ML20205M801
| ML20205M801 | |
| Person / Time | |
|---|---|
| Site: | 07001489 |
| Issue date: | 09/30/1986 |
| From: | Boerner A OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20205M785 | List: |
| References | |
| CON-FIN-A-9076-3 NUDOCS 8811030293 | |
| Download: ML20205M801 (54) | |
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Prepared by gg,aig,,A"*i*d CONFIRMATORY RADIOLOGICAL SURVEY
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Ij'S?'Ociel, OF THE Regulatory commiaion's WINGFOOT LAKE Region ill Office Supported bv ADVANCED TECHNOLOGY CENTER Safeguards and lr'a'n'A': '"
GOODYEAR AEROSPACE CORPORATION Division of
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inspection Programs:
AKRON, OHIO Office of Inspection and Enforcement A.J.BOERNER
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Radiological Site Assessment Program Manpower Education, 9esearch, and Training Division FINAL REPORT
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SEPTEMBER 1986 E
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CONFIRMATORY RADIOLOGICAL SURVEY
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OF THE WINGF00T LAKE ADVANCE 0 TECHNOLOGY CEN*ER GOODYEAR AEROSPACE CORPORATION
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AKRON, OHIO Prepared 'oy A.J. BOERNER I
Radiological Site Assessment Program Manpower Education, Research, and Trainf.ng Division
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Oak Ridge Associated Univeretties L
Oak Ridge, Tennessee 37831 1117 Project Staff J.D. Berger A.S. Masvidal R.D. Condra R.C. Rookard M.R. Dunsmore C.F. Weaver M.J. Laudeman i
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Prepared for Safeguards and Materials Programs Branch
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Division of Inspection Programs U.S. Nuclear Regulatory Cotmaission Region III Office
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FINAL REPORT September 1986
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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 f
performs complementary work under contract number DE-AC05-760R00033 with the L
U.S. Department of Energy.
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t TABLE OF CONTENTS fage,,
ii List of Figures.
List of Tables
............................. iii I
Introduction 1
Site Descrf ption l
S u rvey Proce d u re s............................
2' 6
Results l
Comparison of Results with Guidelines 9
Summary.................................
10 29 f
References 1
j Appendices l
l Appendix At Major Analytical Equipment 1
Appendix B: Measurement and Analytical Procedures Appandix C: Standard Review Plan for Termination of Special Nuclear l
Material Licenses j
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LIST OF FIGURES ram FIGURE 1: Akron, Ohio Area Indicating the Locatiott of the Goodyear Aerospace Corporation Wingfoot Lake Advanced II Technology Center
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FIGURE 2: General Floor Plan of the Wingfoot i.ake Advanced i
12 Technology Center
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FIGURE 3: Areas Associated with the Centrifuge Operation..
13 FIGURE 4: Exterior View of the Wingfoot Facility Showing l
14 I
Casings Storage Area FIGURE 5: Grid Systems Established for Survey Reference 15 FIGURE 6 Locations of Exposure Rate Measurements 16 17 FIGURE 7 Drain Sampling Locations FIGURE 8: Locations of Soil Samples Collected Adjacent to the 18 Casings Storage Area FIGURE 9: Areas of Contamination Identified by the Walkover 19 Surface Scan FIGURE 10: Areas of Concrete Removal During Remedial Action
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?IGURE 11 locations of Subsurface Soil !?mples Collected From Excavated Areas Fdlowing Recoval of 21 Co.
.d. 71 voting 1
FIGURE 12: Location of Contaminated Drain Line Which was Removed.
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l LIST OF TABL2S TABLE 1: Summary of Surf ace Contamination Levels Measured in the 23 Wingfoot Facility TABLE 2: Contamination Levels Measured at Locations Identified by the Surfee Scans 26 TABLE 3 Uraniute-238 Concentrations in Subsurf ace Soil Samples Collected Following Removal of Concrete Flooriaig......
27 TABLE 4: Uranium-238 Concentrations in Soil Sssples Collected Following Removal of Contaminated Drain Lines 28
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CONFIRMATORY RADIOLOGICAL SURVEY OF THE WINGF00T LAKE ADVANCED TECHNOLOGY CENTER C00DYEAR AEROSPACE CORPORATION AKRON, OHIO INTRODUCTION From 1974 to
- 1985, the Goodyear Aerospace Corporation conducted garformance testing on developmental gas centrifuges at the company's Wingfoot
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Lake Advanced Technology Center in Akron, Ohio.
The work was performed through funding by the Department of Enerc' (DOE) and under Nuclear Regulatory Commission (NRC) license SNM-1461. The license authorized the use of slightly enriched (to 1%) UF. Depleted uranium (to 0.5%) was also produced during the 6
testing process.
Following termination of the
- project, Goodyear decontaminated the building and equipment used in the centrifuge operation.
Contaminated materials were drummed and shipped of f sitn.
Subsequently, the l
licensee filed a report (January 1986) with the NRC indicating that the f aci,lity Latisfied the NRC guidelines for release from licensing restrictions.I At the request of the liuelear Regulatory Commission's Region 11I Office, the Radiological Site Assessmant Program of Oak Ridge Associated Universities (ORAU) conducted a confirmatory survey of the Goodyear Wingfoot facility. This report presents the procedures and results of that survey.
SITE DESC 'PTION The Goodyear Wingfoot Lake Advanced Technology Center is located approximately 13 kilometers east of Akron, Ohio on Wingfoot Lake road (Figure 1).
Centrifuge testing and storage of equipment and materials were restricted to the southern portion of a large hangar used for air ship storage and mintenance (Figure 2).
The main floor area where centrifuge operations 2
were conducted contains approximately 1735 m.
Associated areas used for storage of equipment, parts and waste materials comprise an additional 2
1330 m. Several small rooms, consisting primarily of laboratories and office areas, are located adjacent to the min process area. Ceiling heights range from <10 meters to approximately 30 meters in portions of the main process and o},en hangar areas.
1 a
Areas directly and indirectly involved in the centrifuge operation inc1#.eds the fabrication tower, the mass spectrometer laboratory, hood and J
cut off saw room, rotor and column cut up areas, power hacksaw location, UF6 cylinder storage ar.d decontamination areas, a "pit" consisting of five subsurface levels where the actual centrifuge testing took place, and storage areas for vaste and parts (Figure 3).
Centr!fuge casings and contaminated floor amterials and soil were stored outside on a large concrete pad (Figure 4).
Most of the individual components and equipment used in the operation, in addittoa to control er.haust ventilation systems, were removed prior to May 1986.
SURVEY PROCEDURES During the period of May 13-16, 1986, ORAU personnel conducted a i
l confirmatory radiological survey of the Wingfoot 1.ake Advanced Technology Center.
The purpose of the survey was to verify the adequacy of the licensee's final survey and confirm the radiological condition of the facility l
relative to decommissioning criteria.
Obj ec t ive s The objectives of the survey were tot 1.
measure exposure rate levels in the Wingfoot facility; 2.
esasure total and transferable surface contamination levule on floors, wadis, overhead supports, piping and miscellaneous fixtures, ductwork, equipernt and drains in the facility; and 3.
determine radionuclide concentrations in soil and water samples.
Procedures A.
Indoor Areas Gridding A 2 m x 2 m grid pattern was established on the floor (Figure 5) using the southeast corner of the building as the baseline coordinate (A,0).
2
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Alphabetical designations were incremented from east to west; the nume rical portion of the grid was deterafned along a north to south directional.
The grid was extended to include the hood and cut-of f saw h
room and the power hacksaw arcas.
Rooms adjacent to the main processing area 1.e.,
laboratories and of fice areas, we re not gridded.
Based on
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negative survey findings, the grid was not extended beyond the column cut up and power hacksaw areas.
However, measurements taken outside the gridded area (including the waste storage and parts staging areas) were referenced back to existing building features. Measurements in the "pit" area and on lower and upper walls and horizontal su'rfaces were referenced to the floor grid or building landmarks.
Surface Measurements 1.
Main Floor area Floor areas were scanned with alpha and be ta-gasuna floor monitors and NaI(T1) gamma scintillation detectors.
1.ocations inaccessible to the floor monitors were scanned with hand-held alpha scintillation and beta-gamma "pancake" probes.
- Alpha,
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beta-gamma, and gamma scanning was performed on lower walls.
Upper wall and overhesd surface scanning on
- ledges, beams,
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piping, fixtures, equipment and ducteork was conducted using hand-held alpha and beta-gamma probes.
Elevated areas were noted for additional, followup measurements.
Total measurements of alpha and beta-gamma contamination levels on floor and lower wall grid blocks were performed at the center and four equidistant
- points, midway between the center and block corners.
Smears for removable alpha and beta contamination were performed at the location in each grid block whera the highest total
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measurement was obtained.
Total and removable contamination taeasurements were also performed on upper walls and on ledges, piping, and ungridded horizontal and vertical surfaces.
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2.
Pit Area Floor surfaces, lower walls, and equipment were scanned on each
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of five subsurface sections of a pit where centrifuge testing was conducted.
Hand-held alpha, beta-gamma, and gamma detectors were j
used for the scans.
Total and removable contamination levels were determined at representative locations.
3.
Laboratories, Of fice Areas, and Service Mezzanine Alpha, be ta-gamma and gamma scanning was performed on the floor, lower walla and other surfaces in laboratories and office areas
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adjacent to the main process area.
Floor areas were scanned in the service mezzanine.
Alpha and beta-gamma total and removable contamination levels were measured at s11 locations.
4.
Waste Storage Area Floor and wall areas were scanned with alpha, beta-gamma and gamma detectors.
Total and removable contamination levels were measured.
i 5.
Parts Storage Area Floor and equipment surfaces were scanned with portable alpha, beta-gamma, and gasuna detectors for indications of elevated activity.
Exposure Rate Measurements Camma exposure rates at 1 m above the floor were measured at seven locations in the facility, using a pressurized ionization chamber (Figure 6).
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Drain Sampling t
Camma and beta-gamma scanning, using N4I(TI) and pancake G-M detectors respectively, was performed at two drain sampling locations in the
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decontamination area and at one sump location (Figure 7).
The detectors were lowered into the uncovered drain openings for indications of elevated activity.
Water samples were collected from both drains.
One residue sample was collected f rom one of the drains using a towelette attached to a plumber's "snake."
Residue was also collected from a sump in grid block C42.
B.
Outside Areas Surface Measurements 1.
Transportation Routes Walkover surface scans, using gamma scintillation detectors, were performed at transportation entrances into the facility where i
equipment and parts were teceived.
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2.
Casings Storage Area i
Alpha, and beta-gamma scanning was performed on accessible portions of a concrete pad (Figure 4) where centrifuge casings and contaminated building materials were stored.
Total contamination levels were measured at representative locations.
Soil Sampling Two soil samples were collected adjacent to the er,ncrete pad (Figure 8).
Sample Analysis and Interpretation of Data Smears were counted to determine gross alpha and beta activity.
Water and residue samples were counted for gross alpha and beta levels.
Soil samples were analyzed by gamma spectrometry for uranium-238 and any other 5
identifiable photopeaks. Major analytical equipment used for this survey is listed in Appendix A.
Appendix B contains a description of the measurement and analytical procedures applicable to this survey.
Results were compared with guidelines established by the Nuclear 2
Regulatory Commission, for relense of facilities for unrestricted use. These guidelines are presented in Apperdix C.
Total uranium surface contamination 2
2 limits are 15,000 alpha dp /100 cm maximum and 5,000 alpha dpm/100 cm when 2
averaged over an area of 1
m.
The guideline for removable alpha 2
contamination levels for uranium is 1,000 dpm/100 cm.*
The guideline level for residual uranium contamination in soil, established by the NRC for this site, is a total of 35 pCi/g for all uranium isotopes.
Water results were compared to gross alpha (15 pCi/1) and gross beta (50 pCi/1) guideline values established by the Environmental Protection Agency (EPA) for comunity drinking water systems.3
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RESULTS l
l Indoor Areat Surface Scans Alpha and beta-gamma scanning of buildind surfaces identified isolated and general areas of elevated floor activity limited to the decontamination and UF6 cylinder storage areas.
Increased gamma radiation levels were also identified by the walkover scan at isolated locations in the decontamination area.
Surface Contamination Levels Table I summarizes the results of surface contamination measurements performed in the facility.
Isolated and general areas of contamination were
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found in the main proceu area (High Bay) and are described in detail below.
Table 2 presents the results of measurements taken in these areas prior to, ar.d is11owing cleanup activities. Each of the individual rooms, laboratories, and of fice areas surveyed were f ree of contamination.
Measurements taken in the service m ezanine, around a sealed HEPA filter exhaust, showed no elevated 6
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activity.
In addition, surveys conducted in the cent ri.fuge testing area (pit), the "Low Bay" area, containing the column cut up and power hacksaw areas, and in the waste storage area indicated alpha and beta-gamma levels well below the release criteria.
1.
Decontamination Area Highest levels of contamination were located in grid block E48, E50. E52, F50, and G50, (Figure 9).
Site personnel indicattsd that a drum containing contaminated waste water had been acciddntally spilled in this area during earlier decontamination efforts.
Eleveted activity in grid blocks F50 and G50 was associated with the impression of a barrel on the concrete.
Single point measurements taken throughout the decontamination area identified numerous locations of elevated activity.
In particular, contamination was identified around a support I-beam (J52 block) and in an isolated location adjacent to the southeast corner of a sink in grid block J54.
Maximum alpha and beta-gamma levels measured around the 2
2 beam were 16700 dpr/100 cm and 113,000 dpm/100 cm, respectively.
Visual inspection of the area around the beam identified cracks in the concrete; elevated alpha, beta-gamma and gamma levels were noted at these 2
locations.
Near the sink, alpha levels of 13900 dpia/100 cm and 2
beta-gamma levels of 196,000 dpm/100 cm were found.
Elevated activity was noted at an isolated location on an outer shower wall in grid block K54.
Maximum alpha anu beta-gamma levels were 2
2 34'30 dpm/100 cm and 7960 dpm/100 cm, respectively.
I 2.
UF6 Cylinder Storage Area i
l Although UF6 cylinders were also stored in the decontamination area, the small area specifically designated for cylinder storage (Figure 3) was considered separately for the purposes of this survey.
This cylinder storage area included grid bloens H54 and 154.
Elevated levels of alpha 2
and beta gacna contamination, ranging to 31,000 and 190,000 dpm/100 cm,
respectively, were noted in these grid blocks.
The highest levels were 7
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f recorded over a small area of residual uranium dust.
Removable contamination at this location was also significantly elevated.
Exposure Rates
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Exposure rate measurements, taken at representative locations in the facility, ranged from 8 to 9.5 WR/h. These exposure rates are consistent with normal background levels.
Radionuclide Concentrations in Drain Samples A water sample, collected from a drain in the decontamination area grid block F54 contained gross alpha levels of 11.6 1 4.4 pCi/1; gross beta l
1evels were 40.6 1 5.7 pCi/1.
No detectable activity was found on the towelette, used to collect residue from this drain.
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1 Water collected from the sink drain (grid block K54) cuntained gruss j
alpha and gross beta concentrations of 94.2 2 6.7 pCi/1 and 81.3 1 4.6 pCi/1, respectively.
No elevated radiation lovels were detected by scanning of the sump and the residue sample f rom the sump contained no detectable activity.
Outside Areas I
1 Surface Measurements No' locations of elevated direct radiation levels were identified by the gamma scans of the main transportation routes into the facility.
Alpha a nti beta-gaenna measurements on the casings storage area pad, also did not identify residual contamination.
Soil Samples t
Soil samples collected adjacent to the concrete storage pad contained l
uranium concentrations of 0.82 1 0.79 and 0.38 1 0.94 pCi/g.
These concentrations are in the range of normal baseline levels.
8
C0HPARIS0N OF RESULTS WITH GUIDELINES The survey findings indicated total residual contamination exceeding NRC guidelines at grid block locations E50 E52, G50, F48, H54, 154, J52, and J54 Levels measured in blocks E48 and I'50 were near, although below, the 2
5000 alpha dpm/100 cm guideline.
The licensee performed surface cleaning of these areas; however, due to the relative inef fectiveness of these ef forts the J
licensee chose to completely remove portions of the concrete flooring from grid blocks E50 E52, F50, 152, 154, J52, J54, K52, and K54 (Figure 10).
Debris was temporarily placed on the casings storage pad, outside the hanger building and later sent for disposal.
A sink in the K54 grid block was also removed.
i Followup measurements were performed on June 19, 1986.
The licensee's cleanup eliminated the areas of elevated alpha, beta-gamma, and gamma activity l
noted by ORAU around the I-beam in grid block J52 and adjacent to the sink l
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(J54).
Surface cleanup in the H54 and 154 grid blocks (UF6 cylinder storage area) removed residual alpha and beta activity as verified by surface scanning. Levelt of contamination were remeasured on remaining floor surf aces and found to be within guidelines (Table 2).
Soil samples were collected from the area exposed by concrete removal (Figure 11).
Levels of uranium in these samples, presented in Table 3, are within the NRC criterion for this site.
l The highest concentration of U-238 in these samples was 9.24 pCi/g, which, assuming a natural or very slightly enriched isotopic abundance is equivalent to about 20 pCi/g of total uranium.
Scans of an open drain, exposed by removal of the sink at grid block K54, indicated elevated beta-gamma radiation levels. Further investigations by the licensee revealed that a section of this drain, a connecting shower drain, and a small section of the main sever line (Figure
- 12) contained uranium contamination.
These drain lines were removed, and on August 28,
- 1986, additional followup surveys were performed by ORAU.
Scans indicated no residual areas of contamination and soil samples from the excavated areas (Table 4) were in the range of typical baseline concentrations.
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9
SUMMARY
At the request of the Nuclear Regulatory Commission, Region III Of fice, ORAU conducted a confirmatory radiological survey of the Goodyear Aerospace Wingfoot Lake Advanced Technology Center located in Akron, Ohio.
The survey was performed on May 13-16, 1986. The purpose of the survey was to verify the radiological status of the facility relative to release for unrestricted use.
Radiological information collected included exposure
- rates, surface contamination levels, concentrations of uranium and thorium in soil and radionuclide concentrations in water samples.
The survey identified isolated and general areas of residual contamination, concentrated in the decontamination and UF6 cylinder storage areas.
The licensee performed further decontamination of these areas, and followup surveys by ORAU in June and August 1986 confirmed that cleanup had been effective.
Based on the final results of the survey, it is ORAU's opinion that the I
Goodyear Aerospace Wingfoot 1.ake Advanced Technology Center has been remediated to the existing NRC guidelines and therefore satisfies the l
requirements for release for unrestricted use by the general public.
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a AKRON MAIN PLANT ADVA86ED TECHNOLOGY M
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Goodyear Aerospace Corporation Wingfoot Lake Advanced Technology Center i
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TABLE I SiseWtv 0F SURFACE CONTApsim4 TION LEVELS 8EASURED in THE elesGFAT FACILITY GOODYEAft AEROSPACE CDIPURA710N AnRON, Otl0 Location
- h aber of Total Omteminetton stemoweble Omteminetten geo. of Grid Grid Blocks Alphe But Alphe Beta W Blocks Survoyed (den /100cm )
(drm/100s )
(den /I
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(dyn/100ca')
ExceedIag Avn. 8tenge abr. 54enge Avg. Range shr. 8tenge Orlierte Hlgh Boy insin Process Aree)
Floor 47
<28-10100
<28-19200 4650-22300 4 50-98000
<2-1400
<S-1570 3
Lower teolls 5
<28-780
<28-3430 655-2250 680-7960
<2-57
<S-89 0
b
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<49-70
<650-900 (2-3
<S 0
Upper teolls 12 6
<49-90
<6M
<2
<S-7 0
Equipment 13 l
cu High Boy (Pit) b
<43
<S30
<2
<S-6 0
Floor S
b
<49-130
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REFERENCES 1.
Goodyear Aerospace Corporation. Termination of NRC Source Material and Special Nuclear Material License No. SNM-1461 for _ Goodyear /.eruspace Corporation Advanced Technology Center (ATC), January 16, 1986.
2.
U.S.
Nuclear Regulatory Commission.
Policy and Guidance Directive FC 83-3:
Standard Review Plan (SRP) for Termination of Special Nuclear f
Material Licenses of Fuel Cycle Facilities, March, 1903.
3.
Title 40, Code of Federal Regulations, Part 141, Interim Primary Drinking Water Standards, Federal Register, July 1976.
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APPENDIX A MAJOR ANALYTICAL EQUIPMENT
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APPENDIX A Major Analytical Equipment The display or description of a specific product is not to be construed that product or its manufacturer by the authors or their as an endorsement of employer.
A.
Direct Radiation Measurements
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Eberline "RASCAL" Portable Ratemeter-Scaler
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Model PRS-1 I
(Eberline, Sante Fe, NH)
Eberline PRM-6 f
Portable Ratemeter (Eberline, Sante Fe, NM) f Ludium Alpha Floor Monitor Model 239-1 (Ludium, Sweetwater, TX)
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Eberline Alpha Scintillation Probe Model AC-3-7 (Eberline, Sante Fe, NM)
Eberline Beta-Gemma "Pencake" Probe Model HP-260 (Eberline, Sante Fe, NM) 2 Victoreen Beta-Gamma "Pancake" Probe Model 489-110
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(Victoreen, Inc., Cleveland, OH)
Reuter-Stokes Pressurized Ionization Chamber
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Model RSS-111 (Reuter-Stokes, Cleveland, OH)
Victoreen NaI Gamma Scintillation Probe Model 489-55 (Victoreen, Inc., Cleveland, OH)
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B.
Laboratcry Analyses Low Background Alpha-Beta Counter Model LB5110-2080 (Tennelec, Inc., Oak Ridge, TN)
A-1
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Ge(Li) Detector Model LGCC2220SD, 23% efficiency (Princeton Gamma-Tech, Princeton, NJ)
Used in conjunction with:
Lead Shield, SPG-16
( Applied Physical Technology, Smyrna, GA) f High Purity Germanium Detector Model GMX-23195-S, 23% efficiency (EC&G ORTEC, Oak Ridge, TN)
Used in conjunction with:
l Lead Shield, G-16 (Canana Products Inc., Palos Hills, IL)
I Multichannel analyzer ND-66/ND-680 System f
(Nuclear Data, Inc., Schaumburg, IL) l
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l APPENDIX B HEASUREMENT AND ANALYTICAL PROCEDURES 1
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L APPENDIX B Measurement And Analytical Procedures Alpha and Beta-gamma Measurements Measurements of total and transferable alpha radiation levels were performed using Eberline Model PRS-1 portable scaler /ratemeters with Model AC-3-7 alpha scintillation probes.
Measurements of total and transferable beta-gamma radiation levels were performed using Eberline Model PRS-1 portable
)
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scaler /ratemeters with Model HP-260 thin-window "pancake" G-H probes. Count 2
rates (cpm) were cor.ierted to disintegratio. rates (dpm/100 cm ) by dividing the net rate by the 4w efficiency and correcting for active area of the 2
2 detector. Effective window areas were 59 cm for the ZnS detectors and 15 cm l
for the G-M detectors.
Background count rates for ZnS alpha probes averaged I
approximately I cpm; the average background count rate was 41 cpm for the G-M probes.
Surface Sean I
Surface scans of grid blocks in the Wingfoot f acility were performed by passing the probes slowly over the surface.
The distance between the probe and the eurface was maintained at a minimum nominally about 1
cm.
Identification of elevated levels was based on increases in the audible signal from the recording or fndicating instrument.
Alpha scans of large surface areas on the floor of the facility were accomplished by use of a gas 2
proportional alpha floor monitor, with a 600 cm sensitive area.
The inst rument is slowly moved in a systematic pattern to cover 100% of the accessible area.
Beta-gamma scans were conducted using Victoreen pancake G-M 2
probes (15 cui ef fective area) attached to an audible ratemeter. Combinations of detectors Snd instruments for the scans were:
I B-1
. _ _ _ _ _ _ _ _ _ _ _ _ ___________j
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Beta Gamma - G-M probe with PRM-6 ratemeter.
Beta Gamma - G-M probe with "RASCAL" scaler /ratemeter.
Gamma
- NaI scintillation detector (3.2 cm x 3.8 cm crystal) with PRM-6 ratemeter.
Alpha
- ZnS probe with "RASCAL" scaler /ratemeter.
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Alpha
- Gas proportional floor monitor with PRM-6 ratemeter.
Gamma Exposure Rate Measurements Measurements of gamma exposure rates were performed using a Reuter-Stokes pressurized ionization chamber.
The chamber was placed I a above the surf ace at seven locttions throughout the Wingfoot facility.
The average of several
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readings was determined at each location.
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Removable Contamination Measurements Smears for determination of removable contamination levels were collected on numbered filter paper disks 47 nun in diameter, then placed in indiv'. dually labeled envelopes with the location and other pertinent information recorded.
The smears were counted on a low background alpha-beta counter.
Soil Sample Analysis Soil st.mples were dried, mixed, and a portion sealed in 0.5-liter Marinelli beaker.
The quantity placed in cach beaker was chosen to reproduce
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the calibrated counted geometry and ranged from 400 to 900 g of soil.
Net soils waights were determined and the samples counted using C4(Li) and intrinsic germanium detectors coupled to a Nuclear Data Model ND-680 pulse height analyzer system.
Background and Compton stripping, peak search, peak
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identification, and concentration calculations were performed using the computer capabilit.es inherent in th9 analyzer system.
Thi energy peak used
(
for determination of U-113 was t U-238 - 0.094 MeV from Th-234*
- Secular equilibrium was assumed.
B-2
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Water Sample Analysis Water samples were rough-filtered through Whatman No. 2 filter paper.
i Remaining suspended solids were removed by subsequent filtration through 0.45 um membrane filter.
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 be ta usinF a Tennelec Model LB-5110 low-background proportional counter.
Errors and Detection Limits The errors associated with the analytical data presented in the tables of this report, represent the 95% (20) confidence levels for that data.
These errors were calculated based on both the gross sample count levels and the associated background count lavels.
When the net sample count was less than l
the 20 statistical deviation of the background count, the sample concentration was reported as less than the minimum detectable activity (<MDA).
This means that the radionuclide was not present, to the best of our ability to measure
~
it, utilizing the analytical techniques described in this appendix.
Because of variation in background levels, caused by other constituents in the
(
samples, the MDAs 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 As.sessment Program.
With the exception of the measurements conducted with portable gamma scintillation survey meters, ins t rument s were calibrated with NBS-traceable 7
I 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 i Jeruments included daily background and chock-source measurements to confirm.quipment operation within acceptable statistical fluctuations. The ORAU laboratory participates in the EPA and EML Quality Assurance Programs.
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APPENDIX C STANDARD REVIEW PLAN FOR TERMINATION OF SPECIAL NUCLEAR MATERIAL LICENSES
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STANDARD REVIEW PLAN FOR TERMINATION OF SPECIAL NUCLEAR MATERIAL LICENSES
- 1. Intreduction
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This Standard Review Plan (SRP has been developed to provide guidance to the staff engaged in reviewing) applications for the termination of special nuclear material licenses and the release of facilities for unrestricted
(
use.
This plan includes a discussion of NRC policy and technical review criteria for termination of a license.
This plan does not address the following issues:
[
o Onsite disposal of residual radioactive material (other than residual concentrations in soil determined acceptable for unrestricted release).
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o Contamination levels higher than those specified for release for unrestricted use.
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o Possession of discrete quantities of SNM in excess of critical mass quantities.
o Determination of Sici holdup in the facility.
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HQ technical assistance and guidance should be requested.if these issues should arise during the review.
II. Policy It is policy of NRC prior to termination of a material license to possess and use ShM that facilities and grounds shall be decontaminated to such levels so that they can be released for unrestrictec use.
111. Review Procedure Under current NRC regulations, each licensee is required to notify the
(
Commission, in writing, when the licensee decides to permanently discontinue activities involving special nuclear material.
Prior to license termination, the licensee is required to:
o Submit Form NRC-314 that describes the disposal of licensed materials.
o Conduct a final radiological survey.
o Submit a radiological survey report which describes the scope of the Survey, general procedures followed, and presents the survey results.
Appendix ! to this SRP, "Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of License for Byproduct, Source, or Special Nuclear Material," July 1982, provides additional guidance as to what should De containec in the survey report.
C-1
in evaluating an application for termination, the NRC must verify that:
o A reasonable effort has been made to decontaminate the facility to levels below those specified in Table 1, "Acceptable Surface I:ontamination Level s" of Appendix. l.
o Residual soil contamination (including any radioactive material buried onsite by the licensee in accordance with 10 CFR 20.302 or 10 CFR 20.304) shall not exceed the following concentration levels:
Soil Concentration Level Kind of Material (DCi/gm of soill for unrestricted area i)
+ U-234) with daughters present and in equilibrium l
ii) Depleted Uranium or Natural 35 l
Uranium that has been sepa-I rated from its daughters Soluble or Insoluble i
l tii) Enri:hed Uranium 30 Soluble or Insoluble, l
iv) Plutonium (Y) or (W) com-25 pounds v)
/n-241 (W) compounds 30 o Guidance for evaluating radioactivity in surface and groundwaters can be found in footnote 5 of Appendix B to 10 CFR 20 ano in IPA's National Interim Primary Drinking Water Regulations (EPA 570/9-76-003).
If contamination levels are higher than those specified above, HQ guidance should be requested prior to deciding whether those levels are acceptable for unrestricted release.
Definitiens A. Facili ty_
For the purpose of this procedure, "f acility" is defined as buildings, grounds, equipment, instruments, furniture, vehicles, scrap and appur-tenances thereto, and, if necessary, groundwater.
B. A discrete quantity is defined as measurable quantities of SNM in a liquid or solid form that can be accumulated into a single identifiable i
mass or volume.
This does not refer to SNM in the form of contamination on facilities or equipment.
C-2 a
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b C. Critical mass quantities are as specified below:
>350 g U-235, or T200 g U-233, or T200 g Pu, or g U-235 10-233 g Pu
>l for mixtures, or 350 200 200
>450 g Pu as sealed sources Review Criteria Although the NRC's review of an application for license termination centers around the final radiological survey report which the licensee submits.in support of the application, the reviewer should also' review the ope ~ rating f
history of the facility to assess the potential for residual contamination at the site.
This should include a review of the licensing files, inspection reports, prior NRC and other survey reports, if applicable, and facility incident reports.
The review of the licensee's close-out survey report should include an evaluation to assure:
o proper use of radiaton detection instruments, j
l o overall adequacy of the survey, and o that residual contamination levels in the facility are less than NRC's criteria for release of unrestricted use.
A.. Instrumentation
- 1. All instruments used in the survey should have been calibrated by qualified personnel, using accepted practices under the license.
7
- 2. Instruments should have sufficient detection ser.sitivity so that the
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measured data can be used to verify comoliance with acceptable con-tamination levels.
Guicance concerning the detaction sensitivity fer different types of radiation detection instruments is included in Chapter i of NUREG/CR-2082 and NCRP Report No. 50. Environ = ental Radiation Measurements.'
B. Scoce of Surveys All indoor areas (such as floors, walls, ceilings) of the building and outdoor areas (such as roofs, ground area, etc.) should be surveyed for f
radiation contamination levels and reported in the prcper itslin the applicant's survey report.
Prior to surveying the facility it should be divided into specific areas suitable for surveying. Guidance concerning the
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choice of grid sizes and the total sample size required can be found in Cnapter 3 of NUREG/CR-2082.
C-3
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- 1. For Indoor Areas in each survey block which is formed by the grid, the folicwing set of neasurements'should'be conducted and reported:
a), Direct readings for al,oha 'and beta-gamma:
Theaverageandthemaximumcontgminationlevelsatthesurface l
should be reported in dpn/100 cm for the alpha co0nting mode; and in dpm/100 cm' and urads/hr for the beta-gamma counting mode, b) Smear testing for determining alpha and beta-gamma repvable contamination levels should.be reported in dpm/1,00 cm
- 2. For Outdoor Areas l
For the outdoor area radiation survey, the following sample neasure-ments should be conducted and reported:
l a) Direct reading for beta-gama measurement at 1 cm above the surface ano for gamma measurement at 1 meter above the surface
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at eacn gr1e point.
l Both measurements should be expressed in urads/hr.
(The l
external exposure rate at 1 meter above the surface should i
be less than 10 urads/hr above the background level.)
b) Surface _ soil samoles (0-5 cm):
The average soil concentration of radioactivity in the facility may be measured by either taking systematic soil samples at all blocks of the grid system or by taking randomly selected samples l
for an unb!ased estimate.
Guidance concerning the number of samples required for an unbiased estimate may be found in Chapter 3 of NUREG/CR-2082.
If there is any reason to suspect (such as frcm the site record indicating any radioactive spill incident or any elevated external exposure level, etc.) that certain discrete areas may contain extra-l ordinary contamination, soil sarcles shoulc te collected f rom these areas.
The radioactivity in all soil sar ples shall be reported in picoeuries per gram of dry soil, c) Subsurface soil samples:
Subsurface soil sample measurements are required if there is any retson to suspect that subsurface contanination existi in the outdoor area or under the building.
Standard core sampling tqchniques may be used in the suspected contaminated area to asse:s the subsurface soil contanination as a function of depth.
The existence of any of the following conditions may require analysis of subsdrface samples:
(1)
Record showing that radioactive caterial nas been buried at that area.
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(ii)
Radioactive material (such as dry or liquid wastes) ha'd been stored in the areas, underground, or in a pond.
(iii) Any unexplainable, elevated.. direct survey reading in the area.
(iv)
Creeks, streams or undergrcund transfer pipes were used I
as a pathway for contaminated liquid ef fluent release.
d)Watersamoles:
Samples shall be taken from each source of potable water, surface water, and groundwater on the site, including water found in core holes drilled for subsurface soil samples.
l Additional onsite and of fsite groundwater samples may be required if there is any reason to suspect that subsurface contamination I.
exists.
The result of measured water samples shall be reported in pCi/1.
l Sediment samples from streams or ponds into which liquid effluents are released shall be sampled to measure the undissolved radio-nuclides in the liquid effluents.
The. sample result shall be l
l expressed in pCi/gm of dry weight.
i If after rev. ewing the site history and the' applicant s final radiological survey report the reviewer determines that the residual contamination.
' levels in the facility meets NRC's c'1teria for unrestricted release; the reviewer should perform or have perf ormed a c:nfirmatory survey of the facility to verify the licensee's close-out rurvey.
The results of the confirmatory survey should be comcared to the close-out survey to determine that either:
(1) the facility has been decerttaminated to levels accept-able for unrestricted release and therefore, the f acility may be released for unrestricted use, or (2) that additional specified decontamination is l
required.
If the confirmatory survey indicates that further decontamination is required the reviewer should so inferm the licensee and request that additional decon-l tamination be performed so that the f acility meets levels acceptable for unre.
stricted release.
If the confirmatory survey indicates that the facilities are acceptable for unrestricted release, the reviewer should prepare a safety evaluation report (SER) to support the termination action.
l The SER should clearly reveal the extent of the NRC review and technical basis for the licensing actions.
The format and content of the SER should I
be as follows:
o Background Discussion of the history of use of the facility including a description of the kinos and amount of radioactive material that were used in the I
facility and the type of chemical and physical processing on the radio-active material.
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o Discussion Discuss the instruments and methods of survey used in.the.. licensee's survey of the facility.
The licensee's measured residual contamination for each area and how it compares with NRC's criteria for unrestricted release.
o Ccnfirmatory Survey
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Discuss the results of the confirmatory survey and how it compares with the close-out survey.
o Conclusion Based on the findings in the discussion section of the SER, the conclusion f
may be drawn that the release of the facility for' unrestricted use repre-sents an insignificant risk to the public health and safety and to the environment and therefore, the reviewer may recommend that~ the facility
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be released for unrestricted use and the license be terminated.
After the SER is completed, the reviewer will prepare a license termination
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letter for transmittal to the licensee.
The letter, SER, licensee's survey l
l report and NRC's confirmatory, survey report forms a "license package," to l
support the licensing action.
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The termination. letter'is issued to.the licensee and filed in the docket room with the "license package."
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An example of a license termination letter and supporting SER is attached as Appendix 11.
Ouestions and Answer Section
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This section presents more information about the details of managing a licensing case than was covered in the preceeding sections of this review plan.
It is presented in a question and answer format for ease of reference.
Q. During the decontamination / decommissioning phase of the licensee's
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cperation, the licensee may request amendments to the license.
What dces the reviewer need to do to process those amenenents?
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A. In general, the amendment applications are minor in nature and do not
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require sophisticated technical analysis to support or deny the amend.
ment request.
The types of license amendments that may be involved are as follows:
- 1. possession limit change.
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- 2. Modificattien and/or deletion of one or more of the authorized activities.
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- 3. Mcdification and/or elimination of license conditions associateo with process support systems (i.e., ventilation requirements).
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- 4. Modification and/or elimination of crit.icality control requirements.
- 5. provision f.or interim storage of contaminated ;naterial and/or
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equipment.
- 6. Modification and/or elimination of survey / monitoring requirements as t
related to safety and/or environmental issues that are normally l
associated with an operating facility but not during or near the completion of the D&D phase.
Prior to performing a technical review, the reviewer must assu e that the administrative requirements are mst.
This includes docketing, fee assess-ment and payment, and distribution.
The detait s of this should be checked
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with the appropriate administrativo group in che Regions.
The technical review requires an evaluation of the amendment requtst to
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determine how it affects process safety, rt.diological santty, and environmental impact.
This evaluation shoald be documented in the.'orm 1
of a SER similar to the format presented in Section V of this review l
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plan.
Appendix III provides sevelal SERs vritten fer.:M::es tygas i
of amemdments.
These SERs should also shot.1d tilu:trate the type of issues that were considered in the review of various types.of acc :d.cnt.pplications, f
- 0. A licensee may request a portion of the facility to,be released for unrestricted use.
What requirements does the licensee' need to meet prior to approval of this request and how should the reviewer handle that request?
A. Upon completion of the decentamination of the portion of the facility that is desired to be released for unr estricted use, the licensee shall submit
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a report that assesses the results of the decommissioning activities and che enviror.rwntal impacts of iny ruidual contaminctio..
The raport shall include final contamination survey das for the portion of the facility
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under consideratien ai.4 gruedt that prov ide the basis for unrestricted release.
The reviewer Hil review the request applying the same review criteria
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where applicable as presented in Section V of this review plan.
The parts that would not be, applicable are outdoor survey requirements.
All other segments of the orocedures should apply.
In addition to these review requirements, the reviewer must determine that ongoing decontamination activities or storage of materials in other areas
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will not change the final survey status of the structure (s) to be released.
The licensee must demonstrate positive controls in this regard.
If not the request should not be approved.
Depending on the specifics of the license, the reviewer may be required to issue a license amendment.
If so, the procedures presented in the first part of Section VI are applicable.
If the license amendment is not required, the
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review and documentation of the result! of the review is still recuired.
An exanple of both cases are presented in Apoendix IV.
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O. A licensee may request guidance on the radiological surveys that need to be perforced. What guidance should the reviewer provide?
A. The reviewer should provide the licensee with the following two documents I
as the guidance of the radiological survey!
- 1. NRC's "Guidelines for Decontamination of Facilities,and Equt; ment Prior to Release for Unrestricted Use of Termination.of License for Byproduct, Source, Speci al Nucl ear Materi al,". July 1982.
- 2. Monitoring for C'mpliance with Decommissioning Termination Survey o
Criteria, NUREG/CR-2082.
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O. What is the general interpretation of the acceptable surface contamination levels specifled in Table 1 cf NRC's "Guidelines for Decontamination of i
Facilities and Equipment Prior to Release for Unrestricted Use?"
1 A. In addressing this issue, one has to apply the ALARA concept. The acceptable I
surface contamination levels presented.in Table 1 are minimum goals the licensee should attempt to achieve.
In applying the ALARA concept, it is not enough for the licensee to just meet the defined levels but to do the best job he can to reduce the contamination levels as low as reasonably achievable. However, a case may arise where the licensee cannot meet the defined levels presented but.has applied.the ALARA principle.
Such cases should be referred to HQ for technical assistance.
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GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQUIPMENT PRIOR TO RELEASE FOR UNRESTRICTED USE j
OR TERMINATION OF LICENSES FOR BYPRODUCT, SOURCE
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U.S. Nuclear Regulatory Cotunission l
Division of Fuel Cycle & Materia? Safety Washington, D.C.
20555 July 1982 l
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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 1 prior to the application of the covering.
A reasonable effort must be made to minimize the contamination prior to use of any covering.
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3.
The radioactivity on the interior surfaces of pipes, drain lines, or ductwork shall be determined by making measurements at all traps, and other apprnpriate access points, provided that contamination at these locations is likely to be representative of contamination on the
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interior of the pipes, drain line s,' or ductwork.
Surfaces or premises, equipment, or scrap which are likely to be contaminated but i
are of such size, construction, or location as to make the surface l
innecessible for purposes of measurement shall be presumed to be contaminated in excess of the limits.
4.
Ilpon 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 facilities to a long-term storage or standby status.
Such requests must a.
Provide detailed, specific information describing the premises, l
equipment or scrap, radioactive contaminants, and the nature, extent, and degree of residual surf ace 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
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the survey report shall be flied 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 a.
Identify the premises, b.
Show that reasonable af fort has been made to eliminate residual contamination.
c.
Describe the scope of the survey and general procedures followed.
I 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.
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TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS a
Averageb,c,f Maximum,d,f Removable,e.f b
b Nuclides 2
2 2
U-nat, U-235 U-238, and 5,000 dpa a/100 cm 15,000 dpm a/100 cm 1,000 dpm a/100 cm associated decay products 2
2 2
Transuranics, Ra-226, Ra-228, 100 dpm/100 cm 300 (pm/100 cm 20 dpm/100 cm Th-230, Th-228, Pa-231, Ac-227, I-125, I-129 2
2 2
Th-nat, Th-232, Sr-90, Ra-223 1000 dpm/100 cm 3000 dpm/100 cm 200 dpm/100 cm Ra-224, U-232, 1-126, 1 131, I-133 2
2 2
Beta gamma emitters (nuclides 5000 dpm Sy/100 cm 15,000 dpm Sy/100 cm 1000 dpm Sy/100 cm with decay modes other than alpha emission or spontaneous 7
fission) except Sr-90 and
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others noted above.
a Where surface contamination by both alpha-and beta-gamma-emitting nuclides exists, the limits established for alpha-and beta gamma-emitting nuclides should apply independently.
b 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, ef ficiency, and geometric factors associated with the instrumentation.
c Measurements of average contaminant should not be averaged over more than I square meter. For objects of less surf ace area, the average should be derived for each such object.
2 d The maximum contamination level applies to an area of ngt more than 100 cm,
e The amount of removable radioactive material per 100 cm cf surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radiosctive material on the wipe with an appropriate lastrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
f The average and maximum radiation levels associsted 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, recpectively, measured through not more than 7 milligrams per square centimeter of total absorber.
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