ML20235K323
| ML20235K323 | |
| Person / Time | |
|---|---|
| Site: | Lynchburg Research Center |
| Issue date: | 05/31/1987 |
| From: | Deming E OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20235K282 | List: |
| References | |
| CON-FIN-A-9076-3 NUDOCS 8710050047 | |
| Download: ML20235K323 (59) | |
Text
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Appendix A Prepared by i"d
- l,",',11;,,9" ^$$ CONFIRMATORY RADIOLOGICAL SURVEY d
III'Eucle[r OF THE Regulatory commission's CX-10 AREA, BUILDING A Region ll Office LYNCHBURG RESEARCH CENTER BABCOCK & WILCOX LYNCHBURG, VIRGINIA E. J. DEMING Radiological Site Assessment Program Manpower Education, Research, and Training Division l
l FINAL REPORT MAY 1987
"?A 2888 8?888L a G PDR I
CONFIRMATORY RADIOLOGICAL SURVEY OF THE CX-10 AREA, BUILDING A BABCOCK & WILCOX LYNCHBURG, VIRGINIA Prepared by E. J. Deming Radiological Site Assessment Program Manpower Education, Research, and Training Division Oak Ridge Associated Universities Oak Ridge, Tennessee 37831-0117 Project Staff J.D. Berger G.L. Murphy R.D. Condra M.H. Weick M.R. Dunsmore B.C. Williams R.D. Foley C.F. Weaver A.S. Masvidal Prepared for U.S. Nuclear Regulatory Commission Region II Office FINAL REPORT May 1987 This report is based on work performed under Interagency Agreement DOE No. 40-816-83 NRC Fin. No. A-9076-3 between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy. Oak Ridge Associated Universities performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Enargy.
TABLE OF CONTENTS Page List of Figures . . . . . . . . . . . . .. . . . . . . . . ... . . 11 List of Tables . . . . . . . . . . . . . . . . . . . . .. .... . iii Introduction and Site History . . . . . . . . . . . . . . . . .... 1 Site Description . . . . . .. . . . .. . . . . . . . . . . . ... 1 Survey Procedures . . . . . . .. . . . . .. . . . . . . . . ... . 2 Results . .... . . . . . . . . . . .. . . . .. .. . . . ... . 5 Comparison of Results with Guidelines . . . . . . .. . . . ..... 8 Summary . ... . . . . . . . . . . . . . . . . . . . . . . .. .. . 9 _ .___
Appendices Appendix A: Major 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, Source or Special Nuclear Material i
LIST OF FIGURES Page FIGURE 1: Map of Virginia, Showing Approximate location of Babcock and Wilcox Facility . . . . . . . . . . . . . . .
. 10 FIGURE 2: Map of Area Surrounding Babcock and Wilcox Facility ... . 11 FIGURE 3: Locations of Background Measurements and Baseline Sampling . . . . . . . . ... ... ............ 12 FIGURE 4: Location of Building A within the Lynchburg Research Center . . . . . . . . .... . .. ........... . 13 FIGURE 5: Location of Building A and the Waste Retention Tank Relative to Other Buildings at the Lynchburg Research Center . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 FIGURE 6: Building A Floor Plan -- First Floor . . .......... 15 FIGURE 7: Building A Floor Plan -- Second Floor ........... 16 FIGURE 8: Location of Surface Soil Samples . . . . . . . . . . . . . . 17 FIGURE 9: Location of Boreholes in the Waste Retention Tank Area . . . . . . . ... ...... ....... .. 18 FIGURE 10: Location of Exposure Rate Measurements Outside of the Licensed Areas . . . . .. . ... ............ 19 FIGURE 11: Location of Exposure Rate Measurements Inside of the CX-10 Licensed Area (Second Floor) . ......... ... ... 20 FIGURE 12: Plan View of J Lab Indicating Grid Blocks Surveyed . . . .. 21 FIGURE 13: Plan View of K Lab Indicating Grid Blocks Surveyed . . . . . 22 FIGURE 14: Plan View of L Lab Indicating Grid Blocks Surveyed . . ... 23 FIGURE 15: Plan View of M Lab Indicating Grid Blocks Surveyed . . ... 24 FIGURE 16: Plan View of the Old Chem Lab Indicating Grid Blocks Surveyed . . . . . . . . . . . ............... 25 FIGURE 17: Plan View of the Control Room Indicating Grid Blocks Surveyed . . . . . . . . .. . ...... ......... 26 FIGURE 18: Plan View of the Utility Room Indicating Grid Blocks Surveyed . . . . . . . ... ... ............. 27 11
b LIST OF FIGURES (Continued) {
i Page 1
FIGURE 19: Plan View of the Hallways Connecting the Control Room I With Bay 2 Indicating Grid Blocks Surveyed . . . . . . . . . 28
)
l FIGURE 20: Plan View of the Labyrinth (Second Floor) Indicating ]
Grid Blocks Surveyed . . . . . . .. .. . .. . . . . . . . 29 i l FIGURE 21: Plan View of the Floor and Lower Walls of Bay 2 Indicating Grid Blocks Surveyed . . . . .... . . . . . . 30 FIGURE 22: Plan View of the Labyrinth (First Floor) Indicating l Grid Blocks Surveyed . . . . . . . . . . . . . . . . . . . . 31 l
FIGURE 23: Upper walls of Bay 2 Indicating Location of Measurements . . 32 l 1
1 FIGURE 24: Plan View of the Waste Retention Tank Indicating Grid Blocks Surveyed .. ... ... . ... ... . . . . . .. 33 l
FIGURE 25: Plan View of the Waste Retention Tank Sump Pit Indicating Grid Blockc Surveyed . . . . . .. .. . ... . . .. . . . 34 l
iii 1
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E_ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ J
LIST OF TABLES Page TABLE 1: Direct Radiation Levels and Radionuclides Concentrations Measured at Baseline Sampling Locations . . . . .. . .. . . . . . . . . . . . . 35 TABLE 2: Sunmary of Surf ace Contamination Measurements CX-10 Area, Building A . . . . . .. . . . . . . . . . . . . 36 TABLE 3: Radionuclides Concentrations in Paint Samples CX-10 Area, Building A . . . . . .... . . . . . . . . . . 38 TABLE 4: Direct Radiation Levels and Radionuclides Concentrations in Surface Soil Samples Measured at Locations Around Building A . . . . . . . . . .. ... . . . . . . . . . . . 39 TABLE 5: Radionuclides Concentrations in Subsurface Soil Samples Collected From Boreholes Located Near the Waste Retention Tank . .. . .. . . . . . . . . . . . .
40 iv
CONFIRMATORY RADIOLOGICAL SURVEY OF THE CX-10 AREA, BUILDING A-LYNCHBURG RESEARCH CENTER BABCOCK & WILCOX LYNCHBURG, VA INTRODUCTION AND SITE HISTORY The Lynchburg Research Center (LRC), located on the Babcock & Wilcox Company site east of Lynchburg, Virginia, is a facility used to test and study nuclear cycles. Building A was formerly used as a criticality experiment facility and consists of the original building built in 1956 and three additions built between 1957 and 1963. The original building housed two low power criticality experiment reacto rs , CX-1 and CX-19. Both reactors ceased operation in 1971, and were decommissioned and dismantled in 1973.
The first building addition includes a high bay which housed the CX-10 and CX-12 experimental reactors. All formerly utilized areas in the addition, as well as a waste retention tank north of Building A, are controlled under Nuclear Regulatory Commission (NRC) license CX-10, which was originally issued by the Atomic Energy Commission in 1958. Both reactors have been dismantled, and the areas and support rooms were decontaminated in 1985.
At the request of the NRC, Region II, the Radiological Site Assessment Program of Oak Ridge Associated Universities (ORAU) conducted a radiological survey to evaluate the CX-10 facilities radiological status relative to the NRC guidelines for unrestricted use.
SITE DESCRIPTION The Lynchburg Research Center is located on the Mt. Athos, Babcock &
Wilcox property, west of Route 726, 16 kilometers east of Lynchburg, Virginia (Figures 1 and 2). The f acility is in Campbell County along the James River.
The Babcock & Wilcox company maintains 212.5 hectares of which 5.5 ha are utilized by the Lynchburg Research Center. Other Babcock & Wilcox f acilities operated on the site are the Commercial Nuclear Fuels Plant and the Naval Nuclear Fuels Division (Figure 4).
1
Building A is located on the southwest edge of the Lynchburg Research Center. The relative location of Building A to other buildings at the LRC is shown on Figure 5. The CX-10 area of Building A included a high bay, consisting of three levels (Level 1 in the basement, and two additional levels accessed by catwalks), a control room, chemistry laboratory, electronics shop, several offices. and Rooms J, K, L, and M, which originally served as the sub-assembly room (Figures 6 and 7). During decommissioning, all contaminated materials, equipment, drain lines and ductwork were removed from the facility for disposal.
The waste retention tank and sunn pit are located at the southeast corner of Building D-0. An area fanning out to the north and east of the waste retention tank was excavated, because historical information indicated that the tank had overflowed in the past (Figure 9).
SURVEY PROCEDURES Document Review ORAU reviewed the Babcock & Wilcox final survey report and supporting documentation for the CX-10 area. Approximately 10% of the raw data on a room by room basis was compared to the final report. No discrepancies between the raw data and final report were noted.
Facility Survey Gridding Confirmatory measurements were referenced to the grid systems established by Babcock & Wilcox. A 1 m x 1 m grid was reestablished in most areas while a 3mx 3 m grid was utilized in areas where the probability of contamination was low, based cn history of use (e.g., halls and offices).
2
Exposure Rate Measurements Gamma exposure rates at 1 meter above the floor were measured at 13 locations within the CX-10 area and throughout Building A (Figures 10 and 11),
using NaI(Tl) gamma scintillation detectors cross calibrated onsite with a pressurized ionization chamber.
Surface Scans and Measurement of Total and Removable Contamination Thorough, systematic alpha, beta-gamma, and gamma scans were performed on floors and lower walls (up to 2 m) using a gas proportional alpha floor monitor, zine sulfide alpha detectors, " pancake" GM detectors, and NaI(TI) scintillation detectors coupled to scalers /ratemeters with audible indicators. Representative areas on overhead surfaces (higher than 2 m) such as ledges, beams, pipes, ductwork and miscellaneous equipment were also scanned.
Ninety-seven grid blocks on the floors and lower walls in the high bay, control room, offices, and laboratories were randomly selected for surface contamination measurements (Figures 12-23). Total measurements of alpha and beta gamma contamination levele were systematically performed at the center and four points, midway between the center and block corners. Smears for removable alpha and beta contamination were performed at the location in each grid block where the highest direct reading was obtained. Total and removable contamination levels were also measured at 44 locations on the upper walls, ceilings and miscellaneous overhead objects. Eight smears were obtained from access ports within the CX-10 area. Two large wipes were pulled through the major " hot" drainlines in Building A using an engineers tape.
Paint Sampling Thirteen paint samples were collected from 100 cm 2 areas on the floors and walls, at random locations throughout the CX-10 area, using commercial paint stripper.
3
Roof Beta-gamma and gamma scans were conducted on the Building A roof. A sample of gravel was obtained f rom the roof of Bay 2,1 meter west of an air exhaust vent. Two smears were collected from ventilation systems at the roof exhaust ports or vents in the CX-10 area.
Outside Area Survey Cridding A 10 ft x 10 ft grid system previously established by Babcock & Wilcox was used for reference.
Surface Scans and Measurement of Total and Removable Contamination ORAU scanned the outside area adjacent to Building A, using NaI(TI) gamma scintillation detectors. Alpha and beta gamma scans were performed on the floor and lower walls of the waste retention tank. Eight grid blocks on the floor and lower walls of the waste retention tank and sump were selected for surface contamination measurements (Figures 24 and 25). Additional contamination measurements were made in one grid block on the waste retention tank ceiling. Nine smears were obtained, one f rom each of the grid blocks surveyed, for the measurement of removable contamination.
Exposure Rate Measurements Exposure rate measurements were made at the surface and 1 meter above the surface at soil sampling locations (Figure 7), using NaI(T1) gamma I scintillation detectorn, cross calibrated with a pressurized ionization chamber.
Sampling Twelve surface soil samples were collected from the area around Building A (Figure 8). Eight shallow boreholes (to 1.5 m) were drilled near 4
the waste retention tank (Figure 9). Surf ace and subsurf ace soil samples were collected and the boreholes were scanned using a collimated NaI(T1) scintillation detector. Samples of water were collected from the waste retention tank and sump.
Background Samples and Measurements Samples of surf ace soil were collected and background exposure rates were measured at 6 off site locations (Figure 3) in the area around the Babcock &
Wilcox facility to establish baseline radionuclides soil concentrations and direct radiation levels.
Sample Analysis and Interpretation of Results Soil, sediment, gravel, and paint samples were analyzed by gamma _ . _ _
spectrometry, and the spectra were reviewed for identifiable photopeaks, with particular attention to U-238, U-235, Th-232, Co- 60, and Cs-137. Smears obtained for the dete rmination of removable contamination were analyzed for gross alpha and beta activity. Additional information concerning major instrumentation, sampling equipment, and analytical procedures is provided in Appendices A and B. Results were compared with NRC guidelines, established for release of facilities for unrestricted use (Appendix C).
RESULTS ,,
Document Review In general, the decontamination plan appears to be adequately developed and implemented to ensure the NRC guidelines are met, and the final survey report adequately summarized the radiological status of the site. No significant discrepancies were identified in the documents reviewed.
Background Levels and Baseline Concentrations Background exposure rates and baseline radionuclides concentrations in soil from the vicinity of the Babcock & Wilcox, Lynchburg site are presented in Table 1. Exposure rates ranged from 8 to 11 pR/h. Uranium-235
concentrations were <0.3 pCi/g and U-238 concentrations ranged from <0.4 pCi/g to 2.1 pCi/g. Cobalt 60 and cesium 1 37 concentrations were <0.1 pCi/g and
<0.2 pC1/g; Ih-232 concentrations ranged from 0.4 pCi/g to 2.8 pCi/g.
Facility Survey No contamination of building surfaces was identified by alpha and beta-gamma scanning.
Gamma scans inside and outside of Building A, indicated no elevated exposure rate levels above the background rate of approximately 8 to 11 pR/h (Table 4).
Contamination Measurements . __
Results of total and removable contamination measurements are summarized in Table 2. Alpha and beta-gamma levels were generally well below the release guidelines, and, in many instances, less than the minimum detectable activity. The maximum alpha measurement was 260 dpm/100 cm2and the maximum beta-gamma measurement was 3300 dpm/100 cm2 . The maximum removable alpha and 2 2 beta contamination levels were 5 dpm/100 cm and 53 dpm/100 cm ,
respectively. The highest alpha grid block average was 120 dpm/100 cm 2 and the highest beta grid block average was 1700 dpm/100 cm2, Removable alpha contamination, measured on smears collected from 8 access ports within the licensed area, was <2 dpm/100 cm2 . Removable beta .
2 contamination ranged from <5 to 15 dpm/100 cm . No contamination in excess of guidelines was detected on smear samples collected from drainlines within Building A.
Paint Samples 2
Total radionuclides activities measured in paint samples from 100 cm areas are presented in Table 3. Uranium-2?8 activities ranged from <0.2 to 130 dpm/100 cm , and co-60 r.ctivities ranged from <0.2 to 12 dpm/100 cm2 ,
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Roof l Removable l alpha and beta contamination measurements from roof exhaust 2 2 vents were <2 dpm/100 cm and <5 dpm/100 cm , respectively (Table 2).
i The radionuclides concentrations measured in a gravel sample collected from the roof of Building A directly over Bay 2 were: 4.2 pCi/g of Th-232, and 1.5 pCi/g of : Cs-137. Uranium-235, U-238, and Co-60 concentrations were i below the minimum detectable concentrations for these radionuclides. 4 i
Outside Survey l Surface Scans o
Gamma scans of the area outside of Bui?Aing A did not identify any areas with elevated gamma levels. No significant contamination of the waste retention tank and sump was identified by alpha and beta-garama scanning.
Contamination - Measurement s The res61ts of total and removable contamination measurements made in the waste retention tank and sump are included in Talle 2. Total alpha 2
measurements ranged from <27 to 260 dpm/100 cm , and bc[a measurements ranged f rom <440 to 1800 dpm/100 cm2 . Removable alpha and beta ranges were <2 to 2 2 3 dpm/100 cm cttd <5 to 7 dpm/100 cm , respectively.
i Radionuclides Concentrations in Surf ace Soil y
, i The radionuclides concentrations in twelve , oil samples collected around the outside of Building A are presented in Table 4. .These concentrations are within normal background ranges for the baseline samples from the Lynchburg vicinity.
l Radionuclides Concentrations in Soil from Shallow Boreholes '
The radionuclides concentrations measured in soil samples from boreholes at 8 loca? ions around the waste retention tank (Figure 9) tre presented in 7
-____-__-_____-______-___-_N
Table 5. The levels of Co-60, Cs-137, U-238, U-235, and Th-232 are all within the range'of normal background soil concentrations for these nuclides.
Radionuclides Concentrations in Water Gross alpha and gross beta concentrations' measured in water samples collected from the waste retention tank were 53 pCi/1 and 530 pC1/1, t
respectively. The concentrations measured in the water sample from the sump i were 6.8 pCi/1 gross alpha, and 110 pC1/1 gross beta. Further analysis of the water from the waste retention tank was performed to determine the individual radionuclides concentrations. The measured concentrations were: U-238, 170 pC1/1; U-235, <14 pCi/1; Cs-137, 34 pCi/1; Co-60, 24 pCi/1; and Th-232, I
<8.8 pCi/1. I COMPARISON OF RESULTS WITH GUIDELINES l
NRC surface contamination guidelines for release of iacilities for unrestricted use are presented in Appendix C. Because the principal radionuclides of interest are Th-232, U-238, U-235, Co-60 and Cs-137, the more restrictive criteria for Th-232 have been applied for residual alpha ]
contamination, and they are as follnws:
Total Contamination 3,000 dpm/100 cm2 (maximum in a 100 cm2 area) 1,000 dpm/100 cm2 (averaged over 1 m2) l Removable Contamination 200 dpm/100 cm 2 l
For residual beta-gamma contamination, the NRC guidelines are:
Total Contamination 2
15,000 dpm/100 cm2 (maximum in a 100 cm area) '
5,000 dpm/100 cm2 (averaged over 1 m2 )
Removable Contamination 1,000 dpm/100 cm 2 8
All total and removable alpha and beta-gamma levels as well as removable contamination measurements were'within these guidelines.
Soil- samples were compared to the NRC guideline of 10 pCi/g for natural thorium and 30 pCi/g for uranium enriched in U-235. All of the soil samples (surface and subsurface), collected near the LRC Building A and the waste retention tank, were within guideline levels.
ladionuclide concentrations measured in water from the waste retention l
tank, were several orders of magnitude. below the acceptable release levels in 10 CFR 20 Appendix B, Table II.
I
SUMMARY
On July 14 - 25, 1986, ORAU performed a confirmatory radiological survey of the CX-10 area of Building A at the Babcock & Wilcox Lynchburg Research Center in Lynchburg, VA. The survey included surface alpha, gamma and beta-gamma scans, measurement of direct and removable contamination levels, and the measurement of radionuclides in soil, gravel, water, and paint samples. The findings support the close-out survey performed by the licensee, and confirm that the radiological conditions satisfy the NRC guidelines established for release for unrestricted use.
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h 0 5 METERS FIGURE 23: Upper Walls of Bay 2 Indicating Location of Measurements 32
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J' ,
i,
! TABLE I DIRECT RADIATION' LEVELS AND RADIONPCLIDE CONCENTRATIONS L MEASURED AT BASELINE SAMPLING LOCATIONS .^ . !
.BALCOCK & WILCOX COMPANY LYNCHBURG, VIRGINIA 1- -
=
Locationa Gamna Exposure Rates Radiohkielide Concentrations (pCi/g) l b at I m Above the Surface Co-60 Cs-137 U-238 U-235 Th-232 l (pR/h) '
)
1 11 <0.1 (0.1 2.1 1.8b <0.3 2.8 10.7 i
'l 2 8 <0.1 <0.1 0.310.2 <0.2 0.4 1 0.2 s
3 10 <0.1 <0.2 <2.4 <0.2 0.9 1 0.4 4 10 <0.1 <0.1 <0.4 <0.1 0.5 0.3 5 11 <0.1 (0.1 <1.4 <0.3 1.3 10.5 6 10 <0.1 <0.2 <0.8 (0.2 1.2 0.4 l
aRefer to Figure 3. !
b rrors are 20 based only on counting statistics.
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't:ADIONUCLIDE CONCENTRATIONS IN PAINT SAMPLES t' .{
CX-10. AREA,. BUILDING A j y ,'. BABCOCK & WILCOX COMPANY
.]
'LYNCHBURG,' VIRGINIA
.q Location . Surface' Crid Radionuclides Concentrations dom /10'0 cm2 .
(
-]
BAY 2:-
Level: 3- Wall 7. <0.2 <0 . 2 ' 4.9 2.2a <o,7 <1,1 Level .3< Wall 37 <2.0 ' <0.2 8. 2 ' t. 3.1. <0.9 <5.8 Lovel 3- Wall 59 12 2- <1.8 '130 18 <7.3 <12
- Level 2- : Wall 7 5.1 4.0 <1.1- 31 1 11 <3.1 <20 Level-2 Wall :10 <0.9 <0.4 7.3 i 5.6 < 1. 6 . <2.7 Level 2 Wall 17 9.6 i 1.6 <0.7 14 i 8 <2.7 <8.2 Level I Wall 5 3.11-3.6 <1.1 44 1 13.3 <4.0 <6.7 Level 1 Floor C2 3.112.4 2.2'i 1.1 31 9 <2.4 <4.0 Level 1 Floor- G6. <6.7 5.6. t ' 2. 4. 38 1 16 (4.0 <7.1
. Labyrinth Wall L2 '<2.7 <2.0 62 1 18 <5.3 <8.7; CONTROL ROOM West Wall W, 22' <0.9) <0.4 8.7 5. 6 ' <1.8 <3.1~
LEast Hall' Wall: EH, 44 <0.9 .<0.4 24 1 7 <1.8 <12
' South .W all S, 6 <2.9 <2.2 47 i 24 (6.4 19 t 16-aErrors are' 2a based only on counting statistics.
1 38 i
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ .1
TABLE'4 DIRECT RADIATION LEVELS AND RADIONUCLIDES CONCENTRATIONS IN SURFACE SOIL SAMPLES MEASURED AT LOCATIONS AROUND BUILDING A BABCOCK & WILCOX COMPANY LYNCHBURG, VIRGINIA i
Locationa Camma Exposure Radionuclides Concentrations (pCi/g)
Rates at 1 m Above Co-60 Cs-137 U-238 U-235 Th-232 the Surface (p/Rh) 1 11 <0.1 <0.9 4.5 i 1.5b <0.3 1.9 0.5 2 11 <0.1 0.4 1 0.1 1.7 1 0.7 <0.2 1.2 0.4 3 11 <0.1 0.5 i 0.1 0.8'i 0.6 <0.3 1.6 0.5 4 11 <0.1 <1.6 <0.9 <0.3 1.2 0.4 5 11 <0.1 <0.1 <0.7 <0.2 1.4 1 0.5 6 11 <0.1 <0.2 1.8 1 1.0 <0.3 1.3 0.7 7 10 <0.1 <0.1 <1.0 <0.3 2.2 1 0.8 8 10 <0.1 0.8 1 0.1 1.8 1 1.5 <0.3 1.4 1 0.5 9 10 <0.2 0. 3 i 0.1 (0.8 <0.3 1.4 10.4 l
10 10 <0.1 1.1 i 0. 2 <2.2 <0.2 0.9 0.3 11 10 <0.1 0.3 1 0.1 2.3 1 0.6 <0.2 1.4 1 0.6 12 10 <0.1 <0.2 <0.8 <0.2 0.9 1 0.3 j aRefer to Figure 8.
I bE rrors are 20 based only on counting statistics. )
39
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. 'j ,
APPENDIX A MAJOR ANALYTICAL EQUIPMENT l
i i
______2-___-___ - _ . _ - . _ _ _ .
1 APPENDIX A MAJOR ANALYTICALIEQUIPMENT:
The display or description. of a specific product is not to be construed asL an endorsement.of.that product or its manufacturer byf the. authors or their employer.
'A.' ' Direct : Radiation Measurements
.Eberline " RASCAL" Portable Ratemeter-Scaler Model PRS-1 (Eberline, Sante Fe, NM)
Eberline PRM-6 Portable Ratemeter (Eberline, Sante Fe, NM)
Ludium Alpha Floor Monitor Model- 239-1 (Ludium, Sweetwater, TX)
Eberline Alpha Scintillation' Probe Model AC-3-7
'(Eberline, Sante Fe, NM)
Eberline Beta-Gamma " Pancake" Probe Model HP-260 (Eberline, Sante Fe, NM) -
Victorcen Beta-Gamma " Pancake" Probe Model 489-110 (Victoreen, Inc., Cleveland, OH)
Reuter-Stokes Pressurized Ionization Chamber
'Model RSS-111 (Reuter-Stokes, Cleveland, OH)
Victoreen NaI Camma Scintillation Probe Model A89-55 (Victoreen, Inc., Cleveland, OH)
B. Laboratory Analyses Low Background Alpha-Beta Counter Model LB5110-2080 (Tennelec, Inc., Oak Ridge, TN) i A-1 j
1-I i
Ge(L1) Detector Model LGCC2220SD, 23% ef ficiency (Princeton Gamma-Tech, Princeton, NJ) I 1
Used in conjunction with:
Lead Shield, SPG-16 (Applied Physical Technology, Smyrna, GA)
High Purity Germanium Detector f Model GMX-23195-S, 23% efficiency- ;
(EG&G ORTEC, Oak Ridge, TN) i j
Used in conjunction with:
Lead Shield, G-16 (Gamma Products, Inc., Palos Hills, IL)
High Purity Germanium Coaxial Well Detector l
Model GWL 110210-PWS-S, 23% efficiency (EC&G ORTEC, Oak Ridge, TN)
Multichannel analyzer ND-66/ND-680 System (Nuclear Data, Inc., Schaumburg, IL) 4 A-2
__ -___-___--_a
i l
(
APPENDIX B i MEASUREMENT AND ANALYTICAL PROCEDURES l
1
.i
APPENDIX B Measurement and Analytical Procedures Gamma Scintillation Measurement Walkover surface scans and measurements of gamma exposure rates were performed using Eberline Model PRM-6 portable ratemeters with Victoreen Model 489-55 gamma scintillation probes containing 3.2 cm x 3.8 cm NaI(Tl) scintillation crystals. Count rates were converted to exposure rates (tR/h) by cross-calibrating with a Reuter-Stokes. Model RSS-111 pressurized ionization chamber at six representative onsite locations.
Alpha and Beta-Gamma Measurements Floors were scanned for elevated alpha levels by passing slowly over the i urface with a Ludlum Model 239-1 Gas Proportional Alpha Floor Monitor with a 600 cm 2 sensitive area. Other surfaces were scanned using Eberline Model PRS-1 portable scaler /ratemeters coupled to alpha scintillation probes.
Walkover surface scans were also performed using Eberline Model PRM-6 portable ratemeters with Victoreen Model 489-55 gamma scintillation probes containing 3.2 cm x 3.8 cm NaI(TI) scintillation crystals. Other surfaces were scanned using Eberline Model PRS-1 portable scaler /ratemeter with Model HP-260 thin-w!ndow pancake G-M probes. )
, I l j Measurements of total alpha radiation levels were performed using i
Eberline Model PRS-1 portable scaler /ratemeters with Model AC3-7 alpha ]
scintillation probes. Measurement of direct beta-gamma radiation levels were performed using Eberline Model PRS-1 portable Scaler /ratemeters with Model llP-260 thin-window pancake GM probes. Count rates (cpm) were converted 2
to disintegration rates (dpm/100 cm ) by dividing the net rate by the 4n l
efficiency and correcting for active area of the detector. The effective l 2 2 window area was 59 cm for the alpha detectors and 15 cm for the GM i
detectors. The average background count rate was approximately 2 cpm for the .
alpha probes and 44 cpm for the GM probes.
I
\
j r
B-1 l
. 4
Removable Contamination Measurements Smears for determination of removable contamination levels were collected on numbered filter paper disks 47 mm in diameter, then placed in individually labeled envelopes with the location and other pertinent information recorded.
The smears were counted on a low background gas proportional alpha-beta counter. i i
Soil Sample Analysis Gamma Spectrometry Soil samples and gravel were mixed, and a portion placed in a 0.5 liter Marinelli beaker. The quantity placed in each beaker was chosen to reproduce the calibrated counting geometry and ranged from 600 to 800 g. Net weights were determined, and the samples counted using intrinsic germanium and Ge(Li) detectors coupled to a Nuclear Data Model ND-680 pulse height analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system. Each spectra was scanned for identifiable photopeaks which could be attributed to Babcock & Wilcox operations.
Water Sample Analysis Water samples were rough-filtered through Whatman No. 2 filter paper.
Remaining suspended solids were removed by subsequent filtration through 0.45 um membrane filters. The filtrate was acidified by addition of 10 ml of concentrated nitric acid. A known volume of each sample was evaporated to dryness and counted for gross alpha and gross beta using a Tennelec Model LB-5110 low-background proportional counter.
One water sample was further analyzed by measuring a portion into a 0.5 i
liter Marinelli beaker, determining the net weight, and performing gamma j i
spectrometry by counting the sample on an intrinsic germanium detector coupled )
to a Nuclear Data Model ND-680 pulse height analyzer system.
l B-2 !
I i
Paint Sample Analysis Paint samples were dried, homogenized and placed .in small tubes.
Qualitative analysis was performed by counting the samples in an intrinsic germanium well counter.
Errors and Detection Limits d
The errors associated with the analytical data presented in the tables of :
this report, represent the 95% (2a) confidence icvels for that data. These errors 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 20 statistical deviation of the background count, the sample concentration was reported as less than the minimum detectable concentration (<MDC).
Because of variation in background levels and the effects of the compton continuum caused by other constituents in the samples, the MDC's for specific radionuclides differ from sample to sample.
Calibration and Quality Assurance Laboratory and field survey procedures are documented in manuals developed specifically for the Oak Ridge Associated Universities' Radiological Site Assessment Program.
With the exception of the measurements conducted with portable gamma scintillation survey meters, instruments were calibrated with NBS-traceable standards. The calibration procedures for the portable gamma instruments are performed by comparison with an NBS calibrated pressurized ionization chamber.
Quality control procedures on all instruments included daily background and check-source measurements to confirm equipment operation within acceptable statistical fluctuations. The ORAU laboratory participates in the EPA and EML Quality Assurance Programs.
B-3
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i
l
?
APPENDIK C GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQUIPMENT PRIOR TO' RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BYPRODUCT, SOURCE OR SPECIAL NUCLEAR MATERIAL 1
)
i i
1 I
l
GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQUIPMENT PRIOR TO RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BYPRODUCT, SOURCE OR SPECIAL NUCLEAR MATERIAL l
U.S. Nuclear Regulatory Commission Division of Fuel Cycle 6 Material Safety Washington, D.C. 20555 I
July 1982 k
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 surf aces or premises _ and equipment prior to abandonment or release for unrestricted use. The limits in 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 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 documented, are below the limits specified in Table l ' prior to the application of the covering. A reasonable ,
effort must be _ made to minimize the contamination prior to use of any !
covering.-
- 3. The radioactivity on 'the interior surfaces of pipes, drain lines, or ductwork shall be determined by making measurements at all traps, and i 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, construction, or location as to make the surface inaccessible for purposes of measurement shall be presumed to be contaminated in excess of the limits.
- 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 of premises to another organization continuing work with radioactive materials, or conversion of l 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 surface contamination.
- b. 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.
- 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 C-2
the survey report shall be filed with' the Division of Fuel Cycle and Material Safety, USNRC, Washington, D.C. 20555, and also the Administrator of the NRC Regional Office having jurisdiction. The report should be filed at least 30 days prior to the planned date of
, abandonment. They survey report shall:
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- 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 f acilities to conff rm the survey.
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