ML20045A867

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Confirmatory Survey of Bldg 675,Minnesota Mining & Mfg Co, Twin Cities Army Ammunition Plant,Arden Hills,Mn
ML20045A867
Person / Time
Site: 07000832
Issue date: 06/30/1992
From: Landis M, Vitkus T, Weaver C
OAK RIDGE ASSOCIATED UNIVERSITIES
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
ORISE-92-F-57, NUDOCS 9306150141
Download: ML20045A867 (66)
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{{#Wiki_filter:r \\l ]T)d t ), 1 ^r l g. CONFIRMATORY SURVEY fd h l OF BUILDING 675 -l : MINNESOTA MINING AND -MANUFACTURING COMPANY l " TWIN CITIES ARMY AMMUNITION PLANT ARDEN HILLS, MINNESOTA I T. J. VITKUS l Prepared for the Division of Industrial and Medical Nuclear Safety U.S. Nuclear Regulatory Commission . Region ill Office I I I g +: i s ~ T1 4: ~- 'g{ 4f0DM ^ [ kVf $l OR:TSE lbAdRIDON INS *rrrUTM TOR SCIENCE AND EDUCATION- ' " f Environmental Survey and Site Assessment Program fI..

Energy / Environment Systems Division' g

,g 9306150141 920630 l PDR ADOCK 07000832 C PDR

r ORISE 92/F-57 CONFIRMATORY SURVEY OF BUILDING 675 MINNESOTA MINING AhT I MANUFACTURING COMPAhT TWIN CITIES ARMY AMMUNITION PLANT ARDEN IIILLS, MINNESOTA Prepared by T. J. Vitkus Environmental Survey and Site Assessment Program Energy / Environment Systems Division 1 Oak Ridge Institute for Science and Education Oak Ridge, Tennessee 37831-0117 I Project Staff I T. L. Bright J. L. Payne R. C. Gosslee* E. A. Powell M. J. Laudeman Prepared for Division of Industrial and Medical Nuclear Safety I U.S. Nuclear Regulatory Commission Region III Office FINAL REPORT JUNE 1992 I g This report is based on work performed under an Interagency Agreement (NRC Fin. No. A-9076) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy. r Oak Ridge Institute for Science and Education performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy.

  • Currently with Martin Marietta Energy Systems E

I CONFIRMATORY SURVEY OF BUILDLNG 675 I MINNESOTA hlINING AND MANUFACTURING COMPANY TWIN CITIES ARMY AMMUNITION PLANT ARDEN IIILLS, MINNESOTA [p!/7!92_. Prepared by: e / Date: T. J. Vitkuf roject Leader / / Environmen Survey and Site Assessment Program i Reviewed by: Date: foy C. F. Weaver, Laboratory Manager I Environmental Survey and Site Assessment Program I lo//9og Reviewed by: /.' Date: M. R. I2ildis7 Project Manager ~ I Environmental Survey and Site Assessment Program i I Reviewed by: se

    1. -f Date:

4 9/92 D. Berger,' Program @irector nergy/ Environment Systems Division 1 i I

I TABLE OF CONTENTS PAGE Li st o f Figu res............................................. ii List of Tables ..............................................iv-r Introduction and Site History.................................... 1 Facility Description.......................................... 2 .lM Procedures ...............................................3 Findings and Results .............................7 Discussion ...............................................11 I Comparison of Results With Guidelines.............................. 12 S u m m ary........................................... .... 14 Refereaces ...........................................43 I Appendices: Appendix A: Major Sampling and Analytical Equipment Appendix B: Measurement and Analytical Procedures Appendix C: Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, I Source or Special Nuclear Materials

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E Guidelines for Residual Concentrations of Thorium and Uranium Wastes-in Soil I

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II J IE LIST OF FIGURES PAGE FIGURE 1: Map Of The Minneapolis - St. Paul Area - Location Of The Twin Cities Army Ammunition Plant......................15 FIGURE 2: Map Of The Twin Cities Army Ammunition Plant - Ixcation Of 3M Building 675................. .................16 FIGURE 3: Building 675, Original Offices And Production Rooms - Floor Plan 17 I FIGURE 4: Building 675 - Areas of Specific Radionuclide Usage............18 FIGURE 5: Building 675, Production Area - Excavated Areas 19 FIGURE 6: Building 675 - Nonheast Hall and Adjoining Rooms............. 20 FIGURE 7: Building 675, Exterior - Plot Plan and Reference Grid.. 21 FIGURE 8: Building 675, Exterior - Elevatec Direct Radiation Locations.... 22 FIGURE 9: Building 675, Exterior - Measurement and Sampling Loc:.tions...... 23 FIGURE 10: Building 675, Basement - Measurement and Sampling Locations...... 24 FIGURE I1: Building 675, Main Office Area - Measurement and Sampling Locations. 25 I FIGURE 12: Building 675, Lunch Room / Conference Room - Measurement and Sampling Locati on s....................................... 2 6 FIGURE 13: Building 675, South Office - Measurement and Sampling Locations.... 27 FIGURE 14: Building 675, North Office - Measurement and Sampling IAcations... 28 FIGURE 15: Building 675, Health Physics Room - Measurement and Sampling Locations.......................................29 FIGURE 16: Building 675, Women's Restroom - Measurement and Sampling Locations 30 l FIGURE 17: Building 675, Warehouse - Measurement and Sampling locations..... 31 I FIGURE 18: Building 675 - Production Area Bays...................... 32 l ii

E E .5 LIST OF FIGURES (Continued) PAGE FIGURE 19: Building 675, Production Area, Walls, Columns and Footers - Measurement and Sampling Locations .....................33 FIGURE 20: Building 675, Production Area, Ceiling - Measurement and Sampling Locations ..............................34 FIGURE 21: Building 675, Production Area - Measurement and Sampling locations .................................35 i I g i I I l i I I I I I I I iii I

I j

LIST OF TABLES A PAGE TABLE 1: Radionuclide Concentrations In Soil And Sediments, Exterior Location s....................................... 3 6 l TABLE 2: Exposure Rates.................................... 37 TABLE 3: Summary Of Surface Activity Measurements.................. 38 TABLE 4: Radionuclide Concentrations In Soil Interior Locations............ 40 TABLE 5: Radionuclide Concentrations In Miscellaneous Samples........... 42 l I I 1 I i I I I I-I 9 I a iV

CONFIRMATORY SURVEY OF BUILDING 675 AfINNESOTA MINING AM)' MANUFACTURING COMPANY TWIN CITIES ARMY AMMUNITION PLANT ARDEN IIILLS, MINNESOTA INTRODUCTION AND SITE IIISTORY I i The Minnesota Mining and Manufacturing Company (3M) of St. Paul, Minnesota currently l leases four buildings, designated by 3M as Buildings 573,575,590, and 675, from the U. S. Army's Twin Cities Army Ammunition Plant (TCAAP) located in Arden Hills, Minnesota. These buildings have been utilized by 3M for fabrication and production activities involving a i variety of radionuclides. Previous activities that involved special nuclear, source, and by- { product materials were regulated by the Atomic Energy Commission (AEC), predecessor agency 'l to the Nuclear Regulatory Commission (NRC), under now terminated licenses SNM-789, SNM-106, SNM-958, SNM-764 and SMB-239. The current use of by-product materials is regulated under NRC License No. 22-00057-06 (Docket File No. 030-04951). Building 675, designated by the TCAAP as. Building 113, was utilized for'severa ( the production processes between 1962 and 1991. From 1962 to 1965,3M operated a nuclear fuel ( project involving the chemical bonding of carbon powder to uranium, which had been enriched in U-235, in a ceramic matrix. 3M decontaminated the facility following termination of the U-235 project. ' A release survey was then conducted by the AEC following AEC guidelines current at that time. From 1968 to April 1991, 3M Static Control Systems Division used Building 675 for operations that produced / fabricated radioactive _ sources for use in medical and various other industries. 3M's documentation indicates that the three major radionuclides'used in the manufacturing processes werei polonium-210 for the production of nuclear static eliminators, iodine-125 for the production of radioactive seeds to irradiate cancerous tumors, 'and promethium-147 in microspheres to be incorporated'into radioluminescent products..Other radionuclides used by 3M in Building 675 included ytterbium-169, strontium-90, and yttrium-90.- y n____.___-___=___--____

I 3M has discontinued all production processes in Building 675 and plans to return control of the building back to the government. 3M began decontamination activities, that were necessary to l terminate the building's NRC licensing restrictions, in October 1990 and completed the project in June 1991. At the request of the U.S. Nuclear Regulatory Commission (NRC), Region III Office, the Environmental Survey and Site Assessment Program (ESSAP) of Oak Ridge Associated Universities / Oak Ridge Institute for Science and Education (ORAU/ORISE) performed an independent confirmatory radiological survey of Building 675. I FACILITY DESCRIPTION I The TCAAP occupies an area of approximately 810 hectares (2000 acres) located in the northern one-third of the city of Arden Hills, near the intersection of Interstate 35 W and U. S. Highway 10 (Figure 1). With the exception of those buildings leased to outside ccmpanies, most of the plant facilities are inacdve and are in caretaker status. Building 675 is situated in the western ponion of the property (Figure 2). Building 675 was constructed in 1942, with the exception i of a warehouse area, which was added on at a later date. It is a single story structure with a partial basement located in the northwest end of the building. The construction is predominately masonry with columns supporting a concrete slab roof over-layed with tar and gravel buildup. Interior surfaces include cacrete slab Nors with masonry, frame, and/or gypsum board walls. I The main floor occup'es app chely 1740 m (18,700 ft ) and formerly housed the production 2 2 g areas, offices, laboratodes, warehouse, and employee break areas (Figure 3). The basement, used by 3M for miscellaneous ;,'orage, measures approximately 80 m (860 ft ). 3M designated 2 2 five radiological use areas w thin the building to guide remedial activities and release guideline application (Figure 4). T'ae remediation significantly altered the interior of the building. l Equipment and interior wMis have been removed from the production area, as have all overhead structures such as false coilings, ducts, vents, and electrical conduit. Additionally, the concrete floor has been removed, hereby exposing sub-floor soils. Extensive excavation and trenching ( operations have also been performed to remove contaminated soil, sewer lines, and drains (Figures 5 and 6). Tho basement, warehouse, and offices remained essentially intact with the t exception of te re.noval of some false ceilings, wallboard, and other miscellaneous items.

!I Exterior remedial actions included excavation and restoration of contaminated soils adjacent to the southeast corner of the building and also the main sewer line at the northeast corner. PROCEDURES I OBJECTIVE The objectives of the confirmatory survey were to provide independent document reviews and radiological data, for use by the NRC in evaluating the adequacy and accuracy of the licensee's decommissioning report and the facilities radiological status relative to established guidelines. DOCUMENT REVIEW l 3M's draft decommissioning report was reviewed for thoroughness, accuracy, and compliance with the NRC guidelines for residual surface contamination and radionuclide concentrations in soil.

I SURVEY PROCEDURES - EXTERIOR Reference Grid A 10 m x 10 m reference grid was established on the exterior grounds surrounding the building extending out 20 m from each side (Figure 7). The roof of the building was not gridded.

Measurement locations on the roof were referenced to prominent building features. Surface Scans Walkover gamma radiation surface scans were performed over gridded exterior surfaces and the roof using portable Nal(TI) scintillation detectors. Areas of elevated contact gamma radiation were marked for further investigation (Figure 8). In addition, exterior building walls and ledges, i 3 I

as well as the two manholes located north and west of the building were scanned for gamma, alpha and beta activity using NaI(TI) scintillation, ZnS scintillation, and/or GM detectors. All detectors were coupled to ratemeters or ratemeters-scalers with audible indicators. I Exoosure Rate Measurements I Exposure rate measurements were made at five locations around the building at 1 m (3.3 ft) above surfaces using a pressurized ionization chamber (PIC) (Figure 9). I Soil Samoling I Soil samples were collected from 4 representative locations of elevated direct radiation identified by surface scans (Figure 9). I Miscellaneous Samnie Collection I ~ A sediment sample was collected from the north manhole (Figure 9). There was no sediment contained within the west manhole. SURVEY PROCEDURES: INTERIOR Reference Grid A 2 m x 2 m alphanumeric grid was established on the floors and lower walls (up to 2 m)in the basement, offices, and warehouse area (Figures 10 - 17). Individual grid blocks were subdivided 2 into 1 m areas at locations where residual contamination was suspected. Concrete support l columns, spaced at 5.5 m (18 ft) intervals, were used for reference in the production area. An alphanumeric designation was given to each column, and the floor area between columns was I designated alphabetically by bay (Figure 18). Rooms smaller than 10 m were not gridded. The 2 excavation and resultant soil piles prevented establishment of a grid in the east hall and adjoining l I ^ I

E Measurements on ungridded surfaces were referenced to either the floor grid or to rooms. prominent building features. Surface Scans lI Floors, walls (up to 2 m), sub-floor soils, sewer line excavations, columns, exposed footers, ,I ceilings, and vent penetrations were scanned. The specific instrumentation and scanning techniques utilized were dependent upon the history of radionuclide usage within each area. i Surfaces were scanned for gamma, alpha, and beta activity using NaI(TI) scintillation, ZnS scintillation, gas proportional, and/or GM detectors. All detectors were coupled to ratemeters or ratemeters-scalers with audible indicators. Areas of elevated direct radiation detected by scans were marked for furtherinvestigation and identified to the 3M site representative. ESSAP performed follow-up scans at locations where 3M elected to perform additional remediation. I Surface Activity Measurements I Direct measurements to determine total activity were performed on 27 randomly selected grid blocks, located on the floor and lower walls. Measurements were made at the center and four points equidistant between the center and grid block corners. Thirty-eight single-point measurements were performed on accessible ceiling locations within the production area; in rooms smaller than 10 m ; at locations of direct radiation identified by surface scans; and 2 following additional remediation at selected locations (Figures 10 - 20). Smear samples to determine removable activity, were collected from the highest direct measurement location within each grid block and from each single-point measurement location. Direct measurements were made using gas proportional, ZnS scintillation, or GM detectors coupled to ratemeters-scalers. 1 5 L

I Exnosure Rate Measurement Exposure rates were performed at 9 locations within the building. Measurements were made at 1 m (3.3 ft) above floor surfaces using a PIC (Figures 10 - 21). The main office area did not have a history of radiological use, and the background exposure rate for the building was therefore determined in this portion of the building. I Soil Sampline I Within the production area, 9 sub-floor soil samples were collected from regions of elevated direct radiation, identified by gamma scans (Figure 21). Each area of elevated direct radiation detected was small, i.e. less than 1 m, and did not indicate wide-spread contamination. 2 Therefore, an additional 21 systematic sampler were collected from contiguous 25 m to 100 m 2 2 areas to confirm that the " hot spots" were localized (Figure 21) and one post-remedial action sample was collected. Each of the four radionuclide use areas, as indicated on Figure 4, was included in the sampling scheme. Miscellaneous Samole Collection Two paint samples and one concrete sample were collected from locations of elevated direct radiation within the Production Area, in order to qualitatively confirm the contaminants, j (Figure 19). l SAMPLE ANALYSIS AND DATA INTERPRETATION 'I Samples and field data were returned to ESSAP's Oak Ridge, TN facility for analysis and interpretation. Direct measurements were converted to units of dpm/100 cm. Smears were 2 analyzed using a low-background proportional counter to determine gross alpha and gross beta activity. Soil, sediment, and concrete samples were analyzed individually by gamma I 6

spectrometry for U-235, Cs-137, and I-125. Spectra were also reviewed for any other identifiable photopeaks. The interior soil samples from Areas 1,2, 4, and 5, were composited, ~ according to area, for additional analyses. These composite soil samples were analyzed by wet chemistry methods for Sr-90, Pm-147, or Po-210. Paint and concrete samples were analyzed for Sr-90. Appendices A and B describes the analytical instrumentation and procedures utilized. I FINDINGS AND RESULTS DOCUMENT REVIEW I ESSAP's review of 3M's draft decommissioning report identified report sections where additime information was necessary to clarify the radiological condition of the building. The requested information was provided to the NRC Region III office in a July 25, 1991 correspordence'. a I 3M provided the additional data in a January 27,1992 supplemental data correspondence from 2 3M to the NRC Region III office, EXTERIOR SURVEY l Surhee Scans Gamma surface scans of the exterior grounds adjacent to Building 675 identified 8 localized areas of elevated contact radiation, each measuring < 1 m (Figure 8). Alpha, beta and/or 2 gamma scans of the buildings exterior surfaces and manholes did not identify any areas of direct radiation above ambient background levels. I I

l Exoosure Rates l Background exposure rates in Arden Hills, MN, as determined during a 1982 survey performed by ESSAP (previously known as the Radiological Site Assessment Program) ranged from 7 to 12 uR/h.' Exterior exposure rates, which are summarized in Table 2, ranged from 8 to 9 uR/h. I Radionuclide Concentrations In Soils and Miscellaneous Samoles B I Radionuclide concentrations in exterior soil and sediment samples are summarized in Table 1. I Soil samples contained Cs-137 activity levels up to 860 pCi/g, before additional remediation was performed. The Cs-137 concentration range in post-remedial samples is 0.5 to 0.7 pCi/g. The concentration ranges of U-235 and total uranium were 0.1 to 15.0 pCi/g and 2.4 tt 350 pCi/g I respectively. Total uranium concentrations were determined based on the isotopic uranium analytical results provided by 3M in their January 27, 1992 correspondence. The three composite samples analyzed had an average U-235 to U-234 activity ratio of 1:23. The highest U-235 concentration was at 60N,48.5W. Neither I-125 nor any other gamma emitting radionuclides, other than those occurring in nature were identified in soil samples. For the j sediment sample collected from the north manhole, radionuclide concentrations were as follows: U-235,0.8 pCi/g; Total uranium,18 pCi/g (assuming enriched uranium) and Cs-137, <0.7 l pCi/g. Although there has been no previous documented history of Ra-226 use within Building l 675,25 pCi/g of that nuclide was identified in the sediment. l INTERIOR SURVEY l Surface Scans l Surface scans of the office area, basement and warehouse identified elevated direct radiation at the following locations: a pipe located in the former executive washroom (elevated gamma) (Figure 6), two locations on the basement floor at grid points H, 9.2 (elevated beta-gamma) 8

I and C + 0.3,11 (elevated alpha-beta) (Figure 10), and one location on the warehouse floor at grid point E + 1.4,9.5 (elevated alpha-beta) (Figure 17). Within the production area, scans of sub-floor soils identified elevated gamma radiation in Bays V, X, Y, 00, QQ, SS, and UU (Figure 21). Further investigation of these locations identified a buried section of contaminated drain pipe in Bay QQ and a contaminated column footer (north I side of column B,0)in Bay UU. Surface scans of the footer identified elevated levels of alpha activity. The remaining locations identified were the result of contaminated soils. Surface scans of the production area lower walls, wall sills, columns, exposed column footers I and the ceiling identified several locations with residual activity (Figure 19). Two locations of elevated beta activity were identified within Area 4 on columns E,2 and F,5. Elevated alpha activity was identified on the west wall sill at column A,2; in a pipe recess located on the north wall at column B,8; and on a seam on both the north and south sides of column D,5, where walls had been removed (Figure 20). The area of elevated direct radiation extended on the ceiling 20 to 25 meters south of column D,5. Surface Activity l Direct measurements for total and removable surface activity are summarized in Table.3. All of the measurement data presented represents surface activity following final decontamination. l The highest grid block averages for each area of Building 675 are as follows: offices (Area 3), 2 2 2 50 dpm/100 cm alpha and 340 dpm/100 cm beta; basement,290 dpm/100 cm alpha and 2 2 2 550 dpm/100 cm beta; warehouse, 55 dpm/100 cm alpha and <341 dpm/100 cm beta; 2 2 production area (Arcas 1,2,4,and 5),390 dpm/100 cm alpha and 590 dpm/100 cm beta. The range of total activity measurements for each area of the building are as follows: offices, <37 2 2 to 90 dpm/100 cm alpha and <340 to 450 dpm/100 cm beta; basement, < 37 to 1800 dpm/100 2 2 2 cm alpha and <341 to 1500 dpm/100 cm beta; warehouse, <37 to 55 dpm/100 cm alpha and 2 2 < 326 to 760 dpm/100 cm beta; production area, <37 to 1900 dpm/100 cm alpha and <330 to 2300 beta. I I

I Removable activity levels for the building ranged from <7 to 35 dpm/100 cm for gross alpha 2 2 and < 12 dpm/100 cm for gross beta. Eynosure Rates Interior exposure rate measurements are summarized in Table 2. The building background exposure rate was 12 pR/h. The measurements ranged from 11 to 13 uR/h. I Radionuclide Concentrations in Soils and Miscellaneous Samnles i Radionuclide concentrations in soil samples collected from the production area are summarized in Tables 4 and 5. The concentration ranges of radionuclides in the samples collected from locations of elevated direct radiation, prior to additional excavation, were: Cs-137, < 0.1 to 0.1 pCi/g, U-235,0.1 to 35.4 pCi/g; and total uranium, 2.3 to 800 pCi/g. The post-remedial action sample collected from Bay UU contained 0.1 pCi/g of Cs-137,0.3 pCi/g of U-235, and 6.8 pCi/g of total uranium. The concentration ranges of radionuclides in systematic samples are as follows: Cs-137, <0.1 to 0.1 pCi/g; U-235,0.1 to 0.8 pCi/g; total uranium, 2.3 to 18 pCi/g. lodine-125 was not identified in any of the samples. Radionuclide specific analytical results for composite samples are as follows: Area 1, < 17 pCi/g of Pm-147; Area 2, <0.8 pCi/g Po-210; Area 4, < 0.2 pCi/g Sr-90. Strontium-90 concentrations in paint were determined for qualitati!e purposes only. The concentrations of Sr-90 were 13.1 pCi/g and 0.3 pCi/g for l the paint samples collected from columns F,5 and E,2 respectively. i The concrete sample, also collected for qualitative purposes during the remediation of the wall and ceiling seam, had concentrations of U-235, total uranium, and Sr-90 of 8 pCi/g,180 pCi/g, and < 0.3 pCi/g, respectively. I I I I

I DISCUSSION ESSAP's confirmatory survey identified additional findings that were inconsistent with the reported final radiological status of the building. These findings are discussed below. Investigations of the exterior locations of elevated direct radiation determined that the activity detected at 7 of the 8 locations exhibited characteristics of a discrete point-source. The analysis of soil samples collected by ESSAP confirmed the presence of elevated concentrations of Cs-137. The exception was grid coordinate 60N,48.5 W, which contained elevated concentrations of U-235. The 3M site representative believes that the Cs-137 point-sources were microspheres; which had been manufactured by 3M at a separate building (Building 575) also located at TCAAP, and were inadvertently deposited from personnel traffic between Buildings 575 and 675. These locations were subsequently remediated either during the sampling process or excavated by 3M. Post-remedial action scans and sample analyses determined that residual gamma activity and the radionuclides present were typical of natural background. I The U-235 contamination located at grid coordinate 60N,48.5W was distributed over an area 2 of 1 m and exhibited the highest gamma activity noted in a thin band that extended north the length of the building. 3M excavated the location and performed a follow-up investigation of this area. The investigation included composite sampling and isotopic uranium analysis after the completion of ESSAP's confirmatory survey. The analytical results, which were provided in the January 27,1992 correspondence, indicated that the total uranium concentration in this area 5 l is 6.4 pCi/g. As noted in previous sections of this report, several locations of elevated direct radiation were detected during ESSAP's survey of the interior of the building. 3M performed additional decontamination at these locations. Remedial actions included excavation and removal of contaminated pipes; excavation of soil; scabbling and jackhammering concrete and masonry from floors, walls, ceilings, and footers; and paint removal. Post-remedial action surface scans, .I I n g

g sampling, and/or measurements were performed by ESSAP to confirm that residual contamination had been reduced to acceptable levels.

COMPARISON OF RESULTS WITH GUIDELINES The NRC guidelines for surface contamination and residual concentrations of radionuclides in soils, established for license termination or release of a facility for unrestricted use, are presented in Appendix C. I Radionuclides from three guidelines categories had been used in Building 675. Pm-147, and I uranium are included in the category with the least restrictive guideline levels (Table 1 of Appendix C). Two of the radionuclides, I-125 and Sr-90, are found in Table 1 categories with more restrictive guideline levels. Surface contamination guidelines for Po-210, when present from sources other than the decay of uranium-238, are not specified in Appendix. C. Historically, guidelines more conservative than those for uranium decay products have been applied for Po-210 contamination. 3M has proposed that surface contamination guidelines for uranium and Pm-147 apply to Areas'1, 3, the basement, and warehouse, Po-210 guidelines apply in Area 2 and that Sr-90 guidelines apply to Areas 4 and 5.1-125 which was used in Production Area 5, is in the most restrictive guideline category. 3M indicated that measurements performed for I-125 activity were less that the instrumentation's minimum detectable activity (MDA) of 2 2 2 1300 dpm/100 cm which exceeds the guideline level of 100 dpm/100 cm average over 1 m 2 2 and 300 dpm/100 cm maximum in 100 cm. However, the negative results for I-125 in samples and the short half-life of I-125 (60 days) suggests that sufficient decay has occurred since May l 1991 to effectively reduce contamination to less than guideline levels. l The proposed surface contamination guidelines for Building 675, with the exception of Areas 4 and 5 (Figure 4), are those for Pm-147, Po-210, and uranium. The guidelines for Po-210 and l uranium are:

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i .I Total Activity 5,000 a dpm/100 cm, averaged over 1 m 2 2 2 2 15,000 a dpm/100 cm, maximum in 100 cm j W Removable Activity l 2 1,000 a dpm/100 cm l The guideline vels for Pm-147 are: I Total Activity 2 2 5,000 S-y dpm/100 cm, averaged over 1 m 2 2 15,000 S-y dpm/100 cm, maximum in 100 cm Removable Activity 1,000 S-y dpm/100 cm2 h us'cl<lh l< V<ks E Mo <ce c 2 2 1000 dpm/100 cm, averaged over 1 m 2 2 3000 dpm/100 cm, maximum in 100 cm lI l Removable Activity 2 200 dpm/100 cm Guidelines for residual concentrations of thorium and uranium wastes in soils are presented in l Appendix C. The applicable guideline is the Option 1 surface concentration guideline for enriched uranium which is: Enriched Uranium Maximum Concentration (pCi/g) Soluble 30 Insoluble 30 i I 13

1 I Allinterior post-remedial action samples are within this guideline. Soil concentration guidelines for the remaining radionuclides have not been established for the site. Cs-137 was the only NRC licensable radionuclide, other than enriched uranium, which was detected in excess of the minimal detectable concentration of the analytical procedure. Exterior locations where Cs-137 was identified were remediated by 3M. Post-remedial action sample concentrations are typical of background levels. One ESSAP sample contained 350 pCi/g of enriched uranium. 3M's I remediation of this location and subsequent investigation of the contiguous areas determined that the average concentration of total uranium had been reduced to 6.4 pCi/g. I Exposure rates for the interior and surrounding exterior of building 675 were all within the range of background radiation and, therefore, below the guideline level of 5 uR/h above background, currently being used by the NRC for evaluating decommissionings. SU3 DIARY From August 5 through 9,1991, the Environmental Survey and Site Assessment Program of the Oak Ridge Institute for Science and Education performed a confirmatory radiological survey of 3M Building 675, located on the Twin Cities Army Ammunition Plant in Arden Hills, MN. The survey included document reviews, surface scans, surface activity measurements, exposure rate measurements, and soil and miscellaneous material sampling. j 'l ESSAP reviewed 3M's draft decommissioning report and supplemental data correspondences as part of the confirmatory activities. In ESSAP's opinion, 3M's documentation provides an adequate description of the radiological condition of Building 675. ESS AP identified several areas of residual contamination were identified on interior surfaces and in interior and exterior soils. Each of these areas were effectively remediated by the licensee. ESSAP's fm' al independent measurement and sample data compiled for Building 675 are within the surface contamination and soil concentration guideline levels for release to unrestricted use. I I

l ' = MMM1 lI il yse ll I ig s 1ll-rn k TWIN CITIES ARMY t 3 AMMUNITION PLANT (e} s to (k \\ ARDEN Nius NEW si BRIGHTON 4 l ROSEVILLE a / b g l l[ MAPLEWOOD lh I l s ST. PAUL p g ,ac MINNCAPOUS / h MILES I N KILOMETERS I FIGURE 1: Map of the Minneapolis-St. Paul Area - Location of the Twin Cities Army Ammunition Plant 15 gW )

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MMM5 I n! / \\ !O N@Z 110N I /[s$ l' ,/ B EW, + SNbN 100N 90N I 7 80N MAEottO 70N ~ BUILDING ~ 675 ASNA_T 60N [ SON -~ l 1 40N .y _+- 3cN ^ 20N T 10N

c 70w 60W 50W 40W 30w 20W 10W OW I

8 t I k RAILROAD X X X FENCE I o 6.0 ME!ERS l I FIGURE 7: Building 675, Exterior - Plot Plan and Reference Grid

MMM5a I V NOT SURVEYED NORTN "0'E O"4-O-+ 110N - r-9 1 } +) sus-100N STADON I m. y 90N 80N T WEST I WANHOLEO O 70N BUILDING ( 675 i ASPHA[.T

E 4

60N O p - ] 50N } 40N I I = a } 30N 3 v 20N TANK FARM i ~ g r I I AREA 5 NOT i l I svRvntD:c f 10N-1

c I

70W 60W 50W 40W TOW 20W 10W OW RAILROAD X A FENCE AREAS OF ELEVATED C DIRECT RADIATION IDENTIFIED 'I BY SURFACE SCANS FEET O 6,0 o io METERS I FIGURE 8: Building 675, Exterior - Elevated Direct Radiation Locations ~~ E

MMM5b '20N . [ om! / i NOT SURVDTD/ NORTH / MANHOLE 110N f~# 1 E / sus-100N sT e oN I p cj i WL e -.1 gan 7 4 .~ 80N wAEotro 8 $ 70N } BUILDING [ 675 ASPHALT a 60N g? ~ 3 '1 50N ~ I l i 40N 30N V i d 20N TANK FARM i f i AREAS NOT ! I SVRVEYED :d 4__.I [ f 10N 1 I 70W 60W S0W 40W 30W 20W 10W OW I s MEASUREMENT / SAMPLING LOCATIONS i RAILROAD 4 EXPOSURE RATE Y rENCE 8 SOIL l' $ stoiMcNT 0 60 rect METERS l FIGURE 9: Building 675, Exterior - Measurernent and Sampling Locations 23 { J

l lE MMMIS I I J I r/' n i j -=- 12 'l' 'l ' ' ' ' ' ~ 0). / cHwNEY 10 /..f,/ wa .^,'/ O 8 ' 't ~~ ~ ....0 d-e ihg o; SUPPORT s o ,<,,y u 0 """ $*"$ ppd 2 r EE I i TRANSmON A C E G H +.75 I l s l MEASUREMENT / SAMPLING k LOCATIONS \\ l n g SINGLE-POINT 06 N MD ROOR INACCESSIBLE AREA I GRID BLOCK O 12 4 EXPOSURE RATE 6 g uE1tRS g nouse,0: emiiere, e7s. seeememi - ueceerememi eee somnites 'occito"= 24

MMM9 I-I I . 7 -- A 5.6 4 +- ~ ~ d 2' 2.- I - TO EAST 3 l l HALLWAY y f' A f C E C K M Ok+.2 p ] I I I N hb MEASUREMENT / SAMPLING LOCATIONS I ^ GRID BLOCK g . Exeosuac axre o mes 4 l FIGURE 11: Building 675, Main Office Area - Measurement and Sompling Locations 25 l

MMM7 i i 1 A I E / 4 / 4 = A /. ./'.., /., ~ /.,. : _/.. ,., /. :,. ', (.,. ', ( ' ^ ' O-A C E G 1 i+ 1.6 g 1 II I I i lI s k M EASUREMENT/SAMPUNG LOCATIONS 3 L ] CRID BLOCK SO t EXPOSURE RATE O 12 6 A l METERS FIGURE 12: Building 675. Lunch Room / Conference Room - Measurement and Sampling Locctions 26 lI

I 1 l \\ ,E lI l t E 'g a I lI I i E B I l st i 'g MEASUREMENT / SAMPLING LOCATIONS l g SINGLE-POINT LOWER WALLS AND FLOOR l b CEILING VENT FEET O 6 m 2 uETcas i l 27 )

mum 10 ) 5 I i i I I l/ I n A C E E + 1.4 I i I I I i \\, MEASUREMENT / SAMPLING LOCATIONS l CRID BLOCK "U 0 6 o i METERS FIGURE 14: Building 675, North Office - Measurement and Sampling Locations c 28 g

a I I t I N I N .V_ _4 E ', :~w' g-_c J ~ / A 'l /A ~~ I ( / V/ y! A E z C E E <, I i i I 1 I I I t MEASUREMENT / SAMPLING LOCATIONS GRID BLOCK 7 INACCESSIBLE AREA $ EXPOSURE RATE O 6 .i METERS lI " " " " " " " " " " ' " " " " " " " " ' * ' " " " - " " " " ~ ' " ' " " " ~ " " " " ' " " " " 29

l MMM11' i I i I i I i I O O ! 4 -t r -j A i y + O' - A C E E+1 5 I I I s I kg i g MEASUREMENT / SAMPLING LOCATIONS l GRID BLOCK 0 6 m METERS t FIGURE 16: Building 675, Women's Restroom - Measurement and Sampling Locations l 30 1

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uuu4 l l l l l-I A B C D E F e -g 1p a-c = -u /? a J z 7l 3 3 0.. .-O O 2,. ~as: i E ^ 51.- 9,E ~~---- 61J O Og. t 1 su N 5lNy- 'O - Eh l? EE l 00f I, T

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  • 1I i

s I k h,n MEASUREMENT / SAMPLING v LOCATIONS O COLUMN 4 EXPOSURE RATE h Soll PILES E SOIL TRENCH I N 8 SOIL-ELEVATED DIRECT RADIATION o 30 F:=8"""", j uETERs ' l g g n o u a c 2,: e uiiern o s75. Production Area - Measurement and Sornpling Locations 35 l I l

I TABLEI RADIONUCLIDE CONCENTRATIONS IN SOIL AND SEDISIENTS I EXTERIOR LOCATIONS 3M BUILDING 675 ARDEN IIILLS, MINNESOTA Radionuclide Concentration (pCi/g) I Cs-137 U-235 Total U 6 52N, 50W 860 6 1.3 1.2 31' 52N, 50W (Post RA) 0.5 0.3 0.3 0.1 7.2 57N, 63W 8.4 3.7 0.8 0.8 19 57N, 63W (Post RA) 0.5 0.1 0.1 0.1 2.4 d 60N, 48.5W 0.7 0.1 11.2 1 0.7 270 72N, 60.5W 730 i 3 1.0 I 0.7 24 72N, 60.5W (Post RA) 0.7 0.1 0.1 0.1 2.4 North Sanitary Sewer <0.7 0.8 0.2 19

  • Refer to Figure 9.

6 Uncertainties represent the 95% confidence level based on counting statistics only; additional I laboratory uncertainties of 6 to 10% have not been propagated into these data.

  • Calculated based on an U-235 to U-234 activity ratio of 1:23.

d Location remediated by licensee following completion of the confirmatory suney. I I I I I I I 36

I TABLE 2 I EXPOSURE RATES 3M BUILDING 675 ARDEN IIILLS, MINNESOTA I Exposure Rate Location' R/h Exterior 20N,10W 9 70N, 0W 9 110N, 30W 8 80N, 60W 9 10N, 55W 8 Interior Production Area, Section Q 12 Production Area, Section X 12 Production Area, Section JJ 11 Production Area, Section LL 12 Production Area, Warehouse 11 Health Physics Room 12 Main Office 12 Lunch Room 12 Basement 13

  • Refer to Figures 9-21.

I I I I 37 j

M M M M M M M M M M M M M M M M M M M TABLE 3

SUMMARY

OF SURFACE ACTIVITY MEASUREMENTS 3M BUILDING 675 ARDEN IIILLS, MINNESOTA

  1. of Measurements liighest Grid Illock Total Activity Renmvable Activity Area R<mm locatkm*

Average dpm/100 cm dpm/100 cm dpm/I00 cm Guideline 2 2 2 Ave / Mas / Grid likicks Single-l'oints a a Range

  1. Range a Range
  2. Range Renmvahic liasement Ihr 3

2 290 550 < 37-1800 < 340-1500 < 7-35 < l2 5/15/1 lower Walls 2 60 < 340 <37-100 <340 <7 <l2 5/I5/1 Main Office Fhwr 2 50 <340 <37-90 <340- 400 <7 <l2 5/15/1 lower Walls 2 <37 <340 <37 < 340 <7 < t2 5/15/1 North ihr 1 <37 < 330 <37 <330- 400 <7 <l2 5/15/1 y OITire lower Walls I <37 <330 <37 <330 <7 < t2 5/15/1 Samth Office Ihr 1 N/A N/A <37 < 350 <7 < l2 5/15/1 lower Walls 4 N/A N/A <37 < 330 <7 < 12 5/15/1 Ceiling I N/A N/A <37 < 330 <7 <!2 5/15/l Conference Floor 2 <37 <330 <37-60 <330- 410 <7 < l2 5/I5/1 Rimm lower Walls I <37 <330 <37 < 330 <7 <12 5/15/1 Ilealth Fkmr 1 <37 < 330 <37 <330- 450 <7 < l2 5/15/1 l'hysics Room lower Watts I < 37 <330 < 37 < 330 <7 < l2 5/15/l 1 _.__._m._____,______-_2_m,- ~

M M M M M M M M M M M M M M M M M M TAllLE 3 (Contintied)

SUMMARY

OF SURFACE ACTIVITY MEASUREMENTS 3M llUILDING 675 ARDEN IIILLS, MINNESOTA

  1. of Measurements Ifighest Grid Bhwk Total Activity Removable Activity Area 2

2 8 Room laation* Average dpm/100 cm dpm/100 cm dpm/IDO cm Guideline Ave / Max / Grid likwks Single-l'oints a 4 a Range 4 Range a Range $ Range Removable Rest Room ihr 1 45 340 <37 65 <330- 460 <7 <l2 5/15/l lower Walls I <37 <330 <37 <330 <7 < t2 5/15/1 Warehouse Ihr 7 <37 <330 < 37 <330 <7 < t2 5/15/1 lower Walls 2 <37 <340 <37 <340 <7 <l2 5/15/1 I'roduction Columns 1I N/A N/A <37 430 <330- 360 <7 <l2 5/I5/1 gc Area 1 Ceiling 3 N/A N/A INo. 645 <330- 590 < 7-9 < 12 5/I5/1 l'roduction Column I N/A N/A 230 <7 <l2 5/15/1 Area 2 Footer Ceiling 3 N/A N/A 70-1200 <330-2300 <7-11 < t2 5/15/1 l'roducekm Column 3 N/A N/A < 83 <910 <7 <l2 5/15/l Area 4 Ceiling 2 N/A N/A 70-145 < 330 <7 < 12 5/15/l l'rmiuction Column 3 N/A N/A <37-1900 <330-950 <7 < t2 5/15/1 Area 5 Ceiling 2 N/A N/A 220- 350 < 330 <7 <l2 5/15/1 tewer Walls 2 390 590 < 37-1600 <340-1500 <7 < 12 5/15/1 - Refer to Figures 10 20. - No measurement made. N/A Not Applicable. l

I TABLE 4 RADIONUCLIDE CONCENTRATIONS IN SOIL I INTERIOR LOCATIONS 33I BUILDING 675 ARDEN IIILLS, SIINNESOTA Location

  • Radionuclide Concentrations (pCi/g)

Sample Area Bay Cs-137 U-235 Total U6 1 1 K < 0.1 0.1 0.1 2.4 2 1 K <0.1 0.1 0.1 2.4 3 1 P < 0.1 0.1 0.1 2.4 4 1 U < 0.1 0.1 0.1 2.4 5 1 V < 0.1 0.4 0.1 9.6 g 6d 2 00 0.1 0.1 5.2 0.1 130 7 2 00 < 0.1 0.1 0.1 2.4 8 2 00 < 0.1 0.3 0.1 7.2 9 2 00 < 0.1 0.1 0.1 2.4 10d 2 00 < 0.1 28.5 0.2 680 11 2 UU < 0.1 0.2 0.1 4.8 12 2 UU < 0.1 0.1 0.1 2.4 13 2 UU < 0.1 0.4 i 0.2 9.6 I 14d 2 UU 0.1 0.1 30.3 1.1 730 15' 2 UU 0.1 0.1 1.9 0.1 46 16 2 UU (Post-0.1 i 0.1 0.3 0.1 7.2 RA) I 17 2 QQ < 0.1 0.1 i 0.1 2.4 1 88 2 SS 0.1 i 0.1 7.6 i 0.3 180 198 2 WW < 0.1 2.0 0.1 48 '. l 20 2 WW < 0.1 0.2 0.1 4.8 J l 40

I TABLE 4 (continued) I RADIONUCLIDE CONCENTRATIONS IN SOIL INTERIOR LOCATIONS 3M BUILDING 675 ARDEN IIILLS, MINNESOTA Location

  • Radionuclide Concentrations (pCi/g)

Sample # Area Bay Cs-137 U-235 Total U6 21 2 XX < 0.1 0.8 i 0.1 19.2 22 4 HH < 0.1 0.1 0.1 2.4 l 23 4 II < 0.1 0.1 i 0.1 2.4 24 4 II < 0.1 0.1 i 0.1 2.4 25 4 MM < 0.1 0.1 0.1 2.4 26 4 NN < 0.1 0.1 0.1 2.4 27 5 S < 0.1 0.1 i 0.1 2.4 28 5 T < 0.1 0.4 0.3 9.6 29 5 T/S <0.1 0.1 i 0.1 2.4 30 5 X < 0.1 0.2 0.2 4.8 318 5 Y < 0.1 35.4 0.5 850 ' Refer to Figure 21. 6 Calculated based on an U-235 to U-234 activity ratio of 1:23. I ' Uncertainties represent the 95% confidence level, based only on counting statistics; additional laboratory uncertainties of 6 to 10% have not been propagated into these data. dLocation remediated by licensee. I I I . I I 41

I I TABLE 5 g RADIONUCLIDE CONCENTRATION IN MISCELLANEOUS SAMPLES E 3M BUILDING 675 ARDEN IIILIS, MINNESOTA l Radionuclide Concentration pCi/g Location = Sample Type U-235 Sr-90 Column D5-D6, Ceiling Composite Concrete 8.1 0.7 <0.3 6 Column F5 Paint 13.1 1 1.2 Column E2 Paint 0.3 0.2 ' Refer to Figures 18 and 19. 6 Uncertainties represent the 95% confidence level based on counting statistics only; additional laboratory uncertainties of 6 to 10% have not been propagated into these data. -Analysis not performed. I I I I I I I 42

I REFERENCES I Minnesota Mining and Manufacturing Company, Draft Chronological Report of 1. Decontamination Efforts - Bldg. I13 (3hi Blde. 675L June 1991. I T. J. Vitkus, Letter to D. Sreniawski, July 25, 1991 Nuclear Regulatory Commission 2. Region III, Glen Ellyn, IL. i I 3. M. R. Peters, Ixtter to D. G. Wiedeman, January 27, 1992 Nuclear Regulatory Commission Region III, Glen Ellyn, IL. j 4. J. D.' Berger, Environmental Survey of the Static Control Systems Department l Afinnesota. Minine and Manufacturine Co. New Brichton. MN, March 1982. j l 5. D. G. Wiedeman, Memorandum to T. Vitkus, June 15, 1992 Oak Ridge Institute for Science and Education, Oak Ridge, TN. I I I I I I I I .E I 43 g

!g

3-t

) t iI ~ i j s 1 -1 ^ 4 1 APPENDIX A 1 i MAJOR SAMPLING AND ANALYTICAL EQUIPMENT 1 1 'O 1 4 I l I i d b l l i I i 1 g I

I APPENDIX A MAJOR SAMPLING AND ANALYTICAL EQUIPMENT I The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the authors or their employers. I DIRECT RADIATION MEASUREMENT Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM) Eberline " RASCAL" Ratemeter-Scaler Model PRS-1 (Eberline, Santa Fe, NM) I Ludlum Ratemeter-Scaler Model 2220 (Ludlum Measurements Inc., Sweetwater, TX) Ludlum Ratemeter-Scaler Model 2221 (Ludlum Measurements Inc., Sweetwater, TX) Detectors Bicron NaI(TI) Scintillation Detector Model G5 "Fidler" (Bicron, Corporation, Newbury, Ohio) Victoreen NaI(TI), Scintillation Detector I Model 489-55 3.2 cm x 3.8 cm Crystal (Victoreen, Cleveland, OH) Eberline GM Detector Model HP-260 2 Effective Area,15.5 cm q (Eberline, Santa Fe, NM) A-1

ll Ludlum Gas Proportional Detector Model 43-37 2 Effective Area,550 cm i (Ludlum Measurements Inc., Sweetwater, TX) Used on conjunction with: Ludlum Floor Monitor Model 239-1 (Ludlum Measurements, Inc., Sweetwater, TX) 1 Ludlum Gas Proportional Detector Model 43-68 2 Effective Area,100 cm (Ludlum Measurements Inc., Sweetwater, TX) Eberline ZnS Scintillation Detector Model AC-3-7 2 Effective Area,59 cm (Eberline, Santa Fe, NM) Reuter-Stokes Pressurized Ion Chamber Model RSS-111 (Reuter-Stokes, Cleveland, OH) LABORATORY ANALYSES EQUIPMENT I.ow Background Gas Proportional Counter Model LB-5110 (Tennelec, Oak Ridge, TN) Tri-Carb Liquid Scintillation Analyzer I Model 1900CA (Packard Instrument Co., Meriden, CT) High-Purity Germanium Detector . Model GMX-23195-S,23% Eff. (EG&G ORTEC, Oak Ridge, TN) Used in Conjunction with: Lead Shield Model G-16 (Gamma Products, Palos Hills, IL)

I High-Purity Germanium Coaxial Well Detector Model GWL-110210-PWS-S,23% Eff.

[ (EG&G ORTEC, Oak Ridge, TN) Used in conjunction with: Lead Shield Model G-16 (Applied Physical Technology, Atlanta, GA) A-2 9

Multichannel Analyzer ND66/Microvax (Nuclear Data, Schaumburg, IU I Digital Equipment, Maynard, MA) j Multichannel Analyzer I-Series 30 (Canberra, Meridian, CT) I Used in conjunction with: Alpha Spectrometry System Tennelec Electronics Model (Tennelec, Oak Ridge, TN) Used on conjunction with: Surface Barrier Detectors (EG&G ORTEC, Oak Ridge, TN) I I I I I I I A-3

LI I I I I I APPENDIX B SURVEY AND ANALYTICAL PROCEDURES I I I I I I I I

APPENDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES l Surface Scans Surface scans were performed by passing the probes slowly over the surface; the distance between the probe and the surface was maintained at a minimum - nominally about 1 cm. A large surface area, gas proportional floor monitor, with a 550 cm sensitive area detector, was 2 used to scan the floors of the surveyed areas. Other surfaces were scanned using small area 2 2 2 (15 cm,59 cm or 100 cm ), hand-held detectors. Identification of elevated levels was based on increases in the audible signal from the recording and/or indicating instrument. Combinations of detectors and instruments used for the scans were: Alpha Gas proportional detector with ratemeter-scaler. I ZnS scintillation detector with ratemeter-scaler. I Beta Gas proportional detector with ratemeter-scaler. GM detector with ratemeter-scaler. E Gamma NaI(TI) scintillation detector with ratemeter. I NaI(TI) scintillation FIDLER detector with ratemeter-scaler. I I I B-1

I Surface Activitv Measurements Measurements of total alpha and beta activity levels were performed using gas proportional, ZnS scintillation, and/or GM detectors with portable ratemeters-scalers. Count rates (cpm) were converted to activity levels (dpm/100 cm ) by dividing the net rate by the 4 r efficiency and 2 correcting for the active area of the detector. Beta surface activity was determined by correcting the measurement data, performed with the gas proportional detectors, for alpha contribution. The background count rates for the gas proportional detectors averaged approximately 1 cpm and 370 cpm for the alpha and alpha-beta voltage plateaus. Efficiency factors were 0.20 and 0.27-0.28 for the gas proportional detectors at the respective alpha and alpha-beta plateaus. The background count rates and efficiency factors were 2 cpm and 0.19 for the ZnS scintillation detector and 50 cpm and 0.26 for the GM detector. Active face areas for the gas proportional, ZnS scintillation, and GM detectors were 100 cm,59 cm and 15.5 cm, respectively. 2 2 2 Removable Activity Measurements Smears for determination of removable activity were performed using numbered filter paper disks, 47 mm in diameter; smears were placed in labeled envelopes with the location and other pertinent information recorded. The smears were counted on a low background gas-proportional counter for gross alpha and gross beta activity. I Gamma Exoosure Rate Measurements I Measurements of gamma exposure rates were performed using a pressurized ionization chamber (PIC). Soil Samoline Approximately I kg of soil was collected at each exterior and interior sample location. Surface samples were 0-15 cm in depth. B-2

I ANALYTICAL PROCEDURES Radiolocical Analyses Gamma Spectrometry I Soil, sediment, and concrete samples were dried, mixed and/or crushed then placed in an appropriate container, chosen to reproduce the calibrated counting geometry. The net weights I were determined and the samples counted using a high-purity intrinsic germanium detector coupled to a Nuclear Data Model ND-66/ Micro VaxII pulse height analyzer system. Background and Compton striping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system. Energy peaks used for determination of radionuclides of concern were: Cs-137 0.662 MeV U-235 0.143 MeV or 0.185 MeV Ra-226 0.609 MeV from Bi-214* " Secular equilibrium assumed. Spectra were also reviewed for other identifiable photopeaks. I l Strontium - 90 in Soil, Concrete and Paint Aliquots of soil, concrete, and paint were dissolved using a pyrosulfate fusion in which strontium was precipitated as a sulfate. Successive treatments with EDTA preferentially removed lead and excess calcium and returned the strontium to solution. Ferric and other insoluble hydroxides were precipitated at a pH of 12 to 14. Strontium was reprecipitated as a sulfate. Barium was removed as a chromate using DTPA. The final precipitate of strontium carbonate was counted using a low-background Tennelee proportional counter. I B-3

I I Polonium-210 in Soil I Samples of soil were decomposed by digestion with hydrofluoric acid and nitric acid in the presence of a lead carrier. Polonium was co-precipitated with lead sulfide from a dilute acidic I solution separating it from calcium, iron, and other interferences. The sulfide precipitate was dissolved in dilute hydrochloric acid and polonium was spontaneously deposited on a nickel disc I and counted by alpha spectrometry. I Promethium-147 in Soil I Neodymium carrier was added !n ashed soil and rare earths were dissolved in acid. The rare earths were then precipitated as oxalate, which was dissolved in EDTA and radioassayed by liquid scintillation counting. I Uncertainties and Detection Limits I The uncertainties associated with the analytical data presented in this report represent the 95 % l confidence levels for that data. These uncertainties were calculated, based on both the gross sample count levels and the associated background count levels. When the net sample count was less than the 95% statistical deviation of the background count, the sample concentration was reported as less than the detection limit of the measurement procedure. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides in samples, the detection limia differ from sample to sample and instrument to instrument. Additional uncertainties of i 6 to 10%, associated with laboratory procedures, have not been propagated into the data presented in this report. I I I e4 I

I Calibration and Ouality Assurance h 5 Analytical and field survey activities were conducted in accordance with procedures from the following documents: I - Survey Procedures Manual Revision 6 (February 1991) - Laboratory Procedures Manual Revision 6 (April 1991) - Quality Assurance Manual Revision 4 (April 1991) The procedures contained in these manuals were developed to meet the requirements of ANSI /ASME Nuclear Quality Assurance-1 (NQA-1,1989 ed). Calibration of all field and laboratory instrumentation is based on standards traceable to NIST, when such standards are available. In cases where they are not available, standards of an industry recognized organization are used. Calibration of pressurized ionization chambers is performed by the manufacturer. Quality control procedures include: Daily instrument background and check-source measurements to confirm that l equipment operation is within acceptable fluctuations. Participation in EPA and EML laboratory Quality Assurance Programs. Training and certification of all individuals performing procedures. Periodic internal and external audits. lI l B-5

I I I I I APPENDIX C GUIDELINES FOR DECONTAMINATION OF FACILITIES AhB EQUIPMENT PRIOR TO RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BY-PRODUCT, SOURCE OR SPECIAL NUCLEAR MATERIALS AND i GUIDELINES FOR RESIDUAL CONCENTRATIONS OF THORIUM AND URANIUM WASTES IN SOIL I I I I I I 1 I

I GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQUIPMENT PRIOR TO RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BYPRODUCT, SOURCE, 1 OR SPECIAL NUCLEAR MATERIAL I I U.S. Nuclear Regulatory Commission Division of Fuel Cycle & Material Safety I Washington, D.C. 20555 I I May 1987 I I I I I C-1 I: w_

I The instructions in this guide, in conjunction with Table 1, specify the radionuclides and radiation exposure rate limits which should be used in decontamination and survey of surfaces or premises and equipment prior to abandonment or release for unrestricted use. The limits in I Table 1 do not apply to premises, equipment, or scrap containing induced radioactivity for which the radiological considerations pertinent to their use may be different. The release of such facilities or items from regulatory control is considered on a case-by-case basis. 1. The licensee shall make a reasonable effort to eliminate residual contamination. ~ 2. Radioactivity on equipment or surfaces shall not be covered by paint, plating, or other covering material unless contamination levels, as determined by a survey and 3 documented, are below the limits specified in Table 1 prior to the application of the 5 covering. A reasonable effort must be made to minimize the contamination prior to use of any covering. .I 3. The radioactivity on the interior surfaces of pipes, drain lines, or ductwork shall be determined by making measurements at all traps, and other appropriate access points, provided that contamination at these locations is likely to be representative of contamination on the interior of the pipes, drain lines, or ductwork. Surfaces or premises, equipment, or scrap which are likely to be contaminated, but are of such size, l construction, or location as to make the surface inaccessible for purposes of measurement, shall be presumed to be contaminated in excess of the limits. l 4. Upon request, the Commission may authorize a licensee to relinquish possession or control of premises, equipment, or scrap having surfaces contaminated with materials in excess of the limits specified. This may include, but would not be limited to special circumstances such as razing of buildings, transfer from premises to another organization continuing work with radioactive materials, or conversion of facilities to a long-term storage or standby status. Such requests must: a. Provide detailed, specific information describing the premises, equipment or scrap, radioactive contaminants, and the nature, extent, and degree of residual .I. surface contamination. b. I Provide a detailed health and safety analysis which reflects that the residual amounts of materials on surface areas, together with other considerations such as prospective use of the premises, equipment, or scrap, are unlikely to result in an unreasonable risk to the health and safety of the public. I I 'I C-2

I 5. Prior to release of premises for unrestricted use, the licensee shall make a comprehensive radiation survey which establishes that contamination is within the limits specified in Table 1. A copy of the survey report shall be filed with the Division of Fuel Cycle, Medical, Academic, and Commercial Use Safety, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, and also the Administrator of the NRC Regional Office I having jurisdiction. The report should be filed at least 30 days prior to the planned date of abandonment. The survey report shall: a. Identify the premises. b. Show that reasonable effort has been made to eliminate residual contamination. c. Describe the scope of the survey and general procedures followed. d. State the findings of the survey in units specified in the instruction. Following review of the report, the NRC will consider visiting the facilities to confirm the survey. I I I I t I I I C-3

TAllLE 1 ACCEPTAllLE SURFACE CONTAMINATION LEVELS Nuclides' Average ^' Maximum

  • dJ Remevable*^'

6 U-nat, U-235, U-238, and 2 2 2 associated decay products 5,000 dpm a/100 cm 15,000 dpm a/100 cm 1,000 dpm a/100 cm Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, 2 2 2 Ac-227, I-125, I-129 100 dpm/100 cm 300 dpm/100 cm 20 dpm/100 cm Th-nat, Th-232, Sr-90, Ra-223, 2 2 2 Ra-224, U-232,1-126, I-131,1-133 1,000 dpm/100 cm 3,000 dpm/100 cm 200 dpm/100 cm Ileta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 2 2 2 o and others noted above. 5,000 dpm #7/100 cm 15,000 dpm #y/100 cm 1,000dpm #7/100cm L

  • Where surface contamination by both alpha-and beta-gamma-cmitting nuclides exists, the limits established for alpha-and beta-gamma-emitting nuclides should apply independently.

6 As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation.

  • Measurements of average contaminant should not be averaged over more than I square meter. For objects of less surface area, the average should be derived for each such object.

d 2 The maximum contamination level applies to an area of not more than 100 cm. 2

  • The amount of removable radioactive material per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removabic contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.

' The average and maximum radiation levels associated with surface contamination resulting from beta-gamma emitters should not exceed 0.2 mrad /h at I cm and 1.0 mrad /h at I cm, respectively, measured through not more than 7 milligrams per square centimeter of total absorber.

Guidelines for Residual Concentrations of Thorium and Uranium Wastes in Soil I On October 23,1981, the Nuclear Regulatory Commission published in the Federal Register a notice of Branch Technical Position on " Disposal or Onsite Storage of Thorium and Uranium Wastes from Past Operations." This document establishes guidelines for concentrations of uranium and thorium in soil, that will limit maximum radiation received by the public under various conditions of future land usage. These concentrations are as follows: I Maximum Concentrations (pCi/g) for various options Material l' 26 3' 4' I Natural Thorium (Th-232 + Th-228) with daughters present and in l equiiibrium 10 50 500 Natural Uranium (U-238 + U-234) with daughters present and in equilibrium 10 40 200 Depleted Uranium: Soluble 35 100 1,000 Insoluble 35 300 3,000 Enriched Uranium: Soluble 30 100 1,000 ) I Insoluble 30 250 2,500 lI ' Based on EPA cleanup standards which limit radiation to 1 mradlyr to lung and 3 mradlyr to bone from ingestion and inhalation and 10 R/h above background from direct external exposure.

  • Based on limiting individual dose to 170 mrem /yr.
  • Based on limiting equivalent exposure to 0.02 working level or less.
  • Based on limiting individual dose to 500 mrem /yr and in case of natural uranium, limiting exposure to 0.02 working level or less.

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