ML20056H444

Confirmatory Survey of Univ of Kansas Lawrence,Ks, Final Rept

ML20056H444
Person / Time
Site:	05000148
Issue date:	07/31/1993
From:	Ansari A, Beck W, Landis M OAK RIDGE ASSOCIATED UNIVERSITIES
To:
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ML20056H386	List: ML20056H385 ML20056H444
References
ORISE-93-G-8, NUDOCS 9309090327
Download: ML20056H444 (42)

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CONFIRMATORY SURVEY -

OFTHE UNIVERSITY OF KANSAS TRAINING REACTOR UNIVERSITY OF KANSAS LAWRENCE, KANSAS

[ DOCKET 50-148]

A.J. ANSARI AND J. L. PAYNE Preparea fcr the U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation

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ORISE 93/G-8 0 CONFIRMATORY SURVEY OF TIIE UNIVERSITY OF KANSAS TRAINING REACTOR UNIVERSITY OF KANSAS LAWRENCE, KANSAS l Prepared by l A. J. Ansari and J. L. Payne Environmental Survey and Site Assessment Program Energy / Environment System Division Oak Ridge Institute for Science and Education Oak Ridge, Tennessee 37831-0117 I

Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1993 FINAL REPORT I

J I

i This report is based on work performed under an Interagency Agreement (NRC Fin. No.

l A-9093) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy.

l- Oak Ridge Institute for Science and Education performs complementary work under contract l number DE-AC-05-760R00033 with the U.S. Department of Energy.

i tamiry or r- Tmise unhty s.1993 i

i

CONFIRMATORY SURVEY OF TIIE UNIVERSITY OF KANSAS TRAINING REACTOR UNIVERSITY OF KANSAS LAWRENCE, KANSAS Prepared by: , Date: 7 / 8 / 73 A. J. Ansari, f(oject Leader Environmental Survey and Site Assessment Program

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Reviewed by: Date:# 7 2[95 W. C. Beck," Acting Laboratory Manager Environmental Survey and Site Assessment Program Reviewed by: o Date:

N Af. R. Landis, Project Manager Environmental Survey and Site Assessment Program Reviewed by: Date:

A. T. Payne, Quality Assurance Officer Environmental Survey and Site Assessment Program Reviewed by: Date: 2/f/93 J. Berger, Program $[ rector /'

E ironmental Survey and Site Assessment Program um.tv-im mu s.ms r

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

FIELD STAFF J. M. Burton LABORATORY STAFF R. D. Condra J. S. Cox M. J. Laudeman S. T. Shipley CLERICAL STAFF T. T. Claiborne D. A. Cox R. D. Ellis M. S. Perry K. E. Waters ILLUSTRATOR E. A. Powell i

/

t Univemay of Kanses Tsuming Reactor-July 8,1993

TABLE OF CONTENTS PAGE List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii List of Tables ..............................................iii Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv Acronym s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Site Description ............................................. 2 Obj ectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Document Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Procedures ................................................ 3 Sample Analysis and Data Interpretation . . . . . . . . . . . ............ 5 Findings and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 S um mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 References ...............................................16 Appendices:

Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors.

AND Guidelines for Residual Concentrations of Thorium and Uranium Wastes in Soil.

Univerany of Kansas Training ReactorJ.dy 8.1993 i

LIST OF FIGURES PAGE FIGURE 1: State of Kansas - Location of the University of Kansas at Lawrence, Kansas ........................ 9 FIGURE 2: Location of Burt Hall (the Former Nuclear Reactor Center) on the Campus of the University of Kansas . . . . . . . . . . . . . . . . . . . 10 FIGURE 3: Burt Hall, First Floor - Location of the Former Training Reactor Facility ...................................11 FIGURE 4: Reactor Room - The Reference Grid System . . . . . . . . . . . . . . . . . . 12 FIGURE 5: Reactor Room - Measurement and Sampling Locations . . . . . . . . . . . 13 i

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LIST OF TABLES PAGE TABLE 1: Exposure Rate Measurements . . . . . . , , , , , , , , , , , , , , , , , , , , 34 TABLE 2: Radionuclide Concentrations in Miscellaneous Samples ............ 15 l

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Unm# of h Trairdng ReessorJuh 8,3993 ((j

ABBREVIATIONS cm2 square centimeter epm counts per minute dpm/100 cm2 disintegrations per minute /100 square centimeters ft foot GM Geiger-Mueller kg kilogram m meter m2 square meter NaI Sodium Iodide pCi/g picoeurie per gram PIC Pressuriz.ed Ionization Chamber R/h microroentgen per hour ZnS Zine Sulfide i

University of Kanses Tmining ReactorJuly 8.1993 IV

ACRONYMS ASME American Society of Mechanical Engineers CWM Chemical Waste Management, Inc.

EPA Environmental Protection Agency EML Environmental Measurement Laboratory ESSAP Environmental Survey and Site Assessment Program KU Kansas University KUTR Kansas University Training Reactor MDA Minimum Detectable Activity NIST National Institute for Standards Technology NRC Nuclear Regulatory Commission ORISE Oak Ridge Institute for Science and Education-Universky of Kansas Traming ReactorJuly 8,1993 V

(..- ___ _ _ __ _ _ _ _ _

l CONFIRMATORY SURVEY OF THE UNIVERSITY OF KANSAS TRAINING REACTOR UNIVERSITY OF KANSAS LAWRENCE, K!.NSAS INTRODUCTION AND SITE HISTORY The Kansas University Training Reactor (KUTR) was a light water-moderated and light water-cooled pool-type reactor. It was designed and built by the Bendix Aviation Corporation and started operations in June 1961. Although initially authorized to operate at power levels up to 10 kw, the reactor was authorized in 1971 to operate at 250 kw for short periods. The reactor f

was operated under U.S. Nuclear Regulatory Commission (NRC) License R-78 (Docket 50-148),

and was used primarily for the purpose of research, training, and demonstrations. The reactor core contained 2.5 kg 93% enriched uranium. The reactor ceased all operations in June 1984.

There are no records of any major incidents involving releases of radiological material during the operating history of the reactor.

As part of the decommissioning activities, in January and February 1986, all unirradiated nuclear fuel element assonblies were shipped to the Oak Ridge National Laboratory in Oak Ridge, Tennessee. All irradiated nuclear fuel elements were shipped to the Savannah River Plant in

, Barnwell, South Carolina. The entire core support structure and the source and fission chambers were also removed. At this point, the residual radioactivity at the facility was in the activated (

components of the reactor including the reactor aluminum tank wall, portions of the core support structure, and the shielding concrete. The activated control rods and sheaths, grid plate assembly, some aluminum angles, and stainless steel bolts were stored in a shielded vault in a laboratory designated for radioactive materials storage.

In December 1990, the University requested authorization to dismantle the research reactor facility in accordance with a decommissioning plan submitted to the NRC. In September 1991, the NRC issued the order authorizing the dismantling and disposincn of component parts of the facility. Components of the reactor that were outside the biological concrete shield of the reactor (e.g., control console with electronics, drive motors, etc.) were surveyed by the University of Kanans Trahing Reactor-My 8,1993

University of Kansas Radiation Safety Service and recycled. In September and October 1992,.

The Chemical Waste Management, Inc. (CWM), Nuclear Remedial Services, contracted by the University of Kansas, removed the reactor tank, the activated concrete, the activated portions of the beam ports and pneumatic tubes, and other miscellaneous materials including control rod tubes, polyethylene shielding, lead bricks, graphite moderator blocks, and assorted bagged waste materials.

The licensee performed a final radiological survey in November 1992 and provided a final report to the NRC.' The primary contaminants at this facility, based on the licensee's analysis, were  !

Co-60 and Eu-152 - products of neutron activation due to reactor operations. The U.S. Nuclear Regulatory Commission, Region IV Office, requested that the Environmental Survey and Site  !

Assessment Program (ESSAP) of Oak Ridge Institute for Science and Education (ORISE) perform an independent confirmatory survey of this facility. This report summarizes the procedures and results of that survey.  ;

SITE DESCRIPTION Lawrence is situated on the banks of the Kansas River in the northeast corner of the state of Kansas (Figure 1). The facility is located in Burt Hall (formerly, the Nuclear Reactor Center) on the western edge of the main campus of the University of Kansas at Lawrence (Figure 2).

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The reactor room has about 200 m' of floor space. The middle of the concrete floor of the reactor room has been excavated. Adjacent to the reactor high bay area, there are six rooms that are offices, classrooms, and a storage room in which radioactive materials are stored (Figure 3). None of these adjacent rooms, or any area outside Burt Hall were within the scope of the decommissioning activities. However, at the request of the NRC, limited survey activities were performed in these areas.

University of Kannes Traming Reactor July 8,1993 2

OBJECTIVES The objectives of the confirmatory process were to provide independent document reviews and radiological data, for use by the NRC in evaluating the adequacy and accuracy of the licensee's radiological status repon, relative to established guidelines.

DOCUMENT REVIEW As pan of the confirmatory activities ESSAP reviewed the licensee's radiological survey data.'

Analytical procedures and methods utilized by the licensee were reviewed for adequacy and appropriateness. The data were reviewed for accur.cy, completeness, and compliance with guidelines.

l PROCEDURES On April 8 and 9,1993, ESSAP performed a confirmatory survey of the University of Kansas f

Training Reactor on the campus of the University of Kansas at Lawrence. The survey was conducted in accordance with a survey plan which was submitted to and approved by the NRC Region IV Office.2 REFERENCE GRID The reference grid system established by the licensee was used by ESSAP to reference measurement and sampling locations (Figure 4).

SURFACE SCANS Surface scans for alpha, beta, and gamma activity were performed on floors and lower walls (up to 2 m), using NaI and large area gas proportional detectors coupled to ratemeters and ratemeter-scalers with audible indicators. Three drains in the reactor room and one drain in the hot lab, Unhersi y of Kansas Training ReactorJuly 8.1993 3

\

located to the northwest corner of the teactor room were also scanned. Cursory scans were performed in rooms adjacent to the reactor room.

SURFACE ACTIVITY MEASUREMENTS Direct measurements to determine total alpha and total beta surface activity were performed in 36 randomly selected grid blocks on the floor and lower walls. These measurements were performed using ZnS scintillation and thin-window GM detectors, coupled to ratemeter-scalers.

Measurements were performed at the center and at four points equidistant from the center and grid block corners. Four single-point measurements were performed on upper wall surfaces.

A smear sample for determining removable activity was obtained from each grid block, at the location corresponding to the maximum total activity and frcm each single-point measurement location. Smear samples for determining H-3 and C-14 activity were collected from the center point of 10 of the randomly selected grid blocks. Measurement and sampling locations for total and removable activity are illustrated in Figure 5.

EXPOSURE RATE MEASUREMENT Background exposure rate measurements were performed at 3 locations within Burt Hall.

One measurement was made in Room 201 and two measurements in Room 200. Both rooms have similar construction as the reactor facility. Room 201 has no history of radioactive material use and is used as a conference room. Room 200 has a history of use of low energy beta-emitting isotopes. However, this room has been surveyed, remodeled and is currently being used as student offices.

Exposure rate measurements were performed at I m above the interior building surfaces at 5 locations in the reactor room and 2 locations in the adjacent areas using a pressurized ionization chamber (PIC). Measurement locations in the reactor room are shown in Figure 5.

University of Kannes Training Rescur4uly 8.1993 4

MISCELLANEOUS SAMPLING 1

Six soil samples were collected from this facility. Three of these samples were taken from the excavated area in the middle of the reactor room. One soil sample was taken at the center of the area where the demineralizer was formerly located. Two soil samples were taken from the crawlspace underneath the reactor room.

A concrete sample was taken at the south end of the excavated area in the reactor room.

Sampling locations are shown in Figure 4.

SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and survey data were returned to the ESSAP Oak Ridge labomtory for analyses and interpretation. Smears v: e analyzed for gross alpha and gross beta or H-3/C-14 activity. Soil and concrete samples were analyzed by gamma spectrometry. Spectra were reviewed for U-235, U-238, Co-60, and Eu-152 and any other identifiable photopeaks. Direct measurement and 2

smear data were converted to units of disintegrations per minute per 100 cm (dpm/100 cm),

2 and exposure rate measurements were reported in microroentegens per hour ( R/h). Soil and concrete sample results were reported in units of picoeuries per gram (pCi/g). Additional information concerning major instrumentation, sampling equipment, and analytical procedures is provided in Appendices A and B. Results were compared to NRC guidelines which are provided in Appendix C.

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FINDINGS AND RESULTS

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DOCUMENT REVIEW ESSAP reviewed the licensee's radiological survey data and comments were provided to the NRC.' In response to ESSAP's comments, the licensee provided additional information and clarification, regarding site conditions, survey procedures, and data." In ESSAP's opinior, the University of Kennen Training ReactorJuly 8.1993 5

licensee's documentation provides an adequate description of the radiological condition of the facility, relative to the NRC guidelines for release to unrestricted use.

SURFACE SCANS Surface scans for alpha, beta, and gamma activity on the floor and lower walls did not identify any areas of elevated direct radiation.

Three drains, in grid blocks F43, F75, and F115, and one drain in the hot lab, located to the northwest corner of the reactor room, were scanned. No locations of elevated direct radiation were noted.

I SURFACE ACTIVITY LEVELS I

Results of total activity measurements were all less than the detection limits of the procedure l which were 110 dpm/100 cm2 and 2400 dpm/100 cm2 for alpha and beta, respectively.

Removable activity levels were less than the minimum detectable activity of the procedure which l

2 were 12 dpm/100 cm for alpha,17 dpm/100 cm 2

for beta, 6 dpm/100 cm 2 for H-3, and 5 dpm/100 cm2 for C-14.

EXPOSURE RATES 1

Background exposure rates ranged from 11 to 13 uR/h and averaged 12 rR/h. Exposure rate measurements from the 7 locations in the reactor facility ranged from 11 to 13 uR/h (Table 1).

MISCELLANEOUS SAMPLES Radionuclide concentrations for the soil and concrete samples, collected from the reactor facility, are presented in Table 2. The concentration of Co-60 in all samples was s 0.1 pCi/g. The concentration of Eu-152 ranged from <0.2 to 0.8 pCi/g. The concentration of U-235 was l

0.1 pCi/g in all samples. The concentration of U-238 ranged from 0.4 to 1.4 pCi/g.

I University of Karmas Traudng Reactor-July 8,1993 6 L . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

COMPARISON OF RESULTS WITII GUIDELLNES The NRC guidelines for surface contamination and residual concentrations of radionuclides in soil, established for license termination or release of a facility for unrestricted use, are presented in Appendix C.*3 The major contaminants identified were Co-60 and Eu-152. The applicable guidelines are those for beta-gamma emitters (radionuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 are:

Total Activity 5,000 dpm S-y/100 cm2 , averaged over a 1 m2 area 15,000 dpm B y/100 cm2 , maximum in a 100 cm2 area Removable Activity 1000 dpm # y/100 cm2 The surface contamination guidelines for uranium are:

Total Activity 5,000 dpm a/100 cm2 , total, averaged over a 1 m2 area 15,000 dpm a/100 cm2 , total, maximum in a 100 cm2 area Removable Activity 1,000 dpm a/100 cm2 All surface activity measurements were within the guideline levels.

l The NRC guideline for exposure rate at 1 m above surface is 5 uR/h above background.'

Exposure rates measured in the reactor room were all within this limit. The soil concentration guideline for enriched uranium is 30 pCi/g.' Based on a conservative U-234: U-235 ratio of 30 to 1, the highest total uranium concentration in the samples collected (sample #5, southwest corner of crawlspace) is 4.5 pCi/g which is well below the 30 pCi/g limit.

University of Kansas Training ReactorJuly 8, IW3 7

There are no specific concentration guidelines for Co-60 and Eu-152 in soil.

SUMMARY

On April 8 and 9,1993, ESSAP performed a confirmatory survey of the University of Kansas Training Reactor on the campus of the University of Kansas at Lawrence, Kansas. Survey activities included document reviews, surface scans, measurements of tot.?1 and removable surface activity, exposure rate measurements, and soil and concrete sampling.

In ESSAP's opinion, the licensee's documentation provides an adequate description of the radiological condition of the reactor facility. The ESSAP confirmatory measurements support the licensee's conclusion that the facility satisfies NRC's guidelines for release to unrestricted use.

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University of Kansas Trainirs Reuur-July 8.1993 }3

TABLE 1 EXPOSURE RATE MEASUREMENTS UNIVERSITY OF KANSAS TRAINLNG REACTOR UNIVERSITY OF KANSAS LAWRENCE, KANSAS Location Exposure Rate (uR/h)

Facility Exposure Rates Reactor Room, Grid Block F38 12 Reactor Room, Grid Block F81 13 Reactor Room, Grid Block F116 11 Reactor Room, Grid Block F183 12 Reactor Room, Pit Area 11 Engineering Research Lab 11 Hallway adjacent to Hot Lab 13 Background Exposure Rates Room 200, Second Floor, Burt Hall 12 North End Room 200, Second Floor, Burt Hall 11 !

South End j Room 201, Second Floor, Burt Hall '13 u%.rx 7.ww.uy s im 14 f

t _ - _ _ _ - _ _ - - - - _ _ - _ _ _ _-__ - _ _ _ -

TABLE 2 RADIONUCLIDE CONCENTRATIONS LN MISCELLANEOUS SAMPLES UNIVERSITY OF KANSAS TRAINLNG REACTOR UNIVERSITY OF KANSAS LAWRENCE, KANSAS Radionuclide Concentrations (pCi/gP Co-60 Eu-152 U-235 U-238

1. 1-East Excavation Pit 0.1 i 0.1 0.8 0.2 0.1 i 0.1 0.4 0.5

1. 2-West Excavation Pit < 0.1 0.3 0.1 0.1 i 0.1 1.1 0.7

1. 3-South End Dividing Wall of Excavation < 0.1 0.6 0.2 0.1 i 0.1 0.5 i 0.4

1. 4-Center of Demineralizing Pit < 0.1 < 0.2 0.1 i 0.1 0.8 0.7

1. 5-Southwest Corner of Crawlspace < 0.1 < 0.2 0.1 0.1 1.4 0.8

1. 6-West Foundation of Crawlspace < 0.1 <0.2 0.1 i 0.1 1.2 i 0.7 Concrete <0.1 0.7 i 0.3 0.1 0.1 1.0 0.5

' Refer to Figure 4.

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

I I

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L w.4 a x T, n, m ciora iy i. i993 15

REFERENCES

1. University of Kansas, Lawrence, Kansas, " Final Report for Kansas University Training Reactor Decommissioning", December 8,1992.

2. Oak Ridge Institute for Science and Education, " Confirmatory Radiological Survey Plan for the University of Kansas Training Reactor, Lawrence, Kansas", March 25, 1993.

{

3. Letter from A. Jaberaboansari (ORISE) to B. Murray (NRC Region IV),

Reference:

" Final Report for Kansas University Training Reactor Decommissioning," March 8,

[ 1993.

4. Letter from J. Stevens (CWM) to M. Lemon (KU), dated March 23,1993.

{

5. Letter from M. Lemon (KU), provided to W. Holley (NRC) and A. Jaberaboansari (ORISE),

Reference:

" Clarifications, Corrections, and Additional Documentation in

{

Response to Questions and Comments by ORISE/NRC", April 8,1993.

6. U.S. Nuclear Regulatory Commission, " Termination of Operating Licenses for Nuclear

[

Reactors," Regulatory Guide 1.86, Washington D.C., June 1974.

[ 7. U.S. Nuclear Regulatory Commission, " Disposal or On-site Storage of Thorium and Uranium Wastes from Past Operations," 46 FR 52061, Washington, D.C., October 23, 1981.

8. U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards,

" Review Plan: Evaluating Decommissioning Plans for Licenses Under 10 CFR Parts 30,

[ 40, and 70," Washington, D.C.1991.

[

[

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Univernay of Karanas Trairung Reactor-July 8,1993 16 r

APPENDIX A MAJOR INSTRUMENTATION University of Kansas Training AwJOy 8.1993

APPENDIX A MAJOR INSTRUMENTATION 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)

Eberline " Rascal" Ratemeter-Scaler Model PRS-1 (Eberline, Santa Fe, NM)

Ludlum Floor Monitor Model 239-1 (Ludlum Measurements, Inc.,

Sweetwater, TX)

Ludlum Ratemeter-Scaler Model 2221 (Ludlum Measurements, Inc.,

Sweetwater, TX)

Detectors Eberline GM Detector Model HP-260 Effective Area,15.5 cm2 I (Eberline, Santa Fe, NM) i Eberline ZnS Scintillation Detector Model AC-3-7 Effective Area,59 cm2 (Eberline, Santa Fe, NM)

Ludlum Gas Proportional Detector Model 43-37 Effective Area,550 cm 2 (Ludlum Measurements, Inc.,

Sweetwater, TX) vs wyorx T % w ay . ms A-1

Ludlum Gas Proportional Detector Model 43-68 Effective Area,100 cm2 (Ludlum Measurements, Inc.,

Sweetwater, TX)

Reuter-Stokes Pressurized Ionization Chamber Model RSS-Ill (Reuter-Stokes, Cleveland, OH)

Victoreen NaI Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal (Victoreen, Cleveland, OH)

LABORATORY ANALYTICAL INSTRUMENTATION High Purity Extended Range Intrinsic Detectors i

Model No: ERVDS30-25195 (Tennelec, Oak Ridge, TN)

Used in conjunction with:

Lead Shield Model G-ll (Nuclear Lead, Oak Ridge, TN) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT)

High-Purity Germanium Detector Model GMX-23195-S,23% Eff. 1 (EG&G ORTEC, Oak Ridge, TN)

Used in conjunction with:

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

Low 3ackground Gas Proportional Counter Model LB-5100-W (Oxford, Oak Ridge, TN)

Tri-Carb Liquid Scintillation Analyzer Model 1900CA (Packard Instrument Co., Meriden, CT) vs,.o or r- T% ww uy . im A-2

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( APPENDLX B SURVEY AND ANALYTICAL PROCEDURES

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Univ mityorr.m Tndahs Rsackw-My 8.1993

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APPENDLX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES Surface Scans Strface 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. A

{

large surface area, gas proportional floor monitor was used to scan the floors of the surveyed areas. Other surfaces were scanned using small area (100 cm') hand-held detectors.

[

Identification of elevated levels was based on increases in the audible signal from the recording and/or indicating instrument. Combinations of detectors and instniments used for the scans were:

Alpha-Beta -

gas proportional detector with ratemeter-scaler

[

Gamma -

NaI scintillation detector with ratemeter Surface Activity Measurements Measurements of total alpha and beta activity levels were performed using ZnS scindllation and GM detectors with ratemeters-scalers.

[- Count rates (epm), which were integrated over 0.5 minute in a static position, were converted to activity levels (dpm/100 cm') by dividing the net rate by the 4r efficiency and correcting for the active area of the detector. The alpha activity background countrates for the ZnS scintillation detectors averaged I cpm for each detector. Alpha efficiency factors ranged from 0.18 - 0.19 for the ZnS scintillation detectors. The betn activity background count rates for the GM detectors averaged 57 epm. Beta efficiency factors ranged from 0.15 - 0.17 for the GM-c e -ayor w w e , w w w . m3 B-1

f

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detector. The effective windows for the ZnS scintillation and GM detectors were 59 cm' and f 15.5 cm2 , respectively.

[ Removable Activity Measurements

( Removable activity levels were determined using numbered filter paper disks, 47 mm in diameter. Moderate pressure was applied to the smear and approximately 100 cm' of the surface

{

was wiped. Smears were placed in labeled envelopes with the location and other pertinent information recorded. At some measurement locations, smear samples for H-3/C-14 analysis

{

were obtained.

[

Exposure Rate Measurements Measurements of gamma exposure rates were performed at I m above surface using a pressurized ionization chamber (PIC).

(

( Miscellaneous Samoles

( Soil Sampling Approximately I kg of soil was collected at each sample locatior. Collected samples were placed in a plastic bag, sealed, and labeled in accordance with ESSAP survey procedures.

Concrete Sampling The concrete sample was taken by chipping material from approximately 100 cm2 of surface.

The sample was placed in a plastic bag, sealed, and labeled in accordance with ESSAP survey procedures.

v%ax 7% way s. im B-2

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ANALYTICAL PROCEDURES Removable Activity Gmss Alpha / Beta Smears were counted on a low background gas proportional system for gross alpha and gross beta activity.

Liquid Scintillation Smears were counted in a liquid scintillation counter for low-energy beta activity to determine H-3 and C-14 activity.

Miscellaneous Samples Gamma Spectrometry

( Solid Samoles Samples of solid material (soil and construction material) were dried, mixed, crushed, and/or homogenized as necessary, and a portion sealed 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 u:ing b 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. Energy peaks used for determination of radionuclides of conec 3re:

Co-60 1.173 MeV Eu-152 0.344 MeV University of Kansas Traming Rendor-July 8,1993 B-3

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

• Secular equilibrium assumed.

[ Spectra were also reviewed for other identifiable photopeaks.

UNCERTAINTIES AND DETECTION LIMITS The uncertainties associated with the analytical dats presented in the tables 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 repon.

Detection limits, referred to as minimum detectable activity (MDA), were based on 2.71 plus 4.66 times the standard deviation of the background count. 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

{

from other radionuclide in samples, the detection limits differ from sample to sample and instrument to instrument. .

l l

CALIBRATION AND QUALITY ASSURANCE l

Analytical and field survey activities were conducted in accordance with procedures from the following ESSAP documents:

Survey Procedures Manual, Revision 7 (May 1992)

Laboratory Procedures Manual, . Revision 7 (April 1992)

Quality Assurance Manual, Revision 5 (May 1992) ]

i l

University of Kansas Trmning Renaar4uly 8.1993 3-4

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 a.;sess processes during their performance.

{

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 was used. Calibration of pressurized ionization chambers was performed by the manufacturer.

Quality control procedures include:

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

Participation in EPA and EML laboratory Quality Assurance Progra ns.

Training and certification of all individuals performing procedures.

Periodic internal and external audits.

(

)

s 1

I University of Kansas Trsming RamoorJuly 8,1993 B-5 l _ _ _ _ _ _ _ - _ _ _ _ - _ - - - _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ . .

APPENDLX C REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS AND 1

GUIDELINES FOR RESIDUAL CONCENTRATIONS OF THORIUM AND URANIUM WASTES IN SOIL 4

University of Kansas Tmining ReactorJuly B,1993

U.S. ATOMIC ENERGY COMMISSION Juno 1974 REGULATORY GUIDE DIRECTORATE OF REGULATORY STANDARDS

{ REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION A licensee having a possession-only license must retain, with the Part 50 license, authorization for Section 50.51, " Duration oflicense, renewal," of 10 special nuclear material (10 CFR Part, 70, "Special CFR Part 50,

• Licensing of Production and Utilization Nuclear Mateuai"), byproduct material (10 CFR Part

[

Facilities," requires that each license to operate a 30, ' Rules of General Applicability to Licensing of L production and utilization facility be issued for a Byproduct Material"), and source material (10 CFR specified duration. Upon espiration of the specified Part 40,

• Licensing of Source Material'), until the period, the license may be either renewed or terminated fuel, radioactive components, and sources are removed f by the Commission. Section 50.82, "Appheations for from the facility. Arpropriate administrative controls termination oflicenses," specifies the requirements that and facility requirements are imposed by the Part 50 must be satisfied to terminate an operating license, license and the technical specifications to assure that including the requirement that the dismantlement of the proper surveillance is performed and that the reactor facility and disposal of the component parts not be facility is maintained in a safe condition and not inimical to the common defense and security or to the operated.

health and safety of the public. This guide describes methods and procedures considered acceptable by the A possession-only license permits various options Regulatory staff for the termination of operating and procedures for decommissioning, such as licenses for nuclear reactors. The advisory Committee mothba!!ing, entombment, or dismantling. The on Reactor Safeguards has been consulted concerning requirements imposed depend on the option selected.

this guide and has concurred in the regulatory position.

SectNn 50.82 provides that the licensee may

[ B. DISCUSSION dismantis and dispose of the component parts of a t nuclear reactor in acco &re with existing regulations.

When a licensee decides to terminate his nuclear For research reactors and critical facilitier, this has reactor operating license, he may, as a first step in the usually meant the disassembly of a reactor and its process, request that his operating license be amended shipment organization for further use. The site from to restrict him to possess but not operate the facility. which a reactor has been removed must be The advantage to the licensee of converting to such a decontaminated, as necessary, and inspected by the possession-only license is reduced surveillance Commission to determine whether unrestricted access requirements in that periodic surveillance cf equipment can be approved. In the case of nuclear power important to the safety of reactor operation is no longer reactors, dismantling has usually been accomplished by required. Once this possession-only license is issued, shipping fuel offsite, mahng the reactor inoperable, reactor operation is not permitted. Other activities and disposing of some of the radioactive components.

from the reactor and placing it in storage (either onsite or offsite) may be continued.

USAEC REGULATORY GUCES c,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,y,,3,,,,,

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Rcdiocctive components may be eiths shipped Four alterretives for retirement of nuclear reactor off-site for burial at an authorized burial ground or facilities are ccasidered acceptable by the secured on the site. Those radioactive materials Regulatory staff. These are:

remmmng on the site must be isolated from the public by physical barriers or other means to prevent public a. Mothballing. Mothballing of a nuclear reactor access to hazardous levels of radiation. Surveillance is facility consists of putting the facility in a state of necessary to assure the long term integrity of the protective storage. In general, the facility may be barriers. The amount of surveillance required depends left intact except that all fuel assemblies and the upon (1) the potential hazard to the health red safety of radioactive fluids and waste should be removed the public from radioactive material remammg on the from the site. Adequate radiation monitoring, site and (2) the integrity of the physical barriers. environmental surveillance, and appropriate security Before areas may be released for unrestthted use, they procedures should be established under a

} must have been decontaminated or the radioactivity possession-only license to ensure that the health and l must have decayed to less than prescribed limits safety of the public is not endangered.

(Table 1).

b. In-Place Entombment. In-place entombment The hazard associated with the returned facility is consists of sealing all the remaining highly evaluated by considering the amount and type of radioactive or contammated components (e.g., the remaining contamination, the degree of confinement of pressure vessel and reactor internals) within a the remaining radioactive materials, the physical structure integral with the biological shield after security provided by the confinemert, the susceptibility having all fuel assemblies, radioactive fluids and to release of radiation as a result of natural phenomena, wastes, and certain selected components shipped f and the duration of required surveillance. offsite. The structure should provide integrity over the period of time in which significant quantities C. REGULATORY POSITION (greater than Table I levels) of radioactivity remain

{ with the material in the entombment. An

1. APPLICATION FOR A LICENSE TO POSSESS appropriate and continuiag surveillance program BUT NOT OPERATE (POSSESSION-ONLY should be established under a possession-only LICENSE) license.

A request to amend an operating license to a c. Removal of Radioactive. Components and possession-only license should be made to the Director Dismantlir.g. All fuel assemblies, radioactive fluids of Licensing, U.S. Atomic Energy Commission, and waste, and other materials having activities Washington, D.C. 20545. The request should include above :ccepted ~*ted setivity levels (Table 1) the following information: should be ren oved from the site. The facility owner may then have unrestricted use of the site

a. A description of the current status of the facility. with no requirement for a license. If the facility owner so desires, the remainder of the reactor

b. A description of measures that will be taken to facility may be dismantled end all vestiges removed prevent uitice.lity or reactivity changes and to and dispused of, minimize releases of radioactivity from the facility.

d. Convmion to a New Nuclear System or a

c. Any proposed changes to the technical Fossil Fuel System. This altemative, which applies specifications that reflect the possession-only facility only to nuclear power plants, utilizes the existing status and the necessary disassembly / retirement turbine system with a new steam supply system.

activities to be performed. The original nuclear steam supply system should be separated f om the electric generating system and

d. A safety analysis of both the activities to be disposed of in accordance with one of the previous accomplished and the proposed changes to the three retirement alterr.atives.

technical specifications.

3. SURVEILLANCE AND SECURITY FOR TILE

e. An inventory of activated materials and their RETIREMENT ALTERNATIVES WIIOSE location in the facility. FIN A L STATUS REQUIRES A POSSESSION-ONLY LICENSE

2. ALTERNATIVES FOR REACTOR RETIREMENT A facility which has been licensed under a possession-only license may contain a significant amount of radioactivity in the form of activated and l

Note: Section electronicaHy reproduced from photocopy. C-2

contmmmated hardware end structural materials. g. The following reports s'nould be made:

Surveillance and commensurate security should be provided to assure that the public health and safety are (1) An annual report to the Director of not endangered. Licensing, U.S. Atomic Energy Commission,

a. Physical security to prevent inadvertent exposure Washington, D.C. 20545, describing the results of the of personnel should be provided by multiple locked environmental and facility radiation surveys, the status barriers. The presence of these barriers should make of the facility, and an evaluation of the performance of it extremely difficult for an unauthorized person to gain security and surveillance measures.

access to areas where radiation or contamination levels exceed those specified in Regulatory Position C.4. To (2) An abnormal occurrence report to the prevent inadvertent exposure, radiation areas above Regulatory Operations Regional Office by telephone 5 mR/hr, such as near the activated primary system of 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 a power plant, should be appropriately marked and occurrence. 'Ibe abnormal occurrence will also be should not be accessible except by cutting of welded reported in the annual report described in the preceding closures or the disassembly and removal of substantid item.

structures and/or shielding material. Means such as a remote-readout intrusion alarm system should be h. Records or logs rdative to the following items provided to indicate to designated personnel when a should be kept and retained until the license is physical barrier is penetrated. Security personnel that terminated, aiter which they must be stored with other provide access control to the facility may be used plant records:

instead of the physical barriers and the intmsion alarm systems. (1) Environmental rurveys, f

b. The physical barriers to unauthorized entrance (2) Facility radiation surveys,

[ into the facility, e.g., fences, buildings, welded doors, l and access openings, should be inspected at least (3) Inspections of the physical barriers, and quarterly to assare that these barriers have not deteriorated and that locks and locking apparatus are (4) Abnormal occurrences.

intact.

c. A facility radiation survey shou'd be performed 4. DECONTAMINATION FOR RELEASE FOR at least quartrly to verify that no radioactive material UNRESTRICTED USE f is escaping or being transported through the containment barriers in the facility. Sampling should If it is desired to terminate a license and to

[ be done along the most probable path by which eliminate any further surveillance requirements, the

{ radioactive material such as that sto ed in the inner facility should be sufficiently decontaminated to prevent containment regions could be transported to the outer risk to the public health and safety. After the regions of th~e facili't iiiid idiliiiately to the envirors. decontamination is satisfactorily accomplished and the f d. An environmental radiation survey should be site inspected by the Commission, the Commission may authorize the license to be terminated and the facility performed at least semiannually to verify that no abandoned or released for unrestricted use. The significant amounts of radiation have been released to licensee should perform the decontamination using the the environment from the facility. Samples such as following guidelines:

soil, vegetation, and water should be taken at locations l for wh'ch statistical data has been established during a. The licensee should make a reasonable effort to reactor operations. eliminate residual contamination.

e. A site representative should be designated to be b. No covering should be applied to radioactive responsible for controlling authorized access into and surfaces of equipment of structures by paint, plating, or movement within the facility other covering material until it is known that contamination levels (determined by a survey and

[ f. Administrative procedures should be established documented) are below the limits specified in Table 1.

l. for the notification and reporting of abnormal in addition, a reasonable effort should be made (and occurrences such as (1) the entrance of an unauthorized documented) to further minimize contamination prior to person or persons into the facility and (2) a significant any such covering.

( change in the radiation or contamination levels in the facility or the offsite environment. c. The radioactivity of the interior surfaces of pipes, drain lines, or ductwork should be detemnned Note: Section electronically reproduced from photocopy. C-3 i

I

r by maHng measurements et all traps and other After review of the repor2, the Commission may

[ appropriate access points, provided contamination at inspect the facilities to confirm the survey pnor to these locations is likely to be representative of granting approvd for abandonment.

2contammation on the interior of the pipes, drain lines, or ductwork. Surfaces of premises, equipment, or f scrap which are likely to be contaminated but are of 5. REACTOR RETIREMENT PROCEDURES such size, construction, or location as to make the surface inaccessible for purposes of measurement As indicated in Regulatory Position C.2, several

( should be assumed to be contaminated in excess of the alternatives are acceptable for reactor facility permissible radiation limits, retirement. If minor disassembly or 'mothballing" is planned, this could be done by the existing operating f d. Upon request, the Commission may authorize a licensee to relinquish possession or control of premises, and maintenance procedures under the license in effect.

Any planned actions involving an unreviewed safety equipment, or scrap having surfaces contaminated in question or a change in the technical specifications excess of the limits specified. " Itis may include, but is should be reviewed and approved in accordance with

( not limited to, special circumstances such as the the requirements of 10 CFR i 50.59.

transfer of premises to another licensed organization that will continue to work with radioactive materials. If major structural changes to rndioactive

[ Requests for such authodzation should provide: components of the facility are planned, such as reaoval of the pressure vessel or major components of the (1) Detailed, specific information describing the primary system, a dismantlement plan including the equipment, scrap, and radioactive

( premises, contammants and the nature, extent, and degree of information required by i 50.82 should be submitted to the Commission. A dismantlement plan should be residual surface contamination. submitted for all the alternatives of Regulatory Position C.2 except mothba!!ing. However, minor disassembly

[ (2) A detailed health and safety analysis indicating activities may still be performed in the absence of such that the residual amounts of materials on surface areas, a plan, provided they are permitted by existing together with other considerations such as the operating and maintenance procedures. A I f prospective use of the premises, equipment, or scrap, are unlikely to result in an unreasonable risk to the dismantlement plan should include the following:

health and safety of the public. a. A description of the ultimate status of the facility

e. Prior to release of the premises for unrestricted b. A description of the dismantling activities and use, the licensee should make a comprehensive the precautions to be taken.

radiation survey establishing that contamination is

[ within the limits specified in Table 1. A survey repod c. A safety analysis of the dismantling activities should be filed with the Director of Licensing, U.S. including any effluents which may be released.

Atomic Energy Commission, Washington, D.C. 20545, f with a copy to the Director of the Regulatory Operations regmaa! O!Tre having jurisdiction. The

d. A safety analysis of the facility in its ultimate status, report should be filed at least 30 cays prier to the planned date of abandonment. The survey report Up:n satisfactory review and approval of the

[ should: dismantling plan, a dismanding erder is issued by the Commission in accordance with i 50.82. Whcn (1) Identify the premises; dismantling is completed and the Commission has been

[ notified by letter, the appropriate Regulatory (2) Show that reasonable effort has been made to Operations Regional Office inspects the facility and reduce residual contamination to as low as practicable verifies completion in accordance with the levels;

( dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the Commission may (3) Describe the scope of the survey and the general terminate the license. If possession-only license under procedures followed; and which the dismantling activities have been conducted

[- or, as an alternative, may make application to the State (4) State the finding of the survey in units specified (if an Agreement State) for a byproduct materials in Table 1. license.

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TABLEI

( ACCEPTABLE SURFACE CONTAMINATION LEVELS Nuclidea Average" Maximum" Removable" U-nat, U-235, U-238, and associated decay products 5,000 dpm a/100 cm' 15,000 dpm a/100 cm' 1,000 dpm a/100 cm2 Transtmmics, Ra-226, Ra-228, Th-230 Th-228, Pa-231, Ac-227, I-125, I-129 100 dpm/100 cm' 300 dpm/100 cm2 20 dpm/100 cm 2 Th-nat, Th 232, Sr-90, Ra-223, Ra-224, U-232, I-126,1-131. I-133 1,000 dpm/100 cm2 3,000 dpm/100 cm2 200 dpm/100 cm2

( Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except St-90 and others noted above. 5,000 dpm #y/100 cm2 15,000 dpm Sy/100 cm' 1,000 dpm #7/100 cm'

{

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

[ '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 instmmentation.

(

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

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

'The amount of removable radioactive material per 100 cm2 of 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 removab!c contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.

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Guidelines for Residual Concentrations of Thorium

( and Uranium Wastes in Soil

( On October 23,1981, the Nuclear Regulatory Commission published in the Federal register a notice of Branch Technical Position on " Disposal or Onsite Storage of Thorium and Uranium Wastes from Past b Operations." This document established guidelines for concentrations of uranium and thorium in soil, that will limit maximum radiation received by the public under various conditions of future land usage.

These concentrations are as follows:

Maximum Concentrations (pCI/g)

Material f r var US Ptions l' 2' 3* 4' b Natural Thorium (Th-232 + Th-228) 10 50 -

500 with daughters present and in equilibrium Natural Uranium (U-238 + U-234) 10 -

40 200

_ with daughters present and in equilibrium Depleted Uranium:

Soluble 35 100 -

1,000

{ Insoluble 35 300 --

3,000 Enriched Uranium:

b Soluble 30 100 -

1,000 Insoluble 30 250 -

2,500

[ Based on EPA cleanup standards which limit radiation to 1 mrad /yr to lung and 3 mrad /yr to bone from ingestion and inhalation and 10 R/h above background from direct external exposure.

' Based on limiting individual dose to 170 mrem /yr.

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• Based on limiting equivalent exposure to 0.02 working level or less.

' Based on limiting individual dose to 500 mrem /yr and in case of natural uranium, limiting exposure to 0.02 working level or less.

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