ML20148S165

Final Followup Confirmatory Survey of Phase I Decommissioning Former Waste Processing Facility Ga Technologies San Diego,Ca

ML20148S165
Person / Time
Site:	07000734
Issue date:	03/31/1988
From:	Cotten P OAK RIDGE ASSOCIATED UNIVERSITIES
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
Shared Package
ML20148S150	List: ML20148S165
References
CON-FIN-A-9076 ORAU-88-C-129, NUDOCS 8804180187
Download: ML20148S165 (31)

Text

OR AU 88/C-129 I

' IO l

FOLLOW-UP l

erenared by oa? Ridge Associated CONFIRMATORY SURVEY Universities Prepared for Qp U.S. Nuclear

$*"m'issI!n s I

Region V Office PHASE I DECOMMISSIONING m

Sponsored by FORMER WASTE PROCESSING FACILITY I

Division of Industrial and GA TECHNOLOGIES Medical Nuclear Safety I

SAN DIEGO, CALIFORNIA P.R.COTTEN I

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I Radiological Site Assessment Program Manpower Education, Research, and Training Division s

I FINAL REPORT MARCH 1988 I

I nt"sm 898as C

PDR 1

OR AU 88/C 129 FOLLOU-CONFIRMATORY VE OF PHASE I DECOMMis 'JNING I

FORMER UASTE PROCESSIL FACILITY GA TECHNCLOGIES SAN DIEGO, CALIFORNIA Prepared by P.R. COTTEN I

Radiological Site Assessment Program Manpower Education, Research, and Training Division Oak Ridge Associated Universities Oak Ridge, TN 37831-0117 Project Staff J.D. Rerger R.C. Rookard I

R.D. Condra C.F. Ueaver G.L. Murphy Prepared for Division of Industrial and Medical Nuclear Safety U.S. Nuclear Regulatory Commission Region V Office Final Report March 1988 I

This report is based on work performed under Interagency Agreement DOE No. 40-816-83 NRC Fir No. A-9076 between the U.S. Nuclear Regulatory Commission I

and the U.S.

Department of Energy.

Oak Ridge Associated Universities performs complementary work under contract number DE-AC05-760R00033 with the U.

S.

Department of Energy.

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TABLE OF CONTENTS Page i

List of Figures 11 List of Tables iii

.I Introduction I

Procedures I

Results 2

Sucinary.

3 References 17 I

Appendices I

Appendix A:

Major Sampling and Analytical Equipment Appendix B:

Measurement and Analytical Procedures Appendix C:

Decommissioning Guidelines for the CA Technologies Waste Processing Facility I

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I LIST OF FIGURES Page FIGURE 1: Map of San Diego Area. Indicating the Location of the I

CA Technologies Facilities.

4 FIGURE 2:

Area of GA Technologies Plant, Illustrating the Phase I Decommissioning Area.

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FIGURE 3: Waste Processing Facility Area, Indicating the Grid System W

Used for Survey Reference 6

FIGURE 4:

Locations Uhere December 1985 Survey Indicated Soil I

Concentrations Exceeding Guideline Levels.

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FIGURE 5:

Sampling Locations - Phase I Followup.

8 FIGURE 6: Locations of tieasurement and Sampling Locations Along the Canyon Floor.

9 FIGURE 7: Locations of Background !!easurements and Baseline Soil Samples from the Vicinity of GA Technologies.

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I LIST OF TABLES Page TABLE 1 A:

Background Radiation Levels 11 TABLE IB:

Baseline Radionucl J2 Concentrations in Soil.

12 TABLE 2:

Exposure Rates at Sampling Locations 13 I

TABLE 3:

Radionuclide Concentrations in Soil From Remediated Areas.

14 TABLE 4:

Radionuclide Concentrations in Soil From the Canyon Floor.

15 TABLE 5:

Radionuclide Concentrations in Composite Soil Samples.

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!I FOLLOV-UP lg CONFIRtiATORY SURVEY 3

0F PHASE I DEC0FDlISSIONING FORMER UASTE PROCESSING FACILITY jg CA TECHNOLOGIES lg SAN DIEGO, CALIFORNIA INTRODUCTION In mid 1984, CA Technologies, Inc. (CA) of San Diego, California, initiated Phase I decommissioning activities of the Former Uaste Processing Facility (Figures 1-3).

Phase I includes the Solar Evaporation Pond Area, the areas immediately surrounding the Former Uaste Processing Facility and Incinerator Pad, a previous burial site for contaminated asphalt, the hillside and canyon below the I

waste handling facilities, and undeveloped land surrounding the Uaste Processing i

Facilities.

During December 10-17, 1985 a confirmatory survey of Phase I remediation was performed by the Radiological Site Assessment Program of Oak Ridge Associated Universities (ORAU).

The survey identified 49 small isolated areas (Figure 4) of residual contamination; these areas vere primarily east and north of the Uaste Processing Facility, and in the v'.cinity of the f ormer Solar Evaporation Ponds.I lg During 1987, GA Technologies performed additional remedial actions ;o remove lE contamination identified by the December 1985 ORAU survey.

A report, prepared by GA indicates that this remedial action was effective in reducing residual contamination to within the guidelines established for the site.2 At the request of the Nuclear Regulatory Commission's Region V Office, a follovup survey of these recleaned areas was performed by ORAU during September 1987.

This report describes the procedures and results of that survey.

lI PROCEDURES l

1.

The licensee's grid system was reestablished at 30 ft (9.1 m) intervals to lI provide reference points for measurements and sampling.

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I 2.

A walkover surface gamma scan was conducted at 1-2 m intervals throughout the remediated area, using portable countrate instruments with NaI(Tl) gamma scintillation detectors and audible indicators.

A scan of the canyon d rainage area, southeast of the Waste Processing Facilities, was also performed.

I 3.

Exposure rates were measured at the surface and 1 m above the surface at seven locations (Figure 5),

where additional remedial action had been I

performed.

These locations represented those areas which were noted by the 1985 survey to have higher levels of contamination.

Measurements were also performed at four locations in the Canyon (Figure 6).

4 Surface soil samples were collected at locations of exposure rate measurements.

I 5.

Samples and data were returned to Oak Ridge, Tennessee for analyses and evaluation.

Appendices A and B contain additional information regarding I

equipment and procedures.

Results we re compared to guidelines established for decommissioning of this facility (Appendix C).

RESULTS Walkover gamma scans did not identify any locations of significantly elevated direct radiation levelr in the remediated area or along the canyon floor.

Camma exposure rates measured in these areas are presented in Table 2.

In the I

remediated area these rates ranged from 16 to 23 LR/h at surface contact and from 15 to 18 uR/h at I m above the surface.

Me a su re me nts along the canyon floor ranged from 15 to 18 uR/h at the surf ace and f rom 14 to 16 LR/h at I m above the surface.

For comparison, the background exposure rates in the vicinity of the CA Technologies f acility averages about 9.7 uR/h at 1 m above the surface (Table IA). The guideline for decommissioning requires that the average exposure rate be less than 10 uR/h above barkground, which would be a total of 19.7 LR/h.

All exposure levels measured at 1 m above the surface during this survey were less than 19.7 uR/h and therefore this guideline is satisfied.

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8 Tables 3 and 4 present the concentrations of gamma emitting radionuclides, measured in surface soil collected from the remediated and canyon floor areas.

Ranges of concentrations in these samples were Co-60, <0.05 to 1.78 pCi/g; Cs-137, 0.28 to 11.9 pCi/g; Ra-226, 0.83 to 1.57 pCi/g; U-235, <0.23 to 1.69 pCi/g; U-238,

<0.78 to 3.45 pCi/g; Th-228, 1.03 to 9.57 pCi/g; Th-232, 0.98 to 7.46 pCi/g.

Concentrations of Sr-90 and isotopic uranium in two composite

samples, representing the remediated a rea and canyon area, are listed in Table 5.

The Sr-90 concentrations are 0.32 and 2.20 pCi/g; the highest uranium levels are I

U-238, which are 3.29 and 4.42 pCi/g.

On the basis of the U-234/U-238 ratios, it appears that the uranium is depleted in the U-235 and U-234 isotopes.

Uith exception of the total thorium (Th-228 and Th-232) concentration, in the sample from location 105B, all radionuclide levels were below the guideline values in Appendix C and most were in the range of baseline concentrations (see Table Ib).

The thorium concentration in sample 105B was 17.55 pCi/g, or 15.26 pCi/g a bove the a ve ra ge background level.

This is slightly higher than the guideline value of 10 pCi/g above background.

Surface scans in this area did not identify I

significantly elevated direct radiation levels, and sampling during the 1985 survey indicated that soils at grid intersections in the vicinity of this location were well within the guideline levels.

Contamination at grid coordinate 7695N, 9550E is therefore an isolated small area, and averaging over adjacent soil will result in a concentration which satisfies the 10 pCi/g guideline.

SUMMARY

During September

1987, Oak Ridge Associated Universities performed a

radiological survey of areas within the Phase I Decommissioning activities of CA Technologies in San Diego, California.

The survey included locations, which had I

been recediated, following their identification by a December 1985 ORAU survey, and a section of canyon area in the drainage pathway from the Waste Processing Facility.

Survey activities consisted of walkover gamma scans, exposure rate measurements, and soil sampling and analyses.

Findings identified no areas exceeding the decommissioning guidelines, authorized by the Nuclear Regulatory Commission for this site.

Based on these results it is ORAU's opinion that the i

'I radiological data, as presented by the licensee, is adequate and accurate and that the radiological conditions satisfy the established guidelines for release for unrestricted use.

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I TABLE lA BACKGROUND RADIATION LEVELS I

GA TECHNOLOGIES SAN DIEGO, CALIFORNIA B

Gamma Exposure Rates Ganma Exposure Rates Locationa at 1 m Above the Surface at the Surface

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TABLE 18 BASELINE RADIONtCLIDE CONCENTRATIONS IN SOIL CA TEONot0GIES SAN DIEGO, CALIFORNIA 8

Location Radionuclide Concentrations (pCl/q)

Co-60 Cs-137 Ra-226 U-235 U-238 Th(228 & 232)

K-40 b

1

<0.03

<0.02 0.59 t 0.14

<0.17 f.6 t 1.2 1.34 1 0.46 14.0 i 1.7 l

2

<0.05 0.16 1 0.11 0.53 1 0.22

<0.20 1.6 1 1.5 1.98 1 0.86 25.0 1 3.3 3

<0.04

<0.04 0.7910.20 0.39 1 0.24 1.1 1 0.5 2.24 1 0.62 10.4 1 1.7 4

<0.08

<0.05 1.20 1 0.29

<0.32

<l.1 3.08 1 0.79 29.0 1 3.4 5

<0.05

<0.05 1.23 1 0.22 0.69 0.55 1.3 1 0.6 3.20 1 0.80 24.5 1 2.7 6

<0.05

<0.05 0.65 1 0.16

<0.22 1.0 1 0.9 1.92 1 0.78 30.2 1 2.9 RAN3E

<0.03 to <0.08

<0.02 to <0.16 0.53 to 1.23

<0.17 to 0.69 1.0 to 1.6 f.34 to 3.20 10.4 to 30.2 N

AYERAGE

<0.05

<0.06 0.83

<0.33 1.5 2.29 22.2

• Refer to Figure 7 buncertainties represent the 95% confidence levels, based only on counting statistics; additional laboratory uncertaintles of 6 to 10% have not been propagated in these data.

. 8 TABLE 2 8

EXPOSURE RATES AT SAMPLING LOCATIONS PHASE I FOLLOW-UP I

GA TECHNOLOGIES SAN DIEGO, CALIFORNIA Grid Location 8 Coordinate Exposure Rate (uR/h)

ID N

E Contact 1 m Above Surface 100B 7160 9620 16 15 101B 7240 9646 16 15 102B 7532 9660 16 16 103B 7432 9645 20 18 104B 7611 9541 23 16 8

105B 7688 9550 20 15 106B 7528 9544 20 16 270B Canyon Flcor 18 16 271B Canyon Floor 16 15 272B Canyon Floor 15 14 2738 Canyon Floor 16 14 I

aRefer to Figures 5 and 6 I

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W TABLE 3 RADIONCULIDE ChCENTRATIONS IN S0lt FROM REMEDI ATED AREAS FHASE I FOLLOW-UP GA TECir40LOGIES SAN DIEGO, CAllFORNIA Radionuclide Concentrations (pCl/q)

Sample a

No.

Location Co-60 Cs-137 Ra-226 0-235 U-238 Th-228 Th-232 1008 7I60N,9620E

<0.05 0.28 1 0.12 1.17 1 0.25

<0.25

<0.8 1.59 1 0.42 1.61 1 0.47 b

1018 7290N,9646E 1.78 i 0.27 9.9710.44 1.17 1 0.40

<0.38 3.5 1 0.9 2.10 1 0.42 1.96 1 0.55 1020 7532N,%60E

<0.08 1.78 1 0.19 1.50 1 0.33

<0.32

<0.9 1.95 1 0.48 2.03 1 0.70 1038 7432N,9645E 0.52 1 0.15 2.92 1 2.51 1.3010.2.

<0.28

<0.8 2.76 1 0.45

1. % 1 0.55 1043 7611N,9541E

<0.06 1.52 1 0.17 1.57 i 0.23

<0.26 3.0 1 0.7 1.62 1 0.48 2.06 1 0.50 5

1050 7688N,9550E

<0.06 0.32 1 0.14 1.07 1 0.32^

1.69 1 0.86

<l.5 9.57 1 0.81 7.40 1 0.87 1068 752CN,9544E 1.50 1 0.28 11.85 0.50 0.83 1 0.33

<0.31

<0.9 1.26 1 0.48 1.40 1 0.57 aRefer to Elgure 5 buncertainties represent the 951 confidence levels based only on counting statistics; additional laboratory uncertainties of 6 to 10% have not been propagated Into these data.

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M TABLE 4 RADIONCULIDE arJCENTRATIONS IN Soll FROM THE CANYON FLOOR FHASE I FOLLOW-OP GA TEOf40LOGIES SAN OIEGO, CALIFORN1A s

Sample Radlonuclide Concentrations (pCI/g>

No.

Co-60 Cs-137 Ra-226 U-235 U-238 Th-228 Th-232 D

2700 0.15 t 0.10 0.97 1 0.16 1.0 t 0.2 0.57 1 0.12 1.7 1 0.6 f.62 1 0.44 2.00 1 0.57 271B 0.18 1 0.20 0.68 1 0.15 1.4 1 0.3 0.27 1 0.14

<1.0 1.03 1 0.50 0.98 1 0.37 2720

<0.05 0.15 1 0.09 1.4 1 0.3 0.20 1 0.06 2.4 1 0.67 1.50 i O.28 1.50 1 0.55 2738 0.17 1 0.12 1.5 i 0.2 1.3 1 0.2 0.41 1 0.13 0.94 1 1.7 f.64 1 0.39 1.60 1 0.47 C

aRefer to Figure 6 buncertainties represent the 95% confidence levels based only on counting statistics; additional laboratory uncertaintles of 6 to 10% have not been propagated into these data.

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I TABLE 5 R ADICNUCLIDE COiCENTRATIONS IN (DtPOSITE Soll SAMPLES PHASE I FOLLCV-UP I

GA TECFfiOLOGIES SV4 Ot EGO, CALIFORNI A I

Saaple Radionuelfde Concentrations (pCf/q) 10 Sr-90 U-234 U-235 0-238 Composite Aa 0.32 t 0.12 1.85 t 0.26 0.13 1 0.08 4.42 1 0.40 0

I Composite B 2.20 1 0.30 1.74 1 0.24 0.07 1 0.06 3.29 t 0.34 I

aSa-ple identification numbers:

I C yposite A (270B; 271B; 2729; 2738)

Composite Bi (72974, 9646E; 7432N, 9645E; 7611N, 9541E; 7528N, 9544E) buncertainties represent the 951 confidence levels, based only on counting statistics; additional laboratory uncertainties of t 6 to 10f have not been propagated into these data.

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. I REFERENCES I

1.

"Confirmatory Survey of Phase I Decommissioning Former Uaste Processing Facility," CA Technologies, San Diego, California, Oak Ridge Associated Universities, July 1986 2.

Letter from K.E. Asmussen (CA Technologies Inc.) to R. R. Thomas (U.S. Nuclea r Regulatory Commission, Region V), Reference "License SNM-696, Docket 70-1734" August 12, 1987 I

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g smNm A MAJOR SMiPLING AND ANALYTICAL EQUIPMENT I

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i APPENDIX A E

j MAJOR SAMPLING AND ANALYTICAL EQUIPMENT The display or description of a specific product is not to be construed as an endorsement of that product or its manufacturer by the authors or their employer.

A.

Direct Radiation Measurements Eberline "RASCAL"

'l Portable Ratemeter-Scaler

'W ffodel PRS-1 i

(Eberline, Sante Pe, Mi) i Eberline PR!!-6 Portable Ratemeter (Eberline, Sante Fe, N't) iI Victorcen NaI Scintillation Detector

'todel 489-55 (Victoreen, Cleveland, OH)

Reuter-Stokes Pressurized Ionization Chamber Model RSS-Ill

. I (Reuter-Stokes, Cleveland, OH) 8.

Laboratory Analyses Automatic low-background Alpha-Beta Counter tiodel LB5110-2080 (Tennelec, Inc., Oak Ridge, TN) i High-Purity Germanium Detector ig Model CMX-23195-S, 23% efficiency ig (EC&G ORTEC, Oak Ridge, TN) l Used in conjunction with:

i Lead Shield, C-16 I

(Camma Products Inc., Palos Hills, IL)

{g High Purity Cermanium Coaxial Well Detector

'E Model CVL-110210-PWC-S, 23% Efficiency (EC&C ORTEC, Oak Ridge, TN)

Used in conjunction with:

Lead Shield Model G-16 (Applied Physical Technology, Atlanta, CA)

High Purity Germanium Detector

{

Model ICC25, 25% Efficiency (Princeton Camma-Tech, Princeton, NJ)

A-1

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Used in conjunction with:

Lead Shield (Nuclear Data, Schaumburg, IL) itultichannel Analyzer ND-66/ND-680 System g

(Nuclear Data Inc., Schaumburg, IL)

3 Alpha Spectrometry System Tennelec Electronics (Tennelec, Oak Ridge, TN)

Surface Barrier Detectors (EC&G ORTEC, Oak Ridge, TN)

Multichannel Analyzer Model ND-66 I

(Nuclear Data, Schaumburg, IL)

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APPENDIX B

I MEASUREMENT AND ANALYTICAL PROCEDURES 1

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l APPENDIX B Measurement and Analytical Procedures Camma Surface Scans I

Walkover surface scans were performed at approximately 1-2 m intervals using Ebetline Model PRM-6 portable ratemeters with Victoreen Model 489-55 gamma scintillation probes containing 3.2 cm X 3.8 cm NaI(T1) acintillation crystals.

Relative count rates were monitored using earphones and increased rates above the ambient background levels were noted.

I Exposure Rate Measurements I

Measurements of gamma exposure rates were performed using an Eberline PRM-6 po rtc.ble ratemeter with a Victoreen Model 489-55 gamma scintillation probe containing a 3.2 cm x 3.8 cm Nal(T1) scintillation crystal.

Count rates were converted to exposure rates (uR/h) by cross-calibrating with a Reuter Stokes model RSS-Ill pressurized ionization chamber.

Soil Sample Analysis Gamma Spectroscopy I

Soil samples were dried, mixed, and a portion sealed in 0.5-liter Marinelli beaker. The quantity placed in the beaker was chosen to reproduce the calibrated counting geometry and typically ranged from 600 to 800 g of soil. Net soil weights were determined and the samples counted using intrinsic germanium and Ge(Li) detectors coupled to a Nuclear Data Model ND-680 pulse height analyzer system. Background and Compton st ripping, peak search, peak identification, and concentration calculations were perfcrmed using the computer capabilities inherent in the analyzer system. Energy peaks used for d* termination of I

radionuclides of concern were:

I I

B-1 I

I Co 1.173 MeV Cs-137 - 0.662 MeV Ra-226 - 0.609 MeV from Bi-214 (secular equilibrium assumed)

U-235 - 0.144 MeV U-238 - 0.094 MeV f rom Th-234 (secular equilibrium assumed)

B Th-228 - 0.583 MeV from T1-209 (secular equilibrium assumed)

Th-232 - 0.911 MeV from Ac-228 (secular equilibrium assumed)

The spectra were also reviewed for the presence of other radionuclides.

I Strontium-90 Analysis I

Aliquots of soil were dissolved by pyrosulfate fusion and the strontium precipitated as a sulfate.

Successive treatments with EDTA preferentially I

removed lead and excess calcium and returned the strontium to solution.

Ferric and other insoluble hydroxides was precipitated at a pH of 12 to 14.

Strontium was reprecipitated as a sulfate.

Barium was removed as a chromate using DTPA.

The final precipate of strontium carbonate was counted using a low-1:ackground Tennelee alpha-beta proportional counter.

Alpha Spectrometry for Isotopic Uraniun Aliquots of soil were dissolved by pyrosulf ate fusion and precipitated by I

barium sulfate.

The barium sulfate precipitate was redissolved and uranium was separated by liquid-liquid extraction.

The uranium was then precipitated with a cerium fluoride carrier and counted using surface barrier detectors (ORTEC),

alpha spectrometers (Tennelec), and an ND-66 Multichannel Analyzer (Nuclear Data).

Uncertainties and Detection Limits The uncertainties associated with the analytical data presented in the tables of this report, represent the 95% confidence levels for that data. These uncertainties were calculated based on both the gross sample count levels and the associated background count levels.

When the net sample count was less than the 95% statistical deviation of the background count, the sample concentration was reported as less than the detection capability of the measurement procedure.

B-2 I

I Because of variations in background

levels, sample voluees or weights, measurement efficiencies, and Compton contributions from other radionuclides in samples, the detection limits differ from sample to sample and instrument to ins t rume n t. Additional uncertainties of 2 6 to 10%, associated with sampling and laboratory procedures, have not been propagated into the data presented in this I

report.

Calibration and Quality Assurance Laborato ry and field survey procedures are documented in the following manuals, developed specifically for the Oak Ridge Associated Universities' Radiological Site Assessment Program:

"Survey Procedures Manual," Revision 3,

I May 1987; "Laboratory Procedures ?!anuol", Revision 3,

itay 1987 and "Quality Assurance ?tanual", Revision 1. June 1987.

I With the exception of the measurements conducted with portable gamma scintillation survey meters, instruments were calibrated with NBS-traceable standards.

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

Quality control procedures on all instruments included daily background and check-source measurements to confirm equipment operation within acceptable statistical fluctuations.

The ORAU laboratory participates in the EPA and EML Quality Assurance Programs.

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APPENDIX C i g DECOTtISSIONING CUIDELINES FOR THE

' 3 CA TECHNOLOGIES WASTE PROCESSING FACILITY I

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I APPENDIX C I

Decommissioning Guidelines for the CA Technologies Waste Processing Facility Target criteria for unrestricted release of the CA Technologies' Uaste Processing Facility and surrounding areas are presented in the licensee's final report and are as follows:

I External Radiation The gamma exposure rate at I m above the ground surface shall not exceed 10 UR/h above background for an area of greater than 30 ft (9.1 m) x 30 ft (9.1 m) and shall not exceed 20 pR/h above background foranydiscretearea(i.e.lessthan30ft (9.1 m) x 30 ft 9.1m)].

I Inhalation and Ingestion Concentrations of radionuclides in soil shall be such Gat inhalation and ingestion are not expected to result in annual dose equivalents ev.ceeding 20 mrem to the lung or 60 mrem to the bone.

Limiting soll concentrations were derived to satisfy these external and internal target criteria.

The concentration limits are presented in the

]

following Table.

I Radionuclide Concentration Limit Above Background (pCi/g)

I Depleted Uranium 35 Enriched Uranium 30 Thorium (Natural) 10 Co-60 8

Cs-137 15 Sr-90 1.8 x 10 I

Where core than one radionuclide is present, the sum of the ration of the individual radionuclide concentrations to their respective concentration limits I

shall not exceed 1.

C-1