Text

Draft Rept, Confirmatory Survey for Repower Area Fsv Platteville,Co
	ML20084F393
Person / Time
Site:	Fort Saint Vrain
Issue date:	05/16/1995
From:	Abelquist E; OAK RIDGE ASSOCIATED UNIVERSITIES
To:	; NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML20084F374	List: ML20084F369; ML20084F393;
References
	CON-FIN-A-9076 NUDOCS 9506020184
	Download: ML20084F393 (43)
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DR*AFT REPORT LCONFIRMATORY SURVEY FOR THE REPOWER AREA

~ FORT ST. VRAIN

PLATTEVILLE, COLORADO

[ DOCKET NO. 50-267]

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l CONFIRMATORY SURVEY FOR TIIE REPOWER AREA FORT ST. VRAIN PLATTEVILLE, COLORADO l 1 l 1 l Prepared by l

l E. W. Abelquist l Environmental Survey and Site Assessment Program Energy / Environment Systems Division Oak Ridge Institute for Science and Education Oak Ridge, TN 37831-0117 Prepared for the Division of Waste Management U.S. Nuclear Regulatory Commission Headquarters' Office DRAFT REPORT MAY 1995 This report is based on work performed under an Interagency Agreement (NRC Fin. No. A-9076) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy.

Oak Ridge Institute for Science and Education performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy.

This draft report has not been given full review and patent clearance, and the dissemination of f its information is only for official use. No release to the public shall be made without the -

l approval of the Office of Information Services, Oak Ridge Institute for Science and Education.

l Fort St. Vrain-Plattevilk, CO . May 16, 1993 h:\esnap\ reports \s',vrain\ fort vra.001

ACKNOWLEDGEMENTS The author would like to acknowledge the significant contributions of the following staff members:

FIELD STAFF G. R. Foltz LA110RATORY STAFF R. D. Condra R. L. Epperson M. J. Laudeman S. T. Shipley CLERICAL STAFF D. A. Adams R. D. Ellis K. E. Waters ILLUSTRATOR T. D. Herrera Fort St vrain Ptanevitic, CO htay 16, 1995 h \cunp\ reports \nt,vrain\ fort _vra 001

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TABLE OF CONTENTS l PAGE List o f Figu res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii List of Tables

..............................................iii Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Site Description

............................................. 2 Obj ec t ives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Docu men t Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Procedures

................................................ 3 Findings and Results ............. .

.......................... 5 Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Summary . . ...

.......................................... 8 References ..

....................................,.......21 Appendices:

Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors Fort St. Vrain.Planeville, CO May 16,1995 i h:\essapireporta\at_vrain\ fort _vra.001

LIST OF FIGURES i

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PAGE FIGURE 1: Location of the Fort St. Vrain Site-Platteville, Colorado . . . . . . . . . . . 9 FIGURE 2: Plot Plan of the Fort St. Vrain Nuclear Station ................ 10 FIGURE 3: Fort St. Vrain-Repower Area

..........................11 FIGURE 4: Repower Area, Miscellaneous Concrete and Metal Surfaces-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 12 FIGURE 5: Repower Area, Evaporative Cooler Building-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 13 FIGURE 6: Repower Area-Exposure Rate Measurement Locations . . . . . . . . . . . 14 FIGURE 7: Repower Area-Soil Sampling Locations . . . . . . . . . . . . . . . . . . . . 15 FIGURE 8: Fort St. Vrain-Background Soil Sampling and Exposure Rate Measurement Locations

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LIST OF TABLES PAGE TABLE 1: Summary of Surface Activity Measurements . . . . . . . . . . . . . . . . . . 17 TABLE 2: Exposure Rates . . . . . . . . . . . . . . . . . . . . . ..............18 TABLE 3: Background Exposure Rates and Radionuclide Concentrations in Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 TABLE 4: Radionuclide Concentrations in Soil Samples . . . . . . . . . . . . . . . . . . 20 Post St. Vrain Planeville, CO May 16, 1995 iii h \cesap\reportaist_vrain\ fort vra.001

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l ABBREVIATIONS AND ACRONYMS ;

i ASME ' American Society of Mechanical Engmcers ,

em centimeter cm 2 square centimeter cpm counts per minute ~

DOE Department of Energy ,

dpm/100 cm 2 disintegrations per minute /100 square centimeters l EML Environmental Measurement Laboratory EPA Environmental Protection Agency

.ESSAP Environmental Survey and Site Assessment Program l FSV Fort St. Vrain !

-HTGR High Temperature Gas-Cooled Reactor l kg kilogram j m meter .i m2 square meter 'i mm millimeter ,

MeV million electron volts MWe Megawatts electric ;

Nal sodium iodide ;

NIST National Institute of Standard and Technology NRC Nuclear Regulatory Commission ORAU Oak Ridge Associated Universities -

ORISE Oak Ridge Institute for Science and Education pCi/g picocuries per gram PSC Public Service Company of Colorado i

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Fort St. Vrain-Planeviue, CO . May 16.1995 iV h:\essap\ reports \st_vrain\ fort _vra.001 l

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CONFIRMATORY SURVEY FOR THE REPOWER AREA FORT ST. VRAIN PLATTEVILLE, COLORADO INTRODUCTION AND SITE IIISTORY Fort St. Vrain (FSV) was a 330 MWe High Temperature Gas-Cooled Reactor (HTGR) owned and operated by Public Service Company (PSC) of Colorado. The site consists of 6995 hectares (2798 acres) owned by PSC, of which approximately one square mile was designated as the exclusion area during plant operation. The licensee maintained complete control over this area.

The basic installation included a reactor building, turbine building, cooling towers, and an electrical switchyard.

FSV was permanently shutdown in August 1989, with the decision to decommission the facility made during December 1989. On November 23,1992, the Nuclear Regulatory Commission (NRC) issued the Order to Authorize Decommissioning of Fort St. Vrain and Amendment No.

85 to Possession Only License No. DPR-34. During the period 1989 to 1991, a radiological characterization of the FSV site was performed. Currently, the FSV decommissioning is approximately 70 % complete, with completion expected early in 1996 (excluding the final site survey).

PSC has committed to the Colorado Public Utilities Commission to resume electrical generation at FSV through the installation of gas turbines. In order to perform this project, a small section ofland in the southwest area of the site has been cleared in preparation for the repower effort. h This area is referred to as the repower area, where PSC plans to install natural gas-fired combustion turbines and heat recovery boilers to repower the facility.

During plant operations, pre-fabricated steel buildings were located in the repower area. These buildings accommodated a construction workshop, a quality control facility that performed radiography, a small warehouse, and a flammable storage building.

Fort St. Vram Pintteville. CO - May 16. 1995 hiessap\ reports \st,vrain\ fort _vra.001

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l The repower area does not have a known history of radioactive contamination. This was reaffirmed by the evaluation of the characterization results for the repower area, which did not identify radioactive contamination due to licensed activities. The repower area was therefore classified as an unaffected area. However, elevated levels of Cs 137 were identified in surface soil collected from a localized area outside, but adjacent to the repower area. This area had previously been used for temporary storage of spent fuel shipping casks.

At the request of the NRC's Division of Waste Management, Headquarters' Office, the Environmental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) performed an independent confirmatory radiological survey of the repower area at the Fort St. Vrain site in Platteville, Colorado. '

SITE DESCRIPTION The FSV facility is located approximately 56 kilometers (35 miles) north of Denver and 5.6 kilometers northwest of the town of Platteville, in Weld County, Colorado (Figure 1). The site consists of 6995 hectares owned by PSC of which approximately one square mile was designated as the exclusion area during plant operation.

The repower survey area is located within the restricted area of the FSV facility on the east side of the turbine building, north of the electrical switchyard (Figure 2). This location is approximately 12,225 square meters in size. The area has been isolated from the balance of the restricted area by a chain link fence and locking gates controlled by FSV Security. The boundaries of the repower area include portions of the original restricted area fence on the south and east sides and newly erected fence on the west and north sides (Figure 3). The boundary also includes the east, and a portion of the south exterior walls of the evaporative cooler building.

Fort St Vram-PLueviuc, CO - htay 16. l995 2 h.\essap\reportsist,vrain\ fort _vra 001

1 OBJECTIVES The objectives of the confirmatory survey were to provide independent document reviews and radiological data for use by the NRC in evaluating the adequacy and accuracy of the licensee's procedures and final status survey results. ,

DOCUMENT REVIEW .

. t ESSAP has reviewed the licensee's final survey report and radiological survey data.5 Procedures and methods utilized by the licensee were reviewed for adequacy and appropriateness. The data were reviewed for accuracy, completeness and compliance with guidelines.

PROCEDURES i t

During the period March 20 through 22,1995, ESSAP performed a confirmatory survey at the l Fort St. Vrain site in Platteville, Colorado. The survey was conducted in accordance with a ;

survey plan dated March 17,1995, submitted to and approved by the NRC's Division of Waste Management, Headquarters' Office.2 This report summarizes the procedures and results of the f

survey. Additional information concerning major instrumentation, sampling equipment, and analytical procedures is provided in Appendices A and B.

SURVEY PROCEDURES The licensee's final status survey of the repower area included two general categories of survey units: surfaces and structures, and open land areas. The area was further divided into survey units. The surface and structure survey units within the repower area included the Valve Pit, Miscellaneous Metal Surfaces, Concrete Slab at the Security Fence, Miscellaneous Concrete Surfaces, Evaporative Cooler Building (east and south walls below 2 m), and Evaporative Cooler Building (east and south foundation walls), The open land area survey units included the general soil area within the repower area, leach field soil area, septic system, and monitoring wells.

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Reference System ESSAP selected specific measurement and sampling locations from each of the survey units.

Because survey maps illustrating the measurement locations were not provided by the licensee, ESSAP requested that the licensee identify some of their measurement and sampling locations for confirmatory measurements. Measurement and sampling locations were referenced to prominent site features and recorded on survey maps.

Sprface Scans Soil surfaces were scanned for gamma radiation using Nal scintillation detectors. Approximately 10% of the soil in the general and leach field soil areas was scanned. Structure surfaces were also scanned with gas proportional detectors over the 0.5 m2 area surrounding each direct measurement location. Particular attention was given to cracks and joints in the surfaces and walls, ledges, drains, and other locations where material may have accumulated. All detectors were coupled to ratemeters or ratemeter-scalers with audible indicators. Locations of elevated direct radiation detected by scans were marked for further investigation.

S!!rface Activity Measurements Direct measurements for total beta activity were performed at 38 locations, representing each of the survey units within the repower area. Measurements were performed using gas proportional detectors, coupled to portable ratemeter-scalers. Smear samples, for determining removable activity levels, were collected from each direct measurement location. Measurement and sampling locations are shown on Figures 4 and 5. ;

EXUnsure Rate Measurements Exposure rate measurements were performed within the general soil area of the repower area.

Exposure rates were measured at I m above surfaces at 10 locations using a pressurized ionization chamber (PIC). Background exposure rate measurements were performed using a PIC Tort SL Vram Platteville, CO - May 16.199) 4 h.\essap\ reports \st,vrain\ fort _vra 001

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at five locations within a 0.5 to 10 km radius of the site. Measurement locations are shown on Figures 6 and 8.

' Soil Samoling 6

Five background soil samples were collected from each of the external background exposure rate measurement locations.

A total of ten soil samples were collected randomly from the general soil area within the repower area. Soil sampling locations are shown on Figures 7 and 8.

i SAMPLE ANALYSIS AND DATA INTERPRETATION l Samples and data were returned to ESSAP's laboratory in Oak Ridge, Tennessee for analysis and interpretation. . Soil samples were analyzed by solid state gamma spectrometry. Spectra ;

were reviewed for Co-60, Cs-137, and any other identifiable photopeaks. Soil sample results were reported in units of picoeuries per gram (pCi/g). Smear samples were analyzed for gross alpha and gross beta activity using a low background gas proportional counter, and the results converted to dpm/100 cm2 . Direct measurements for surface activity were converted to units 2

of disintegrations per minute per 100 square centimeters (dpm/100 cm). Exposure rates were reported in units of microroentgens per hour ( R/h). Results were compared with the licensee's documentation and NRC guidelines established for release to unrestricted use, which are l provided in Appendix C.

FINDINGS AND RESULTS DOCUMENT REVIEW ESSAP reviewed the licensee's final status survey report, including the final status survey data and provided comments to the NRC.3 The survey instrumentation and procedures used, including the assessment of background contributions to surface activity measurements, were discussed at Fort St. Vrainti neville, CO . May 16.1995 5 se-.e report.s.i_

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length. Guidelines for surface contamination and exposure rates were clearly stated, however, radionuclide concentrations in soil that correspond to the 10 mrem per year site-specific soil guideline were not specified. The site operational history, decommissioning activities, and final survey results provided sufficient information on the radiological status of the repower area.

SURVEY RESULTS S.urface Scans Surface scans for beta activity on structure surfaces did not identify any locations of elevated direct radiation. Surface scans for gamma activity within the general soil area also did not result in the identification of any locations of elevated direct radiation.

Sudace Activity Measurements Surface activity measurements for total beta activity are summarized in Table 1. Total beta activity levels for all measurement locations ranged from <340 to 710 dpm/100 cm 2, Removable activity levels were all less than the minimum detectable activity of the procedure 2

which was 12 dpm/100 cm for gross alpha and 16 dpm/100 cm2 for gross beta.

Ihp9sure Rates Site exposure rates are summarized in Table 2. Background exposure rates ranged from 15 to 17 R/h, and averaged 16 pR/h (Table 3). Exposure rates in the repower area ranged from 16 to 26 pR/h at 1 m above the surface.

i Radlonuclide Conctntrations in Soil Samples Radionuclide concentrations in background samples are summarized in Table 3 and were <0.2 pCi/g for Co-60, < 0.1 to 0.2 pCi/g for Cs-137,1.4 to 1.8 pCi/g for Th-228,1.5 to 2.2 pCi/g for Th-232, < 0.1 pCi/g for U-235, and < 2.1 pCi/g for U-238.

Fort St. VraimPlatteville, CO - May 16, 1995 6 h \essap\ reports \st,vrain\ fort _vra 001

Concentrations of radionuclides in surface soil samples collected randomly from the repower -

area summarized in Table 4. Radionuclide concentration ranges are as follows: <0.2 pCi/g for Co-60, < 0.1 pCi/g for Cs-137,1.1 to 2.0 pCi/g for Th-228,1.1 to 1.8 pCi/g for Th-232, < 0.1 pCi/g for U-235, and <2.3 pCi/g for U-238.

COMPARISON OF RESULTS WITII GUIDELINES The primary contaminants of concern for this site are beta-gamma emitters resulting from the operation of the FSV facility. The applicable NRC guidelines for beta-gamma emitters in unaffected areas are provided in Regulatory Guide 1.86.4The guidelines are:

Total Activity 5,000 dpm/100 cm2 , averaged over a 1 m2 area 15,000 dpm/100 cm2 , maximum in a 100 cm2 area Removable Activity 1,000 dpm/100 cm2 Surface activity measurements for total and removable activity were all within the surface contamination guidelines.

The guideline values for radionuclide concentrations in soil are the radionuclide-specific concentrations which could result in an average annual total effective dose equivalent (TEDE) of 10 mrem to an individual in a population group exposed to radioactive material following decommissioning. These values may be determined in accordance with the methodology contained in NUREG/CR-5512, Volume 1 and as presented in NUREG-1500.5,6 Concentrations of radionuclides in soil samples are comparable to the concentrations measured in background samples (Tables 3 and 4). Therefore, compliance is demonstrated by the fact that soil samples collected from the repower area are indistinguishable from background levels.

Fort St. Vram-Platteville, CO May 16, 1995 7 h \empWrodMst,vrainWn,vraM1

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The exposure rates guideline, measured at 1 m above the surface, is 5 R/h above background.7 With the exception of one elevated exposure rate measured in the northwest corner of the repower area (Figure 6, #5), all exposure rates were within the guideline. This elevated exposure rate measurement was due to dismantlement activities to remove the core support floor (significantly activated concrete) in the reactor building. These activities have temporarily affected exposure rates in the repower area, most notably in areas closest to the reactor building.

Soil sampling in the areas affected by these increased exposure rates resulted in no indication of soil activity in excess of background levels.

SU51 MARY During the period March 20 through 22, 1995, at the request of the NRC's Division of Waste Management, Headquarters' Office, the Environmental Survey and Site Assessment Program of ORISE of CRISE performed a confirmatory survey at the Fort St. Vrain site in Platteville, Colorado. Survey activities included document reviews, surface scans, surface activity measurements, exposure rate measurements, and soil sampling.

The confirmatory survey identified one location within the repower area that exhibited an elevated exposure rate measurement. This location within the repower area was influenced by elevated radiation from dismantling activities on the core support floor in the repower area. Soil sampling in this area confirmed that the elevated exposure rate measurement was not the result of elevated soil concentrations. The confirmatory survey results are consistent with those obtained by the licensee and support the licensee's conclusion that residual activity levels in the repower area satisfy the guidelines for release to unrestricted use, l

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Fort St. Vrain - Background Soil Sampling and Exposure Rate Measurernent Locations Foit st, vinin-Plattevale, CO - May 16, 1995 16 h \essap\ reports \st,vrain\ fort _vra 001 l l

4

.. . l TABLE 1

SUMMARY

OF SURFACE ACTIVITY MEASUREMENTS REPOWER AREA FORT ST. VRAIN PLATTEVILLE, COLORADO Number of Range of Removable Location" Measurement Range of Total Beta Activity (dpm/100 cm ) 2 Locations Activity (dpm/100 cm 2) l Concrete Slab 6 < 420 < 12 < 16 (at security fence)

Miscellaneous Concrete 3 < 420 < 12 < 16 Valve Pit 3 <420 - 710 < 12 < 16 Miscellaneous Metal 4 < 340 < 12 < 16 Evaporative Cooler Bldg.

11 < 340 < 12 < 16 (East and South Walls)

Evaporative Cooler Bldg.

11 < 420 < 12 < 16

, (East and South Foundation)

Refer to Figures 4 and 5.

1 1

)

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Fat St. Vra'ubPlattevdle, CO . May 16. IM5 17 h:\cssap\reportaist_vrainVort_vra.001 i

1 TABLE 2 i EXPOSURE RATES REPOWER AREA l FORT ST. VRAIN PLATTEVILLE, COLORADO Location

Exposure Rates ( R/h) at 1 m Above Surface l 1 18 l 2 18 j l

3 20 i

4 18 )

5 26 6 21

~

7 19 8 17 9 16 10 16

' Refer to Figure 6.

l Fort St Vram-Platteville, Co - May 16,1993 18 h : \e ssap\ reports \st,v r a in\ fort,v ra ,001

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TABLE 3 BACKGROUND EXPOSURE RATES AND RADIONUCLIDE CONCENTRATIONS IN SOIL j REPOWER AREA i FORT ST. VRAIN PLATTEVILLE, COLORADO Radionuclide Concentration (pCi/g)

R e R/h) l Location" at 1 m Co-60 Cs-137 Th-228 Th-232 U-235 U-238 Above l

Surface 1 < 0.1 < 0.1 1.8 i 0.2 6 1.7 i 0.4 < 0.1 0.9 1.4 17 2 < 0.2 < 0.1 1.5 0.2 1.6 0.4 < 0.1 < 2.1 16 3 < 0.1 < 0.1 1.4 0.1 1.8 0.4 < 0.1 < 1.5 15 4 < 0.1 < 0.1 1.6 0.1 1.5 0.5 < 0.1 < 1.9 - 17 5 < 0.1 0.2 i 0.1 1.7 i 0.2 2.2 i 0.6 < 0.1 < 1. 8 16 -

" Refer to Figure 8. i b(JnCertainticS repicSent the 95% confidence level, based only on counting statistics.

{

I l

Fort St. Vrain-PlaueviJte. CO . M.y 16.1995 19 h:se...r s,erortis.i_v,.intrott_vr..co l

b

TABLE 4 RADIONUCLIDE CONCENTRATIONS IN SOIL SAMPLES REPOWER AREA FORT ST. VRAIN PLATTEVILLE, COLORADO Radionuclide Concentration (pCi/g)

Co-60 Cs-137 Th-228 Th-232 U-235 U-238 1 < 0.1 <0.1 1.6 0.26 1.5 i 0.5 < 0.1 < 1.7 2 < 0.1 < 0.1 1.7 0.2 1.6 i 0.5 < 0.1 < 2.1 3 < 0.1 < 0.1 1.5 0.1 1.6 0.4 < 0.1 0.9 1.3

< 2.3 I 4 < 0.2 < 0.1 2.0 0.2 1.8 0.4 < 0.1 5 < 0.1 < 0.1 1.5 0.1 1.6 0.4 < 0.1 1.0 i 0.9 6 < 0.1 < 0.1 1.7 t 0.1 1.6 0.4 < 0.1 <1.4 7 < 0.2 < 0.1 2.0 i 0.2 1.8 0.4 < 0.1 < 2.2 8 < 0.1 < 0.1 1.9 0.2 1.8 i 0.6 < 0.1 < 1.7 9 < 0.2 < 0.1 1.6 0.2 1.4 i 0.4 < 0.1 < 2.1 10 < 0.1 < 0.1 1.1 i 0.1 1.1 0.4 < 0.1 < 1.4

Refer to Figure 7.

b Uncertainties represent the 95% confidence level, based only on counting statistics.

Port St Vrain Platieville. CO hisy 16, 1995 20 wangwpomwi_minvui_m.coi

REFERENCES

1. Public Service Company of Colorado," Final Status Survey Plan and Report," Cintichem, Inc., December 5,1994.

2. Oak 1Udge Institute for Science and Education " Confirmatory Survey Plan for the Repower Area, Fort St. Vrain, Platteville, Colorado (Docket No. 50-267)," March 17, 1995.

3. Oak Ridge Institute for Science and Education, letter from E. W. Abelquist to D.

Fauver, NRC/NMSS, " Document Review - Final Survey Report for Release of the Repower Area, Fort St. Vrain, Platteville, Colorado (Docket No. 50-267)," March 13, 1995.

4. U.S. Nuclear Regulatory Commission, " Termination of Operating Licenses for Nuclear Reactors," Regulatory Guide 1.86, Washington, D.C., June 1974.

5. NUREG/CR-5512, " Residual Radioactive Contamination from Decommissioning,"

Volume 1, October 1992.

6. NUREG-1500, " Working Draft Regulatory Guide on Release Criteria for Decommissioning: NRC Staff's Draft for Comment," August 1994.

7. NUREG-0586, " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," August 1988.

Test St. Vrain-Platteville. CO - May 16, 1995 21 h se. .psierortos i_vr.;nstori_vr..ooi

APPENDIX A ,

MAJOR INSTRUMENTATION l

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Fort St Vrain~Plattevale, CO May 16,1995 h:\essap\ reports \st_vrain\ fort,vra.001

I APPENTIX A l I

MAJOR INSTRUMENTATION 1

)

The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the authors or their employers.

DIRECT RADIATION MEASUREMENT

- Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM) 1 Ludlum Ratemeter-Scaler Model 2221 (Ludlum Measurements, Inc.,

Sweetwater, TX)

Detectors Ludlum Gas Proportional Detector Model 43-68 Effective Area,100 cm2 ,

(Ludlum Measurements, Inc.,

Sweetwater, TX)

Reuter-Stokes Pressurized Ion Chamber Model RSS-111 (Reuter-Stokes, Cleveland, OH) e Victorcen Nal Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal (Victorcen, Cleveland, OH)

LABORATORY ANALYTICAL INSTRUMENTATION ,

High Purity Extended Range Intrinsic Detectors !

Model No: ERVDS30-25195 ,

(Tennelec, Oak Ridge, TN) l Used in conjunction with: l Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, TN) and Fort St. Vrain-Planeville, CO - May 16,1995 A-1 h:se...psreportissi_vr insrort_vr..ooi I

q l

Multichannel Analyzer 3100 Vax Workstation )

(Canberra, Meriden, CT) l High-Purity Germanium Detector Model GMX-23195-S,23% Eff.

(EG&G ORTEC, Oak Ridge, TN)

Used h conjunction with:

Lead Shield Model G-16 (Gamma Products, Palos Hills, IL) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT)

Low Background Gas Proportional Counter Model LB-5110-W ,

(Oxford, Oak Ridge, TN) l l

l l

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roe 3,. v,. . n. ... co . m.y 16. im A-2 h se .psreportos.i_vr instort_vra cot

APPENDIX B SURVEY AND ANALYTICAL PROCEDURES l

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l Fort St Vrain Platteville, Co . May 16, 1995 h.\c nnap\ reports \st_v rain \ fort,v ra .001

., t

t. APPENDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURFS

' S_urface Scami Surface scans were performed by passing the probes slowly over the surface; the distance between the probe and the surface was maintained at a minimum-nominally about I cm. Hand-held gas proportional detectors were used to scan the structural surfaces. Identification of elevated !cvels was based on increases in the audible signal from the recording and/or indicating instrument. Combinations of detectors and instruments used for the scans were:

Beta -

gas proportional detector with ratemeter-scaler Gamma -

Nal scintillation detector with ratemeter

~

Surface Activity Measurements i Measurements of total beta activity leveIs were performed using gas proportional detectors with l ratemeter-scalers. Count rates (cpm), which were integrated over 1 minute in a static position, !

were converted to activity levels (dpm/100 cm2 ) by dividing the net rate by the 4r efficiency and j correcting for the active area of the detector. The beta activity background count rate for the gas proportional detectors was 507 and 761 cpm on metal and concrete surfaces, respectively.

The beta efficiency factors ranged from 0.24 to 0.25 for the gas proportional detectors calibrated to Tc-99. The effective probe area for the gas proportional detectors is 126 cm2, l l

l Soll Samoling Approximately 1 kg of soil was collected at each sample location. Collected samples were placed in a plastic bag, scaled, and labeled in accordance with ESSAP survey procedures.

Fort St. Vrain Planeville. CO . May 16.199$ B-1 Me ssap\ reports \st,v rain \ fort _vra .001

ANALYTICAL PROCEDURES Gamma Soeettonalty Samples of soil materials were dried, mixed, crushed, and/or homogenized as necessary, and a portion scaled in 0.5-liter Marinelli beaker or other appropriate container. The quantity placed in the beaker was chosen to reproduce the calibrated counting geometry. Net material weights were determined and the samples counted using intrinsic germanium detectors coupled to a 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.

All photopeaks associated with the radionuclides of concern were reviewed for consistency of activity. Energy peaks used for determining the activities of radionuclides of concern were:

Co-60 1.173 MeV Cs-137 0.662 MeV Th-228 0.239 MeV from Pb-212*

Th-232 0.911 MeV from Ac-228*

U-235 0.186 MeV U-238 0.063 MeV from Th-234*

Secular equilibrium assumed.

Spectra were also reviewed for other identifiable photopeaks.

t f

u I

r- si v..a n.m.m. co . m, ie. m3 n-2 no..gwn- tmm.m wi

l .

UNCERTAINTIES AND DETECTION LIMITS The uncertainties associated with the analytical data presented in the tatles of this report represent the 95 % confidence level for that data. These uncertainties were calculated based on both the gross sample count levels and the associated background count levels. Additional uncertainties, associated with sampling and measurement procedures, have not been propagated into the data presented in this report.

Detection limits, referred to as minimum detectable activity (MDA), were based , on 2.71 plus 4.65 times the standard deviation of the background count [2.71 + (4.65/BKG)]. When the activity was determined to be less than the MDA of the measurement procedure, the result was reported as less than MDA. Because of variations in background levels, measurement efficiencies, and contributions frem other radionuclide in samples, the detection limits differ from sample to sample and instrument to instrument.

CALillRATION AND QUALITY ASSURANCE Calibration of all field and laboratory instrumentation was based on standards / sources, traceable to NIST, when such standards / sources were available. In cases where they were not available, standards of an industry recognized organization were used. Calibration of pressurized ionization chambers was performed by the manufacturer.

Analytical and field survey activities were conducted in accordance with procedures from the following documents of the Environmental Survey and Site Assessment Program:

Survey Procedures Manual, Revision 8 (December 1993)

Laboratory Procedures Manual, Revision 9 (January 1995)

Quality Assurance Manual, Revision 7 (January 1995) om si w.. n.nn.n., co . un is. im B-3 h wmrwenxuwi_ man _m mi

i

. l The procedures contained in these manuals were developed to meet the requirements of DOE Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to assess processes during their performance. l Quality control procedures include:

Daily instruraent background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.

Participation in EPA and EML laboratory Quality Assurance Programs.

Training and certification of all individuals performing procedures.

Periodic internal and external audits.

Fort St Ytain-Platteville, CO May 16,1993 E-4 h\emp\repatsut_mm\fut,m 001

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= ;

APPENDIX C REGULATORY GUIDE 1.86, TERMINATION OF OPERATING .

LICENSES FOR NUCLEAR REACTORS l

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l Fm1 St. Vras Ptaneville, Co . May 16, 1995 h \esemp\ reports \st vrain\ fort,vra 001 1

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j U.S. ATOMIC ENERGY COMMISSION June 1974 REGULATORY DIRECTORATE OF REGULATORY STANDARDS GUIDE REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION important to the safety of reactor operation is no longer required. Once this possession-only license is issued, j Section 50.51, " Duration of license, renewal," of 10 reactor operation is not permitted. Other activities CFR Part 50, " Licensing of Production and Utilization from the reactor and placing it in storage (either onsite Facilities,' requires that each license to operate a or offsite) may be continued.

production ar,d utilization facility be issued for a ,

specified duration. Upon expiration of the specified A licensee having a possession-only license must I period, the license may be either renewed or terminated retain, with the Part 50 license, authorization for ,

by the Commission. Section 50.82, " Applications for special nuclear material (10 CFR Part, 70, "Special l Nuclear Material"), byproduct material (10 CFR Part l termination oflicenses," specifies the requirements that must be satisfied to terminate an operating license, 30, " Rules of General Applicability to Licensing of including the requirement that the dismantlement of the Byproduct Material"), and source matehal (10 CFR facility and disposal of the component parts not be Part 40, " Licensing of Source Material"), until the j inimical to the common defense and security or to the fuel, radioactive components, and sources are removed l health and safety of the public. This guide describes from the facility. Appropriate administrative controls j methods and procedures considered acceptable by the and facility requirements are imposed by the Part 50 Regulatory staff for the termination of operating license and the technical specifications to assure that licenses for nuclear reactors. The advisory Committee proper surveillance is performed and that the reactor on Reactor Safeguards has been consulted concerning facility is maintained in a safe condition and not this guide and has concurred in the regulatory position. operated.

B. DISCUSSION A possession-only license permits various options and procedures for decommissioning, such as When a licensee decides to terminate his nuclear mohalling, entombment, or dismantling. The ,

reactor operating license, he may, as a first step in the requirements imposed depend on the option selected.

process, request that his operating license be amended to restrict him to possess but not operate the facility. Sectior4 50.82 provides that the licensee may The advantage to the licensee of converting to such a dismantle and dispose of the component parts of a possession-only ,license is reduced surveillance nuclear reactor in accordance with existing regulations.

requirements in that periodic surveillance of equipment For research reactors and critical facilities, this has USAEC REGULATORY GUIDES c p.a .# pubbered es.d m.y b. obt n.d by . u i indicatino the div i.a dund t. me u.S. At.mi. Enrgy C.mme .n. W h egt.n, o.C. 20646.

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compn.nc. . m is. noi .que..d. Mem d. .nd .wution. def'=.at fr.m th . ..I .ut in the gud.. Wdl b. .cc.pt.b6. sf th.y pr.wid. . b t. f or the hnding. e.gw..t. t. tee m.v.nc. .r continu.nc .f . p.rm.t .t be.n.e by tre

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-...~,....tn..m,.,m.s.,..r.n... ;; =.~, -g,1,,, , =i a-Note: Section electronically reproduced frorn photocopy. C-l

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usually meant the disassembly of a reactor and its minimize releases of radioactivity from the facility, shipment organization for further use. The site from which a reactor has been removed must be c. Any proposed changes to the technical ,

decontaminated, as necessary, and inspected by the specifications that reflect the possession-only facility Commission to determine whether unrestricted access status and the necessary disassembly / retirement !

can be approved. In the case of nuclear power activities to be performed.

reactors, dismantling has usually been accomplished by shipping fuel offsite, maing the reactor inoperable, d. A safety analysis of both the activities to be and disposing of some of the radioactive components, accomplished and the proposed changes to the technical specifications. t Radioactive components may be either shipped off-site for burial at an authorized burial ground or e. An inventory of activated materials and their secured on the site. Those radioactive materials location in the facility, remaining on the site must be isolated from the public by physical barriers or other means to prevent public 2. ALTERNATIVES FOR REACTOR access to hazardous levels of radiation. Surveillance is RETIREMENT necessary to assure the long term integrity of the barriers. The amount of surveillance required depends Four attematives for retirement of nuclear reactor !

upon (1) the potential hazard to the health and safety of facilities are considered acceptable by the the public from radioactive material remaining on the Regulatory staff. These are:

site and (2) the integrity of the physical barriers.

Before areas may be released for unrestricted use, they a. Mothballing. Mothballing of a nuclear reactor must have been decontaminated or the radioactivity facility consists of putting the facility in a state of must have decayed to Ices thu prescribed limits protective storage. In general, the facility may be !

(Tai'- 1). left intact except that all fuel assemblies and the radioactive fluids and waste should be removed The hazard associated with the returned facility is from the site. Adequate radiation monitoring, evaluated by considering the amount 'md type of environmental surveillance, and appropriate security remaining contamination, the degree of confinen nt of procedures should be established under a the remaining radioactive materials, the physical possession-only license to ensure that the health and security provided by the confinement, the susceptibility safety of the public is not endangered.

to release of radiation as a result of natural phenomena, and the duration of required surveillance. b. In-Place Entombment. In-place entombment consists of sealing all the remaining highly

s. C. REGULATORY POSITION radioactive or contaminated components (e.g., the pressure vessel and reactor internals) within a

1. APPLICATION FOR A LICENSE TO POSSESS stmeture integral with the biological shield after BUT NOT OPERATE (POSSESSION-ONLY having all fuel assemblies, radioactive fluids and LICENSE) wastes, and certain selected components shipped offsite. The structure should provide integrity over A request to amend an operating license to a the period of time in which significant quantities possession-only license should be made to the Director (greater than Table l levels) of radioactivity remain of Licensing, U.S. Atomic Energy Commission, with the material in the entombment. An Washington, D.C. 20545. The request should include appropriate and continuing surveillance program ,

the following information: should be established. under a possession-only license. ,

1

a. A description of the current status of the facility. l

c. Removal of Radioactive. Components and j

b. A description of measures that will be taken to Dismantling. All fuel assemblies, radioactive fluids l

prevent criticality or reactivity changes and to and waste, and ot% materials having activities Note: Section electronically reproduced f rom photocopy, C-2 l

. ** o .

' above accepted unrestricted activity levels (Table 1) and access openings, should be inspected at least I should be removed from the site. The facility quarterly to assure that these barriers have not owner may then have unrestricted use of the site deteriorated and that locks and locking apparatus are l with no requirement for a license. If the facility intact. l owner so desires, the remainder of the reactor l facility may be dismantled and all vestiges removed c. A facility radiation survey should be performed and disposed of. at least quarterly to verify that no radioactive material 4 is escaping or being transported through the l

d. Conversion to a New Nuclear System or a containment barriers in the facility. Sampling should ,

Fossil Fuel System. This alternative, which applies be done along the most probable path by which !

I only to nuclear power plants, utilizes the existing radioactive material such as that stored in the inner turbine system with a new steam supply system. containment regions could be transported to the outer The original nuclear steam supply system should be regions of the facility and ultimately to the environs.

separated from the electric generating system and disposed of in accordance with one of the previous d. An environmental radiation survey should be three retirement alternatives. performed at least semiannually to verify that no significant amounts of radiation have been released to

3. SURVEILLANCE AND SECURITY FOR Tile the environment from the facility. Samples such as RETIREMENT ALTERNATIVES WIIOSE soil, vegetation, and water should be taken at locations FIN AL STATUS REQU1RES A for which statistical data has been established during POSSESSION-ONLY LICENSE reactor operations.

A facility which has been licensed under a e. A site representative should be designated to be possession-only license may contain a significant responsible for controlling authorized access into and amount of radioactivity in the form of activated and movement within the facility.

contaminated hardware and structural materials.

Surveillance and commensurate security should be f. Administrative procedures should be established provided to assure that the public health and safety are for the notification and reporting of abnormal not endangered, occurrences such as (1) the entrance of an unauthorized

a. Physical security to prevent inadvertent exposure person or persons into the facility and (2) a significant of personnel should be provided by multiple locked change in the radiation or contamination levels in the barriers. The presence of these barriers should make facility or the offsite environment.

it extremely difficult for an unauthorized person to gain access to areas where radiation or contamination levels g. The following reports should be made:

exceed those specified in Regulatory Position C.4. To prevent inadvertent exposure, radiation areas above (1) An annual report to the Director of 5 mR/hr, such as near the activated primary system of Licensing, U.S. Atomic Energy Commission, a power phmt, should be appropriately marked and Washington, D.C. 20545, describing the results of the should not be accessible except by cutting of welded environmental and facility radiation surveys, the status closures or the disassembly and removal of substantial of the facility, and an evaluation of the performance of stmetures and/or shielding material. Means such as a security and surveillance measures.

remote-readout intrusion alarm system should be provided to indicate to designated personnel when a (2) An abnormal occurrence report to the physical barrier is penetrated. Security personnel that Regulatory Operations Regional Office by telephone provide access control to the facility may be used within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery of an abnormal instead of the physical barriers and the intrusion alarm occurrence. The abnormal occurrence will also be systems. reported in the annual report described in the preceding item.

b. The physical barriers to unauthorized entrance !

ir.to the facility, e g., fences, buildings, welJed doors, h. Records or logs relative to the following items Note: Section electronicaHy reproduced from photocopy. C-3 )

should be kept and retained until the license is licensee to relinquish possession or control of premises, terminated, after which they must be stored with other equipment, or scrap having surfaces contaminated in plant records: excess of the limits specified. This may include, but is not limited to, special circumstances such as the (1) Environmental surveys, transfer of premises to another licensed organization (2) Facility radiation surveys, that will continue to work with radioactive materials.

Requests for such authorization should provide:

(3) Inspections of the physical barriers, and (4) Abnormal occurrences. (1) Detailed, specific information describing the premises, equipment, scrap, and radioactive

4. DECONTAMINATION FOR RELEASE FOR contaminants and the nature, extent, and degree of UNRESTRICTED USE residual surface contamination.

If it is desired to terminate a license and to (2) A detailed health and safety analysis indicating eliminate any further surveillance requirements, the that the residual amounts of materials on surface areas, facility should be sufficiently decontaminated to prevent together with other considerations such as the risk to the public health and safety. After the prospective use of the premises, equipment, or scrap, decontamination is satisfactorily accomplished and the are unlikely to result in an unreasonable risk to the ,

site inspected by the Commission, the Commission may health and safety of the public. ,

authorize the license to be terminated and the facility abandoned or released for unrestricted use. The e. Prior to release of the premises for unrestricted licensee should perform the decontamination using the use, the licensee should make a comprehensive following guidelines: radiation survey establishing that contamination is within the limits specified in Table 1. A survey report

a. The licensee should make a reasonable effort to should be filed with the Director of Licensing, U.S.

eliminate residual contamination. Atomic Energy Commission, Washington, D.C. 20545, with a copy to the Director of the Regulatory

b. No covering should be applied to radioactive Operations regional Office having jurisdiction. The surfaces of equipment of structures by paint, plating, or report should be filed at least 30 days prior to the other covering material until it is known that planned date of abandonment. The survey report contamination levels (determined by a survey and should:

documented) are below the limits specified in Table 1.

In addition, a reasonable effort should be made (and (1) Identify the premises; documented) to further minin.ize contamination prior to l any such covering. (2) Show that reasonable effort has been made to ;

reduce residual contamination to as low as practicable

c. ' Die radioactivity of the interior surfaces of levels; pipes, drain lines, or ductwork should be determined by making measurements at all traps and other (3) Describe the scope of the survey and the general appropriate access points, provided contamination at procedures followed; and these locations is likely to be representative of contamination on the interior of the pipes, drain lines, (4) State the finding of the survey in units specified or ductwork. Surfaces of premises, equipment, or in Table 1.

scrap which are likeiy to be contaminated but are of such size, construction, or location as to make the After review of the report, the Commission may '

surface inaccessible for purposes of measurement inspect the facilities to confirm the survey prior to should be assumed to be contaminated in excess of the granting approval for abandonment, permissible radiation limits.

d. Upon request, the Commission may authorize a 5. REACTOR RETIREMENT PROCEDURES Note: Section electronically reproduced f rom photocopy. C-4

- As indic:ted in Regulatory Position C.2, several ;

attematives are acceptable for reactor facility !

retirement. If minor disassembly or "mothballing" is ;

planned, this could be done by the existing operating l and maintenance procedures under the license in effect. !

Any planned actions involving an unreviewed safety ,

question or a change in the technical specifications i should be reviewed and approved in accordance with the requirements of 10 CFR i 50.59.

If major structural changes to radioactive components of the facility are planned, such as removal of the pressure vessel or major components of the primary system, a dismantlement plan including the information required by { 50.82 should be submitted to the Commission. A dismantlement plan should be submitted for all the alternatives of Regulatory Position C.2 except mothballing. However, minor disassembly activities may still be performed in the absence of such a plan, provided they are permitted by existing operating and maintenance procedures.

A dismantlement plan should include the following:

a. A description of the ultimate status of the facility

b. A description of the distrantling activities and the precautions to be taken.

c. A safety analysis of the dismantling activities including any effluents which may be released.

d. A safety analysis of the facility in its ultimate status.

L Upon satisfactory review and approval of the ;

dismantling plan, a dismantling order is issued by the Commission in accordance with I 50.82. When dismantling is completed and the Commission has been notified by letter, the appropriate Regulatory Operations Regional Office inspects the facility and vedfies completion in accordance with the dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the Commission may terminate the license. If possession-only license under

, which the dismantling activities have been conducted or, as an alternative, may make application to the State (if an Agreement State) for a byproduct materials license. i i

i Note: Section electronicany reproduced from rihotocopy. C-5

e '. ' '

I

. 4 l

I TABLE 1 i ACCEI' FABLE SURFACE CONTAMINATION LEVELS l Nuclide* Average'" Maximuned Removable6 '

U-nat, U-235, U-238, and associated decay products 5,000 dpm a/100 cm2 15,000 dpm a/100 cm2 1,000 dpm a/100 cm2 Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, Ac-227,1-125, I-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm 2 Th nat, Th-232, Sr-90, Ra-223, Ra 224, U 232,1-126, I-131, I-133 1,000 dpm/100 cm2 3,000 dpm/100 cm2 200 dpm/100 cm2 )

i Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except St-90 and others noted above. 5,000 dpm Sy/100 cm2 15,000 dpm Sy/100 cm2 1,000 dpm Sy/100 cm2

'Where surface contamination by both alpha- and beta-gamma-emitting nuclides exists, the limits established for alpha- and beta- gamn.4-emitting nuclides should apply independently.

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

' Measurements of average contaminant should not be averaged over more than 1 square meter. For objects ofless surface area, the average should be derived for each such object.

'The maximum contamination level applies to an area of not more than 100 cm2 .

'The amount of removable radioactive material per 100 cm 2of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.

Note: Section electronically reproduced from photocopy. C-6

ML20084F393

Text

SUMMARY

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