0.0277 0.5 aDER = normalized absolute difference (refer to page 11 for explanation). A DER less than 3, indicates that the results are in agreement at the 99% confidence level.

bErrors represent the total propagated uncertainties at the 95% confidence level cErrors are reported at the 95% confidence level Ford Reactor Survey Report C-4 5176-SR-01-1

Table C-5. SOFs for Soil Samples Submitted for Inter-Laboratory Comparison Sample ID SOFa ORAU UM ORAU UM 5176S0003 FNR-SOIL-2-007 0.01 0.02 5176S0004 FNR-SOIL-2-008 0.21 0.19 5176S0005 FNR-SOIL-2-009 0.47 0.41 5176S0006 FNR-SOIL-2-013 0.08 0.07 5176S0007 FNR-SOIL-2-014 0.12 0.10 aNegative values were not included in the SOF calculation Ford Reactor Survey Report C-5 5176-SR-01-1

APPENDIX D MAJOR INSTRUMENTATION Ford Reactor Survey Report 5176-SR-01-1

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 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)

Ford Reactor Survey Report D-1 5176-SR-01-1

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)

Ford Reactor Survey Report D-2 5176-SR-01-1

APPENDIX E SURVEY AND ANALYTICAL PROCEDURES Ford Reactor Survey Report 5176-SR-01-1

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 2012c; 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 were traceable to National Institute of Standards and Technology (NIST).

Field survey activities were conducted in accordance with procedures from the following Independent Environmental Assessment and Verification Program documents:

  • Survey Procedures Manual (ORAU/ORISE 2013a)
  • Laboratory Procedures Manual (ORAU/ORISE 2013b)
  • Quality Program Manual (ORAU 2012b)

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

Quality control procedures include:

  • Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations
  • Training and certification of all individuals performing procedures
  • Periodic internal audits Ford Reactor Survey Report E-1 5176-SR-01-1

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 and CsI) 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 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 1998). 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 was 0.14. The average concrete background for the detectors was around 390 counts per minute (cpm). The minimum detectable count rate (MDCR) and scan MDC was calculated as:

Bi = (390)(2 s)(1 min/60 s) = 13 counts MDCR = (1.96)(13 counts)1/2[(60 s/min)/2s] = 212 cpm MDCRsurveyor = 212/(0.5)1/2 = 300 cpm Scan MDC = (300)/(0.14*1.26) = 1,700 dpm/100 cm 2 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 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 0.14 (calibrated with Tc-99). ORAU determined Ford Reactor Survey Report E-2 5176-SR-01-1

construction material-specific background for each surface type encountered for determining net count rates. However, the material-specific background was used for the sole purpose of determining an a prior static MDC. 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 FNR was 540 dpm/100 cm2.

E.3.3 SOIL SAMPLING 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 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.

Ford Reactor Survey Report E-3 5176-SR-01-1

Radionuclide TAPa (MeV) MDC (pCi/g)

Co-60 1.173 0.06 Cs-137 0.662 0.05 Ag-108m 0.434 0.04 Ag-110m 0.658 0.04 aSpectra were also reviewed for other identifiable TAPs that would not be expected at this site.

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.

Ford Reactor Survey Report E-4 5176-SR-01-1

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