ML20205M805

Confirmatory Radiological Survey of Wingfoot Lake Advanced Technology Ctr,Goodyear Aerospace Corp,Akron,Oh, Draft Rept.Record of 860821 Telcon Encl

ML20205M805
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Site:	07001489
Issue date:	09/30/1986
From:	Boerner A OAK RIDGE ASSOCIATED UNIVERSITIES
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
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ML20205M785	List: ML20205M780 ML20205M790 ML20205M794 ML20205M801 ML20205M805 ML20205M807 ML20205M809 ML20205M811 ML20205M912 ML20205M921
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CON-FIN-A-9076-3 NUDOCS 8811030294
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CONFIRMATORY RADIOLOGICAL SURVEY rs es Prepared for OF THE U.S. Nuclear Regulatory commission's WINGFOOT LAKE Region ill Office Supponed by ADVANCED TECHNOLOGY CENTER Safeguards and 7,'a'n';. "' '"'

GOODYEAR AEROSPACE CORPORATION n

Division of Inspection Programs:

AKRON, OHIO Office of Inspection and Enforcement A.J.BOERNER

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l Radiological Site Assessment Program Manpower Education, Research, and Training Division

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DRAFT REPORT SEPTEMBER 1986

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DRAFT CONFIRMATORY RADIOLOGICAL SURVEY OF THE WINGF00T LAKE ADVANCED TECHNOLOGY CENTER C00DYEAR AEROSPACE CORPORATION AKRON, OHIO Prepared by A.J. B0ERNER Radiological Site Assessment Program Manpower Education, Research, and Training Division Oak Ridge Associated Universities Oak Ridge, Tennessee 37831-0117 Project Staff J.D. Berger A.S. Masvidal R.D. Condra R.C. Rookard M.R. Dunsmore C.F. Weaver M.J. Laudeman Prepared for Safeguards and Haterials Programs Branch Division of Inspection Programs U.S. Nuclear Regulatory Commission Region III Office DRAFT REPORT September 1986 This report is based on work performed under Interagency Agreement DOE No.

40-816-83 NRC Fin.

No.

A-9076-3 be t wee n the U.S.

Nuclear Regulatory Commission and the U.S. De pa r t me n t of Energy.

Oak Ridge Associated Universities pe r f o r ms complementary work under contract number DE-AC05-760R00031 with the U.S. Department of Energy.

This draft report has not been given full review and patent clearance, and the dissemination of its information is only for of ficial use.

No telease to the public shall be made without the approval of the Of fice of Information Services, Oak Ridge Associated Universities.

i DRAFT TABLE OF CONTENTS

.T E*K*

11 List of Figures.

List of Tables

.iii

.i

-Introduction 1

i I

Site Description L

Survey Procedures....................

2 Results..

6 Comparison of Results with Guidelines..................

9 Summary................................. 10 References 29 Appendices j

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Appendix At Major Analytical Equipment Appendix B: Measurement and Analytical Procedures i

Appendix C: Standard Review Plan for Termination of Special Nuclear Material Licenses t

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DRAFT f

LIST OF FIGURES P,a g FIGURE 1?

Akron, Ohio Area Indicating the Location of the I

Goodyear Aerospace Corporation Wingfoot Lake Advanced Technology Center 11 FIGURE 2:

General Floor Plan of the Wingfoot Lake Advanced Techeslogy Center 12 FIGURE 3 Areas Associated with the Centrifuge Operation..

13 FIGURE 4:

Exterior View of the Wingfoot Facility Showing 14 E

Casings Storage Area FIGURE 5: Grid Systems Established for Survey Reference 15 FIGURE 6:

Locations of Exposurs Rate Measurements 16 FIGURE 7 Drain Sampling Locations 17 FIGURE 8:

Locations of Soil Samples Collected Adjacent to the Casings Storage Area 18 FIGURE 9: Areas of Contamination Identified by the Walkover Surface Scan 19 i

FIGURE 10: Areas of Concrete Removal During Remedial Action 20 l

FIGURE 11: Locations of Subsurface Soil Samples Collected I

From Excavated Areas Following Removal of f

21 Concrete Flooring l

i FIGURC 12: Location of Contaminated Drain Line Which was Removed...

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DRAFT LIST OF TABLES Page TABLE 1: Summary of Surf ace Contamination Levels Measured in the 23 Wingfoot Facility TABLE 2:

Contamination Levels Measured at Locations Identified by the Surface Scans 26 TABLE 3: Uranium-238 Concentrations in Subsurface Soil Samples Collected Following Removal of Concrete Flooring.

27 TABLE 4: Uranium-238 Concentrations in Soil Samples Collected Following Removal of Contaminated Drain Lines 28 111 k

DRAFT i

CONFIRMATORY RADIOLOGICAL SURVEY OF THE l

WINGF00T LAKE ADVANCED TECHNOLOGY CENTER i

COODYEAR AEROSPACE CORPORATION AKRON, OHIO l

INTRODUCTION From 1974 to

1985, the Goody g,-

Aerospace Corporation conducted performance testing on developmental 37, centrifuges at the company's Wingfoot Lake Advanced Technology Center f>

Nr.ron, Ohio.

The work was performed through funding by the Department of f.nergy (00E) and under Nuclear Regulatory Commission (NRC) license SNM-1461..The license authorized the use of slightly enriched (to 1%) UF. Depleted u*snium (to 0.5%) was also produced during the 6

testing process.

Following termination of the proj e ct,

Goodyear decontaminated the building and equipment used in the centrifuge operation.

Contaminated materials sere drummed and shipped of fsite.

Subsequently, the licenses ed a report (January 1986) with the NRC indicating that the facility

.atisfied the NRC guidelines for release from licensing restrictions.I At the request of the Nuclear Regulatory Commission's Region III Office, the Radiological Site Assessment Program of Oak Ridge Associated j

l j

Universities (ORAU) conducted a confirmatory survey of the Goodyerr Wingfoot l

facility. This report presents the procedures and results of that survey.

l SITE DESCRIPTION

[

l The Gocdyear Wingfoot Lake Advanced Technology Center is located approximately 13 kilometers east of Akron, Ohio on Wingfoot Laku 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

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and maintenance (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 f rom <10 meters to approximately 30 meters in portions of the main process and open hangar areas.

I

DRAFT Areas directly and indirectly involved in the centrifoge operation included:

the fabrication tower, the mass spectrometer laboratory, hood and cut off saw room, rotor and column cut up areas, power hacksr.w location, UF6 cylinder storage and decontamination areas, a "pit" consisting of five subsurface levels where the actual centrifuge testing took place, and storage areas for waste and parts (Figure 3).

Centrifuge casings and contaminated floor materials and soil were scored outside on a large concrete pad (Figure 4).

Most of the individual components and equipment used in the operation, in addition to control exhaust ventilation systems, were renoved prior to May 1986.

SURVEY PROCEDURES During the ; e riod of Hay 13-16, 1986 ORAU personnel conducted a confirmatory radiological survey of the Wingfoot 1ske Advanced Technology Center.

The purpose of the survey was to verify the ade;m cy of the licensee's final survey and confirm the radiological condition of :he facility relative to decommissioning criteria.

Obj e ct ive s The objectives of the survey were tot 1.

measure exposure rmte levels in the Wingfoot facility; 2.

measure total and transferable surface contamination les a is on floors, walls, overhead supports, piping and miscellaneous fixtures, l

ductwork, equipment and drains in the facility; and l

3.

determine radionuclide concentratians in woll and water samples.

Pteaedures A.

Indoor Areas Cridding A2mx 2 m grid pa t t e rn was establimbce on the floot (Figure 5) using the southeast corner of the building.. e the ba seline coordinate (A,0).

2

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DRAFT Alphabetical designations were increts ted from east to west; the nume rical portion of the grid was de te rmir.cd along a north to south directional.

The grid was extended to include the hood and cut-off saw room and the power hacksaw areas.

RoorA adjacent to the main processing area 1.e.,

laboratories and c fice areas, were not gridded.

Based on negative survey findings, the grid was not extended beyond the column cut up and power hacksaw areas.

However, measu rement s taken outside thc gridded area (including the waste atorage and par ts staging areas) were referenced beck to existing building features. Measurements in the "pit" area and on lower and upper walls and horizontal surfaces were referenced to the floor grid or building landmarks.

Surface Measurements 1.

Main Floor area Floor areas were scanned with alpha and beta-gamns 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, beta-gamma, and gamma scanning was performed on lower walls.

Upper wall and overhead surface scanning on

ledges, beams, piping, fixtures, equipment and ductwork was conducted using i,a n d -

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 we?e performed at the location in each grid block where the highest total meas u re me nt was obtained.

Total and removable contamination measurements were also p'

s ned on upper walls and on ledges, piping, and ungridded hor'.

21 and vertical surfaces.

3 f

DRAFT 2.

Pit Area Floor surfaces, lower walls, and equipment were scanned on each of five subsurface sections of a pit whe re centrifuge testing was conducted.

Hand-held alpha, beta-gamma, and gamma detectors were used for the scans.

Total and removable contamination levels were determined at representative locations.

3.

Laboratories. Of fice Arecs, and Service Mezzanine Alpha, beta-gamma and ganma scanning Gas performed on the floor, lower walls and other surfaces in laboratories and office areas adjaceat 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 all locations.

4.

Waste Storage Area Floor and wall areas were scanned with alpha, be ta-gamma and gamma detectors.

Total and removsble contamination levels were measured.

5.

Parts Storage Area Floor and equipment surfaces were scanned with portable alpha, beta-gamma, and gamma detectors for indications of elevated activity.

Exposure Rate Measurements Gamma exposure rates at I m above the floor were measured at seven locations in the facility, using a prescurized ionization chamber (Figure 6).

i 4

DRAFT Drain Sampling Gamma and beta-gamma scanning, using NaI(TI) and pancake G-M detectors respectively, was performed at two drain sampling locations in the 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 f rom both drains.

One residue sample was collected from 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.

Trausportation Routes l

Walkover surface scans, using gamma scintillation detectors, were performed at transportation entrances into the facility where equipment and parts were received.

2.

Casings Storage Area Alpha, and beta-gamma scanning was perforced on accessible portions of l

a concrete pad (Figure 4) where centrifuge casings and contaminated j

building materials were stored.

Total contamination levels were measured at representative locations.

Soil Sampling Two soil samples were collected adjacent to the coacrete pad (Figure 8).

Sample Analysis and Interpretation of Data Smears were counted to determine gross alpha and be t a activity.

Wa*er and residue samples were counted for gross alpha and beta levels.

Soil samples were analyzed by gaena spectrometry for uranium-238 and any other 5

DRAFT 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 Regulatory Commission, for release of f acilities for unrestricted use2 These guidelines are presented in Appendix C.

Total uranium surface contamination 2

2 limits are 15,000 alpha dpm/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 h'RC 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 community drinking water systems.3 RESULTS Indoor Areas Surface Scans Alpha and be t a-gamma scanning of building surfaces identified isolated and general areas of elevated floor activity limited to the decontamination i

and UFf cylinder storage areas.

Increased gamma radiation levels we re also l

identified by the walkover scan at isolated locations in the decontamination l

area.

1 l

Surf ace Contamination Levels Table 1 sun.ma ri ze s the results of surface contamination measurements performed in the facility.

Isolated and general areas of contamination were found in the main process area (High Bay) and are described in detail below.

Table 2 presents the results of measurements taken in these areas prior to, and following cleanup activities.

Each of the individual rooms, laboratories, and of fice areas surveyed were free of contamination.

Measurements taken in the service mezzanine, around a sealed REPA filter exhaust, showed no elevated 6

DRAFT.

l activity.

In addition, surveys conducted in the centrifuge testing area (pit), the "Low Bay" area, containing the colum cut up and power hacksaw areas, and in the waste storage area indicated alpha and beta-gamma levels well below the release criteria.

t 1.

Decontamination Area Highest levels of contamination were located in grid block E48, E50, E52, P50, and 050, (Figure 9).

Site personnel indicated that a drum i

containing contaminated waste water '. tad been accidentally spilled in this area during earlier decontamination ef forts.

Elevated activity in grid blocks F50 and 050 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, l

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 dpm/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 l

2 locations.

Near the sink, alpha levels of 13900 dpm/100 cm and 2

beta-gamma levels of 196,000 dp /100 cm were found.

Elevated activity was noted at an isolated location on an outer shower wall in grid block K54.

Maximum alpha and be ta-gamma levels were 2

2 3430 dpa/100 cm and 7960 dpm/100 cm, respectively.

l 2.

UF6 Cylinder Storage Area l

Although UF6 cylinders were also stored in the decontamination area, the small area specifically designattd for cylinder storage (Figure 3) was considered separately for the purposes of this survey.

This cylinder storage area included grid blocks H54 and 154.

Elevated levels of alpha 2

and beta-gamm contamination, ranging to 31,000 and 190,000 dpm/100 cm,

respectively, were noted in these grid blocks.

The highest levels were 7

DRAFT recorded over a snall area of residual uranium dust.

Removable contamination at this location was also significantly elevated.

Exposure Rates Exposure rate measurements, taken at representative locations in the f acility, ranged f rom 8 to 9.5 WR/h. These exposure rates are consistent with normal background levels.

Radionuclide Concentrations in Drain Samples A water sample, collected f rom a drain in the decontamination area grid block F54 contained gross alpha levels of 11.6 1 4.4 pCi/1; gross beta levels were 40.6 1 5.7 pCi/1.

No detectable activity was f ound on the towelette, used to collect residue f rom this drain.

Water collected from the sink drain (grid block K54) contained gross alpha and gross beta concentrations of 94.2 1 6.7 pC1/1 and 81.3 1 4.6 pC'./1, respectively.

No elevated rsdiation levels were detected by scanning of the sump and the residuo sample f rom the sump contained no detectable activity.

Outside Areas Surface Measurements No locations of elevated direct radiation levels were identified by the gamma scans of the main transportation routes into the facility.

Alpha and beta-gam:u measurements on the casings storage area pad, also did not identify residual contamination.

Soil Samples Soil samples collected adjacent to the concrete storage pad contained 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

DRAFT COMPARISON 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 F50 were near, although below, the 2

5000 alpha dpm/100 cm guideline.

The licensee performed surf ace cleaning of these areas; however, due to the relative ineffectiveness of these efforts the licensee chose to completely remove portiens of the concrete flooring from grid blocks ESO, 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.

Followup measurements were performed on June 19, 1986.

The licensee's cleanup eliminated the areas of elevated alpha, beta-gamma, and gamma activity noted by ORAU around the I-beam in grid block J52 and adjacent to the sink (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.

Levels of contamination were remeasured on remaining floor surf aces and found to be within guidelines (Table 2).

Soil samples were collected f rom 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.

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-gamns 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 sewer line (Figure

12) contained uranium contamination.

These drain lines were re moved, and on August 28,

1986, additional followup surveys were performed by ORAU.

Scand indicated no residual areas of contamination and soil samples from the excavated areas (Table 4) were in the range of typical be.seline concentratic,as.

9 b

DRAFT

SUMMARY

At the request of the Nuclear Regulatory Commission, Region III Office, 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 infurcation 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 decor.tamination 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.

s Based on the f f nal results of the survey, it is ORAU's opinion that the Goodyear Aerospace Wingfoot Lake Advanced Technology Center has been remediated to the existing NRC guidelines and therefore satisfies the requirements for release for unrestricted use by the general public.

10

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TABLE I SLM*45RY OF SURFACE CONTAMIMATION LEVELS MEASWED IN THE WINGTOOT FACILITY GOODYEAR AEROSPACE COWORATION ArRON, OHIO Location

%mber of Total tontamination Removable Contaminatica No. of Grid e

Grid Blocks Al pha Oste-Gamme AlphaRasp Eheta Range Blocks (dpe/IO0c2)

(dpm/100c2)

(dpm/IOOct )

(dps/100c# )

Exceeding Surveyed Avg. Range Mss. Range Avg. Range Ken Range

& lterla High Bay (Main Process Ves)

Floor 47

<28-10100

<28-19200

<650-22300

<650-98000

<2-1400

<5-1570 3

Lower walls 5

<28-780

<18-3430 655-2250 680-7960

<2-57

<5-89 0

upper useIs 12 C

449-70

<650-900

<2-3

<5 0

b Equipment 13

<49-90

<650

<2

<5-7 0

b PJ High Bay IPSf) b

<49

<530

<2

<5-6 0

F I oor 5

b

<49-150

<530

<2-3

<5-6 0

tower walls IO Shss Spectrometry Lab Floor 2

<28

<28-30

<530-F20 (530-3120

<2

<5-6 0

D

<28-70

<5 50

<2

<5-6 0

tower walls 2

fboas 1-6 b

<28-70

<530-560

<2

<5 0

Floor 6

b

<28-30

<530-870 (2-3

<5-6 0

tower Walls 6

0 23>n

-4

TAHLE I (Contineed)

SweeAR7 0F StftFACE ODNTapelseATION LEVELS MEASURED sef THE WINGFOOT FACILITY GOW) YEAR AEROSPACE COlFORATION Aset0N, OHIO Location stumber of Total Contesination Roseveble Conteminetton No. of Grid

& Id 81ochs Alphe Bete-Ca===

AIpine Menge 8sta Range IMochs Serveyee (dpe/100cm )

(den /100cm )

(ope /100cm )

(dpe/IO0cm )

Exceeding Ave. Renes seen. Range Ave Renes seen. Renes

&lterle Hood / Cut Of f See floon Floor 2

98-140 210-350 570-580 740

<2-3

<5-6 0

D

<28-30

<530

<2

<5 0

tower me1Is 2

2050 5

<3 0

1090 Hood 1

Office Arees b

Floor 20

<28

<650-830

<2-5

<5-8 0

Service senzrentne D

<650-1550

<2-3

<5-7 0

<28-70 floor 3

Low Boy Aree Floor 4

29-110 30-230

<650-860

<650-1550

<2-7

<5-7 0

Lower me1Is 2

33 50

<650-670

<650-760

<2

<5 0

D

<650

<2

<5 0

50 Upper meiIs S

D

<49-50

<650

<2-7

<5-10 0

Equipment 3

0

D>

'11 H

f

\\

TABLE I (CDatinued)

SiseewtY OF SURFACE CouTAptmAT10N LEVELS MEASURED IN THE WINGFOOT FACILITY G0mWEMt AEROSPACF. C0sPORATION Asm04, OHIO Location shenber of Total Contaminetton masonable conte Instion leo. of Grid Grid Blocks Alphe Befeh Alphe Range Bote Range Blocks serveyed (ope /looc=2, g,,,f,,,c,23 g,,,f,gac,z)

(ope /800c=2, g,c,,,,,g Ave. mange seen. Range Ave. Renee se. Range

&lterte weste storage gree Floor 2

32-38 S0-90

<%0 560-610

<2-7

<S-6 0

Lo.or lasiIs I

<28,

30

<%0

<560

<2

m H

f

_~

]

TAel.E 2 CONT 4petNATION LEVELS seEA5(stED AT LOCATIONS IDENTIFIED BY THE StstFACE SCAfe5 GOODYEAR AEROSPACE Cofr0 RAT 10m AMR0es, OHIO b

D Roon/ pres Location Grid Surface BEFORE DE00NT4petNATIOu AFTER DECONTApetNATION a

Identification Alp u Beta-Gemme Alpha Bota-Gemme (den /300c2) (dem/100cb (den /100c2) (ope /300M) 0.cnntami net ton 1

E48 Floor 1610 4550 140 935 Area 2

E50 Floor 7100 18900 12dC 770c 3

ES2 Floor 10100 17200 120 1820 4

F50 Floor (Borrel Rlag) 1960 4030 405 268 6 c

5 G50 Floor (Berrel Ring) 2290 22300 78 1200 6

J52 Floor (8-Boan)

Il90d8 10400d C

C 8

c

_c 7

J52 Floor (1-Beam) 1670d8 II300d 8

J52 Floor (1-Bosal 1650d 918 @

C c

9 J54 F1oor 1390d Id C

C UF CVIInder Storage 1

66 4 Floor 347d 2600M 6

Area 2

66 4 Floor 3100d 19000d8 3

154 Floor 1360d 1950 8 stein Floor i

F48 Floor IM 1800-4500 GuideiIne 5000 teamImise l*JX10 Average f

• Refer to Figure 9.

bFive point moesurement unless otherwise Indicated.

CConcrete removed - refer to Table 3 for soll sagling results, done point asesorament only.

• No locations of elevated activity noted by surf ace scans following cleenop.

D>

m

~4

TAnLE 3 Ufumettse-23e CouCENTRATIows Im SUBSURFACE S0lt SAsrLES COLLECTED FOLL0mleG RDEDWA1. OF CouCRETE FLOORING GOODYEAR AEROSPACE C0fr0 RATION AKRON, OHIO b

Grid identification Location

• Depth U-238 RadionucIlde Coocentrations (pCI/g)

(ce)

C E*0.5, 50+2 1

0-15 9.24 1 2.46 F+2, 50*3 2

0-15 0.4610.73 K+0.7, 54+0.5 3

0-15 1.22 1 0.30 K+0.I, 52+0.2 4

0-15 0.00 i 1.01 I+ 1.5, 54+0.8 5

0-15 1.83 1 0.79 J+0.5, 53+0 6

0-13 2.37 1 1.09 I+3.5, 52+0 7

0-15 1.22 1 0.53 U

(;eldellae 35 (Total trantia) sRefer to Figure 11 bDepth below concrete /soll Interface.

"Errors are 23 based on counting statistics.

0

D>

m 1

-4 f

I I

J

TABLE 4 URANitM-238 CONCENTRATIONS IN SOf L SAWLES COLLECsED FOLLOslNG REMOVAL OF (INTAMahATED ORAIN LINES COJDYEAR AEROSPACE CORPORATION APRON, OHIO Grid identification *

()-238 Radionuclide Concentrations (pCl/g) 1+1, 50+l,1 0.6310.63*

1+1.3 SO+l.2 1.12 1 0.99 J+0.e, 30+1.1

<0.49 J+1.2, S0+1.2 1.26 1 0.84 J*1.3, 50+t.4 0.43 1 0.73 K+1.4, 50+1.1 1.05 i 1.10 K+0.3, 52+0.5 0,84 1 0.71 K+1.8, S2+1.4

<0.57 East End of Excavation

<0.40 Center of Escavation 0.57 1 0.S2 west End of Excavation 0.99 i 0.45 Guideline 35 (Total Uranium)

"Errors are 25 tmsed on counting statestics.

O 32 3>

m

-4

DRAFT REFERENCES 1.

Goocyear Acrospace Corporation. Termination of NRC Source Material and Special Nuclear Material License No.

SNM-1461 for Goodyear Aerospace Corporation Advanced Technology Center ( ATC), January 16, 1986.

2.

U.S.

Nu c le.s r Regulatory Commission.

Policy and Guidance Directive FC 83-3:

Standard Review Plan (SRP) f or Termination of Special Nuclear Material Licenses of Fuel Cycle Facilities, March, 1983.

.1.

Title 40, Code of Federal Regulations, Part 141 Interim Primary Drinking Water Standards. Federal Register, July 1976.

l 29

DRAFT APPENDIX A MAJOR ANALYTICAL EQUIPMENT I

.r_-.-_.

6 DRAFT APPENDIX A Ma;;or Analytical Equipment The display or description of a specific product is - not to be construed

. as an endorsement of that product or its manuf acturer by the authorn 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 l

Model 489-110 (Victoreen, Inc., Cleveland, OH)

Reuter-Stokes Pressurized Ionization Chamber i

Model RSS-ill (Reuter-Stokes, Cleveland, O!!)

Victoreen Na1 Gamma Scintillation Probe Model 489-55 (Victoreen, Inc., Cleveland, OtO l

B.

Laboratory Analyses Lew Background Alpha-Beta Cov.ter Model LB5110. 2080 (Tennelec, Inc., Oak Ridge, !!'.

l l

A-1

DRAFT Ce(Li) Detector Model LGCC2220SD, 23% efficiency (Princeton Gamma-Tech, Princeton, NJ)

Used in conjunction with:

Lead Shield, SPG-16 (Applied Physical Technology, Smyrna, GA)

High Purity Germanium Detector Model GMX-23195-S, 23% efficiency g-(EG6G ORTEC, Oak Ridge, TN) s Used in conjunction with:

l Lead Shield, G-16 l

(Gamma Products Inc., Palos Hills, IL) i l

Multichannel analyzer g

ND-66/ND-680 System I-(Nuclear Data, Inc., Schaumburg, IL) 4 1

k I

l I

l A-2

DRAFT i

APPENDIX B MEASURF. MENT AND ANALYTICAL PROCEDURES I

l l

l l

\\

DRAFT I

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 scalertratemeters with Model HP-260 thin-window "pancake" G-M probes. Count 2

rates (cpm) were converted to disintegration rates (dpm/100 cm ) by dividing the net rate by the 4x efficiency and correcting for active area of the 2

E l

detector. Effective window areas were 59 cm for the ZnS detectors and 15 cm for the G-M detectors.

Background count rates for ZnS alpha probes averaged l

approximately 1 cpm; the average background count rate was 41 epm for the G-M probes.

Surface Sean 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 nominally about I

cm.

and the eurface was maintained at a minimum Identification of clevated levels was based on increases in the audible signal from the recording or indicating 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 instrument is slowly moved in a systematic pattern to cover 100% of the accessible area.

Beta-gacna scans were conducted using Victorcen pancake G-M 2

probes (15 cm ef fective area) attached to en audible ratemeter. Combinations of detectors and instruments for the scans were:

l l

B-1

DRAFT Beta Gamma - G-M probe with PRM-6 ratemeter.

Beta Gamma - G-M probe with "RASCAL" scaler /ratemeter.

Gamma

- Na1 scinttilation detector (3.2 cm v 3.8 cm crystal) with PRM-6 ratemeter.

Alpha

- ZnS probe with "RASCAL" scaler /ratemeter.

Alpha

- Gas proportional floor monitor with PRM-6 ratemeter.

J' Gamma Exposure Rate Measurements Measurements of gamma exposure rates were performed using a Reuter-Stokes pressurized ionization chamber.

The chamber was placed )( 1 m above the surface at seven locations throughout the Wingfoot facility.

The average of several readings was determined at each location.

l l

l 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 envel. opes with the location and other pertinent information recorded.

The smears were counted on a low background alpha-beta counter.

Soil Sample Analysis Soil samples were dried, mixed, and a portion sealed in 0.5-liter Marinelli beaker.

The quantity placed in each beaker was chosen to reproduce the calibrated counted geometry and ranged from 400 to 900 g of soil.

Net soils weights were determined and the samples counted using Ge(Li) and intrinsic ge rmanium 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.

The energy peak'/ used

\\

for determination of U-238 0 w was:

at 1

U-238 - 0.094 tieV f rom Th-234*

l

• Secular equilibrium was assumed.

B-2

DRAFT 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 p m 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 beta using a Tennelec Model LB-5110 low-background proportional counter.

u Errors and Detection Limits l

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 levels.

When the net sample count was less than the 2b 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 dif fer f rom sample to sample.

Calibration and Quality Assurance Labo ra tory 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 ce t e r s, instruments were calibrated with NBS-traceable standards.

The calibration procedures for the portable gacnt instruments are performed by comparison with an NBS calibrated pressurized ionization chamber.

l Quality control procedures on all ins t rume n t s included daily background and check-source measurements to con'irm equipment operation within acceptable statistical fluctuations. The ORAU labnratory participates in the EPA and EML i

Quality Assurance Programs.

B-3

b APPENDIX C STANDARD REVIEW PLAN FOR TERMINATION OF SPECIAL NLCLEAR MATERIAL LICENSES h

l s

l STANDARD REVIEW PLAN FOR TERMINATION OF SPECIAL NUCLEAR MATERIAL LICENSES

1. Intreduction This Standard Review Plan (SRP) has been developed to provide guidance to the staf f engaged in reviewing applications for the termination of special nuclear material licenses and the release of facilities for unrestricted use.

This plan includes a discus 3 ion 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).

l o Contamination levels higher than those specified for release for unrestricted use.

o Possession of discrete quantities of Stim in excess of critical r, ass quantities.

o Determination of Stai holdup in the facility.

.HO technical assistance and guidance should be requested.if these issues should arise during the review.

II. Policy It is policy of NRC prior to terminatien of a material license to possess and use ShM that facilities and grounds snall 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 ter.ination, the licensee is required to:

o Submit Form NCC-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 to'.

survey, general procedures followed, and presents the survey resulth.

Appendix 1 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," Jul) 1982, provides additional guidance as to what should oe containec in tne survey report, c-?

In evaluating an application for termination, the NRC must verify that:

A reasonable ef fort has been made to decontaminate the facility to levels o

below those specified in Table 1, "Acceptable Surf ace Contamination Level s" of Appendix 1.

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 erceed the fo11nwino concentration levels:

Soil Concentration Level Kind of Material (oCi/gm of soil) for unrestricted area i)

Natural Uranium (U-238 10

+ U-234) with daughters present and in equilibrium

11) Depleted Uranium or Natural 35 Uranium that has been sepa-rated f rom its daughters Soluble or Insoluble 30 tii) Enriched Uranium Soluble or Insoluble iv) Plutonium (Y) or (W) com-25 i

pounds v)

An-241 (W) compounds 30 S Guidance for evaluating radioactivity in surface and croundwaters 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 l

for unrestricted release.

Definitions l

l A. Facility 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 mass or volume.

This does not refer to SilM in the form of contamination on f acilities or equipment.

(

l C-2

l C. Critical mass quantities are as specified below:

>350 g U-235, or T200 g U-233, or T200 g Pu, or 9 U:235 9 U-233 g Pu

>l for mixtures, or a50 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 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 i

evaluation to assure:

1 o proper use of radiaton detection instruments, 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.

2. Instruments should have suf ficient detection ser.sitivity so that the measured data can be used to verify ccmoliance with ac:eptable con-tamination levels. Guicance concerning the detection sensitivity fcr dif ferent types of radiation detection instrume.ts is included in Chapter 4 of NUREG/CR-2032 and NCRP Repor' No. 50 Environmental Radiation Measurements.

B.,5coce 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 radiation contamination levels and reported in the prcper units lin the applicant's survey report.

Prior to surveying the facility it should be divided into specific areas suitable for surveying. Guidance concerning the choice of grid sizes and the total sam;le size required can be found in Chapter 3 of NUREG/Ck-2082.

c-3

1. For Indoor Areas In each survey block which is formed by the grid, the following set of measurements'should be conducted and reported:

a) Direct readings for aloha'and beta-gamma:

The average and the maximum centgmination levels at the surface should be reported in dpm/100 cm for the alpha counting mode; and in dpm/100 cm' and urads/hr for the beta-gamma counting mode.

b) imear testing for determining alpha and beta-ganma re vable contamination levels should be reported in dpm/100 cm

2. For Outdoor Areas l

For the outdoor area radiation survey, the following sample measure-j ments should be conducted and reported:

I a) Direct reading for beta-gaerna measurement at I cm above the l

surf ace ano f or gamma measurement at I r eter aoove the surf ace at eacn grio point:

1 Both measurements should be expressed in urads/hr.

(The I

external exposure rate at 1 meter above the surface should 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 randenly selected samples fcr an unt: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-ordinary contamination, soil samcles shcule te collected from these areas.

The radioactivity in all soil samples shall be reported in picoeuries per gram of dry soil, c) Subsurface soil samples:

Subsurface soil sample measurements are required if there is any reason to suspect that subsurface contamination exists in the outdoor l

area or under the building.. Standard core sampling techniques may be used in the suspected contaminated araa to assess the subsurface soil contamination as a function of depth.

The existence of any of the following conditions.nay require analysis of subsurface samples:

l (i)

Record showing that radioactive material has been buried at that area.

C-4

(ii)

Radioactive material (such as dry or liquid wastes) had been stored in the areas, underground, or in a pond.

(iii) Any unexplainable, elevated, direct survey reading in the area.

(iv)

Creeks, streams or underground transfer pipes were used as a pathway for contaminated liquid ef fluent release, d) Water samoles:

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.

Additional onsite and offsite groundwater samples may be required if there is any reason to suspect that subsurface contamination exists.

The result of measured water samples shall be reported in pCi/1.

Seoiment 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 expressed in pCi/gm of dry weight.

If after reviewing 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 criteria for unrestricted release; the reviewer should perform or have performed a confirmatory survey of.the facility to verify the licensee's close-out survey.

The results of the confirmatory survey should be compared to the close-out survey to determine that either:

(1) the f acility has been decontaminated 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 required.

j If the confirmatory survey indicates that further decontamination is required l

the reviewer should so inform the licensee and request that additional decon-tamination be performed sc, that the f acility mee.s levels acceptacle for unre.

j 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.

The SER should clearly reveal the extent of the NRC review and technical basis for the licensing actions.

The fornat and content of the SER should 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 f acility and the type of cheatical ano pnysical processing on the radio-active material.

C-5

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 Cenfirmatory Survey 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 may be drawn that the release of the f acility 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 be released for unrestricted use and the license be terminated.

l Af ter the SER it. completed, the reviewer will prepare a license termination letter for transmittal to the licensee.

The letter, SER, licensee's survey report and NRC's confirmatory. survey report forms a "license package," to I

support the licensing action.

l l

The termination letter is issued to the licensee and filed in the docket I

room with the "license package."

An example of a license termination letter and supporting SER is attached as Appendix 11.

Questions and Answer Section 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 /deconmissioning phase of the licensee's operation, the licensee nay request amencments to the license.

What dces the reviewer need to do to process those amenenents?

A. In general, the amendment applications are minor in nature and do not 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.

2. Modificattion and/or deletion of one or more of the authorized activities.

3. Modification and/or elimination of license cenditions associatea with proces s support systems (i.e., ventilation requirements).

C-6

4. Modification and/or elimination of criticality control requirements.

5. Prov>

n for interim storage of contaminatec' material and/or equi pn.nt.

6. F. edification and/or elimination of survey / monitoring requirer.ents as related to safety and/or environmental issues that are normally 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 assure that the administrative requirements are met.

This includes docketing, fee assess-ment and payment, and distribution.

The details of this should be checked with the appropriate administrative group in the Regions.

The technical review requires an evaluation of the amendment request to determine how it affects process safety, radiological safety, and environmental impact.

This evaluation should be documented in the form of a SER similar to the format presented in Section V of this review pl an.

Appendix 111 provides several SERs written for various types of amendments.

These SERs should also should illustrate the type of issues that were considered in the review of various typer.of amendment applications.

Q. 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. Upen comoletion of the decontamination of the portion of the facility that is desired to be released for unrestricted use, the licensee shall submit a report that assesses the results of the decommissio.:ing activities and che envirormental impacts of any residual contaminctior..

The raport shal'.

include final :entamination survey data for the portion of the facility under consideration and grounds that ' provide the basis for unrestricted release.

/

The reviewer shall review the request applying the same review criteria where applicable as presented in Section V of this review plan.

The parts tnat would not be applicable are outdoor survey requirer.ents.

All l

other segments of the procedures should apply.

In addition to these review requirements, the reviewer must determine that ongoing decontamination activities or storage of materials in other areas l

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 amencment.

If so, the procedures presented in the first part i

of Section VI are applicable, if the license amencment is not requireo, the l

review ano documentation of the results of the review is still recuired.

L An example of both cases are presented in Appendix IV.

C-7

-i

0. A licensee nay request guidance on the radiological surveys that need to be perforced. What guidance should the reviewer provice?

A. The reviewer should provide the licensee with the following two documents as the guidance of the radiological surveyi

1. NRC's "Guidelines for Decontaminaticn of Facilities.and Equiprent o ier to Reir se for Unrestricted Use of Terminaticn.of Licenst for r

Byproduct, Source, Special Nuclear Material," July 1982.

2. Monitoring for Compliance with Decer.missioning Termination Survey Criteria,PNREG/CR-2082.

Q. What is the general interpretation of the acceptable surface contamination levels specified in Table 1 of NRC's "Guidelines for Decentamination of Facilities and Equipment Prior to Release for Unre.stricted Use?"

A. In addressing this issue, one has to apply the ALARA concept. The acceptable 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.

e f

l C-d

GUIDELINES FOR DECONTAMINATION OF FACILITIES Ai.'D EQUIPMENT PRIOR TO FILEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BYPRODUCT, SOURCE OR SPECIAL NUCLEAR MATERIAL p

/

1 U.S. Nuclear Regulatory Commission Division of Fuel Cycle & Material Safety Washington, D.C.

20555 t

/

July 1982 1

C-9

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 prcaises 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 t h('r une may be different. The release of such facilities or items f rom regulatory control is considered on case-by-case basis.

1.

The licensee shall make a reasonable effort to eliminate residual contaminstion.

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.

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, construction, or location as to make the surface inaccessible for purposes of reasurement 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, transf er 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, equipment or scrap, radioactive contaminants, and the nature, extent, and degree of residual surface contamination.

)

l b.

Provide a detailed health and safety analysis which reflects that l

the residual anounts of materials on surface areas, together with l

other considerations such as prospective use of the premises.

l equipeent 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-10

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 h7C Rcgional 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 offort has been made to eliminate residual e

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 confirm the survey.

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C-11

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TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS Nuclides Averaged *C'I Maximum 'd'f Removableb e,f b

a 2

2 2

U-nat. U-235, U-238, and 5,000 dpm a/100 cm 15,300 dpm a/100 cm 1,000 dpm a/100 cm associated decay products 2

2 2

Transuranics, Ra-2 26 Ra-228, 100 dpm/100 cm 300 dpm/100 cm 20 dpm/100 cm Th-230 Th-228, Pa-231, Ac-227, I-125, I-129 2

z 2

Th-nat, Th-232, Sr-90, Ra-223 1000 dpm/100 cm 3000 dpm/100 ca 200 dpm/100 cm Ra-224, U-232, 1-126, 1-131, 1-133 2

2 2

Beta-gamca emitters (nuclides 5000 dpm Sy/190 cm 15,000 dpm By/100 cm 1000 dpm By/100 cm with decay modes other than alpha emission or spontaneous fission) except Sr-90 and n1 others noted above.

I PJ a Where surface contamination by both alpha-and beta-gamma-emitting nuclides exists, the limits l

established for alpha-and beta-gamma-emitting nuclides should apply independently.

b As used in this table, dpa (disintegrations per minute) mea

• 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. F2r objects of less surface area, the average should be derived far each such object.

d The maximum contamination level applies to an area of ngt more than 100 cm2,

• The amount of removable radioactive material per 100 cm of surface area should be ostermined 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 removable conta mination 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 associated with surface contamination resulting f rom 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.

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