ML11346A300

Final Status Survey Report and Request for License Termination for the University of Arizona Research Reactor

ML11346A300
Person / Time
Site:	05000113
Issue date:	12/01/2011
From:	Offerle R Univ of Arizona
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
UA-MCP-FS-01, Rev 0
Download: ML11346A300 (91)

Text

Lngineering Building (20)

'I THE UNIVERSITY P. O. Box 210020 Tucson, Arizona 85721-0020

. OF ARIZONA. Tel: (520) 621-6205 Fax: (520) 621-8096 Nuclear Reactor Laboratory hrtp://reactor.arl.arizona.edu December 1, 2011 U.S. Nuclear Regulatory Commission 10 CFR 50.82(b)

ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Final Status Survey Report and Request for License Termination for the University of Arizona Research Reactor, Facility License No. R-52, Docket No. 50-113 Gentlemen:

This letter provides our final status survey report. Upon review and agreement by the U.S.

NRC, we ask you specify our former nuclear reactor site suitable for unrestricted release and that you terminate our reactor license.

ENERCON Services and NRC/ORISE surveyed our Nuclear Reactor Laboratory rooms, the Engineering Building, and the campus surroundings in August and September, respectively.

Based upon the data provided by the enclosure, the University of Arizona meets the radiological criteria for license termination.

Cordially, Robert A. Offer&

Acting Director Enclosure University of Arizona Final Status Survey Report Copy to:

Mr. John B. Hickman, NRC FSME/DWMEP/DURLD U.S. NRC Region IV Dr. Leslie Tolbert, Vice President for Research Director, Arizona Research Laboratories Director, Office of Radiation, Chemical & Biological Safety University of Arizona Research Reactor

0 ENERCON FINAL STATUS SURVEY REPORT University of Arizona Nuclear Reactor Laboratory Nuclear Regulatory Commission Facility Operating License R-52 UA-MCP-FS-01 Revision 0 November 23, 2011 Prepared by LVI Environmental Services, Inc.

12 Oak Drive Shawnee, OK 74801 And Enercon Services, Inc.

4490 Old William Penn Highway Murrysville, PA 15668 Prepared for The University of Arizona Decommissioning of the Nuclear Reactor Laboratory LVI Project Number 341015

Wif S7RVCES FJ University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report FINAL STATUS SURVEY REPORT University of Arizona Nuclear Reactor Laboratory Nuclear Regulatory Commission Facility Operating License R-52 UA-MCP-FS-01 Revision 0 November 23, 2011 Prepared by:

Dustin G. Miller Certified Health Physicist Reviewed by:

Kevin E. Taylor Certified Health Physicist Approved by:

Corey E.DeWitt Project Manager

(§d14 -

Approved by:

Robert Offrerle Director. Nuclear Reactor Labora/

Approved by:

Daniel Silvain Chairman, Reactor Committee

University of Arizona nI SERWCES 0 Nuclear Reactor Lab D&D Final Status Survey Report

SUMMARY

OF CHANGES Revisions to this report will be tracked when revisions are issued. Changed sections will be identified by special demarcation in the margin. A summary description of each revision will be noted in the following table.

Revision Date Description of Change Number 0 November 23, 2011 Initial issue

AM. University of Arizona Nuclear Reactor Lab D&D SERVICES Final Status Survey Report TABLE OF CONTENTS ACRO NY M S AND ABBREVIA TIO N S .................................................................................... iii 1.0 EXECUTIVE SU M M A RY ............................................................................................... 1 2.0 Introduction ........................................................................................................................ 2 2.1 Purpose and Objective .......................................................................................... 2 2.2 Project Background ................................................................................................ 2 2.3 Decom m issioning Activities .................................................................................. 4 3.0 FINAL STATUS SURVEY METHODOLOGY ........................................................ 8 3.1 Release Criteria ...................................................................................................... 8 3.2 Classification and Sam ple Size ............................................................................. 10 3.2.1 Sam ple and Location Identification .......................................................... 11 3.3 Types and M ethods of Surveys ............................................................................. 11 3.4 Survey Instrum entation ......................................................................................... 12 3.4.1 Instrum ent M odels .................................................................................... 12 3.4.2 Instrum ent Calibration ............................................................................... 13 3.4.3 Pre-Operational Checks ............................................................................. 13 3.4.4 Instrum ent Efficiency ................................................................................. 14 3.4.5 M inimum Detectable Concentration ........................................................ 14 4.0 FINAL STATUS SURVEY RESULTS .................................................................... 17 4.1 Reactor Tank ........................................................................................................ 17 4.1.1 SteelTank Liner ........................................................................................ 17 4.1.2 Concrete Portion of the Reactor Tank ...................................................... 18 4.2 Building Surfaces .................................................................................................. 20 4.2.1 Reactor Control Room ................................................................................ 20 4.2.2 Reactor Room ........................................................................................... 21 4.2.3 Equipm ent Storage Room ........................................................................ 21 4.2.4 Second Floor Storage Room ...................................................................... 22 4.3 Outside Concrete Pads ........................................................................................ 23 4.4 Storage Pit Results and Inventory ......................................................................... 23 4.5 M iscellaneous Equipm ent .................................................................................... 24 4.6 Sink Drain ................................................................................................................. 24 5.0 CONCLUSIO N S ........................................................................................................ 25

6.0 REFERENCES

................................................................................................................. 26 i

V SERTCES O0 Nuclear University of Arizona Final Status Reactor SurveyLabReport D&D LIST OF TABLES Table 3-1: NRC License Termination Screening Levels for Surfaces ........................... 9 Table 3-2: NRC License Termination Screening Levels for Soils ................................ 9 Table 3-3: Survey Design Param eters ............................................................................ II Table 3-4: Survey Instrum ent M DCs ............................................................................. 16 Table 4-1: Static Measurement Results - Reactor Tank Liner ..................................... 18 Table 4-2: Concrete Analysis Results Summary .......................................................... 19 Table 4-3: Concrete Analysis Results per Sample ........................................................ 20 Table 4-4: Static Measurement Results - Reactor Control Room ............................... 20 Table 4-5: Static Measurement Results - Reactor Room ............................................ 21 Table 4-6: Static Measurement Results - Equipment Storage Room .......................... 22 Table 4-7: Static Measurement Results - Second Floor Storage Room ....................... 23 Table 4-8: Static Measurement Results - Storage Pits ................................................. 24 LIST OF APPENDICES APPENDIX A - Technical Description of Instruments APPENDIX B - Calibration Records APPENDIX C - Daily Source Checks APPENDIX D - Radiological Survey Forms APPENDIX E - Laboratory Data Package ii

University of Arizona SERVICES 0 Nuclear Reactor Lab D&D Final Status Survey Report LIST OF ACRONYMS AND ABBREVIATIONS WtR/hr microRoentgen per hour ALARA As Low As Reasonably Achievable cpm Counts per minute D&D Decontamination and Decommissioning DOE Department of Energy DP Decommissioning Plan dpm Disintegrations per minute Ei Instrument efficiency EPA US Environmental Protection Agency Es Surface efficiency Et Total instrument efficiency FSS Final Status Survey HEPA High Efficiency Particulate Air MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC Minimum Detectable Concentration MDCR Minimum Detectable Count Rate MOU Memorandum of Understanding Nal Sodium Iodide NRC Nuclear Regulatory Commission NRL Nuclear Reactor Laboratory NORM Naturally Occurring Radioactive Materials ORISE Oak Ridge Institute for Science and Education OSRP Off-Site Source Recovery Project pCi/g picocuries per gram RCO Radiation Control Office SCA Single Channel Analyzer SOF Sum of Fractions TEDE Total Effective Dose Estimate UA University of Arizona VSP Visual Sample Plan iii

LVf SERICES 1O University of Arizona Nuclear Reactor Final Status LabReport Survey D&D 1.0 EXECUTIVE

SUMMARY

The University of Arizona (UA) Research Reactor was a TRIGA pool-type reactor designed and constructed by General Atomic Division of General Dynamics Corporation. The reactor operated under US Nuclear Regulatory Commission (NRC) license R-52 and was located within the Nuclear Reactor Laboratory (NRL) on the 391-acre campus of the University of Arizona in Pima County Arizona in the city of Tucson.

The University ceased operation of the facility on May 18, 2010 and the reactor fuel was removed by the Department of Energy on December 23, 2010, with the fuel being delivered to Idaho National Laboratories. The NRL underwent decommissioning activities from May 2011 through September 2011 followed by Final Status Surveys (FSS) using calibrated instruments to measure Total (Static) Beta activity and to perform radiological scan measurements. Smears were also collected for tritium and beta at every measurement location plus several items based on professional judgment. A total of 142 measurement locations and six concrete sampling locations were used to assess the final radiological status of the facility.

All Total Beta measurements were less that the release criteria for the most restrictive isotope (Co-60 at 7,100 disintegrations per minute per 100 square centimeters [dpm/100cm 2]) and all scan measurements were less than the investigation levels. The maximum result surveyed on a building surface was 4,204 dpm/100cm 2 on a brick wall of the Equipment Storage Room (Room 216); however, material specific background radiation levels were not subtracted from the gross measurements. It was evident in the survey that the NRL's exterior brick walls contained detectable levels of Naturally Occurring Radioactive Materials (NORM).

Based on the results described in Section 4.0 of this FSS report, the NRL meets the requirements for unrestricted release specified in 10CFR20, Subpart E, Radiological Criteria for License Termination, by meeting the UA FSS Plan, UA-MCP-FS-001 and the UA Decommissioning Plan.

I

University of Arizona V

SERVICES O Nuclear Reactor Final Status LabReport Survey D&D 2.0 Introduction 2.1 Purpose and Objective This report describes the purpose, scope and technical approach used during the FSS of the four rooms and the reactor pit at the UA NRL.

The purpose of the FSS activities is to demonstrate that the radiological conditions at the UA NRL satisfy the release criteria and the NRC license may be terminated. For the purpose of this demonstration, each survey unit is independently evaluated. The objective of the survey is to demonstrate at a 95% minimum level of confidence that the radiological release conditions have been met.

2.2 Project Background The UA research reactor was a TRIGA pool-type reactor designed and constructed by General Atomic Division of General Dynamics Corporation. The reactor was located within the NRL on the 391-acre campus of the University of Arizona in Pima County Arizona in the city of Tucson as shown in Figure 1. The University is about 65 miles north of the Mexican Border at Nogales, Arizona, 110 miles south east of Phoenix, Arizona and 120 miles from the western border of New Mexico. The campus is centrally located in the city of Tucson, and is bounded roughly by East Speedway Boulevard, North Campbell Avenue, East Sixth Street, and North Park Avenue.

2

University of Arizona SERVCES 0 Nuclear Reactor Lab D&D Final Status Survey Report Figure 1: University of Arizona, Nuclear Reactor Laboratory Location 4

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The reactor was constructed in 1958 and went into operation on December 6 of that year. The licensed power was 10 kW thermal with operations of 30 kW possible for short times. The original core loading consisted of 61 aluminum-clad fuel elements and two additional aluminum-clad fuel elements were obtained. The stainless steel-clad fuel replaced the aluminum clad fuel in1973. The facility license was amended to allow operations to 1llkW and pulses to 1,100 MW. There was no history of major accidents or spills. In 1976, radiation monitors indicated a leaking fuel element. The damaged element was identified as Element # 4058.

Operators removed # 4058 from service and isolated it in the reactor pool. No additional release of fission products or nuclear fuel from this fuel element or any other element occurred.

The NRL is located on the first floor of the north wing of the Engineering Building (Building 20) on the University's Main Campus as shown in Figure 2. The reactor core, defueled in December 2010, was located in a reactor tank that was 21 feet deep and 6.5 feet in diameter, located below grade in Room 124. The tank contained approximately 5,000 gallons of demineralized water.

Three rooms, shown in Figure 2, in the Engineering Building were established as the Nuclear Reactor Laboratory and are designated a controlled access area. These are: Room 122, the 3

University of Arizona SERMET 0 Nuclear Reactor Lab D&D Final Status Survey Report Reactor Control Room; Room 124, the Reactor Room; and Room 124A, Equipment Storage and Experiment Setup Room. Room 216, the Second Floor Storage Room located directly above Room 124, was also designated a controlled access area.

Figure 2: NRL Room Drawing ROOM 164 NUCLEAR CHEMISTRY LABORATORY ROOýM 163 2.3 Decommissioning Activities During the calendar year 2011, the UA NRL underwent full decontamination and decommissioning (D&D) activities to support the termination of NRC license R-52. D&D activities were conducted by LVI Services, Inc. (LVI) and its partner Enercon Services, Inc.

(ENERCON) in accordance with the NRC approved UA NRL DP (ENERCON, 2009). The LVI team began project planning in January 2011 with the development of all plans and procedures needed to perform the decommissioning activities. Following owner review and approval of the plans and procedures, the LVI team mobilized personnel to prepare the site for D&D activities.

4

University of Arizona SERVCES MO Nuclear Reactor Lab D&D Final Status Survey Report Radiological D&D activities began in early May 2011 and were conducted through the summer of 2011.

The D&D activities consisted of disposing or free releasing miscellaneous equipment from the NRL, removing the control rods, drives, bridge, cooling coils, thermalizer and reactor assembly from the reactor tank, draining the tank water, removing the gunite layer from the reactor tank and then removing the activated sections of reactor tank steel liner and concrete. The Filtration/

Ion Exchange System, Cooling Compressor and associated piping were removed and disposed based upon survey results. After all items were removed from the pool, the water was analyzed by the UA Radiation Control Office (RCO) with the results less than NRL licensed release limits. The water was released to the sanitary sewer via the authorized release point in Room 124A.

Following removal of reactor internals and pool water, a concrete core drill was advanced through the sidewalls and the tank floor to collect samples for analysis to investigate the depth of activation in the concrete reactor tank. The core drill was advanced horizontally through the gunite, steel, and concrete at three locations at the core mid plan and at one location vertically downward at the axial centerline of the reactor tank. Two additional cores were advanced at an elevation 5.5 feet above the reactor tank bottom where surface radiological activity measurements were background. Each core was then segmented by material (i.e. gunite, steel, concrete) and then the concrete was cut into 2-inch segments. Some of the gunite, steel, and concrete segments were initially screened to the UA RCO to be analyzed by gamma spectroscopy. Some samples were sent to the offsite radioanalytical laboratory, Teledyne Brown Engineering, in Knoxville, TN, for gamma spectroscopy and hard to detect radiological analysis.

The analysis of the cores revealed the following three items that resulted in modifications to the decommissioning activities and the original FSS Plan:

" On average, the concrete tank walls were two to three times thicker than indicated in the design drawings.

• Low but detectable levels of neutron activation products, i.e. Cobalt-60 and Europium-152 were identified horizontally outward in concrete at the mid plane but not in the 5

SERVICESw LW O University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report surrounding soil. The only concrete with concentrations of radionuclides above the screening values were located in a small area directly behind the reactor's thermal column. Cobalt-60 or Europium-152 was not identified at the coring locations 5.5 feet above the reactor floor.

The only hard to detect radionuclide identified in the concrete samples was tritium.

The increase in concrete tank thickness raised concerns over the safety of activities required to remove all lower tank concrete to background levels as originally planned in the DP (ENERCON 2009). After discussions with the NRC, the University submitted a request for revision to the DP to apply the NRC Soil Screening criteria in NUREG 1757, Volume 1, Appendix B, Table B.2 to residual concrete in the reactor tank. Based upon the analytical results of tank concrete core samples, it was believed that the concrete remediation in the reactor tank could be conducted to meet these soil screening criteria without requiring a structural assessment and reinforcement of the residual reactor tank.

The gunite lining the reactor tank wall, also thicker than expected, was removed from all surfaces of the reactor tank. During removal of the gunite, LVI/ENERCON used personal respirators with High Efficiency Particulate Air (HEPA) filters as well as negative air machines with HEPA filters to mitigate the spread of airborne activity into Room 124.

An oxyacetylene torch was used to cut away all of the inner steel liner located in the bottom of the reactor tank. This included all steel below 5.5 foot elevation core locations where Cobalt-60 and Europium-152 activity were not identified. All this steel was sized and packaged as radiological waste.

On the west side of the reactor pit, we found the builders had installed a 6 foot wide by 2 foot high by 2 foot deep rectangular metal box in the wall of the reactor tank behind the thermalizer block. In the design drawings, the area was labeled a "potential" experimental facility; however, historical records indicated it had never been built. The rectangular box was not accessible and was filled with red clay brick. The brick and metal were removed and disposed of with the rest of the activated concrete and steel tank liners as radiological waste.

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University of Arizona L VN SERVICES 1O Nuclear Reactor Lab D&D Final Status Survey Report The concrete in the area behind the thermal column was removed up to 12 inches in depth to meet the NRC soil screening levels. Other core samples in the bottom section of the reactor pit indicated residual radioactive material concentrations in the concrete were less than the screening levels. Additional concrete, two to four inches in depth, was removed in the potential neutron flux areas (i.e., the pit floor and core mid-plane) for As Low As Reasonably Achievable (ALARA) purposes.

Approximately 13 cubic yards of concrete material and gunite were removed from reactor tank.

This and all other radiological wastes generated during the D&D project were packaged and transported to appropriately licensed disposal facilities upon completion of D&D activities. This includes the Americium-Beryllium start up source which was transferred to NSSI facility in Houston, Texas through the DOE's Off-Site Source Recovery Project (ORSP).

Following the removal of all radiological wastes, the FSS was conducted in the four controlled NRL rooms and inside the reactor pit using the methodology described in the following section.

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SERWCES MO University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report 3.0 FINAL STATUS SURVEY METHODOLOGY The FSS provided data and information to demonstrate that the NRL facility and site complies with the NRC annual dose limit criteria of 25-mrem/yr Total Effective Dose Estimate (TEDE),

by meeting the radiological criteria listed in the DP approved for use by the NRC in Technical Specifications Amendment 20 for the UA NRL Possession-Only License R-52.

The FSS was designed in accordance with the NRC's NUREG 1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), and the NRC-approved FSS Plan, UA-MCP-FS-01, in order to demonstrate that the NRL facility and site comply with the radiological release criteria stated in Section 3.1. The FSS Plan described the process for survey preparation, survey design, data collection, data evaluation, and documentation of survey results The FSS included the walls, ceiling, and floors of the four rooms of the NRL; residual surfaces of the reactor tank liner; the concrete pad beneath the cooling system; the concrete pad beneath the Ion Exchange/filtration system; and the exposed concrete at the bottom of the reactor pit and fuel storage tubes.

3.1 Release Criteria For the UA FSS, the NRC License Termination Screening Levels listed in NUREG 1757, Volume I, Appendix B (Tables B. 1 and B.2) were approved as the release criteria for structure surfaces (Table B. 1) and soil/concrete (Table B.2). However, in Table B.2, the Eu- 152 screening value is greater than the US Environmental Protection Agency (EPA) and NRC Memorandum of Understanding (MOU) consultation trigger of 7.0 picocuries per gram (pCi/g). Therefore, the Eu-152 screening level was reduced to the value contained in Table 1, Consultation Triggers for Residential and Commercial/Industrial Soil Contamination, of the EPA/NRC MOU. The NRC-approved screening criteria for UA NRL radionuclides of concern are listed in Table 3-1 for building surfaces and Table 3-2 for soils and concrete.

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0 University of Arizona Nuclear Reactor Lab D&D SERVICES Final Status Survey Report Table 3-1:

NRC License Termination Screening Levels for Surfaces Radionuclide Acceptable Screening Levels for Unrestricted Release (dpm/1 00cm 2)

Tritium (H-3) 1.2E+08 Carbon- 14 3.7E+06 Manganese-54 3.2E+04 lron-55 4.5E+06 Cobalt-60 7.1E+03 Nickel-63 1.8E+06 Technetium-99 1.3E+06 Cesium- 137 2.8E+04 Generally, to perform surveys that address multiple radionuclides, the expected distribution of the radionuclides must be known. However, characterization data for walls and surfaces at the UA NRL showed no contamination on the walls and surfaces which would support the derivation of a radionuclide distribution Therefore, a radionuclide distribution was not calculated and, as a conservative assumption, all detectable beta/gamma surface activity was assumed to be Co-60 since it has the lowest most conservative screening level of 7,100 dpm/1 00cm2 .

Table 3-2: NRC License Termination Screening Levels for Soils Radionuclide Default DCGL (pCi/g)

Cobalt-60 3.8 Tritium (H-3) 110 Europium- 152 7*

Europium- 154 8

• Adjusted to NRC/EPA MOU value 9

University of Arizona L V SERVICES 1O Nuclear Reactor Final Status LabReport Survey D&D Compliance with the volumetric contamination screening criteria was demonstrated with a combination of surveys and volumetric materials sampling and analysis.

3.2 Classification and Sample Size Area classification in accordance with MARSSIM protocols ensures that the number of measurements and the scan coverage is commensurate with the potential for residual contamination to exceed the approved release criteria. The classification of an area is based on characterization measurements and historical use of the area. Areas that have no reasonable potential for residual contamination because there was no known impact from site operations are classified as non-impacted areas. Non-impacted areas are not required to be surveyed beyond what is completed as a part of site characterization to confirm the area's non-impacted classification.

Areas that may contain residual radioactivity from licensed activities are considered impacted areas. Based on the levels of residual radioactivity present, impacted areas are further divided into Class 1, Class 2 or Class 3 designations. Class I areas have the greatest potential for residual activity while Class 3 areas have the least potential for residual contamination. Each classification will typically be bounded by areas classified one step lower to provide a buffer zone around the higher class. Exceptions occur when an area is surrounded by a significant physical barrier that would make transport of residual activity unlikely from one area to the adjacent area. In such cases, each area will be classified solely on its own merit using the most reliable information available.

Table 3-3 lists the survey units and classification for the areas of the NRL included in the FSS.

The Reactor Pit and storage tubes were considered Class 1. Class 2 areas included the lower surfaces of the Reactor Room and the Equipment Storage Room, Room 124 and Room 124A respectively. Class 3 areas included the upper surfaces of Rooms 124 and 124A, all of the Control Room, and the second floor storage area, Room 216. The building exterior and surrounding land area was considered non-impacted except for concrete surfaces beneath the filtration/ion exchange system and cooling compressor located outside and adjacent to the north wall of the building. The two exterior concrete pads were included in the survey of the reactor room lower surfaces.

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Lin 0 SERTiCES University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report Table 3-3: Survey Design Parameters MARSSIMCharacterization Data Survey Area Cassif Area Relative Number of (if) Mean Beta Std Dev Shift Locations Lower Surfaces (concrete) 1 29 N/A N/A 3 14 Reactor Pit Upper Surface (tank) 1 25 N/A N/A 3 14 Floor and Walls <2m Exterior concrete pads 2 102 253 1625 3 14 Room 124 Walls >2m and Ceiling 3 100 253 1625 3 14 Floor and Walls <2m 2 61 2293 742 3 14 Room 124A Walls >2m and Ceiling 3 45 2293 742 3 14 Room 122 All surfaces 3 107 320 438 3 14 Room 216 All surfaces 3 182 891 910 3 14 Note: Elevated readings in Room 124A were attributed to shine from four (4) tons of natural uranium contained in the Subcritical assembly.

3.2.1 Sample and Location Identification Measurement and sampling locations were plotted using Visual Sample Plan (VSP) and numbered sequentially using a two digit number, e.g., 01, 02, 03, and 04. The inputs for the VSP software are shown in Table 3-3 above based on the characterization data obtained in 2009 and 2010. The locations in the reactor pit were identified with "Pit" preceding the two digit number.

The locations outside of the reactor pit were numbered sequentially across the entire NRL.

Therefore, there are no duplicate measurement location numbers between survey units. In addition to the locations plotted by VSP, several locations were chosen by professional judgment. These locations are identified by a description of the item or area and annotated with "PJ" on the survey form.

3.3 Types and Methods of Surveys Survey measurements and sample collection were performed by personnel trained and qualified in accordance with the FSS Plan (ENERCON 2011). FSS measurements included surface scans, direct surface measurements, removable radionuclide measurements (beta and tritium), and 11

University of Arizona Nuclear Reactor Lab D&D SERWCES Final Status Survey Report gamma spectroscopy of volumetric material samples. On-site and off-site lab facilities were used for gamma spectroscopy and liquid scintillation counting.

3.4 Survey Instrumentation Instruments were selected that provided an adequate response to radionuclides of concern and capable of having a minimum detectable concentration (MDC) less than 50% of the applicable release criteria for the survey. MDC was calculated for each model. The MDC was calculated using formulae contained in MARSSIM and shown in Section 5.2.3. The MARSSIM contains the most current guidance on recommended methods of calculating survey instrument MDC.

The background rate and detector efficiency used in the MDC are averages calculated from the daily operational check of each instrument. Table 3-4 shows the instrument models selected and summarizes the typical observed background counts in counts per minute (cpm), typical observed detector efficiency, and a typical observed MDC in dpmr/100 cm 2 for each instrument model. Other objectives in selecting instruments included special features such as digital displays to provide a more accurate reading than conventional analog displays.

3.4.1 Instrument Models The instruments selected for this survey are described in the following sections. Technical description data sheets for each instrument and detector used are provided in Appendix A.

3.4.1.1 Ludlum Model 2221 with 43-68 Gas Proportional Detector This is a scaler/ratemeter instrument with gas proportional detector that has an active detection 2

area of 126 cm2. This combination was used for all static and scan measurements.

3.4.1.2 Ludlum Model 2221 with 44-9 "Pancake" GM Detector This is a scaler/ratemeter instrument used for beta/gamma static and scan measurements in areas inaccessible to the gas proportional detector. The detector has an active area of 15 cm 2 .

3.4.1.3 Ludlum Model 2221 with Model 44-10 Sodium Iodide (Nal) 2"x2" Detector This is a scaler/ratemeter instrument was used for gamma radiation scans. The detector is a 2"x2" Nal detector with a sensitivity of approximately 900 cpm per microRoentgen per hour (p[R/hr) for Cs-137.

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University of Arizona L V I SERVICES 10Nuclear Reactor Lab D&D Final Status Survey Report 3.4.1.4 Packard Tricarb Liquid Scintillation Counter This instrument, owned and operated by the UA RCO, was used for the analysis of tritium smears.

3.4.1.5 Ludlum Model 19 MicroR This is a ratemeter instrument that was used for exposure rate surveys. It is generally accepted throughout the nuclear industry that the MDC of this instrument is equivalent to the background readings in gR/hr.

3.4.1.6 Ludlum Model 3030E with 43-10-1 Sample Counter This instrument is a dual channel scaler with a Model 43-10-1 dual phosphor detector sample tray for alpha and beta smear counting.

3.4.2 Instrument Calibration All field instrumentation was calibrated by MJW Technical Services located in Olean, NY. The survey instruments were calibrated prior to initiation of the FSS. ENERCON maintains original instrument calibration data and certificates for the calibration sources at the calibration lab and provided copies to file on-site. Appendix B provides the calibration records for the instruments used.

3.4.3 Pre-Operational Checks All background readings for counting instruments were conducted daily during instrument use.

Any instrument that used a battery had a daily battery check performed before instrument operation. An appropriate initial source response, as determined by a comparison to calibration information, to the appropriate on-site check source was obtained prior to the instrument being placed in service for the project.

Each day an instrument was used, the instrument received an operational check that consisted of a background reading (observed cpm for ratemeters or 1-minute static count for scalers) and a count of a known check source (observed cpm for ratemeters; 1-minute static count for scalers) with the detector in contact with the source in its holder to provide a consistent source to detector distance. If at any time an instrument exceeded its calibration due date or fell outside of the 13

L VI University of Arizona SERWCES 10Nuclear Reactor Final Status LabReport Survey D&D

+/-10 percent of the established initial source check, the instrument was recalibrated, and a new initial source check was performed. This information was recorded on a Daily Source Check Log form. Appendix C provides the results of the source check measurements made for instruments used during the FSS.

3.4.4 Instrument Efficiency The efficiency of the detection capabilities for each applicable instrument was calculated using the following formula:

Ei = (C%- CO)

Instrument Efficiency (EB) S Where:

C, = Measured source count in cpm Cb = Measured background count in cpm S = 2 pi source activity value in dpm Total Instrument Efficiency (Et) E, = E,

• E, Where:

Ei = Instrument efficiency Es = Surface efficiency (0.25 for beta emitters < 400 keV f3ma)

The total instrument efficiency (Et) is a product of the instrument efficiency (E1 ) and the surface efficiency (E,) and is used to convert the raw instrument counts into the standardized unit of dpm. The surface efficiency utilized follows the recommendations in ISO-7503-1, which makes recommendations for default surface efficiencies. A surface efficiency of 0.25 is recommended for beta emitters with maximum energies less than 400 keV (NUREG 1507); therefore, the surface efficiency used for the FSS was 0.25 based a beta energy maximum 318 keV for Co-60.

3.4.5 Minimum Detectable Concentration The instrument and detector combination selected for the wall and floor scans had a MDC well below the release criteria specified in Section 3.1 as shown in Table 3-4. The MDC is the 14

University of Arizona L V SERWTCES IO1Nuclear Reactor Final Status LabReport Survey D&D concentration of radioactivity that an instrument can be expected to detect at a 95 percent confidence level. For instruments performing direct and/or scan measurements, the MDC goal was 10 to 50% of the applicable release criteria. The MDC goals were met for the FSS instruments and information used to calculate the MDC for instrumentation used during the FSS were documented on the radiation survey form.

For static (direct) surface measurements, with conventional detectors, the MDC was calculated using the formula:

MDC (dpm/IlOOcm 2) = [3+3.29(RbXTXl +T,/Tb)]

Variables:

Rh = Background count rate (cpm)

Tb = Background count time (min)

T,. = Sample Count Time (min)

E, = Total Instrument Efficiency (MARSSIM Section 6.6.1)

The MDCscan for beta-gamma measurements was calculated by determining the Minimum Detectable Count Rate (MDCR). The MDCR was determined by first determining the minimum detectable net source counts using Formula 6-8 in the MARSSIM as below.

Minimum number of detectable source counts: si =d'ýb-,

Where:

d' = value from MARSSIM Table 6.5 for applicable true and false positive rates (1.38 for a true positive proportion of 0.95 and a false positive proportion of 0.60) bi = Number of background counts in a given time interval The MDCR is calculated from Formula 6-9 in the MARSSIM:

,60 Minimum detectable count rate: MDCR = st *6 i

15

University of Arizona SERWCrS 0 Nuclear Reactor Lab D&D Final Status Survey Report Where:

i = Observed time interval (seconds) = the inverse of the detector scan speed in detectors-widths per second Finally, applying the detection efficiency correction resulted in an MDCscan in standardized units (dpm/1 00-cm2) from this formula:

MDCR Scan MDC: MDC,,,,, =

• probearea 100cm 2 Where:

Ei = Instrument efficiency E, = Surface efficiency 2

probearea = total area of the detector face in cm p = Surveyor efficiency (value from a range between 0.5 and 0.75)

Note: The value for p was developed in Draft NUREG/CR-6364 and NUREG -1507. It is a percentage estimate of the likelihood a surveyor will reliably detect an elevated count rate.

The a priori MDCs for beta/gamma radiation for the instruments used in the FSS are listed in Table 3-4 below and are recorded on the Radiological Survey forms in Appendix D. Scan speeds were equal to one detector-width per second and the surveyor efficiency was assumed to be 0.5. Background count times were 10 minutes, direct measurement count times were 30 seconds, and swipe sample count times were 1 minute.

Table 3-4: Survey Instrument MDCs Serial Probe 2-pi BKGD P/y MDC 0/y MDCscan Instrument and Detector Number Area Efficiency (cpm) (dpm/100 cm 2) (dpm/100 cm 2)

Ludlum Model 2221 with 178100 126 Model 43-68 Gas with cm2 25% 257 778 1539 Proportional Probe PR094759 Ludlum Model 2221 with 178100 Model HP-210 Geiger- with 029 15 cm 2 73% 192 1913 3816 Mueller Probe 16

University of Arizona LIN SERVICES 1Nuclear Reactor Final Status LabReport Survey D&D 4.0 FINAL STATUS SURVEY RESULTS The following sections describe the FSS survey techniques employed in each survey unit and are organized by survey area, i.e. room or reactor pit. The data collected in the FSS is summarized in each section with each radiological survey record provided in Appendix D. Data results were calculated with instrument background subtraction only. Material specific background, i.e.,

NORM in brick and concrete, were not subtracted from the gross results.

4.1 Reactor Tank The reactor tank was surveyed by two different methodologies because of the type and level of residual radioactive materials. The lower portion of the tank liner and surrounding concrete were impacted by the neutron flux of the reactor. The tank and some portions of the concrete were removed in this lower section and residual radioactive materials exist in volumetric concentrations. The upper portion of the tank was not impacted by the neutron flux; therefore, only the surface of the tank would be impacted by residual radioactive materials.

4.1.1 Steel Tank Liner The upper portion of the steel tank liner was determined to be not impacted by the neutron flux from the reactor. The determination was made by collecting two core samples from the liner at a distance above the tank floor based on the WMG Activation Analysis report. The metal cores from approximately six feet above the tank floor were analyzed by gamma spectroscopy at the UA RCO and no activation products, i.e., Eu-152 and Co-60 were detected in the samples. The reactor tank liner was cut a few inches above the core sample using an oxyacetylene torch and the activated portions removed.

The remaining tank liner was designated a Class 1 area with fourteen (14) locations plotted by VSP on a systematic grid placement method. Each location was surveyed for tritium, loose beta/gamma, and total beta/gamma. The upper tank was scanned using a Ludlum Model 2221 with 43-68 gas proportional probe over 100% of its surface area at a rate of one detector width per second. The MDCsCa, was calculated at 1558 dpm/100cm 2 which is 22% of the screening value for Co-60. Table 4-1 presents a summary of the data for the Tank Liner.

17

University of Arizona Nuclear Reactor Lab D&D SERVICES Final Status Survey Report Table 4-1: Static Measurement Results - Reactor Tank Liner Tritium Smear Loose Beta Total Beta (dpm/1OOcm2) (dpm/100cm2) (dpm/10Ocm2)

Number of Std Std Std Measurements Mean Max Dev Mean Max Dev Mean Max Dev 14 1 15 5 -12 50 53 -848 -167 335 4.1.2 Concrete Portion of the Reactor Tank The exposed concrete wall and floor at the bottom of the reactor tank was determined by concrete coring to have residual radioactive materials in volumetric concentrations and was managed as a single Class I survey unit. As a conservative measure, the residual radioactive material concentrations in the concrete at the base of the reactor tank are being compared to the NRC screening values for soil. Thus, the concrete was sampled in accordance with soil sampling and gamma scanning protocols.

The wall and floor concrete in the bottom of the reactor tank was activated to a depth of several inches with measurable concentrations of gamma-emitting Co-60 and Eu-152. After removable of the concrete to a depth of no detectable radiation, the concrete was surveyed as if it were volumetrically contaminated soil with scans using a Nal 2"x2" detector.

The scans were performed using a Ludlum Model 2221 Single Channel Analyzer (SCA) with a Ludlum Model 44-10 NaI detector. The SCA was set for the Co-60 and the Eu-152 gamma energies in order to block out as much background radiation as possible. Co-60 has two gammas per decay, one at 1173 keV and the other at 1332 keV, with the average at 1253 keV. Eu-152 has a gamma photon energy of 1408 keV. The SCA window should be set to encompass the two Co-60 gamma energies and the Eu-152 gamma plus 25 keV on each side of the peaks. Therefore the window would have a width of 235 keV (1149 keV to 1408 keV).

Volumetric sampling of media was implemented in the lower portion of the reactor tank where the reactor tank had been removed. Volumetric samples were analyzed by gamma spectroscopy 18

University of Arizona SERVICES 0 Nuclear Reactor Lab D&D Final Status Survey Report for gamma emitting radionuclides and liquid scintillation counting for tritium by a qualified off-site laboratory.

Samples of approximately 1,500 grams were collected from the surface layer (0 to 2 inches) of the exposed concrete using procedure UA-MCP-RC-06, Sample Collection Procedure. Because the activation of the concrete is dependent on the horizontal distance from the reactor core, samples were collected to a depth of approximately two (2) inches in the fourteen (14) locations.

Table 4-2 presents a summary of the data results and complete laboratory report is included as Appendix D.

Table 4-2: Concrete Analysis Results Summary Results shown in pCi/g No. Standard Screening Fraction Analysis Radionuclide Samples > Mean Maximum Deviation Criteria of Procedure MDC Criteria Liquid Scint Tritium 12 2.99 8.01 2.35 110 0.03 Gamma Spec Co-60 5 0.10 0.45 0.13 3.8 0.03 Gamma Spec Cs-137 0 -0.01 0.00 0.01 11 0.0 Gamma Spec Eu-152 12 0.95 4.69 1.15 7 0.14 Gamma Spec Eu-154 0 0.41 1.07 0.27 8 0.05 SOF 0.25 In addition to the survey unit falling within the Sum of Fractions (SOF) calculation, each individual sample was within the SOF calculation. The maximum percentage of the screening level was in the Tank-6 location at 90% as demonstrated in Table 4-3 below.

19

University of Arizona an SERMET 0 Nuclear Reactor Lab D&D Final Status Survey Report Table 4-3: Concrete Analysis Results per Sample Results shown in percent Screening Criteria Tank Location 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Co-60 0% 0% 0% 3% 3% 12% 1% 2% 1% 2% 5% 7% 1% 0%

Cs-137 0% 0% 0% 0% 0% 0% 0% 0% 0% 0% 0% 0% 0% 0%

Eu-152 8% 3% 0% 10% 10% 67% 7% 12% 11% 6% 16% 24% 8% 7%

Eu- 154 5% 1% 0% 6% 6% 4% 4% 6% 4% 4% 10% 13% 6% 3%

Tritium 2% 1% 0% 2% 1% 7% 3% 2% 2% 1% 7% 5% 3% 3%

Percent 15% 5% 0% 21% 20% 90% 15% 22% 18% 13% 38% 49% 18% 13%

4.2 Building Surfaces 4.2.1 Reactor Control Room The Reactor Control Room (Room 122) was designated a Class 3 area with 14 locations plotted using a simple random sampling placement method. In addition, 5 locations were surveyed in and on the reactor control panel. The room was surveyed for tritium, total beta, and removable beta with the data results summarized in Table 4-4 below. Three data results (two tritium and one Total Beta direct measurement) were greater than the calculated MDC, but were less than 10% of their respective screening criteria.

Table 4-4: Static Measurement Results - Reactor Control Room Tritium Smear Loose Beta Total Beta (dpm/1OOcm2) (dpm/1OOcm2) (dpm/100cm2)

Number of Std Std Std Measurements Mean Max Dev Mean Max Dev Mean Max Dev 14 12 64 19 -6 50 34 367 861 310 20

SERTiEST LW O University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report 4.2.2 Reactor Room The Reactor Room (Room 124) was split into two different MARSSIM classifications. Both areas were surveyed for tritium, total beta, and removable beta with the data results summarized in Table 4-5 below. Several data results were greater than the calculated, but all were less than 13% of their respective screening criteria.

The floors and lower walls of the room were designated a Class 2 area with the locations plotted using the systematic grid sampling placement method. Because of the grid size, a 15th location was automatically added by VSP to maintain area coverage. Per the FSS Plan, 50% of the area, primarily the floor, was scanned for hot spots with none found.

The upper walls and ceiling were designated a Class 3 area and the locations plotted with the simple random sampling placement method. Per the FSS Plan, 10% of the area, primarily each wall, was scanned for hot spots with none found.

Table 4-5: Static Measurement Results - Reactor Room Tritium Smear Loose Beta Total Beta (dpm/100cm 2) (dpm/1OOcm 2) (dpm/100cm 2)

Number of Std Std Std Measurements Mean Max Dev Mean Max Dev Mean Max Dev 15 1 26 11 -26 29 31 548 1839 576 15 3 17 9 5 21 17 825 2044 591 4.2.3 Equipment Storage Room The Equipment Storage Room (Room 124A) was split into two different MARSSIM classifications. Both areas were surveyed for tritium, total beta, and removable beta with the data results summarized in Table 4-6 below. Several data results were greater than the calculated, but all were less than 13% of their respective screening criteria and material specific background was not subtracted from the gross measurement.

21

University of Arizona Wf SERMICES 1O Nuclear Reactor Lab D&D Final Status Survey Report The floors and lower walls of the room were designated a Class 2 area with fourteen (14) locations plotted using the systematic grid sampling placement method. Additional measurements were collected using professional judgment on the sink drain and the drain pipe.

Additional scans were conducted in the area of the sub-critical assembly used to be located. Per the FSS Plan, 50% of the area was scanned for hot spots with none found. The scan focused on the floor and the square meter around each wall survey location.

The upper walls and ceiling were designated a Class 3 area and the locations plotted with the simple random sampling placement method. In addition, 4 locations were surveyed using professional judgment, i.e. HEPA duct inside, HEPA duct top-side, jib crane rail, "squirrel cage" exhaust fan, and reactor mirrors. Per the FSS Plan, 10% of the area, primarily each wall, was scanned for hot spots with none found.

Table 4-6: Static Measurement Results - Equipment Storage Room Tritium Smear Loose Beta Total Beta (dpm/100cm 2) (dpm/100cm 2) (dpm/1OOcm 2 )

Number of Std Std Std Measurements Mean Max Dev Mean Max Dev Mean Max Dev 14 19 125 32 -8 41 27 1995 3381 984 14 5 32 12 -27 25 31 1660 3536 1171 4.2.4 Second Floor Storage Room The Second Floor Storage Room (Room 216) was designated a Class 3 area and 14 locations plotted with the simple random sampling placement method. The room was surveyed for tritium, total beta, and removable beta with the data results summarized in Table 4-7 below. One additional location, neutron beam port cap, was surveyed for loose and total beta. The beam port was designed for neutrons to be streamed vertically from the reactor to the second floor. A 12-inch diameter hole in the floor was installed along with a carbon-steel cap. The cap was scanned for beta/gamma radionuclides and a static measurement was collected on the reactor side of the cap with no residual radioactive material indicated in the results.

22

University of Arizona aO C) Nuclear Reactor Lab D&D SERTCES Final Status Survey Report Table 4-7: Static Measurement Results - Second Floor Storage Room Tritium Smear Loose Beta Total Beta (dpm/100cm2) (dpm/100cm2) (dpm/1OOcm2)

Number of Std Std Std Measurements Mean Max Dev Mean Max Dev Mean Max Dev 14 4 61 20 -12 79 43 1580 4204 1162 4.3 Outside Concrete Pads The Chiller and the Demineralizer were located outside against the north wall of the Engineering Building. Both systems and their associated piping were removed in their entirety during the decommissioning. For the FSS, the pads were surveyed for loose tritium, loose beta and total beta with the results included on the Reactor Room (Room 124) survey form in Appendix D.

Both concrete pads, approximately I square meter each, were scanned for hotspots with none found. The highest total beta reading from the two pads was less than 22% of the screening value without material specific background being subtracted from the gross cpm measurement.

4.4 Storage Pit Results and Inventory Six storage pits are located east of the reactor tank. Each pit is approximately 11 inches in diameter and 7 feet deep. Each pit was scanned for radiological activity during the FSS using a Geiger-Mueller "Pancake" probe. The storage pits were originally included with the lower surfaces of the Reactor Room, however, during the decommissioning, elevated activity was found in Storage Pit #2; potentially above the screening criteria. The storage pit was decontaminated using a wire brush drill bit on a long extension. The pit was then wiped out and the waste disposed of with the concrete and gunite. Due to the difficult access to the storage pits, VSP locations were not plotted. Two locations were selected per storage pit to receive a static measurement along with a 100% surface scan. The results from the storage pit survey are summarized in Table 4-8.

23

SERCI(ST V1 University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report Table 4-8: Static Measurement Results - Storage Pits Loose Beta Total Beta (dpm/100cm 2) (dpm/1OOcm 2)

Number of Std Std Measurements Mean Max Dev Mean Max Dev 12 -16 17 24 952 4971 1780 During the confirmatory survey, the NRC contractor, Oak Ridge Institute for Science and Education (ORISE), discovered a very small hot spot in the bottom of Storage Pit #2. When the result from the Ludlum 44-9 15-cm 22 probe was normalized to 100 cm 2 , the result was more than twice the Co-60 screening value. The spot was localized to an areas less than one 15 cm2 and therefore the total activity across the contiguous 100 cm 2 area around the spot was less than the Co-60 screening value. Subsequently for ALARA purposes, the team performed additional decontamination of the spot and sent the cleaning rags to RCO for gamma spectroscopy analysis.

The results indicated the radionuclide to be Cs-137 which has a larger screening value than Co-60. The post-decontamination survey for Storage Pit #2 was performed by collecting seven (7) contiguous measurements over the 100 cm2 using a 15 cm2 GM probe. The total activity of the seven measurements was 2,903 dpm for the 100 cm 2 area with four of the seven measurements less than the instrument MDC of 455 dpm per detector area. The radiological survey form can be found at the end of Appendix D.

4.5 Miscellaneous Equipment Several pieces of equipment were surveyed in the Reactor Room. These items included the HEPA duct, HEPA duct top-side, jib crane rail, squirrel cage exhaust fan, and reactor overhead viewing mirrors. These items were surveyed for tritium, loose beta and total beta. All measurements are less than MDC and are recorded on the survey form for the Reactor Room-Upper Surfaces located in Appendix D.

4.6 Sink Drain The sink drain pipe in the Equipment Room was dismantled for the FSS. The pipe and drain were surveyed for tritium, loose beta and total beta. All measurements were less than MDC and are recorded on the survey form for the Equipment Room, Room 124A Lower Surfaces.

24

University of Arizona LIN SERVICES 1Nuclear Reactor Final Status LabReport Survey D&D

5.0 CONCLUSION

S Based on the results described in Section 4.0 of this FSS report, the NRL meets the requirements for unrestricted release specified in the UA FSS Plan, UA-MCP-FS-001 and the UA DP (ENERCON 2009).

All Total Beta measurements were less that the release criteria for Co-60, the most restrictive isotope of concern. The maximum result for total beta activity was located in the very bottom of Storage Pit #2 with a result of 4,971 dpm/1 00cm 2 . The average beta activity for each survey unit was less than 1,000 dpmr/100cm2. For all building surfaces that will be exposed to personnel after license termination (excludes the Storage Pits), the mean beta activity result was 778 dpm/1 00cm 2 with a median result of 630 dpm/100cm 2 and a maximum result of 4,204 dpm/1OOcm 2 .

The average level of residual radioactive materials on building surfaces is 11% of the screening criteria which would theoretically result in a dose to the average member of the critical group of 2.7 mrem Total Effective Dose Equivalent (TEDE) per year. It should be noted that natural background activity specific to various building materials was not subtracted from the gross activity results; therefore, the true dose to the average member of the public from reactor activities will actually be less than 2.7 mrem TEDE.

The average level of residual radioactive materials contained in the concrete in the bottom of the reactor pit is 25% of unity which will theoretically expose the average member of the critical group to 6.25 mrem TEDE per year. It should be noted that the screening criteria for this FSS assumes the end state is a resident farmer scenario, the residual material is in the top six inches of the soil, and there is an unlimited source term. This scenario is highly conservative for the end state of the NRL because the residual radioactive materials are bound in the concrete monolith almost 20 feet below the surface, the pit will be backfilled with concrete grout, and there is very limited source term. Additionally, the groundwater, vegetation, milk, or meat exposure pathways do not exist for the true end state of the NRL. Therefore, the actual dose to the average member of the critical group will be significantly less than 6.5 mrem TEDE per year.

25

University of Arizona Nuclear Reactor Lab D&D SERO ES Final Status Survey Report

6.0 REFERENCES

1. 10CFR20.1402, Radiological Criteriafor Unrestricted Use

2. 10CFR50.82, Termination of License

3. NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), Revision 1 (August 2002)

4. NUREG-1507, Minimum Detectable Concentrations With Typical Radiation Survey Instrumentsfor Various Field Conditions, December 1997

5. NUREG-1 505, A NonparametricStatisticalMethodology for the Design and Analysis of FinalStatus DecommissioningSurveys, Rev. 1, June 1998 draft

6. NUREG- 1757, ConsolidatedDecommissioning Guidance, September 2006

7. Enercon Services, Inc. Final Status Survey Plan University of Arizona Nuclear Reactor Laboratory,May 2011.

8. Enercon Services, Inc. University of Arizona Nuclear Reactor Laboratory Decommissioning Plan, May 2009.

9. WMG Report 08-125D-RE-122, University of Arizona Activation Analysis and Component Characterization,April 2009.

10. ISO-7503-1, Evaluation of surface contamination, 1988 26

University of Arizona SERVTCrS 0 Nuclear Reactor Lab D&D Final Status Survey Report APPENDIX A TECHNICAL DESCRIPTION OF INSTRUMENTS

Mode 222 Gencal Prpos Raelneer/sale Radiation Detection for a Safer World Features

. Digital Ratemeter with Built-In Scaler

• Dual Analog Meter & Digital LCD Presentation a Rich Array of Discrete Front Panel Controls

- Single Channel Analysis

- RS-232 Data Output

• Wide Range HV rM

-Overload Protection 2 Part Number: 48-2065 Specifications INDICATED USE: field analysis RATEMETER: provides a four-decade analog display that is operational any time the instrument is turned on, the ratemeter can also be presented on the accompanying LCD as a digital value when the LCD is switched to the "Dig. Rate" position SCALER: range from 0-999999 counts TIMER: switch selectable divisions of 0.1, 0.5, 1, 2, 5, 10 minutes or CONT (continuous) for manual timing CONCURRENT USE: scaler or digital ratemeter is active when not selected, allowing for concurrent use CD SUGGESTED DETECTORS: GM, proportional, scintillation HIGH VOLTAGE: adjustable from 400-2400 volts (can be checked on display)

THRESHOLD: adjustable from 100-1000 (can be checked on display)

WINDOW: adjustable from 0-1000 above threshold setting (can be turned on or off) I, GAIN: adjustable from 1.5-100 mV at a threshold setting of 100 OVERLOAD: senses detector saturation, indicated by --- - -" on LCD display and meter going to full scale (adjustable depending on detector selected)

CONNECTOR: series "C" type METER DIAL: 0-500 cpm, 50-500k cpm logarithmic scale (others available)

DIGITAL DISPLAY: 6-digit LCD display with 1.3 cm (0.5 in.) high digits LCD BACKLIGHT: activated by LAMP switch AUDIO: built-in unimorph speaker with volume control CD AUDIO DIVIDE: toggle switch for 1, 10, or 100 events per click AUDIO JACK: for optional headset CONTROLS: 2

" Power Switch: on/off "Scale Selector Switch: xl, xl0, xl00, xlK, Log

• Response Switch: toggles between FAST (4 sec.) or SLOW (22 sec.) from 10% to 90% of final reading (D

• Meter Zero Pushbutton: to zero meter

• Audio Volume Knob: adjusts volume of audio output "Audio Divide Switch: divides audio clicks by either 1, 10, or 100

• Meter Lamp Switch: illuminates meter

" Window Switch: selects wide open window counts or a narrower setting for viewing a specific energy band

• Scaler Count Time Selector Switch: 0.1, 0.2, 0.5, 1, 2, 5, or 10 minutes

• Count Pushbutton: starts or resets scaler count

• Count Hold Pushbutton: stops scaler count and leaves result in the display

• Scaler/Digital Rate Switch: selects which mode is presented on the LCD

• Test Buttons: battery, high voltage, threshold, window

• Calibration Controls: recessed screwdriver adjustments with calibration cover POWER: 4 each "D" cell batteries BATTERY LIFE: typically 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> with alkaline batteries (battery condition can be checked on digital display)

SIZE: 22.9 x 10.9 x 25.4 cm (9 x 4.3 x 10 in.) (H x W x L) including handle WEIGHT: 2.5 kg (5.5 Ib) including batteries S RS-232 OUTPUT OPTION: This option facilitates exporting data to be read as output to a computer or serial printer from either the ratemeter or scaler reading. (Part No. 4261-148)

P.O. Box 810, Sweetwater. Texas 79556 / http:i/www.ludlums.com 800-622-0828 / 325-235-5494 / Fax: 325-235-4672 / E-mail: ludlum awludlums.com 6/15/2010

Mode 43-6 Alpu BtuGolln~ Dlecol Radiation Detection for a Safer Wiorld 4-C Part Number: 47-2005 Specifications INDICATED USE: alpha beta survey SUGGESTED INSTRUMENTS: Models 12, 16, 18, 2000, 2200, 2221, 2224-1, 2241-2, 2350-1, 2360 CA DETECTOR TYPE: rechargeable gas proportional RECOMMENDED COUNTING GAS: P-10 (10% methane, 90% argon)

GAS RECHARGE: will operate on static charge for at least 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> with a 1 m (39 in.) cable; 5-hour static charge if connected to a simultaneous alpha/beta measuring instrument GAS CONNECTORS: double-end quick disconnect EFFICIENCY (4Tr): 20- 2 39Pu; 15%- 14C; 30%-"Tc; 30%-9°Sr/°Y less than 1%-gamma SIMULTANEOUS ALPHA/BETA COUNTING: 17.5%- 239 Pu; 20%- 99 Tc; 20%--KSr/9Y (D

WINDOW: typically 0.8 mg/cm 2 aluminized Mylar ( other thicknesses available)

WINDOWAREA: 126 cm 2 (19.5 in2) active, 100 cm 2 (15.6 in2) open BACKGROUND: alpha: < 3 cpm (when operating at alpha-only plateau region) beta-gamma: typically 350 cpm (10 pR/hr field) 2 OPERATING VOLTAGE: alpha typically 1100-1400 volts; beta-gamma: typically 1600-1800 volts COUNTER THRESHOLD SETTING: typically 2-5 mV =1 SIZE: 10 x 11.7 x 19.8 cm (3.9 x 4.6 x 7.8 in.) (H x Wx L), with handle WEIGHT: 0.9 kg (2 Ib)

Model 43-68 Probe Gas Accessories Part Number Quantity Description 2310017 1 2-stage regulator 13-7836 1 regulator output fitting (1/4 in. FPT to 1/4 in. tube) 48-2146 1 Model 2750 Flow Control Station 22-9689 22-9522 21-9547 3

2 1

plastic fitting nuts for 1/4 in. vinyl tubing fittings for flowmeter (1/8 in. MPT to 1/4 in. tubing) 10 to 100 cc flowmeter (1/8 in. MPT to 1/4 in. tubing)

C) 22-9514 AR x feet 1/4 in. OD vinyl tubing for gas supply P.O. Box 810. Sweetwater, Texas 79556 ! http:/www.ludlums.com 800-622-0828 / 325-235-5494 / Fax: 325-235-4672 / E-mail: ludlumcdludlums.com 9/14/2010

Radiation Detection for a Safer World a-C CD 3

(1)

C Part Number: 47-1540 Specifications INDICATED USE: low-level, wide-energy gamma detection ENERGY RESPONSE: energy dependent "I SUGGESTED INSTRUMENTS: general purpose survey meters, ratemeters, and scalers OPERATING VOLTAGE: 500-1200 volts SCINTILLATOR: 5.1 x 5.1 cm (2 x 2 in.) (Dia x L) Nal (I

SENSITIVITY: typically 900 cpm/pR/hr (137Cs gamma)

BACKGROUND: 9750 cpm RECOMMENDED ENERGY RANGE: 50 KeV-3.0 MeV PHOTOMULTIPLIER TUBE: 5.1 cm (2 in.) diameter, magnetically shielded CONNECTOR: series "C" (others available)

TEMPERATURE RANGE: -20 to 50 °C (-4 to 122 °F), may be certified to operate from

-40 to 65 °C (-40 to 150 OF) C)

CONSTRUCTION: aluminum housing with beige polyurethane paint SIZE: 6.6 x 27.9 cm (2.6 x 11 in.) (Dia x L) S WEIGHT: 1.0 kg (2.3 Ib)

P.O. Box 810. Sweetwater, Texas 79556 / http:/"www. ludlurns.com 800-622-0828 / 325-235-5494 / Fax: 325-235-4672 , E-mail: ludlum{dludlums.com 6/17/2010

A 7aBei 1 a IC

'(111 ,11 Radiation Detection for a Safer World Features

• Alpha Beta Dual Channel Sample Counter

• Simultaneous Alpha & Beta Counting

• 5.1 cm (2 in.) Diameter Sample Tray

• Independent Readouts

• CPM & DPM Modes Alpha/Beta Alarms

• QC Check C 2

• 8-Hour Battery Operation

• Real Time Clock RS-232 Interface

• Includes PC Software Introduction" Part Number: 48-3456 This system joins Ludlum's Model 3030E dual channel scaler with a Model 43-10-1 dual phosphor detector sample tray to produce a complete alpha beta sample counting system. The 3030E electronics incorporates independent backlit LCD readouts to support discriminated alpha and beta sample counting. The system features background subtraction, crosstalk correction, separate alpha/beta alarms, CPM/DPM operating modes, and a pre-scripted QC function with automatic reminder timer.

The instrument supports both 110 and 220 Vac operation, and includes a trickle-charged gel-cell battery for portable offsite use for up to eight hours. Status indicators located along the front panel inform the operator when another QC check is required, if the detector is nonfunctional, if it is operating in DPM or CPM mode, and ifeither an alpha or beta alarm setpoint has been exceeded.

SpecificationsI5 TUBE: 5.1 cm (2 in.) diameter magnetically shielded photomultiplier WINDOW: 0.4 mg/cm2 aluminized Mylar CD ACTIVE and OPEN AREA: 20.3 cm2 SAMPLE HOLDER: aluminum housing with sample tray capable of holding a 2.5 cm (1 in.) or 5.1 cm (2 in.) diameter sample up to 1.02 cm (0.4 in.) thick EFFICIENCY(4Tr):

alpha: 32% for 23°Th; 39% for 23U; 37% for 239Pu beta: 5% for 14C; 27% for "Tc; 29% for 137Cs; 26% for 1°SrI°Y (D

CROSS TALK:

alpha to beta: 10% or less beta to alpha: 1%or less BACKGROUND:

alpha: 3 cpm or less beta-gamma: typically 80 cpm or less (10 pR/hr field)

SOFTWARE: PC based to perform setup and calibration routines including background subtract, crosstalk correction, cpm/dpm modes, daily QC check parameters, alarm levels, and automatic plateaus. All parameters are stored in the instrument in non-volatile memory. The supplied software is capable of logging and storing the following: Sample Number, Sample Date, Sample Time, Alpha Count, Beta Count, Sample Type, Comments ELECTRONICS:

size: 24.1 x 13.5 x 25.4 ( 9.5 x 5.3 x 10.0 in.) (H x Wx D) weight: approximately 2.7 kg (6 Ib)

DETECTOR: 0 size: 23.6 x 11.4 x 23.6 cm (9.3 x 4.5 x 9.3 in.) (H x Wx L) weight: 1.9 kg (4.1 Ib)

P.O. Box 810, Sweetwater. Texas 79556 / http:/:www.ludlums.com 800-622-0828 / 325-235-5494 / Fax: 325-235-4672 ' E-mail: ludlum~ldudlums.com 6/13/2011

Moe 19 Miro Ratemete Radiation Detection for a Safer World Features C a High Sensitivity to Gamma

• 0-5000 pR/hr Range (others available)

• Rugged

• Splashproof Protection for Outdoor Use

• 2000 Hour Battery life ,

0 CE Certified Introduction Part Number: 48-1615 The Model 19 is a highly sensitive gamma The front-panel controls include a rotary switch pR/meter employing an internally housed, 2.5 for selecting the five-decade range and instrument cm diameter by 2.5 cm thick (1.0 x 1.0 in.) Nal detector. The measuring range is 0-5000 pR/

hr. The cast aluminum instrument housing with its shut-off, an audio on/off switch, a fast/slow response switch, and pushbuttons for activating the meter lamp, high-voltage display, CD separate battery compartment and accompany- battery test, and count reset. The Model 19 is ing metal handle offer an industrial robustness a complete turn-key system and includes two and quality that promote long lasting protection "D" cell batteries.

and instrument life.

Specifications C INDICATED USE: low-level (pR) gamma survey MEASUREMENT RANGE: 0-5000 pR/hr DETECTOR: 2.5 x 2.5 cm (1 x 1 in.) Sodium Iodide (Nal[TI]) scintillator SENSITIVITY: s- 175 cpm/pRthr (137 Cs gamma)

CD ENERGY RESPONSE: energy dependent LINEARITY: reading +/-10% of true value METER: 6.4 cm (2.5 in.) arc, 1 mA analog type.

METER DIAL: 0-25 pR/hr, 0-50 pR/hr, BAT TEST, High Voltage Test _ .CD CONTROLS ROTARY KNOB SELECTOR: 0-25, 0-50, 0-250, 0-500, 0-5000 pR/hr, Instrument Off RESPONSE: toggle switch for FAST (4 seconds) or SLOW (22 seconds) from 10% to 90% of final reading AUDIO: built-in unimorph speaker with ON/OFF switch (greater than 60 dB at 0.6 meters [2 ft])

HIGH VOLTAGE: pushbutton to display HV setting on meter BATTERY: pushbutton checks battery voltage RESET: pushbutton to zero meter Co LAMP: pushbutton to activate meter light CALIBRATION CONTROLS: accessible from front of instrument (protective cover provided)

WORKING ENVIRONMENT: splashproof shields for outdoor use TEMPERATURE RANGE: -20 to 50 °C (-4 to 122 *F), may be certified for operation from -40 to 65 °C (-40 to 150 °F)

POWER: 2 each "D"cell batteries (housed in sealed compartment that is externally accessible)

BATTERY LIFE: typically 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> with alkaline batteries (battery condition can be checked on meter)

CONSTRUCTION: cast and drawn aluminum with beige polyurethane enamel paint 0,

SIZE: 15.2 x 8.9 x 21.6 cm (6.0 x 3.5 x 8.5 in.) (H x W x L), without handle WEIGHT: 2.04 kg (4.5 Ib), including batteries

• P.O. Box 810. Sweetwater, Texas 79556 1 http:f'wwwludlums.com 800-622-0828 / 325-235-5494 / Fax: 325-235-4672 / E-mail: ludlum(iýludlums.com 4/5/2010

University of Arizona ERVTCr 0 Nuclear Reactor Lab D&D Final Status Survey Report APPENDIX B CALIBRATION RECORDS

243 Root St.

Suite 100 Glean, New York 14760 Voice: (716) 372-5300 Certificate Of Calibration Fax: (716) 372-5307 This Certificate will be accompanied by Calibration Charts or Readings where Applicable Customer Instrument Customer Name: Enercon Services, Inc. Manufacturer: Ludlum Measurements Address: 4490 Old William Penn Hwy. Model: 3030E Serial Number: 268970 Murrysville, PA 15668 Detector Manufacturer: Ludlum Measurements Contact Name: Dustin Miller Det. Model: 43-10-1 Serial Number: PR285086 Customer P01 or rder Calibration Method: Electronic CC. Number: MLICLVIO07 Number: 2011-2696 instrument Received: 0 Within Tolerance [] Out of Tolerance 0 Repairs required [3 Other (See Comments)

[3 Geotropism D Meter Zero [0 Mech. Ck. 0 HV Readout F- Battery Check 0 Reset 0 Audio [a Window Status Q FS Response [] Linearity Q3 Background Subtract r Alarm Set Temperature: 70.9 F Humidity: 47% Pressure: 28.5 in Hg Altitude: 1450 ft

__________Instrument Calibration Multiplier\Range Calibration Instrument Response Reference instruments and / or Sources Point Before Calibration After Calibration Pulser: 500-2 220100 Alpha 40 cpm 40 cpm 40 cpm Pu239 C7-636 Tc99 C7-641 Alpha 400 cpm 400 cpm 400 cpm Comments Alpha 4000 cpm 3995 cpm 3995 cprn Inst. Voltage: 800 V Isotope Efficiency Distance Alpha 40000 cpm 39947 cpm 39947 cpm Window Status Pu239 39.9% 0 inch Alpha 400000 cpm 399430 cpm 399430 cpm Beta threshold: 4 mV Tc99 25.1% 0 inch Beta-Gamma 40 cpm 40 cpm 40 cpm Beta window: 50 mV Beta-Gamma 400 cpm 400 cpm 400 cpm Alpha threshold: 120 mV Beta-Gamma 4000 cpm 3995 cpm 3995 cpm Beta-Gamma 40000 cpm 39947 cpm 39947 cpm Beta-Gamma 400000 cpm 399452 cpm 399452 cpm Alpha crosstalk in the Beta channel is <10%

Beta crosstalk in the Alpha channel is <1%

Verified QC Check - PASS Statement of Certification MJWVIecnlnlcI Sevices, inc certifies tnat the above Instrument has been calibrated by standards traceable to the National Institute of Standards and Technology, or to the calibration facilities of other International Standards organization members, or have been derived from accepted values of natural physical constants or have been derived by the ratio type of calibration techniques. The calibration system conforms to the requirements of ISO/IEC 17025 and ANSI N323. The Instrument listed above was inspected prior to shipment and it met all the manufacturers published operating specifications. (MJW technical Services is not responsible for dam e during shipment or use of this instrument).

d B/

Instrument /Rve Calibrated By: /' I/* jZ4_ Reviewed B Date Calibration Date: 05/03/2011 Calibration Due: 05/03/2012

Suite 100

-U W* Olean, New York 14760 243 Root Voice: St. 372-5300 (716)

Certificate Of Calibration

'i2IflflJ]S*i~f~ Fax (1716) 372-5307 This Certificate will be accompanied by Calibration Charts or Readings where Applicable

- Customr ,'.i .Initrumnent '-'

Customer Name: Enercon Services, Inc. Manufacturer: Ludlum Measurements Address: 4490 Old William Penn Hwy Model: 2221 Serial Number: 178100 Murrysville, PA 15668 Detector Manufacturer: n/a Contact Name: Dustin Miller Det. Model: n/a Serial Number: nla Customer P0/ Work Order Calibration Method: Electronic CC. Number: SUNYUBO22-NB Number: 2011-2701 C Instrument Received: 0o Within Tolerance I Out of Tolerance j- Repairs required 0 Other (See Comments)

El Geotropism I" Meter Zero 0 Mech. Ck. HV Readout 0 Battery Check 0 Reset El Audio 9 Window Status 0 FS Response fi Linearity IJ Background Subtract 5 Alarm Set Temperature: 71.3 F Humidity: 42% Pressure' 28.6 in Hg Altitude: 1450 ft Reference instruments and / or Sources Multiplier\Range Calibration Point Calibration Response Before Instrument After Calibration Pulser: 500-2 220106 xI 100 cpm 100 cpm 100 cpm . .) 'Th-ý1

ý'Z

,.~lrlft X 1 400 cpm 400 cpm 400 cpm Inst. Voltage: 1000 V Ref. Voltage 1: 1500 V X 10 1 kcpm 1 kcpm 1 kcpm Input Sensitivity: 10 mV Inst. Voltage 1: 1499 V X 10 4 kcpm 4 kcpm 4 kcpm Ref. Voltage 2: 500 V X 100 10 kcpm 10 kcpm 10 kcpm Inst. Voltage 2: 499 V X 100 40 kcpm 40.5 kcpm 40.5 kcpm X 1K 100 kcpm 100 kcpm 100 kcpm Model 2221 currently set for gross counts X 1K 400 kcpm 400 kcpm 400 kcpm High Voltage: 1000V Digital Ratemeter 40 cpm 39 cpm 39 cpm Threshold: 100 (10mV)

Digital Ratemeter 400 cpm 398 cpm 398 cpm Window: OUT Digital Ratemeter 4 kcpm 4.009 kcpm 4.009 kcpm Digital Ratemeter 40 kcpm 40.006 kcpm 40.006 kcpm Digital Ratemeter 400 kcpm 400.158 kcpm 400.158 kcpm Digital Scaler 40 cpm 40 cpm 40 cpm Digital Scaler 400 cpm 400 cpm 400 cpm Digital Scaler 4 kcpm 3.998 kcpm 3.998 kcpm Digital Scaler 40 kcpm 40.009 kcpm 40.009 kcpm Digital Scaler 400 kcpm 400.156 kcpm 400.156 kcpm Log Scale 50 cpm 50 cpm 50 cpm Log Scale 500 cpm 510 cpm 510 cpm Log Scale 5 kcprn 5 kcpm 5 kcpm Log Scale 50 kcpm 50 kcpm 50 kcpm Log Scale 500 kcpm 500 kcpm 500 kcpm

- .Sttement of Certificatinp MJW Technical Sevices, Inc certifies that the above instrument has been calibrated by standards traceable to the National Institute Of Standards and Technology, or to the calibration facilities of other International Standards organization members, or have been derived from accepted values of natural physical constants or have been derived by the ratio type of calibration techniques. The calibration system conforms to the requirements of ISO/IEC 17025 and ANSI N323. The Instrument listed above was inspected prior to shipment and it met all the manufacturer's published operatinacifications. (MJW technical Services is not responsible for damage incurred during shipment or use of this instrument).

Instrument Calibrated By:..

Date 5-3-/ -/

Calibration Date: 05/04/2011

Suite 100 U

~~w R Olean, New York 14760 243 Root St.

Voice: (716) 372-5300 Certificate Of Calibration

%iIifl( B I Fax: (716) 372-5307 This Certificate wili be accompanied by Calibration Charts or Readings where Applicable C'.. - .o Instrument Customer Name: Enercon Services, Inc. Manufacturer: Ludlum Measurements Address: 4490 Old William Penn Hwy Model: 2221 Serial Number: 211772 Murrysville, PA 15668 Detector Manufacturer: Ludlum Measurements Contact Name: Dustin Miller Det. Model: 44-2 Serial Number: PR289775 Customer PO/ Work Order Calibration Method: Electronic CC, Number: SUNYUB022-NB Number: 2011-2724 Instrument Received: Li Within Tolerance Q Out of Tolerance 0 Repairs required 5 Other (See Comments) 0 Geotropism 0 Meter Zero R0 Mech. Ck. 0 HV Readout 0 Battery Check . Reset 0 Audio IJ Window Status 0l FS Response fJ Linearity 5] Background Subtract 5 Alarm Set Temperature: 73.3F Humidity: 40% Pressure: 28.5 in Hg Altitude: 1450 ft ntrumenht

......................... -- ratdLIEI Multiplier\Range Calibration Instrument Response Reference instruments and / or Sources Point Before Calibration After Calibration Pulser: 500-2 220099 X1 100 cpm cpm 100 cpm Cs137 7753CM I Cs137 1130104 XI 400 cpm cpm 400.cpm ~ met "777~~;

X 10 1 kcpm kcpm 1 kcpm Inst. Voltage: 700 V Isotope Efficiency Distance X 10 4 kcpm kcpm 4 kcpm Input Sensitivity: 10 mV Cs137 4.6% 0 inch X 100 10 kcpm kcpm 10 kcpm X 100 40 kcpm kcpm 40 kcpm Ref. Voltage 1: 1500 V X 1K 100 kcpm kcpm 100 kcpm Inst. Voltage 1: 1508 V X 1K 400 kcpm kcpm 400 kcpm Ref. Voltage 2: 500 V Digital Ratemeter 40 cpm 39 cpm 39 cpm Inst. Voltage 2: 501 V Digital Ratemeter 400 cpm 398 cpm 398 cpm Digital Ratemeter 4 kcpm 3.977 kcpm 3.977 kcpm THR = 101 Digital Ratemeter 40 kcpm 39.832 kcpm 39.832 kcpm WIN = 100 (IN)

Digital Ratemeter 400 kcpm 398.108 kcpm 398.108 kcpm Digital Scaler 40 cpm 39 cpm 39 cpm Digital Scaler 400 cpm 398 cpm 398 cpm Cs137 150 kcpm/mR/Hr Digital Scaler 4 kcpm 3.982 kcpm 3.982 kcpm Digital Scaler 40 kcpm 39.859 kcpm 39.859 kcpm Digital Scaler 400 kcpm 398.088 kcpm 398.088 kcpm Log Scale 50 cpm cpm 50,cpm Log Scale 500 cpm cpm 500 cpm Log Scale 5 kcpm kcpm 5 kcpm Log Scale 50 kcpm kcpm 50 kcpm Log Scale 500 kcpm kcpm 490 kcpm

.... " "....; '" state "--Ica Ion Iof ,.

MJW Technical ,SeviceS, Inc ceatifies that the above instrument has been calibrated by standards traceable to the National Institute of Standards and Technology, or to the calibration facilities of other International Standards organization members, or have been derived from accepted values of natural physical constants or have been derived by the ratio type of calibration techniques. The calibration system conforms to the requirements of ISO/IEC 17025 and ANSI N323. The Instrument listed above was inspected prior to shipment and it met all the manufacturer's published operating specifications. (MJW technical Services is not responsible for damage incurred during shipment or use of this instrument).

Instrument Calibrated By:

Calibration Date: /2011 12/2012

University of Arizona am Nuclear Reactor Lab D&D SERTCE"S Final Status Survey Report APPENDIX C DAILY SOURCE CHECKS

OENERCON Fxcel;ence--Evety project. Every day.

Appendix A DAILY SOURCE CHECK LOG Model Model SIN Detector Detector SIN Inst. Efficiency

'~Z2Z _78~/( ()y4q ~'

Date Time BKGD Reading Initials Month/Year Ai rzc//

&/'/z I t-7 V-z_ i1.337 _____

2.6 136-7 Source CalibrationlReference Reading 91/7-1/1t 5 Z7H ( t's 6 Ij I-l2/

(/7 _Z q6__ (L _ _ Win Source Type:

Ai-ýZ1-/ q72 ZV/5 7 - 3/q5-7 ,. Source SIN:

G/7J-t~f !1'T1*6q 7-1 1Z-7& b Source Activity: q Y70 Distance to source:

9/7/" / _o Reading:

,/z 31

/*'-31 4

///L

/z 5 6

7 iZ 304____

13-7f 8 13q_1 _

/z 9 - /* -V 10 3 Y,

_/A Upper Limit /N70 .

Averagee L .1 Lower Limit i .Zo03 Reviewed By /--Z Survey Supervisor Date Note: A new sheet should be used each month and when a new calibration has been performed.

Transfer the Source Reading data to the new sheet when changing sheets at the beginning of a new month.

A-1

OENERCON Excellence-Every project.Every dcoy Appendix A DAILY SOURCE CHECK LOG Model Model SIN Detector etector SIN Inst. Efficiency

_2__ZiI ?177Z n, Iq-9 60-3 2 1 M Date Time BKGD Reading Initials Jg 1 3 /11 OCI-2-c? 5iS3' Month/Year 4 t9LST/cl7ABPWiM Z 12-011

/7-31q 1 60 3 T7/012o_0 iL/i/ .. iLL Source CalibrationlReference Reading

_V Source Type:

Source S/N:

(P&o5 Source Activity:

Distance to source:

Reading:

2 5-4Z 3

4 6"5-- Z 5 q 6 5%//

7 8

9 10 Upper Limit_______

Average 3&

Lower Limit _______

Reviewed By 94S_//

Survey Supervisor Date Note: A new sneet snould be used each montn ano wnen a new calibration has been performed.

Transfer the Source Reading data to the new sheet when changing sheets at the beginning of a new month.

A-1

CýOENERCON Exceltence-Everyproject. Every d*y Appendix A DAILY SOURCE CHECK LOG Mod-(

ModelI _ Model SIN _ Detector ctor SN Inst. Efficiency

('2-? r Date Time BKGD

/ 101/ Reading Initials Month/Year LCi Iý,r Ze, /I Source CalibrationlReference Reading Source Type:

Source SIN: SucS.-4 63 Source Activity:

Distance to source:

A.

Reading:

4 6 -

7 7-77 8 __ __ __

9 35 10 1 35-Ti 0)11, ,.n tL)R b .: '7_1t (,, & .I 7/ ef,, r-'

Upper Limit 1 7-5-7 Average "j .. -

Lower Limit 5 '7 l Reviewed By )

Survey Supervisor Date J

• 4 Note: A new sheet Snould oe used eacn month ana when a new calibration has been pertormed.

Transfer the Source Reading data to the new sheet when changing sheets at the beginning of a new month.

A-1

On 0 SERMiEWS University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report APPENDIX D RADIOLOGICAL SURVEY FORMS

o ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Room 122 - Reactor Control Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP Serial Calibration Probe Area Tritium Tritium Tritium MDA 2s Inst. Beta Beta MDA Instrument Number Due Probe lcml S !cki po Ulciency

, (d41m) Beta kgd . d Date: 8/23/2011 2

Ludium 2221 178100 5/4112 43-68 126 NIA N/A N/A 257 0.25 389 Smear Area 100 cm Ludlum 3030E 268970 5/3/12 43-10-1 N/A N/A N/A N/A 77 0.24 142 MARSSIM Classification:

TriCarb LSC 434238 N/A Liquid Scin. N/A 5 0.34 39 N/A N/A N/A Class 3 MDC_ @ 95% Detection 1558 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A probability 2dnm/10cmZt N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times = 0.5 minute duration, except Trtium smears at I mtm. Material specific background not subtracted from gross cpm.

Removable Tritium Removable Beta Total Beta Beta Scan I By Dose Rate I 1-meter Gen.

VsP Net DPM I Net DPM / DPM / MAX Area Notee LOC. # Description CPM 100cm2 CPM lOOcma Gr(71 CPM 100cm' 111 CPM CPM I IR/hr I I 1 I l I 1 Wall (east) wall board 6 <MDA 71 <MDA 290 <MDA 170 350 2 Wall (north) concrete 5 <MDA 65 <MDA 324 861 210 400 3 Ceiling 10 <MDA 77 <MDA 300 <MDA 250 350 4 Wall (west) concrete 10 <MDA 79 <MDA 272 <MDA 150 350 5 Window (east) glass 5 <MDA 87 <MDA 266 <MDA 150 350 6 Wall (east) wall board 6 <MDA 79 <MDA 306 <MDA 200 380 7 Floor 11 <MDA 67 <MDA 272 <MDA 250 350 8 Ceiling 6 <MDA 89 <MDA 300 <MDA 250 350 9 Ceiling 27 64 69 <MDA 276 <MDA 150 350 10 Cabinet (wood) 7 <MDA 70 <MDA 308 <MDA 200 350 11 Wall (east) wall board 1 <MDA 72 <MDA 316 <MDA 170 330 12 Floor 8 <MDA 66 <MDA 242 <MDA 150 400 13 Floor 7 <MDA 78 <MDA 258 <MDA 150 350 14 Door - metal 10 A1l 87 <MDA 268 <MDA 200 400 PJ Ctl Pnnel -top 72 <MDA 230 <MDA 100 300 PJ Ctl Panel -panel 69 -<MDA 202 <MDA 100 300 PJ CtUPanel -inside W 76 <MDA 188 <MDA 100 300 PJ Ctl Panel -side E 69 <MDA 238 <MDA 100 300 PJ CUPanel -front bottom east 81 <MDA 232 <MDA 100 300 N/A NIA

9

...l 0

0 0

rw

-*)

0ENERCON RADIOLOGICAL SURVEY REPORT Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Room 124 - Lower Surfaces Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP Serial Calibration Probe Axes TTritium itium Tritium MDA 2z last. Beta Bets MDA S Nuber Prob Beta Bk4[d Eft. idpm) Date: 8/24/2011 2

5/4112 43-68 126 N/A N/A N/A 257 0.25 389 Smear Area 100 cm Ludlum 2221 178100 Ludlum 3030E 268970 5/3112 43-10-1 N/A N/A N/A N/A 77 0.24 142 MARSSIM Classification:

TriCarb LSC 434238 N/A Liquid Scin. N/A 5 0.34 39 N/A N/A NIA Class 2 MDC.,. @ 95% Detection 1558 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Probabilitv idnm/n110cmrn: 5 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times = 0.5 minute duration, except Trtium smears at 1 min. Material specific background not subtracted from gross cpm. Starts from South Removable Tritium Removable Beta Total Beta Beta Scan I Rv DoseRate I I * - I* 4------------------

Net DPM / Net DPM / DPM / 1-meter Gen-VSp MAX I Area Nor.

CPU PRlhr ttR/lhr Loc.# Description CPM looc= 2 CPM lOOcm 2 Gross CPM 100cma KIN CPM 15 Floor 8 <MDA 60 <MDA 272 <MDA 200 400 16 Floor 8 <MDA 76 <MDA 302 <MDA 200 400 17 Floor 2 <MDA 64 <MDA 294 <MDA 200 400 18 Floor 1 <MDA 64 <MDA 250 <MDA 200 400 19 Floor 5 <MDA 69 <MDA 252 <MDA 200 400 20 Wall 1 1 <MDA 73 <MDA 322 836 200 400 21 Wall 1 2 <MDA 67 <MDA 282 <MDA 200 400 22 Wall 1 14 <MDA 80 <MDA 344 1119 200 400 23 Wall 2 7 <MDA 67 <MDA 400 1839 300 500 24 Wall 2 12 <MDA 73 <MDA 346 1144 300 500 25 Wall 3 5 <MDA 74 <MDA 334 990 300 500 26 Wall 3 3 <MDA 79 <MDA 326 887 300 500 27 Wal 3 Door 2 <MDA 74 <MDA 264 <MDA 300 500 28 Wall 4 6 <MDA 84 <MDA 248 <MDA 300 500 29 Wall 4 5 <MDA 58 <MDA 258 <MDA 300 500 PJ Chiller Pad (Outside) 2 <MDA 84 <MDA 424 2147 300 600 PJ Demineralizer Pad (Outside) 01 -A) A otni ri NIA INIA N/A NIiA NIA IN/A

Room 124 Lower

+

22

+

t is 23 N

15 16 *R 24 26*

O ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Room 124(A) - Lower Surface Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP Serial Calibration Probe Area 2

Tritium Tritium Tritlm MDA 2x Beta Be8t/ tst. 2 0A Instrument Number Due _ Probe_ cm I Bcklround Efficiency BdpmBeta Bkd __. (dp_ Date: 8/24/2011 2

Ludlum 2221 178100 5/4/12 43-68 126 N/A N/A N/A 257 0.25 389 Smear Area 100 Cm Ludlum 3030E 268970 5/3/12 43-10-1 N/A N/A N/A N/A 77 0.24 142 MARSSIM Classification:

TriCarb LSC 434238 N/A Liquid Scin. N/A 5 0.34 39 N/A N/A N/A Class 2 MDC:_ @ 95% Detection t 1558 N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Prohabilitv idnmllmllOýcm N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times 0.5 minute duration, except Trtium smears at I min. Material specific background not subtracted from gross cpm.

Removable Tritium 1 r Removable Beta 1 V Total Beta

• r 1*

Beta Scan I

I 1-By Dose1 Rate Gen Il VSP Net DPM / Net DPM / DPM / 2 1-meter IAGeane.

Area Nh LOC. # Descrivition CPM 100cm2 CPM 100cm 2 Gross CPM 100cm MIN CPM PR/hr PR/hrI 30 Floor 13 <MDA 72 <MDA 302 <MDA 300 400 31 Floor 7 <MDA 77 <MDA 318 784 300 400 32 Floor 14 <MDA 87 <MDA 370 1453 300 400 33 Floor 7 <MDA 61 <MDA 360 1324 300 400 34 Floor 15 <MDA 79 <MDA 356 1273 300 400 35 Wall 7 <MDA 75 <MDA 474 2790 400 600 36 Wall 9 <MDA 75 <MDA 520 3381 400 600 37 Wall 48 125 71 <MDA 490 2996 400 600 38 Wall 6 <MDA 80 <MDA 460 2610 400 600 39 Breaker Box 7 <MDA 80 <MDA 294 <MDA 200 400 40 Wall 3 <MDA 67 <MDA 474 2790 400 600 41 Wall 6 <MDA 71 <MDA 468 2713 400 600 42 Wall 7 <MDA 74 <MDA 464 1 2661 400 600 43 Wall 7 ~-WAfA 82 <MDA 420 2096 400 600 I + + t PJ Sink drain 85 <MDA 200 400 PJ Drain Pipes Below drain 76 <MDA 200 500 PJ Drain Pipes Start of "P" Trap 69 <MDA PJ Drain Pipes End of "P" Trap 69 <MDA PJ Drain Pipes Start of pipe to sewer 92 <MDA PJ Sub Crit Area (6' Diam) 200 1 400

+ +

Room 124A Floors and Lower Walls 33 N

30 %ii 40 39

0 ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Room 124 - Upper Surfaces Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP Serial Calibration Probe Area Tritium Tritium Tritium MDA 2n Inst. Beta Beta MDA Instrument Number Due Probe leraa Background Efficiency {dpmjl Beta S__d Sff. idpm) Date: 8/24/2001 2

Ludlum 2221 178100 514/12 43-68 126 N/A N/A N/A 257 0.25 389 Smear Area 100 cm Ludlum 3030E 268970 5/3/12 43-10-1 N/A N/A NIA N/A 77 0.24 142 MARSSIM Classification:

TriCarb LSC 434238 N/A Liquid Scin. N/A 5 0.34 39 N/A N/A N/A Class 3 NANAMDC_ @ 95% Datection 1558 2

N/A N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A Prnhahlil/f0dnrrlO0cm V N/A N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times 0.5 minute duration, except Trtium smears at 1 min. Material specific background not subtracted from gross cpm.

- .. ..... T :o -. 1 Removable Tritium Removable Beta Total Beta Beta *Scan I----------------

ri Dose . Rate t 94T DPM / 1-meter Gen VSP Net Net DPM / DPM / Area Not MAX CR/hr Loc. # Description CPM 100cma CPM 100cm 2 Gross CPM 100cm 2

Am CPU CPM I I uR/hbr

............................ ......Y ZL* .........

44 Concrete 6 'MDA 81 <MDA 314 <MDA 200 400 45 Concrete 8 <MDA 73 <MDA 314 <MDA 200 400 46 Ceiling 11 <MDA 75 <MDA 264 <MDA NA/

47 Concrete 6 <MDA 80 <MDA 402 1864 300 500 48 Concrete 4 <MDA 78 <MDA 416 2044 300 500 49 Wall Board 5 <MDA 74 <MDA 300 <MDA 200 400 50 Ceiling 4 <MDA 73 <MDA 350 1196 N 51 Ceiling 1 <MDA 78 <MDA 334 990 N/A N/A 52 Window 11 <MDA 79 <MDA 266 <MDA 200 400 53 Ceiling 8 <MDA 75 <MDA 284 <MDA N 54 Ceiling 2 <MDA 65 <MDA 310 <MDA 55 Ceiling 6 <MDA 74 <MDA 306 <MDA N/A N/A 56 Concrete 11 <MDA 72 <MDA 268 <MDA 200 400 57 Concrete 4 <MDA 78 <MDA 330 939 200 400 58 Wall (East) Concrete 6 <MDA 72 <MDA 360 1324 200 400 PJ Inside Vent Duct (Post HEPA) 9 <MDA 81 <MDA 280 <MDA 200 300 PJ Top of HEPA Duct 3 <MDA N/A N 272 <MDA 200 300 PJ Top of Bridge Crane Beam 8 <MDA 70 <MDA 312 <MDA 200 400 PJ Back Side of Mirrors 4 <MDA 78 <MDA 304 <MDA 200 400 PJ Squirrel Cage Fan 7 <MDA 82 339.0 224 <MDA 200 400

57 49 Room 124 Upper 51 54 45II 55

,5q3

2) :

52:

C ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Room 124A - Upper Surfaces Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP Serial Calibration Probe Area Tritium Tritium Tritium MDA 2x Inst. Beta Beta MDA 2

Instrument Number Due Probe (cm ) Background Efficiency Idn Beta kdff. (dpm) Date: 8/24/2011 2

Ludlum 2221 178100 514/12 43-68 126 N/A N/A N/A 257 0.25 389 Smear Area 100 cm Ludlum 3030E 268970 5/3/12 43-10-1 N/A N/A NIA N/A 77 0.24 142 MARSSIM Classification:

TriCarb LSC 434238 N/A Liquid Scin. N/A 5 0.34 39 N/A N/A N/A Class MDC==, @ 95% Detection 315 N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A N/A proa@ ilit5 IdemJlfOmZ; 1558 N/A N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times = 0.5 minute duration, except Trtium smears at t min. Material specific background not subtracted from gross cpm.

Removable Tritium Removable Beta Total Beta Beta Scan I By Dose Rate VSP Net DPM / Net DPM / DPM / 1-meter MAX Area

= INot LoC.# Description CPM 100cm 2 CPM lOOcm2 Gross CPM loocm2 MIm CPM CPUM pR/hr 59 59 5 <MDA 66 <MDA 296 <MDA 200 400 60 60 Ceiling 8 <MDA 83 <MDA 310 <MDA 200 400 61 61 4 <MDA 64 <MDA 298 <MDA 200 400 62 62 Wa;ll - brick 3 <MDA 56 <MDA 456 2559 400 600 63 63 Wall - brick 5 <MDA 64 <MDA 476 2816 400 600 64 64 Door - wood 5 <MDA 75 <MDA 316 <MDA 200 400 65 65 Ceiling 4 <MDA 71 <MDA 308 <MDA 200 400 66 66 Ceiling 6 <MDA 69 <MDA 308 <MDA 200 400 67 67 Wall - brick 17 <MDA 72 <MDA 532 3536 400 600 68 68 Wall - brick 7 <MDA 82 <MDA 502 3150 400 600 69 69 Wall - brick 6 <MDA 68 <MDA 426 2173 300 500 70 70 Wall - brick 14 <MDA 68 <MDA 482 2893 400 600 71 71 Ceiling 7 <MDA 80 <MDA 286 <MDA 200 400 72 72 Wall - brick A -TUITIA -AsAnA A IA 1 1197 'Infn NIA NIA NIA NIA N/A N/A NIA NIA N/A NIA NIA NIA

62 Room 124A Ceiling and Upper Walls 67 60 N 59 63 70 69

0ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory I Number Room 216 Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP 2

Serial Calibration Probe Area Tritium Tritium Tritium MDA s Inst. Beta Beta MDA Instrument Number me Probe (cmal Background Efficiency fdfrl Beta Bkgd . dpmn Date: 8/25/2011 Ludlum 2221 178100 514112 43-68 126 N/A N/A N/A 257 0.25 389 Smear Area 100 cm 2 Ludlum 3030E 268970 5/3/12 43-10-1 N/A N/A NIA N/A 77 0.24 142 MARSSIM Classification:

TriCarb LSC 434238 N/A Liquid Scin. N/A 5 0.34 39 N/A N/A N/A Class 3 MDC_, @ 95% Detection 1558 2

N/A N/A N/A NIA N/A N/A NIA N/A N/A N/A Probability idnm/10Ocm 1 N:A N/A N/A N/A NIA N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times = 0.5 minute duration, except Trtium smears at 1 min. Material specific background not subtracted from gross cpm.

Removable Tri.*um Kemovable Beta Total Beta Beta Scan I BvDoe Rate I 7 1 1 4- 4-Net DPM / Net DPM / DPM / 1-meter Gen.

VSP MAX Not.

I Area LOC. # Description CPM 100cm 2 CPM 100cm2 Gross CPM 100cm 2

NIN CPU CPM I

.jR/hr 73 Rm 216A Ceiling 11 <MDA 96 <MDA 346 1144 300 500 74 Rm 216A Brick 3 <MDA 79 <MDA 370 1453 300 500 75 Rm216A Brick 6 <MDA 65 <MDA 358 1299 300 500 76 Rm 216A Brick 6 <MDA 62 <MDA 402 1864 300 500 77 Rm216A Base Trim 13 <MDA 73 <MDA 256 <MDA 200 500 78 Rm 216A Ceiling 26 61 76 <MDA 320 810 200 400 79 Rm216 Ceiling 0 <MDA 67 <MDA 376 1530 200 400 80 Rm 216 Floor 2 <MDA 65 <MDA 330 939 200 400 81 Rm216 Floor 0 <MDA 87 <MDA 314 <MDA 200 400 82 Rm 216 Floor 5 <MDA 68 <MDA 308 <MDA 200 400 83 Rm216 Brick 6 <MDA 67 <MDA 484 2919 400 600 84 Rm 216 Brick 3 <MDA 83 <MDA 520 3381 400 600 85 Rm 216 Ceiling 4 <MDA 66 <MDA 350 1196 200 400 A

86 Rm 216 Brick 85 <MDA 584 4204 400 600 4 4 4- 4-PJ Rx Side of Neutron Beam Port Cap 400 300 N/A N/A N/A NIA N/A NIA NIA NIA NIA NIA

• F

-4 I-4 0

0 F3 0

83 Room 216 86 SN N 79 4

0 ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Reactor Pit - Lower (concrete)

Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP serial Calibration Probe Area Tritium Tritium Tritium MDA 2a Inst. Beta Beta MDA Instrument Number Due Probe (cmjl Background Efficiency (dpmj Beta kgd f (dp) Date: 8/23/2011 Ludlum 2221 211772 5/4/12 44-10 2x2 Nal NIA N/A N/A 519 cpm N/A N/A Smear Area 100 cm, Ludlum 19 91563 5/4/12 Internal Nal NIA NIA N/A N/A 23 pR/hr N/A N/A MARSSIM Classification:

N/A N/A N/A N/A N/A N/A NIA N/A N/A N/A N/A Class 1 MDC-- @ 95% Detection N/A 2

N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Prob*bilitv idnmn/0Ocm N N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Ten minute background count in center of lower pit was 5188 counts; equals 518.8 cpm. SCA window set from 1000 keV to 1200 keV.

I Removable Tritium I Removable Beta I Total Beta I Gamma Scan I J By Dose Rate VSP Contact 1-meter Net DPMi/ Net DPM /2 DPM i pRlhr I pRlhr Note.

L. # Description CPM Il00cm2 CPM 100cm oree CPM 100cm 2

IN Pit-1 Concrete Pit-2 Concrete Pit-3 Concrete Pit-4 Concrete Pit-5 Concrete (thermalizer cave)

Pit-6 Concrete Pit-7 Concrete Pit-8 Concrete Pit-9 Concrete Pit-10 Concrete Pit-11 Concrete Pit-12 Concrete Pit-13 Concrete Pit-14 Concrete PJ Bottom of Pit - Center N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

Reactor Pit - Lower (concrete)

• *iiý Fp R4 Fp

-it--ý Pit

• Ykmý

0 ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Reactor Pit - Upper (Steel Liner)

Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP Serial Calibration Probe Area Tritium Tritium Tritium UDA 2z Inst. Beta Beta MDA 2

Instrument Number Due Probe (cm a Background Efficiency fd~L Beta Bkgd Eft. (dpmj Date: 8/24/2011 2

Ludlum 2221 178100 5/4112 43-68 126 N/A NIA N/A 257 0.25 389 Smear Area 100 cm Ludlum 3030E 268970 5/3/12 43-10-1 N/A N/A N/A N/A 77 0.24 142 MARSSIM Classification:

TriCarb LSC 434238 N/A Liquid Scin. N/A 5 0.34 39 N/A N/A N/A Class I MDC._ @ 95% Detection 1558 2

N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Probabilitv (dnm/lOflem l N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times = 0.5 minute duration, except Trtium smears at I min. Material specific background not subtracted from gross cpm.

Removable Tritium Removable Beta Total Beta Beta Scan I 1W Dose Rate Gen.

Vap Net DPM / Net DPM / DPM / 1-meter Not-s pR/hr Area Loc. # Descriotion CPM 100cma CPM 100cm 2 Gross CPM 100cm2 MIN CPU 1.021 Pit-15 SteelLiner 4 <MDA 52 <MDA 154 <MDA 200 300 Pit- 16 Steel Liner 5 <MDA 70 <MDA 208 <MDA 200 300 Pit-17 Steel Liner 5 <MDA 59 <MDA 192 <MDA 200 300 Pit-18 Steel Liner 4 <MDA 88 <MDA 212 <MDA 200 300 Pit-19 Steel Liner 10 <MDA 76 <MDA 194 <MDA 200 300 Pit-20 Steel Liner 4 <MDA 86 <MDA 164 <MDA 200 300 Pit-2o Steel Liner 6 <MDA 89 <MDA 168 <MDA 200 300 Pit-22 Steel Liner 4 <MDA 87 <MDA 164 <MDA 200 300 Pit-23 Steel Liner 6 <MDA 67 <MDA 186 <MDA 200 300 it-24 Steel Liner 5 <MDA 82 <MDA 178 <MDA 200 300 Pit-25 Steel Liner 4 <MDA 60 <MDA 212 <MDA 200 300 Pit-26 Steel Liner 6 <MDA 86 <MDA 170 <MDA 200 300 Pit-27 Steel Liner 5 <MDA 77 <MDA 244 <MDA 200 300 Pit-28 Steel Liner 5; - finA A;n R-ATIA 190 <MDA 200 300 PJ Tank steel above thermal Cave - West Side onn NIA N/A N/A N/A NIA NIA NIA N/A N/A N/A

Reactor Pit - Upper (Steel Liner)

• Pit--2

• _POE _P -A h_ _FPt2 I

OENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory Number Storage Tubes Room124 Surveyed by: Dustin Miller Reviewed By: Kevin Taylor, CHP Serial Calibration Probe Area Tritium Tritium Trittiun MDA Beta Bkgd 2n Inst. Beta Beta MDA Instrument Number Due Probe lcmZi Backround Efficiency idpm) BetaW__rd Eff. (dpm) Date: 8/25/2011 Ludlum 2221 178100 5/4/12 HP-210 15.5 N/A N/A N/A 191 0.73 717 Smear Area 100 cm, Ludlum 3030E 268970 5/3/12 43-10-1 N/A N/A N/A N/A 77 0.24 142 MARSSIM Classification:

N/A N/A N/A N/A N/A 5 0.344 N/A N/A N/A NIA Class I MDC__@ 95% Detection 3816 2

N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Probability (dnmn/1lOcm 3816 N/A N/A N/A N/A N/A N/A W/A N/A N/A N/A NIA Surface Efficiency: 25%

Remarks: Count times = 0.5 minute duration, except Trtium smears at I min. Material specific background not subtracted from gross cpm.

Smear I Removable Tritium I Removable Beta Total Beta Beta Scan I & Doe Rate I 1-meter Gen.

Net °PM!2 Net DPM /2 DPM / Note-taR/hr Area

1. Description CPM 100cm loocm2 I pRlhrI 1 Pit #1 (From South) Bottom 2 Top 3 Pit #2 Bottom 4 Top 5 Pit #3 Bottom 6 Top 7 Pit #4 Bottom 8 Top 9 Pit #5 Bottom 10 Top 11 Pit #6 Bottom 12 Top N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A

0 ENERCON RADIOLOGICAL SURVEY REPORT Room Purpose of Survey: Final Status Survey - University of Arizona Nuclear Reactor Laboratory I Number Storage Tubes Room124 Surveyed by: Dan Jordan Reviewed By: Dustin Miller serial Calibration Probe Area Tritium Tritium T C Beta 2x lst. Beta Beta MDC Instrument Nbutber Due Probe (cat2 Bac d Efficency (dpfI STf. idipm Date: 10/6/2011 2

44-9-18 15 N/A N/A N/A 50 0.25 455 SmearArea N/A cm Ludlum 2241-3 248063 5/31/12 N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A N/A MARSSIM Classification:

N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A N/A Class 1 MDC_ @ 95% Detection NA N/A N/A NIA N/A N/A N/A N/A N/A N/A N/A N/A prbabilit fdnm/100cm,1; NIA N/A NIA N/A N/A N/A N/A NIA N/A N/A N/A Surface Efficiency: 25%

Remarks: Count times 0.5 minute duration, except Trtium smears at 1 min. Material specific background not subtracted from gross cpm.

Removable Tritium Removable Beta Total Beta Beta Scan Dose Rate Net DPM / Net DPM /DPM perl BUotre

1. Description CPM 100cmý CPM 100cmO Grow CPM probe area MIN CPM CPM h R/hr I Pit 2 Highest NA/A/A NA170 1935N/ NA NA NA 2 Pit 2 West N/ N/ NA / 80 484N/ NA NA NA 3 Pit 2 East N/ / / / 80 484N/ NA NA NA 4 Pit 2 bottom 70 <MDA 5 Pit 2 bottom N/ /A NA / 60 <MDAN/ NA NA NA 6 Pit 2 wall N/ / / / 60 < MDAN/ NA NA NA 7 Pit 2 Wall NA/A/A/A60 < MDAN/ NA NA NA 2

N/A NIA Pit 2 Sum of measurements (100 cm ) 2903 This survey for Pit #2 was conducted after two decontamination efforts.

The decontamination efforst were rag from the first decontamination effort was analyzed by gamma spectroscopy with the results indicating 91 dpm of Cs-137 was (7

806 removed.

SERMWCS 0 University of Arizona Nuclear Reactor Lab D&D Final Status Survey Report APPENDIX E LABORATORY DATA PACKAGE

• k*WkTELEDYNE 01~ BROWN ENGINEERING, INC.

A Teledyne Technologies Company 2508 Quality Lane Knoxville, TN 37931-3133 865-690-6819 Dustin Miller/ Corey DeWitt Enercon 4490 Old William Penn Highway Murrysville, PA 15668 Report of Analysis/Certificate of Conformance 09/13/2011 LIMS #: L47414 Project ID#: EN005-3EUOFA- 10 Received: 08/16/2011 Delivery Date: 09/15/2011 P.O.# PER RECEIPT Release #:

SDG#:

This is to certify that Teledyne Brown Engineering - Environmental Services located at 2508 Quality Lane, Knoxville, Tennessee, 37931, has analyzed, tested and documented samples, as received by the laboratory, as specified in the applicable purchase order.

This also certifies that requirements of applicable codes, standards and specifications have been fully met and that any quality assurance documentation which verified conformance to the purchase order is on file and may be examined upon request.

I hereby certify that the above statements are true and correct.

Keith Jeter Operations Manager Cross Reference Table Client ID Laboratory ID Station ID (if applicable) 66i L47414-1 002 003 005 t L47414-3--

L47i-414 .

006 1,47414-6 007 L47414-7 008 L47414-8 Page I of 16

,f' TELEDYNE 01 BROWN ENGINEERING, INC.

A Teledyne Technologies Company 2508 Quality Lane Knoxville, TN 37931-3133 865-690-6819 CrossReference Table Client ID Laboratory ID Station ID _ff4ppjiýccabý 009 L47414-9 1

010 L47414-10 011 L47414-11 012 L47414-12 013 L47414-13

-~ 014 L47414-14 Method Reference Numbers Matrix - Analysis -- Method Reference SD I GAMMA EPA 901.1 This report shall not be reproduced or distributed except in its entirety.

Page 2 of 16

AV O TELEDYNE 01 BROWN ENGINEERING, INC. Case Narrative ATeledyne Technologies Company 09/13/2011 13:33 L47414 EN005-3EUOFA-10 Enercon Services, Inc.

Co-57 key gamma is at 122.06 Key. The key gamma for Eu-152 is at 121.78 Kev. The gamma energies are too close for the gamma system to distinguish one from the other and assigns the counts in the region to both nuclides.

Page 1 of 1

Report of Analysis TELEDYNE 09/13/11 08:58 BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon EN005-3EUOFA-10 Sample ID: 001 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

L1MS Number: L47414-1 Ri e Activity IUncertainty Run i Aliquot Aliquot Reference Count Count Count Radionuclide SOP# Conc 2 Sigma MDC Units # Volume Units Date Date Time Units Flag Values

[-3 1 2003 2.18E+00 6.38E-01 F 9.48E-01 1 pCi/g T 1 2.53 g wet I 109/08/111 60 F M F+ I

,G-108M 1 2007 F -1.35E-02 1 2.80E-02 4.39E-02 pCi/gDry _ F 342.98 F gdry 108/10/1100:00 1 08/26/11f 5400 F Sec Fi I _ j No I

.-40 1 2007 F 2.27E+0 1 1.37E+00 F 3.23E-01 I pCi/gDry T _ 342.98 F gdry 108/10/11 00:00 1 08/26/111 540 Sec +T V __ 1 es [

+

IN-54 1 2007 1 1.89E-02 1 3.70E-02 F 6.30E-02 1 pCi/g Dry _ 1 342.98 F gdry 108/10/1100:00 1 08/26/11i 5400 1 Sec I U

'0-57 2007 N F 1.95E-01 1 5.60E-02 F 4.52E-02 1 pCi/gDry 1 1 342.98 F gdry 108/10/1100:00 1 08/26/111 5400 1 Sec -I I_____

Ycs[

'0-58 1 2007 1 9.48E-03 I 3.83E-02 IF6.41E-02 1 pCi/gDry T _ 342.98 F gdry V08/10/11 00:00 08 /26/11A 5400 T Sec --- _UT_ I No _F

0-60 1 2007 F 7.38E-03 1 3.65E-02 F 6.15E-02 pCi/g Dry T 342.98 1 gdry F08/10/11 00:00 1 08/26/111 5400 1 Sec WU j_ 1No 7

,G-l0IM 1 2007 i 2.17E-02 1 3.24E-02 F 5.63E-02 1 pCi/gDry T I 342.98 T gdry 108/10/11 00:00 1 08/26/11F 5400 T Sec UF 1N-7=_

B-124 1 2007 T 3.53E-03 i 4.45E-02 F 6.71E-02 1 pCi/g Dry _ 1 342.98 gdry 108/10/1100:00 1 08/26/111 5400 T Sec I u I1 No I S-134 1 2007 i 1.09E-02 3.82E-02 F 5.64E-02 1 pCi/gDry T _ 342.98 T gdry 108/10/1100:00 1 08/26/111 5400 1 Sec 1 U I I No F

'S-137 1 2007 T -9.99E-03 I 3.43E-02 F 5.57E-02 1 pCi/g Dry 1 1 342.98 T gdry 108/10/11F00:00 08/26/11I 5400 T Sec F U __ I No I A-140 1 2007 F 2.55E-02 6.92E-02 F 1.20E-01 pCi/g Dry _ 1 342.98 1 gdry 108/10/li 00:00 1 08/26/111 5400 1 Sec I UF I No 1 1

,U-152 1 200;7 T 5.61E-01 1 1.42E-01 I 1.51E-01 1 pCi/gDry 1 1 342.98 1 g dry 108/10/11 00:00 1 08/26/111 5400 T Sec I+ _[ I Yes[

.U-154 1 2007 T 3.96E-01 1 1.14E-01 F 1.29E-01 1pCi/gJDry I 342.98 1 gdry 108/10/11 00:00 1 08/26/11i 5400 1 Sec FU-* ____ 1T-j

,C-228 1 2007 T 8.81E-01 2.27E-01 2.11E-0l 1 pCi/gDry I 342.98

- 1 gdry 108/10/11 00:00 1 08/26/111 5400 1ec S IT+ I Yes 12

--228 1 2007 1 8.03E-01 9.57E-02 I 1.03E-01 I pCi/g Dry I342.98 g dry 108/10/11 00:00: 08/26/1l 5400 I Sec F+ [ Yes I B4-232 1 2007 1 6.33E-01 1.82E-01 i 2.02E-01 1 pCi/gDry I -342.98 1 gdry 1108/10/1100:00 08/26/111 5400 Sec I + 1 i Yes -lag Values j = Compound/Analyte not detected (< MDC) or less than 3 sigma No Peak not identified in gamma spectrum I- = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only)

SJ* = Yes Peak identified in gamma spectrum Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma ligh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported spec MDC exceeds customer technical specification are related only to the samples tested in the condition they Low recovery are received by the laboratory.

+I High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 3 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 'R BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 002 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-2 Activity IUncertainty Run Aliquot Aliquot Reference Count Count Count Radionuclide SOP# Conci 2 Sigma MDC Units # Volume Units Date Date Time Units Flag Values 1-3 ] 2003 7.01E-01 6.63E-01F 1.06E+00 pCi/g i I 2.27 gwet _ _ _09/08/11[ 60 M U

.G-108M 2007 1 8.47E-03 T 2.61E-02 F4.45E-02 pCi/g Dry _ 1 317.11 1 gdry 108/10/1100:00 1 08/26/111 5400 Sec U No 1-40 2007 1 1.65E+01 T 1.47E+00 I 4.74E-01 pCi/gDry 1 I__

1317.11 1 gdry 108/10/11 00:00 1 08/26/111 5400 1 Sec + I Yes IN-54 2007 1 1.47E-02 T 3.64E-02 6.10E-02 F pCi/g Dry 1 _ 317.11 gdry 108/10/11 00:00 1 08/26/11 5400 1 Sec UI UNo1

'0-58 F 2007 1 -3.92E-02 T 3.97E-02 5.67E-02 I pCi/gDry 1 _ 317.11 1 gdry 108/10/11 00:00 1 08/26/111 5400 1 Sec I U I __ -No--*-

ý0-60 _ 2007 1 1.47E-02 T 3.84E-02 F 6.63E-02 I pCi/gDry 1 _ 317.11 1 gdry 108/10/1100:00 1 08/26/111 54001 Sec uU _ 1No [

.G-110M F 2007 1 -1.55E-02 I 3.39E-02 F 5.27E-02 I pCi/gDry 2 _ 317.11 1 gdry 108/10/1100:00 1 08/26/111 5400 Sec IU F 1No 1 B-124

'S-134 F

F 2007 2007 1 -7.45E-03 1 4.57E-02 1-2.83E-02 3.74E-02 F 6.34E-02 F 4.77E-02 F

1 pCi/g Dry pCi/g Dry i

1

[

317.11 317.11 11 gdry 108/10/11 00:00 08/26/11 gdryIO8/10/1100:00 08/26/11!

5400 1 54001 Sec Sec 1 U F_

FU FNo[ No [

'S-137 F 2007 1 -1.77E-03 T 3.56E-02 F 5.77E-02 F pCi/g Dry 1 I 317.11 1 gdry 108/10/1100:00 1 08/26/11 5400 1 Sec IU _ I No 1 A-140 F 2007 1 -5.17E-02 T 8.42E-02 F 1.19E-01 I pCi/g Dry _ i 317.11 1 gdry 108/10/11 00:00 ! 08/26/111 5400 1 Sec FU F I No I U-152 F 2007 1 2.39E-01 3.02E-01 I 1.63E-01 I pCi/gDry 1 1 317.11 1 gdry 108/10/11_00:0.--0F68/26/1 5400 1 See IU I I Yes I U-154 I 2007 1 5.66E-02 T 8.28E-02 F 9.81E-02 I pCi/g Dry 1 [ 317.11 I gdry 08/10/-1100:00 1 08/26/111 5400 1 Sec IU 1 I No 1 H-228 220-0-fI -5.37E-01T 9.57E-02 1.01E-01 F pCi/g Dry 1  ! 317.11 g dry 108/10/1100:00 1 08/26/111 5400 1 See I + 1 I Yes 1 H-232 1 2007 1 6.72E-01 T 1.94E-01 F 1.64E-01 F pCi/gDry 1 1 317.11 1 gdry 108/10/1100:00 1 08/26/111 5400 1 Sec F+ F 1Yes 1

,lag Values 3 Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum

-J* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma igh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported gpec = MDC exceeds customer technical specification are related only to the samples tested in the condition they

-I = Low recovery are received by the laboratory.

High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 4 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 _"P! BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon

)ustin Miller/ Corey DeWitt EN005-3EIJOFA- 10 Sample ID: 003 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-3 Activity Uncertainty Run: Aliquot Aliquot Reference Count Count Count Radionuclide SOP# Conc 2 Sigma MDC Units # Volume Units Date Date Time Units Flag Values

'-3 2003 1 3.19E-01 6.84E-01 T 1.11E+00 pCi/g 2.16 gwet - __ 09/08/11 60 M_ U -

,G-108M 2007 -1.33E-02 2.98E-02 l 4.57E-02 pCi/g Dry I---326.11 gdry 08/10/1100:00 1 08/29/111 4732 ' Sec U1 [ -No

.-40 1 2007 1.70E+01 1.44E+00 1 4.68E-01 pCi/g Dry -- I 326.11 gdry 08/10/1100:00 1 08/29/11! _ 4732 _Sec ---.. +1 __ _,Yes IN-54 2007 2007 3.63E-03 3A.7E-02 5.71E-02 pCi/g Dry -I 326.11 gdry 1008/10/1100:000 08/29/11 4732 S No 2007 1.35E-02 4.08E-02 6.88E-02 pCi/g Dry 1 ~ 326.11 gdry 108/10/11 00:00 1 08/29/11 4732 Sec 71.U _ No 8.56E-03 3.81E-02 6.48E pCi/g Dry I 326.11 326.11

.... g dry g dry-108/10/11 00:00 108/10/11_00:00 08/29/li1_ 4732 1 e 08/29/11 47321 Sec U

lU . __I No No 2007 -.

G-1 10M -2.32E-M02 3.48E-02 5.37E-02 pCi/g Dry ~~1 - 326.11 gdry 108/10/1100:00 08/29/11 4732 M Sec U I No_o B-124 2007 1.41E-02 4.41E-02 6.89E-02 pCi/g Dry 326.11 igdry 1 08/10/11 00:00 08/29/11 4732 7 ec* U* ..... tNo S-434 *1.52E-02 3.86E-02 5.30E pCi/g Dry 326.11 g dry 1 -s/i-0-/i1--- 08/29/11: 4732. Sec 1U  ! No S-137 A-140 S2007 2007 7.1 OE-03

-2.49E-02 3.62E-02 8.36E-02 5.90E-02 i.27E-01o [

pCi/g Dry "v--I 326.11 g dry -1 0/i-1I00:00

- 08/29/171-- 47-32*-- Sec FU No 7 326.11 - d 08/10/11 00:00 08/29/11 4732--sec U _ - No U-152 2007 6.02-E-03 .17E-01 I.75E pCi/g Dry ~1 32611 g dry o108/10/11 00:00 08/29/11-43 Sec IU -- 1-No U-154 2007 3.40E-02 5.36E-02 9.32E-02 pCi/g Dry 1~~~~~ ___ 326.11 gdry 108/10/1100:00I 08/29/11 4732 Secc - Ye A-226 2007 1.31E+00 8.46E-01 1.07E+00 ~~1 _____ 326.4- g dry_ 08/10/1i 00:00 1 08/29/11F 4732 1 See + Yes fI-228 2007 6.28E-01 8.72E-021 9.33E-02 pCi/g Dry H-232 p~/ Dry 4.99E 1.60E-01 2.04E-01 1~ F 326.11 ggdry 108/10/1100:00 1 08/29/11 4732 I Sec + YesI

'lag Values j = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum

= Activity concentration exceeds MDC and 3 sigma; peak identified(gamrna only) Yes = Peak identified in gamma spectrum J* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma

-igh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported

'pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they

= Low recovery are received by the laboratory.

1 = High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 5 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 ~

• ! BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 004 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: -CollectStop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-4 1 Activity Uncertainty 1Run! Aliquot Aliquot I Reference Count Count i Count Radionuclide SOP# Conc 2 Sigma I MDC Units # Volume Units Date Date Time Units Flag Values

[-3 2 1.88E+001 6.67E-01 1.01E+00 pCilg 2.38 gwet 009/08/11 60 -- MA +1

,G-108M -2007 1 -2.11E-02 I 3.09E-021, 4.87E-02 I pCi/g Dry 1 _ 1 331.1 i gdry 08/10/11i00:00 8 /29/11 4759 1 Sec I U 1 Nou1-e-406 -- 2007 I1.83E+01] 1.67E+00 1 5.67E-01 FpCi/gDry 1 1 331.1 1 gdry 08/10/1100:00 08/29/11[ 4759 Sec 1+ 1 -- -C -

IN-54 1 2007 2.14E-021 3.71E-02 1-6.35E-02 F pCilgDry 1 _ 331.1 gdry 108/10/1100:00 1 0G829/11 4759 1- Secc Ul,-_-_ No

'0-57 F 2007 2.22E-01 5.13E-02 pCigDry 1 331.1 gdry 08/10/11 00:00 1 08/29/11 5 Se -+ s e4.83E-02 0-58 F2007 -3.07E-02 1 4.20E-02 . 6.17E-02 pCi/g DryI yr 1T 331.1  ! gdry i 08/10/11 00:00 ! 08/29/-1 4759- Sec 1U ITNo

'0-60 2007 1 1.08E-0I 5.86E-02 1 6.19E-02 I pCi/gDry 1 331.1-1 gdry 108/10/1100:00 1 08/29/* 1* 4759 ec + es G-1 10M 1 2007 1-1.68E-02 I 3.81E-02 1 5.95E-02 iTpCi/gDry 1 [ 331.1 1 gdry 108/10/10/100:00 08/29/111 47591 Sec jU- _ No- -

B-124 -200-7 19.06E-04 I 5.31E-02 1 7.54E-02 F pCi/g Dry 1 _ 331.1 ggdry 108/10/1100:00I 08/29/Ill 4759 Sec 1U 1 1No

S-134

'S-1374

• -2007 2007 1 5.55E-03 I 4.29E-02 1 6.16E-02 F pCi/gDry 1 7.37E-041 3.75E-02 1 6.12E-02-[ pCi/gDry 11 1 331.1 331.1 1 1 gdry gdry 08/10/1100:00 1 08/29/11 108/10/1100:00108/29/il 49 4759 S

-Sec u1 uI No No T1I]

A-140 2007 1 1.91E-02 1 8.29E-02 1 1.42E-01 F pCi/g Dry 1i__ 1 331. 1 gdry 108/10/11000:00 189 4759 I S U 1 1No U152 -2007 I 6.87E-01 1 1.38E-01 1 1.49E-01 1 pCi/gDry 1 1 331.1 gdry 108/10/11 00:00 108/29/11T 4759 1 Sec I+ Y U-154 1 2007 1 4.50E-011 9.78E-02 1 1.48E-01 7pCi/g Dry I 331.1 g dry 108/10/11 00:00 108/29/11! 4759 1Sec U* No H-228 1 2007 5.47E-01 1 .10E-011 1.16E-01 I pCi/gDry 1 1 331.1 1 gdry 0810/1100:00 82 11 47591 Sec 1+1 _ -Yesl H-232 1 2007 i 6.31E-01 1 2.11E-01 1 2.40E-01 I pCi/gDry 1 331.1 1 gdry 108/10/1100:00 1 08/29/111 4759 1 Sec + Yes 7lag Values

= Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum k = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum

• J* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma

'ligh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported 3pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they

= Low recovery are received by the laboratory.

-1 = High recovery MDC - Minimum Detectable Concentration 3olded text indicates reportable value. Page 6 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 " o BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 005 Collect Start: 08/1012011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-5 1 Activity !Uncertainty 'Runf Aliquot Aliquot Reference Count Count Count Radionuclide, SOP# Conc 2 Sigma - MDC Units # Volume Units! Date Date Time Units Flag Values

[-3 -2003 1 1.54E+0 6.68E-01 1.03E+O0F pCi/g I 2.34 gwet '_ 09/08/111 60 _ M + F

,G-108M 2007 1 -2.82E-02 F 2.67E-02 I 3.94E-02 F pCi/gDry 1 358.1 1 gdry 108/10/11 00:00 0608/29/11 4761 F Sec U _ _No__ -

AO40__

{N-54 2007 F 2007 __.6_+_

1 1.98E-02 1.33E__

3.36E-02 1 35.89E-02 I pCi/gDry pCi/gDry D _

1 358.1 358.1 gdry gdry F108/10/1100:00 F8/29/1l 108/10/1100:00 F08/29/11 4761 4761, See Sec I

F + i Fes F -No

.0-57 2007 3,0-01E i 5.18E-02 I 4.40E-02 F pCi/gDry 358.1. gdry Fo0/1/l1000 8/101F F08/2D/ 4761 F 10-58 F 2007 1 -9.55E-03 I 3.65E-02 5.91E-02F pCi/gDry _ F 358.11 gdry 108/10/11 00:00 1 08/29/111 4761 F Sec j U __ FNo 0-60 F 2007 1 1.16E-O1i 5.91E-02 I 4.74E-02 F pCi/gDry FT 358.1 1 gdry 108/10/1100:00 F 08/29/111 47611 NSee 1+1 i FYes I

,G-110M B-124 1 2007 1 -3.01E-03

]20071! 2.87E-02 1

-2.07E-02 I 4.03E-02 6,17E-02 F pCilg 1 4.49E-02 pCi/gDry Dry __ F 358.1 358.111 gdry gdry 00:00 F08/29/1 08/10/1100:00 108/10/11 08/29/11I 4761 - F Sec IU 4761 [ U ____ i -No I No j.S-134 1 2007 I 7.30E-03 I 3.61E-02 i5.16E-02 F pCi/g Dry I_ 358.1 1 gdry F08/10/11/00:00 1 08/29/111 4761 See I U No

S-137 T 2007 -4.11E-02 3.12E-02 358.1 gdry 00 00:00 08/1 Sec U- - -N

.A-140 V-4.02E-02 I 7.62E-02 1 1.13E-01 F pCilgDry F 358.1 gdry F08/10/1100:00 F 08/29/111 4761 -Se U-- __--No1 U-152 F 2007 1 7.26E-013 1.64E-01 [ 1.42E-01 F pCi/gDry 358.1 I gdry F08/10/1100:00 F 08/29/1-* 4761 F Sec *77 F _ -

U-154 F 2007 1 4.53E 7.25E-02 T 1.39E-01 pCi/g Dry I_ 358.1 1 gdry F08/10/1100:00 1 08/29/11 4761 1 Sec iU*F FNo I

,C-228 7 2007 -42.55E+00 1.52E+00 I 2.18E-01 I pCi/gDry I F-__358.1 I gdry F08/10/11 00:00 V08/29/11 4761 1 Sec l+U IYesN -

'11-228 F 2007 - 5.34E-01 I 9.03E-02 1 1.01E-01 F pCi/g Dry I F 358.1 I gdry 108/10/11 00:00 F 08/29/111 4761 Se_ is.

-SeT-- i H-232 1 2007 1 5.63E-01 I 1.67E-01 I 2.08E-01 F pCi/g Dry 1 F 358.1 gdry FP0810/11 00:00 08/29/Tl 4761 Se 1 +Yes

?lag Values J = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum F = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum J* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma Jigh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported 3pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they

= Low recovery are received by the laboratory.

A = High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value. Page 7 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 BROWN ENGINEERING, INC.

ATeledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA- 10 Sample ID: 006 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-6 Activity Uncertainty Run Aliquot Aliquot Reference Count Count Count Radionuclide SOP# Conc 2 Sigma MDC Units # Volume Units Date I Date Time Units Flag Values

[-3 1 2003 1 8.01E+00 8.63E-01 1.08E--00 pCi/g 1 2.23 ] gweti [09/08/111 60 1M [-+- _____

,G-108M 2007 1 -3.08E-03 1 3.34E-02 F5.34E-02 1 pCi/gDry 1 289.48 1 gdry 108/I0/1100:00 1 08/29/111 4761 1 Sec KIU1 1No 1

-40 1200 F 1.89E+011 1.46E+00 3.47E-01 I pCi/gDry 1 289.48 1 gdry 08/10/i100:00 08/29/I 4761 -1 Sec ++I I IYes 4N-54 2007 F 4.24E-02 1 7.30E-02 F7.52E-02 pCi/gDry r 1  ! 289.48 1 gdry 108/10/11 00:00 1 08F29/1- 4761 See IU I IYes_

'0-57 1 2007 F 1.48E+00I 7.23E-02 F 6.1OE-02 [ pCi/g Dry 1 _ 289.481 gdry 108/10/1100:00 1 08/29/111 4761 -Sec 7+ 1 _ Yes_

'0-58 F-2007 F- 6.25E-03 1 5.16E-02 F8.50E-02 1 pCi/g Dry 1 _ 289.48 1 gdry 108/10/1100:00 1 4761 1 Sec FU No 1--

'0-60 200-7 4.47E-01 1 6.69E-02 [6.4 1E-02 1 pCi/g Dry 1 IF 289.48 1 gdry 108/10/1100:00 1 08/29/I11 4761 1 Sec I +F7FI T1Yes-1

,G-1IOM 1 2007 F 2.83E-02 I 4.17E-02 F 7.21E-02 I pCi/gDry 1 I 289.48 1 gdry 108/10/1100:00 1 08/29/11 4761 1 Sec IU I No I B-124 2007 8.97E-02 I 6.39E-02 TF 1.02E-01 1 pCi/g Dry 1 289.48 1 gdry 108/10/1100:001 08/29/111 4761 i Sec 1U F IN0o -

!S-134 [ 2007[ 1.70E-01 1 5.78E-02 F6.91E-02 I pCi/gDry 1 1 289.48 1 gdry 108/10/1100:00 1 08/29/1 11 4761 1 Sec I + _I IYes----

'S-137 1 2007 F -1.57E-02 4.60E-02 F 7.45E-02 1 pCi/g Dry 1 1 289.48 1 gdry 108/10/1100:00108/29/11 476161 Sec F U- _ I Nol A-140 207 -5.50E-03 1 9.74E-02 F 1.56E-01 I pCi/g Dry 1 [ 289.48 1 gdry 108/10/1100:00[ 08/29/11! 4761 1 Sec I w __ I No1 1U-152 1 2007 F 4.69E+001 2.66E-01 F 1.78E-01 1 pCi/g Dry 1 [ 289.48 1 gdry 1 08/10/11 00:00 108/29/11 4761 1 Sec K+ _ - Yes I

,U-154 1 2007 1 3.19E-01 1.65E-01 F 3.01E-01 I pCi/g Dry I 1 289.48 1 gdry 108/10/1100:00 1T08/29/11[ 4761 1 Sec I U* I __ INo-F-11-228 1 2007 I 6.46E-01 I 9.18E-02 I 1.16E-01 pCi/g Dry I 289.48 1 gdry [08/10/1100:00 08/29/11 on 46 Sec F 1YeslI 11-232 1 2007 F 6.78E-01 1 2.22E-01 2.58E-01 i pCi/g Dry I 1 289.48 , gdry 08/10/11 00:00 08/29/111 4761 1 Sec + !YesI Flag Values U = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum I- = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum LJ* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma F-igh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported Spec = MDC exceeds customer technical specification are related only to the samples tested in the condition they L = Low recovery are received by the laboratory.

H = High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 8 of 16

Report of Analysis

• TELEDYNE 09/13/11 08:58 BROWN ENGINEERING, INC.

ATeledyne Technologies Company L47414 Enercon

)ustin Miller/ Corey DeWitt EN005-3EUOFA- 10 Sample ID: 007 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-7 Activity Uncertainty I Run I Aliquot I Aliquot Reference Count Count ' Count Radionuclide SOP# Cone 2 Sigma MDC Units i # Volume Units Date Date Time I Units Flag Values

-3 -S_ .03 E+00- 1.07E+00 pCilg 2.25 T g wet 09/08/11 60 - +--

G-108M 08/

-1.30E-02 2.77E-02 4.27E-02F pCi/g Dry 337.28 I g dry .10/1100:00 68/29 /1 3804 Sec IU I No-

-40 2.20E+01 1.519 P+00 403 EJ-17 1 2007 pCilg Dry 337.28 g dry /i0/1100:00 08/29/I - 3804 7 Sec Yes rN-54___ - I~o2E~7 3.87E-02 p~i/g Dry 337.28 08, 6.46E-02 g-dry- 08/ 10/1100:00 08/29/11,1 38-04 Sec - o 7 0-57 I1.52E-01I 5.30E-02 4.98Ei-02 pCilg Dry . 337.28 - g dry 08/'10/1i 00:00 08/29/11i- 3804 Sec- S-u Yes A1I

' 7 i .24E-02_ 10111i 00-:00 0 82/26if! 3804-'

0-58__ T-2007 3j.99Eý-02 6.3 1E02 pCi/g Dry 1 337.28 g dry i08/'10/1i00i00- 08i298 i. 3804i Sec ,No b- O-M-- 2.48E 3.49E-02 6.29E pCi/g Dry 337.28 08/ 10/11 00:00 ' 08/29/11 M3804 Sec

-gd*y 1 2007 8.33E -03 31.8E 5.56F-02 -i pCi/g Dry 337.28 g dry 08/ Sec B-124 2007 S-1.04E-03 4.5413-02 I 6.55E-02 S337.28

- g dry

'10/lI 00:00 108/29/Il 380 U--

_TDO07 1 dt~g Dry 10/1100:00 08/29/11) -3804 See -No S-134 T 2007 17.17E-02 -i 3.43'E-02 5.17E,02 pCi/g Dry 337.28-1 g dry 108/ '10/1100:00' 08/29/1 -3864 1 Sec No S-137 2007 -8.8 1E-03 3.46E-02 5.60E-02 pCig Dry '10/1i 00:00 j08/29/1ii 3804 Sec No!

A-140 -1.84E'-02 33.28-g dry 08, ti

- 2007 7.4613-0i 1.35E-_0I pCi/g EIry 337.28 g dry F08/'10/11 00:00' 08/29/i-1 3804- 1 See SeC No U-152 _2007 4.89E-01 1.38E-01 I .52E-01 pCi/g Dry- 337.28 -; 08/'10/11 00:00 -1 08/29/11' 3804-] Sec U

• 2007 g dry U-154 - 2007 L6if-_01 V1.36E -pC/g pCi/g Dry 3337.28 'No 2007 -Sec,-

11-228 -

9.61"-1 T 9.48E-602 _9.16_E-02' pCi/g Dry 33-7.28- _g dry 08/ 10/1100:00 -08/29/11- 3804 YesI H-232 --.... 08/ 5cc_

6.69E-01 2.1613-01 2.23E-01 g dry '10/1100:00 '-08/29/11 3804 1 7+

'lag Values j = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum F = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum 3* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma tigh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported 3pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they

= Low recovery are received by the laboratory.

= High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 9 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 Prk~ BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 008 Collect Start: 08/10t/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: 1,47414-8 Rdould py Activity Uncertainty Uis RunI Aliquot 1Aliquot Reference ICount Count fCount Radionuclide SOP# Conc 2 Sigma MDC Units Volume Units I Date Date Time Units Flag Values

[-3 [ 2003 F 2.52E+00F F -1.42E-02-] 7.06E-01 1.04E+00 I pCi/g I 4.78E-02 F pCi/g Dry 1 l 2.3 F 317.34 11 gwet I _ 09/08/11 60 TT M FIFF+ __ F

,G-108M [ 2007 F 2007 2.81E-02 gdry 08/10/11 00:00 F08/29/11 4756 Sec

-40 F 2.18E+01

-1.48E+00 4.22E-01 pCi/gDry 1 F 1 317.34 1 gdry 10 /1100:00 08/29/11 4756 1 Sec F+ ____

___e_

4N-54 F-2007 F 6.28E-04 T 3.82E-02 F 6.69E-02 pCi/gDry j _ 1 317.34 1 gdry 108/10/1100:00 1 4756 1 I08/29/11 Sec iU I No 1

'0-57 1202.38E-01 1 5.38E-02 1 5.00E-02 I pCi/g Dry 1 _ 317.34 1 gdry 108/10/1100:00 F08/29/1IT 47-56 T 5cc -ý-F[-Yes

'0-58 . 2007 -1.88E-02 I 3.60E-02 I 6.06E-02 I pCi/gDry 1 _ 317.34 1 gdry 108/10/1100:00 1 4756 1 Sec -U ___o_

'0-60 F 2007 F 7.50E-02 T 4.24E-02 F 7.74E-02 I pCi/g Dry 1 _ 317.34 1 gdry 108/10/1100:00 08/29/11l-4756 Sec VU _ No 1

,G-110M F2007 F 1.18E-03 I 3.33E-02 F 5.88E-02 1 pCi/gDry 1 F 317.34 1 gdry 108/10/1100:00 F08/29/111 4756 1 SeciF U ___ iNo F-B-124 :2007 F -3.34E-03 I 4.68E-02 I 7.05E-02 1 pCi/gDry 1 F 317.34 1 gdry 108/10/1100:00 6-08/29/i1T 4756 1 Sec I U F FNo 1

S-134 2007

F -8.39E-03 T 3.79E-02 F 1 5.63E-02 pCi/gDry 1 317.34 1 gdry 108/10/11 00:00 1 08/29/11 4756

• Sec KU7- No1

'S-137 F 2007 F -2.04E-02 1 3.69E-02 F 6.20E-02 I pCi/g Dry _ F 317.34 1 gdry 108/10/11 00:00 F 08/29/111 4756 1 Sec I U I -I N--I-

,A-140 F 2007 F -4.79E-02 I 8.41E-02 F 1.47E-01 I pCi/gDry _ F 317.34 1 gdry 108/10/11 00:00 08/29/1I 4756 1 Sec FIFi No- 1 U-152 F 2007 8.07E-01 1 T 1.48E-01 F 1.69E-01 1 pCi/gDry _ F 317.341 gdry 108/10/1100:00 F08/29/11 4756 1 See F + Yes --

-U-154 F 2007 4.82E-01 T1.09E-01 F 1.44E-01 1 pCi/gDry 1_ 317.34 gdry 108/10/1100:00 F08/29/11d 4756 T Sec U*F __ [ No i

,C-228 F 2007 F 1.64E+00 I 9.18E-01 I 2.27E-01 I pCi/gDry 1 _ 317.34 1 gdry 108/10/1100:000 08/29/111 4756 1 Sec +I FYes 1 I 2007 1I F 11 1 t+

228 8.02E-01 I 1.22E-01 I 1.29E-01 I pCi/gDry 317.34 gdry 08/10/1100:00 1 08/29/11 4756 Sec [Yes 1t-232 I 2007 I 9.09E-01 I 2.16E-01 ]2.17E-01 pCi/g Dry I 317.34 gdry 108/10/11 00:00 F 08/29/111 4756 1 Sec [7

___ Yes 1

-lag Values J = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum F- Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma Ii-gh = **** Unless otherwise noted, the analytical results reported Activity concentration exceeds customer reporting value

'pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they Low recovery are received by the laboratory.

High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 10 of 16

Report of Analysis ,A* TELEDYNE 09/13/11 08:58 BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA- 10 Sample ID: 009 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-9 Activity Uncertainty Runj Aliquot Aliquot I Reference J Countj Count Count Radionuclide SOP# Conc 2 Sigma MDC Units # Volume Units Date Date Time Units Flag Values

_-3 2003 F 2.16E+00 7.59E-01 1 1.15E+00 6 pCi/g 1  ! 2.09 1 gwet 1 -1 09/08//11 60 M + I

_G-108M F2007 1.27E 3.28E-02 i 5.51E-02 pCi/gDry 1_ 337.38 gdry 108/10/11 00:00 1 08/9/111 3594 i Sec 7 UFT i No I

-40 2007 2.13E+01 2.03E+00

! 7.34E-01 I pCi/gDry _ 337.38 1 gdry ;08/10/1100:00 108/29/HF 3594 1 Sec I *] -YesiF-IN-54 F 2007 F 2.36E-03 1 4.82E-02 8.05E-02 I pCi/gDry 1 _ 337.38 1 gdry 108/10/11 00:00 1 08/29/11[ 3594- Sec K 7 oi-No 0-57 2007 Fl.56E-1l 6.56E-02 5.44E-02 i pCi/g Dry 1_ 337.38 gdry 108/10/1100:00 1 08/29/111 3594 T Sec + IYes1

.0-58 F 2007 -3.04E-02 I 4.55E-02 1 6.97E-02 pCi/g Dry _ 1 337.381 g dry 08/10/1100:00 1 08/29/111 3594 T Sec F F No [

0-60 F 2007 F 3.60E-02T 5.02E-02 8.93E-02 T pCi/gDry 1 337.38 1 g dry 108/10/1100:00 1 0/229/- 359 Sec U U _ INo

.G-IlOM F 2007 F 1.76E-02 1 3.65E-02 1 6.44E-02 I pCi/gDry __ 337.38 1 gdry 108/10/1100:00 1 08/29/11I 3594 Sec IUF I B-124

ýS-134 F-2 F 2007

-3.62E-03 3.27E-021 5.58E-02 4.72E-02 1 8.17E-02 17.17E-02 II pCi/gDry pCi/gDry 1

1_

337.38 337.38 1 1 gdry gdry 108/10/11 00:00 08/29/11 108/10/1D00:00 I0829/

3594 3594 T Sec I U SecFU--T--_

I No II

-No-I

S-137 A-140 FF 07-F -9.65E-03 1 2007 I -4.40E-02 1 4.06E-02 1.I5E-01 1- 6.67E-02

[1.77E-01 II pCi/g Dry pCi/gDry 1

1__

_ F 337.38 337.38

[g dry 1 gdrydy 1 08/10/11 00:00 08291 108/10/1100:00 0No 08/29/11 3594 T* c 3594 Sec K-U- F ___ No0 U-152 2007 7.63E-01 T 1.87E-012 1 1.33E-01 i pCi/gDry 1 1 337.38 1 gdry 108/10/1-1000:00 0-8/29/11I 3594 ] Sec KN + __ IYes1_1 U-1540 2007 F 3.36E-01 8.37E-021 1.55E-01 I pCi/gDry _ 1 337.38 1 g dry 108/10/1100:00 1 --/29/II 35-9-4 Sec FU* F 7----1N H-228 F 2007 F 8.37E-01 1.76E-01 1 1.23E-01 I pCi/gDry 1 J 337.38 1 gdry 108/10/1100:00 08/29/111 3594 7 Sec F +7 TYes I[

H-232 72007 F 8.12E-01 1 3.OOE-01 I 2.35E-01 pCi/g Dry _ 337.38 1 gdry 108/10/lI 00:00 08/29/11! 3594 See*Ys "lag Values J =

Compound/Analyte not detected (< MDC) or less than 3 sigma No =Peak not identified in gamma spectrum I- =

Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only)

J* =

Yes = Peak identified in gamma spectrum Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma

{igh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported 3pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they are received by the laboratory.

11 = Low recovery High recovery MDC - Minimum Detectable Concentration Solded text indicates reportable value.

Page 11 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 P'~ BROWN ENGINEERING, INC.

ATeledyne Technologies Company L47414 Enercon E)ustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 010 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-10 Activity Uncertainty M Run Aliquot Aliquot Reference I Count Count Count Radionuclide SOP# Co2c 2 Sigma MDC Units Volume I Units Date Date Time Units Flag Values

[-3 2003 F 1.14E+00 1-7.14E-01 I 1.12E+00 pCi g 2.14 gwet _ _ [ 09/08/111 60 i M [ + I I I

,G-108M I 2007 1 1.50E-02 I 2.48E-02 1 7 431E-02 I pCi/gDry 349 1 gdry 108/10/1100:00 T 08/29/11 4761 6 Sec[ U 1No- 0 1-40 S2007 j 1.84E+01] 1.42E+00 I 2.77E-011 pCi/gDry 1 349 1 gdry 108/10/1100:00 T 08/29/11it 4761 1 Sec + 1 [Yes--

1 IN-54 i 2007 1 2.59E-02 3.42E-02

- 1 6.07E-02 I pCi/g Dry I_ 349 1 gdry T08/10/11 00:00 08/29/11T 4761 1 Sec [ U _ [No V

'0-57 1 2007 1 1.49E-01 I 4.38E-02 3.41E-02 I pCi/g Dry 349 gdry 108/10/11 00:00 1 08/29/111 4761 1 Sec I + I i Yes ii' 0-58 I2007 1 3.03E-03 1 3.52E-02 1 5.89E-02 I pCi/g Dry 1 [ 349 gdry 08/10/1100:00 1 29/11 4761 1 See KU _ VNo

0-60 [2007 1 6.82E-02 5.13E-02 i 5.35E-02 1 pCi/gDry 1 _ 349 1 gdry 08/10/11 00:00 08/291/111 4761 1 See U I[Yes LG-I10M [2007 1 7.25E-03 I 2.94E-02 1 5.08E-02 pCi/gDry 1 349 I g dry 108/10/11 00:00 1 08/29/111 476i1 [Sec F U [ No I B-124 20071 -1.89E-02 1 3.87E-02 1 5.93E-02 1 pCi/gDry 1 [ 349 1 gdry 108/10/1100:00 [ 08/29/111 4761 1 Sec T U 1 _ No H

'S-134 1 2007 1-1.55E-02 I 3.60E-02 1 5.33E-02 pCilgDry 1 [ 349 gdry 108/10/1100:00 F08/29/ 11 4761 1 Sec I U 1 - [No T S-137 F 2007 1 -9.38E-03 I 3.21E-02 1 5.27E-02 1 pCi/gDry 1_ 349 1 gdry j08/10/11 00:00 F 08/29 1 4761 1 Sec 1 No I A-140 2007 1 -5.42E-02 1 7.96E-02 1.07E-01 1 pCi/gDry 1 349 gdry 08/10/1100:00 [ 08/29/11 4761 1 Sec FUl [

[No

U-152 F 2007 1 4.25E-0i1 1.29E-01 I 1.37E-01 1 pCi/gDry 1 _ 349 1 gdry 108/10/1100:00 1 08/29/111 4761 1Sec---+/-1 T [Yes

,U-154 F 2007 1 3.01E-01 8.86E-02 1 9.99E-02 I pCi/gDry 1 -- F-349 I gdry 108/10/1100:00 1 08/29/11 4761 1 Sec FU*1 ___ Noo

,C-228 I 2007 1 5.12E-01 1 2.12E-01 i1.80E-01 I pCi/gDry 1 [ 349 1 gdry 108/10/1100:00 F 08/29/11 4761 1 Sec F+ 1 [Yes I 11-228 I 2007 1 6.41E-01 8.47E-02 1 8.33E-02 1 pCi/g Dry 1 F 349 1 g dry 108/10/11 00:00 1 08/29/11F 4761 1 Sec [T+ [ Yes1

-- 232 *20-07 1 5.80E-01 1.70E-0 1 1.83E-01 pCi/gDry 1 [T 3491 gdry 108/10/1100:00 08/29/11F 4761 1 Sec + I_ Yes-1-1 Flag Values U = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum 4- = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum U1* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma Fligh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported Spec = MDC exceeds customer technical specification are related only to the samples tested in the condition they L = Low recovery are received by the laboratory.

H = High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 12 of 16

Report of Analysis P9 TELEDYNE DBROWN ENGINEERING, INC.

09/13/11 08:58 A Teledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 011 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-11 Activity Uncertainty I Run1 Aliquot Aliquot Radionuclide SOP# Cone 2 Sigma MDC~ Units Volume Units 1 2003 1 7.651E+00 1 8.70E-01 I 1.10E+00 pCi/g _ 1 2.18 ' gwet _09/08/111 60 F M +

,G-10 8M 2007 -2.78i E-02 F 3.58E-02 F5.58E-02 1 pCi/gDry F_ 377.22 [gdry 108/10/11 00:00 1 08/29/ F 4759 T S-e U 1-40 2007 1 1.94E 1+01 1.79E+00 F 4.22E-01 1 pCi/gDry F_ 377.22 1 gdry 108/10/1100:00 1 08/29/1F 4759 F Sec IN-54 1 2007 F 8.001E-02 F 5.OOE-02 F 9.23E-02 pCi/gDry _ 1 377.22 1 gdry 108/10/1100:00 I 408/9/1- 4759 F Sec , U

'0-57 V 2007 3.881E-01 F 6.42E-02 1 5.38E-02 I pCi/g Dry F _ 377.22 1 gdry 108/10/1100:00 1 08/29/1l 4759 S 58I 2007 I -2.70E-02 5.28E-02 k8.E-02 V pCi/gDry F 1 377.221 gdry 108/10/11 00:00 08/29/11l 4759 F Sec U F_o__- N

'0-60 [ 2007 j 1.78E-01,T 5.69E-02 7.18E-02 1 pCi/gDry Fl 377.22 - gdry 108/10/1100:00 1 O/2-9/111 4759 F Sec I + F_ Yes_

,G-I10M 1 2007 - 2.93E-02 4.42E-02 F 7.75E-02 I pCi/g Dry _- 377.22 1 gdry 108/10/11 00:00 08/29/111 4759 F Sec [ U F _____ I No B-124 1 2007 1.76E-02 T 6.15E-02 F 9.18E-02 I pCi/gDry F _ 377.22 1 gdryl[8/10/1100:001 08/29/11 4759 F Sec I U FNoj

!S-134 2007 F 3.86E-02 I 5.02E-02 F 7.54E-02 I pCi/g Dry T _ 377.22 1 gdry 108/10/1100:00 1 08/29/11 4759 F Sec IU I No

S-137 I 2007 F -2.21E-02 1 4.73E-02 i7.63E-02 1 pCi/gDry 1 377.22 gdry 108/10/11 00:007 08/29/111 4759 1 Sec IU F 7 -No A-140 AU-152 1 2007 -6.56E-02 2007 F 1.14E+502 E 1.08E-01 1.87E-01 F

F 1.60E-01 E-01 -01 1 pCi/gDry pCi/gDry T 1 377.22 377.22 1

1 gdry 108/10/11 00:00 1 08/29/111 gdry1 4759 F Sec - U No 6Er 10/1/1 0wo wo10800 29/111 4759 1 See IU 1 Y NsT

,U-154 2007 F 7.84E-01 I 1.30E-01 f 1.84E-01 I pCi/g Dry _ I 377.22 1 gdry 108/10/1100:00 1 08/29/1-1[ 4759 F Sec +W* INo ii

,C-228 200 7.55E-011 3.23E-01 7 3.05E-01 I pCi/g Dry 1 1 377.22 gdry 108/10/1100:00 08/29/11, 4759 F Sec F + Fes I

,C-228 T 2007 7.84E-01 [ 1.00E-01 1.01E-01 I pCi/gDry T 377.22 gdry 1 08/10/1100:00 1 08/29/11T 4759 -Sec T + I -- lYes H-232 200( 5.09E-0101 1 2.91E-01 I pCi/gDry If 377.22 gdry 108/10/1100:00 08/-9/Sec +/- I Yes Flag Values U = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum F = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only)

U* = Yes = Peak identified in gamma spectrum Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma High = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported Spec = MDC exceeds customer technical specification are related only to the samples tested in the condition they L = Low recovery are received by the laboratory.

HI =

High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 13 of 16

Report of Analysis qLTELEDYNE 09/13/11 08:58 BROWN ENGINEERING, INC.

ATeledyne Technologies Company L47414 Enercon Dustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 012 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-12 Radionuclide SOP# Activity Cone C Uncertainty 2 Sigma MDC Units Run # Aliquot Volume Aliquot Units Reference Date Count Dat Count Count Units Flag Values Volum Unis__Daeat TimeUn

[-3 1V22003 75.01E+00 7.97E-01 1.09E+00 pCi/g i 1 2.21 F g wet 090_8_ j08/ 60 F M [ +

G-108M 2007 F- 1.75E-02 T 3.44E-02 F5.91E-02 I pCi/g Dry } [343.93 ] gdry 108/10/1100:00 1 08/29/111 3802 1 Sec U No

,-40 27-F 1.89E+01T 1.68E+00 F 5.82E-01 F pCi/gDry _ F 343.93 gdry 1 08110/ 100:00 + s

-Ye08/23Sec-F IN-54 1 2007 F 6.64E-02 F 4.72E-02 F 8.79E-02 1pCi/gDry _ 1 343.93 F gdry 108/10/11 00:00 1 08/29/-1 3802 c -- UFIU I No -

'0-57 12007 2 5.29E-O1 T 7.37E-02 F 4.73E-02 I pCi/gDry 1 1 343.93 1 gdry 108/10/1100:00 1 08/29/111 3802 F Sec F+ F IYes7

0-58 2 t.86E-02 5.17E-02 I 8.86E-02 pCi/g Dry I_ 343.93 1 gdry 1 08/10/1100:00 1 08/29/lU 3802 1 Sec Fu [ _ N

'0-60 2007 - 2.74E-01 1-2007 7.32E-02 F 7.30E-02 I K .03E-02 1 pCi/g Dry 1 11 1 343.93 g dry 108/10/1100:00 1 08/29/il 3802 F Sec + F IYes 1 o F7

,G-110M B-124

-1.93E-02 1 20-207-0 5.19E-02,I 3.76E-02 5.09E-02 F 8.13-2 pCi/g Dry pCi/g Dry ___

343.93 11 343.93 1 gdry g dry 08/10/11 00:00 08/29/11j 108/10/1100:00 1 08/29/11i 3802 3802 Sec I Sec FUU F___7I No1-1 -

'S-134 1 2007 F 5.37E-021 7E0 3.96E-02 F 5.63E-02 I pCi/g Dry 1 F 343.93 1 g dry 108/10/11 00:00 1 08/29/-1lj 3802 1 Sec 1,uKiYYeYs

!S-137 - 2007 F -2.00E-02 I 4.07E-02 F 6.55E-02 I pCi/gDry 1 ] 343.93 1 gdry 108/10/1100:00 1 08/29/111 3802 7-ý-cS . U [I_No A-140 2007 F 4.45E-04 1.11E-01 F 1.81E-01 pCi/gDry D _ 343.93 1 gdry 108/10/11 00:001 08/9/1 3802 ! Sec I UI I No I

,U-152 2007 F 1.6E+0 F2.11E-01 F I.68E-01 pCi/gDry 343.93 gdry 108/10/f100:001[08/29/11, 3802 F Sec F + _____ I Yes1

• U-154 2007 1.07E+00 1.49E-01 F 1.84E-O1Y pCi/g Dry I _ 343.93 1 gdry 108/10/1100:00 008/29/111 3802 F Sec I U* F_ No 1 A-226 - 2007 I 2.261E+00 1.32E+00 i 1.26E+00 1 pCi/gDry 1 _ - 343.93 g dry 108/10/11 00:00 1 08/29/111 3802 S--c F+ F IYes1--

11-228 1 2007 F 7.15E-01-F 9.69E-02 F 9.90E-02 pCi/g Dry 1 [ 343.93 1 gdry 108/10/1100:00 I 08/29/1ll 3802 1 Sec + I

+/- !Yes 1H-232 2007 F I 6.35E-01 2.09E-01 F2.87E-01 I pCi/g3Dry 1 -08/10/1100:00 08/29/11! 3802 _ Sec-_--__

Se343.93+gdry Ys-es_--

Flag Values IJ = Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum 1* = Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma

-igh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported 3pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they Low recovery are received by the laboratory.

J High recovery MDC - Minimum Detectable Concentration Bolded text indicates reportable value.

Page 14 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 P'! BROWN ENGINEERING, INC.

ATeledyne Technologies Company L47414 Enercon

)ustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 013 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

LIMS Number: L47414-13 Activity Uncertainty I Run Aliquot Aliquot Reference Count Count Count Radionuclide SOP# Conc 2 Sigma MDC Units I Volume Units Date Date Time Units Flag Values

-3 1 2003 1 2.91E+001 7.54E-01 1.11E+00 F pCi/g 1 2.17 I gwet 1 09/08/111 60 M 1+ 1 1

.G-108M 1 2007 1.93E-02 F 3.45E-02 5.85E-02 pCi/gDry 1 1 271.73 1 gdry [08/10/11 00:00 T 08/29/111 4756 1 Sec 1U No

-40 1 2007 F 1.71E+01 ! 1.52E+00 5.14E-01 pCilg Dry 1 1 271.73 1 gdry 108/10/1100:00 T08/29/111 4756 7 Sec + I_ Yes IN-54 1 2007 1 -1.53E-02 F 4.25E-02 6.72E-02 pCi/gDry 1 1 271.73 F gdry 108/10/11 00:00 T 08/29/111 4756 1 Sec U 1_ No 0-57 1 2007 1 2.42E-01 F 6.14E-02 I 5.47E-02 F pCi/gDry _ 271.73 1 gdry 108/10/11 00:00 T 08/29/111 4756 ! Sec + I iYesl 0-58 1 2007 1 -2.15E-03 F 4.47E-02 I 7.31E-02 F pCi/gDry F _ 271.73 1 gdry 108/10/11 00:00 T 08/29/I 4756 1 Sec UI 1 No 1 0-60 1 2007 1 3.62E-02 F 4.28E-02 i 7.79E-02 pCi/g Dry F 271.73 g dry 108/10/11 00:00 T 08/29/111 4756 1 See U 1 1 Nol1 G-110M 1 2007 1 -6.82E-03 F 3.68E-02 I 6.01E-02 F pCi/g Dry F _ 271.73 F gdry 108/10/11 00:00 1 08/2-/lI F 4756 Sec U 1 1I No I B-124 1 2007 , 5.44E-02 F 5.29E-021 8.72E-02 F pCi/g Dry F 1 271.73 I gdry 108/10/11 00:00 1 08/29/111 4756 1 Sec IU I [No 1-S-134 1 2007 1 4.93E-02 I 4.57E-02 17.31E-02 F pCi/gDry F _ 271.73 1 gdry [08/10/1100:00 08/29/111 4756 1 Sec U I_ No]

S-137 1 2007 1 3.01E-03 3.96E-02 I 6.63E-02 F pCi/gDry F _ 271.73 1 gdry 108/10/1100:00 1 08/29/111 4756 1 Sec I U 1_ No 1 A-140 1 2007 1 3.05E-02 F 9.60E-02 1 1.66E-01 F pCi/gDry _ 1 271.73 1 gdry F08/10/11 00:00 1 08/29/111 47561-- Sec I U I 1 No I U-152 1 2007 1 5.66E-01 F 2.22E-01 1 2.09E-01 I pCi/g Dry F _ 271.73 1 gdry F08/10/11 00:00 T 08/29/11 4756 1 Sec i + I 1Yes [

U-154 1 2007 1 4.89E-01 I 1.24E-01 1 1.69E-01 F pCi/gDry F _ 271.73 F gdry 108/10/11 00:00 1 08/29/1lI 4756 1 Sec i U* 1 1 No I C-228 1 2007 F 6.18E-01 2.82E-01 I 2.65E-01 F pCi/g Dry F 1 271.73 1 gdry 108/10/11 00:00 -108/29/11 47561 See 1 +! _ I _ s-[

H-228 1 2007 1 6.18E-01 1 1.1OE-0l 1.22E-01 I pCi/gDry I 271.73 1 gdry [08/10/11 00:00 1 08/29/111 4756 1 Sec I + I IYes V H-232 1 2007 1 4.74E-0.1 1 2.21 E-01 2.53E-01 [ pCi/g Dry I I 271.73 I gdry 108/10/1100:00 1 08/29/111 4756 See + 1 Yes I

lag Values

= Compound/Analyte not detected (< MDC) or less than 3 sigma No = Peak not identified in gamma spectrum

= Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes = Peak identified in gamma spectrum J* = Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma tigh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported

pec = MDC exceeds customer technical specification are related only to the samples tested in the condition they Low recovery are received by the laboratory.

I = High recovery MDC - Minimum Detectable Concentration 3olded text indicates reportable value. Page 15 of 16

Report of Analysis TELEDYNE 09/13/11 08:58 BROWN ENGINEERING, INC.

A Teledyne Technologies Company L47414 Enercon

)ustin Miller/ Corey DeWitt EN005-3EUOFA-10 Sample ID: 014 Collect Start: 08/10/2011 00:00 Matrix: Solids (SD)

Station: Collect Stop: Volume:

Description:

Receive Date: 08/16/2011  % Moisture:

JLMS Number: L47414-14 I Activity Uncertainty i Run Aliquot Aliquot Reference Count Count Count

,adionuclide SOP#* Conc 2 Sigma MDC Units # Volume Units Date Date Time Units Flag Values

.3 _ 2006-3 2.87E+001 7.04Ei-01 1.03E+0076 pCi/g t I 2.34 g wet I 09/09/1-1160 TM --- i 7-3-108M 1 2007 4.78E-03 1 3.8 1E-02 I 6.24E-02 pCi/g Dry I _ 322.41 g dry 108/10/ 11 00:00 08/29/11 3803 1 See I U I I No

-40 12007 I 1.73E+0l 1.55E+00 I 5.58E-01 T pCi/gDry 1 _ 322.41 gdry T08/10/i 11 00:00 7 08/29/111 3803 1 Sec 1 + _ 1Yes 1-N-54 12007 2.07E-02 4.50E-02 I 7.70E-02 T pCi/g Dry F __ 322.41 F g dry T 08/10/:[1100:00 F 08/29/111 3803 1 Sec- U _ No

- I

)-57 I 2007 ] 1.30E-01 -6.90E-02 55.35E-02 1 pCi/g Dry I 1- 322.41 gdry T08/10/111 00:00 F 08/29/11 3803 F Sec- F + 1 1Yes 1]

1 2007 F -6.23E-02 F 1-58 4.59E 6.39E-02 1 pCi/g Dry 1 322.41 T gdry 08/10/ 1100:00 1 08/29/111 3803 1 See I U j I No 1 11-60 1 2007 -6.50E-04 I 4.52E-02 I 7.46E-02 I pCi/g Dry I 322.41 F gdry T08/10/11 00:00 F08/29/111 3803 SecI] U V TNo 1 3-110M 2007 I 3.20E-02 4.05E-02 I 7.19E-02 pCi/g Dry I _ 322.41 1 gdry 108/10/11 00:00 1 08/29/111 3803 1 See 1 U I FNo T 3-124 1 2007 F -7.96E-03 5.80E-02 1 8.22E-02 pCi/g Dry I 322.41 I gdry 108/10/11 00:00 I 08/29/111 3803 1 Sec I U I I No 3-134 1 2007 I 1.50E-02 1 4.50E-02 I 6.73E-02 I pCi/gDry 1 1 322.41 1 gdry I08/10/11 00:00 I 08/29/111 3803 I Sec I U I I Not 1 2007 1 -2.15E-02 I 4.46E-02 j 7.1OE-02 [ pCi/g Dry j 322.41 [ g dry I 08/10/11 00:00 1 08/29/111 3803 F Sec I U I TINoF-k-1 40 1 2007 F 1.52E-02 K 1.14E-01 I 1.90E-01 I pCilg Dry I 1 322.41 F gdry T08/10/1100:00 F08/29/11II 3803 1 Sec I U I I No 1 J-152 I 2007 1 4.88E-01 I 1.73E-01 1.98E-01 pCilgDry 1 322.41 1 gdry I 08/10/11 00:00 I 08/25 3803 1 Sec I + I I Yes I 3-154 1 2007 F 2.62E-011 1.39E-01 I 1.49E-01 pCi/g Dry 322.41 1 gdry 08/10/11 w 00:00 F08/21 /1111 3803 1 Sec I U* i I No I 1-228 2007 I 6.53E-01 I L-34E-01 -ý2&0 I pCi/g Dry i 1 322.41 F gdry t08/10/1 00:00 F08/20 3803 1 See I + I I Yes! 232 2007 I 7.88E-01 I 2.23E-01 -- 2,32E-01 pCi/g Dry 1 322.41 ggdry 108/10/11 00:00 1 08/29/111 3803 I Sec I + I IYesi lag Values

= Compound/Analyte not detected (< MDC) or less than 3 sigma No Peak not identified in gamma spectrum

= Activity concentration exceeds MDC and 3 sigma; peak identified(gamma only) Yes Peak identified in gamma spectrum

=* Compound/Analyte not detected. Peak not identified, but forced activity concentration exceeds MDC and 3 sigma

• igh = Activity concentration exceeds customer reporting value **** Unless otherwise noted, the analytical results reported pee = MDC exceeds customer technical specification are related only to the samples tested in the condition they

= Low recovery are received by the laboratory.

I = High recovery MDC - Minimum Detectable Concentration lolded text indicates reportable value.

Page 16 of 16

University of Arizona Nuclear Reactor Lab D&D SERVICES Sample Collection Procedure ATTACHMENT B .7 W vi )-

IW NERCO N Serviced, Inc. Chain of Custody Record Laboratory C-ontart Fpajac: Ma~: s ~ 7 ~cotm't Aw~-~ COC va. 06 T~~V~ar iContact C;=tr.a:/ ~ .

AA 0 Phone: 02fc1j~

Date TSmE 140.tC~._~-Thne~~ef -~

Sipl oneatffica_ _ _ _ _

<1 <______no_

A) C cL ~ ~ That R~e'~.____ ____ ____ ___

2d 11eh=r-ashe6 ta R770 r~a ianm I

IluvilwanC i" dt,,: I ap e'~

D,:py'7I I 12 UA-MCP-RC-06, Rev. Rev. 00 12

LVI SERMES University of Arizona Nuclear Reactor Lab D&D Sample Collection Procedure ATTACHMENT B ENERCON SeryviceS, Inc. Chain of Custody Record Ste ouam

____________________lmaar Mke COC N~O: 00o phime 021 ays DO No.

Sample SAmple Sample Sample Idandiikrtdoii se Tm yiMorxto.m Ncoev Sample spE)cLýz

.rew*nsnon  !~d, 01 fr-2=L 3=R.SOý J=HNO, 6 ýYXXOH 0- Orboer Nme_____________

?PVSX? THaardIdo~oo Son~t DirvoojaffAfr may &Jarej 5 ifmpe ae reai amls~

I swnthj OIWHno,d QORADWOe O38WI4OIad OM1liCG Ou,.a- MR011-Madiaft OotrPCooI5Ub CkOct*~ for V=-.bs /Weeki Dan tcir-Je one)

Ve) A TV~~At 'eT /0 tehq-sý*

qoin. ~'co==yn Dja-me: Conip~mn:

UA-MCPRC-06 Rev.me 0d 12 :CMPM:DuF~a UA-MCP-RC-06, Rev. 0 12

08/17/11 10:27 Teledyne Brown Engineering Sample Receipt Verification/Variance Report SR #: SR28654 Client: Enercon Services, Inc. Project #: EN005-3EUOFA-10 LIMS #L47414 Initiated By: JSIMMONS Init Date: 08/17/11 Receive Date: 08/17/1.1 Notification of Variance Person Notified: Contacted By:

Notify Date:

Notify Method:

Notify Comment:

Client Response Person Responding:

Response Date:

Response Method:

Response Comment Criteria Yes No NA Comment 1 Shipping container custody seals present NA and intact.

2 Sample container custody seals present NA and intact.

3 Sample containers received in good Y condition 4 Chain of custody received with samples Y 5 All samples listed on chain of custody Y received 6 Sample container labels present and y legible.

7 Information on container labels Y correspond with chain of custody 8 Sample(s) properly preserved and in NA appropriate container(s) 9 Other (Describe) NA For Hazardous Materials Only:

10 Paperwork shows TBE and shippers name, NA address and phone number 11 Paperwork shows sample quantity NA information

4WTELEDNQE QC Summary Report for L47414 BRQWN: ENGINEERING ATeledyne Thechno.oafaes Company EN005-3EUOFA-10 09/13/2011 08:59 GAMMA Duplicate Summary 3E Sample ID Radionuclide Matrix Count Date/Time Original Result DUP Result Units RPD Range Qualifier P/F

'G12496-1 K-40 SD 08/29/2011 12:27 2.271E+01 2.097E+01 pCig Dry 8.0 <50 + P 47414-1 GAMMA Associated Samples for WG12496 Sample # Client ID L47414-1 001 L47414-2 002 L47414-3 003 L47414-4 004 L47414-5 005 L47414-6 006 L47414-7 007 L47414-8 008 L47414-9 009 LA7414-10 010 L47414-11 011 L47414-12 012 L47414-13 013 L47414-14 014 Positive Result Compound/analyte was analyzed, peak not identified and/or not detected above MDC

< 5 times the MTDC are not evaluated Nuclide not detected Spiking level < 5 times activity Pass Fail Page 1 0, Not evaluated

4U T:ELEDyNiE QC Summary Report for L47414 .BROWNENGINEER ING A Teledy-ne Trchrwologes Company EN005-3EUOFA-10 09/13/2011 08:59 H-3 Method Blank Summary IE Sample ID Radionuclide Matrix Count Date/Time Blank Result Units Qualifier P/F G12548-1 H-3 WO 09/08/2011 00:21 < 2.300E+00 pCi/Total U P LCS Sample Summary IE Sample ID Radionuclide Matrix Count Date/Time Spike Value LCS Result Units Spike Recovery Range Qualifier P/F G12548-2 H-3 WO 09/08/2011 01:15 2.52E+02 3.070E+02 pCi/Total 121.6 70-130 + P pike ID: 3H-041706-1 pike Conc: 5.05E+02 pike Vol: 5.OOE-01 Duplicate Summary IE Sample ID Radionuclide Matrix Count Date/Time Original Result DUP Result Units RPD Ranze Qualifier P/F G12548-3 H-3 SD 09/08/2011 02:17 < 1.150E+00 < 1.I1OE+00 pCi/g Wet <50 ** NE 17514-1 Positive Result Compound/analyte was analyzed, peak not identified and/or not detected above MDC

< 5 times the MDC are not evaluated Nuclide not detected Spiking level < 5 times activity Pass Fail Page 2 M Not evaluated

AITELEDYNE QC Summary Report for L47414 BROWN ENGINEE'RING, A te dyne Thchdotoaes Comparny EN005-3EUOFA-10 09/13/2011 08:59 H-3 U-3 Associated Samples for WG12548 Sample Client ID A7414-1 001

,47414-2 002

)A7414-3 003

,47414-4 004 47414-5 005

)47414-6 006 A7414-7 007 A7414-8 008 A7414-9 009 A7414-10 010

,47414-11 011

)47414-12 012 A7414-13 013 A47414-14 014 Positive Result Compound/analyte was analyzed, peak not identified and/or not detected above MDC

< 5 times the MDC are not evaluated Nuclide not detected Spiking level < 5 times activity Pass Page 3 Fail IF Not evaluated