ML20056H439
| ML20056H439 | |
| Person / Time | |
|---|---|
| Site: | 05000192 |
| Issue date: | 07/31/1993 |
| From: | Ansari A, Beck W, Landis M OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | |
| Shared Package | |
| ML20056H434 | List: |
| References | |
| ORISE-93-G-9, NUDOCS 9309090320 | |
| Download: ML20056H439 (46) | |
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ORISE 93/G-9 CONFIRMATORY SURVEY OF THE UNIVERSITY OF TEXAS TRIGA REACTOR AUSTLN, TEXAS Prepared by 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 Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1993 FINAL REPORT I This report is based on work performed under an Interagency Agreement (NRC Fin. No. A-9093) 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-AC-05-760R00033 with the U.S. Department of Energy. University of Texas . hty 8.1993 TRIGA Reactor
l CONFIIGIATORY SURVEY I OF TIIE UNIVERSITY OF TEXAS TRIGA REACTOR AUSTIN, TEXAS I 7 8! O Prepared by: Date: A. J. Ansari,gloject Leader Environmentil Survey and Site Assessment Program 1 Reviewed by: >Sd Date: 7!M/9.3 W. L. Beck, Acting Laboratory Manager Environmental Survey and Site Assessment Program I
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Reviewed by: , Date:# 7 d1/ 8 ' M. R. Landis,'Profect Manager Environmental Survey and Site Assessment Program Reviewed by: % m (%#_ Date: )![ b A. T. Payne, Quality Assurande Officer ' ' I Environmental Survey and Site Assessment Program I l Reviewed by: N Date: 7/9/8)
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J. Berger, PrograntfDirector ironmental Survey and Site Assessment Program l L r Universey of Texas - July 8,1993 TRIGA Reactor F
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. L.audeman S. T. Shipley CLERICAL STAFF T. T. Claiborne D. A. Cox R. D. Ellis M. S. Perry K. E. Waters [ ILLUSTRATOR E. A. Powell r University of Tczas - July 8.1993 TRIGA Reactor
TABLE OF CONTENTS PAGE List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii List of Tables ..............................................iii Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv Acronym s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v Introduedon and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Obj ecti ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Document Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Procedures ................................................3 Sample Analysis and Data Interpretation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 f Findings and Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 S u m mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 References ...............................................20 Appendices: Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures Apper. dix C: Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors. Guidelines for Residual Concentrations of Thorium and Uranium Wastes in Soil. l i University of Texas -My 8,1993 TRIGA Reactor i
LIST OF FIGURES PAGE FIGURE 1: Austin, Texas Area - IAcation of the University of Texas ..................................10 FIGURE 2: Taylor Hall, First Floor - Location of the Reactor Room 131 .................................11 FIGURE 3: Room 131, Reactor Room - Measurement and Sam Locations . . . . . . . . . . . . . . . . . . . . . . . . . pling . . . . . . . . . 12 FIGURE 4: Top Edge of the Reactor Pool - Measurement and Sampling ) Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 FIGURE 5: Walls of the Reactor Pool- Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 I FIGURE 6: Floor of the Reactor Pool - Measurement and Sampling Iecations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l University of Texas . My 8,1993 TRIGA Reactor ii
l ' LIST OF TABLES PAGE TABLE 1: Summary of Surface Activity Measurements - Locations of Elevated Direct Radiation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 TABLE 2: Summary of Surface Activity Measurements - Areas Adjacent to the Reactor Room ................................17 I TADLE 3: Exposure Rate Measurements . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 TABLE 4: Radionuclide Concentrations in Crawlspace Soil Samples . . . . . . . . . . l 19 I I I I I I I I i a l I L E l Univerniy of Tczas - My 8,1993 TRIGA Reactor jij
ABBREVIATIONS r em2 square centimeter epm counts per minute dpm/100 cm2 disintegrations per minute /100 square centimeters ft foot i GM Geiger-Mueller kg kilogram u kw ki'owatt m meter g p m 2 square meter NaI Sodium Iodide pCi/g picoeuries per gram PIC Pressurized Ionization Chamber R/h microroentgens per hour ZnS Zine Sulfide 1 I I I University of Texas - July 8,1993 TRIGA Reacw iV
.x ACRONYMS AEC Atomh Energy Commission ASME American Society of Mechanical Engineers EPA Environmental Protection Agency EML Environmental Measurement Laboratory ESSAP Environmental Survey and Site Assessment Program MDA Minimum Detectable Activity NETL Nuclear Engineering Teaching Laboratory NIST National Institute for Standards Technology NRC Nuclear Regulatory Commission ORISE Oak Ridge Institute for Science and Education TRIGA Training, Research, Isotopes, Gen ral Atomics I
l I I l I I I I L r University of Texas - July 8,1993 TRIGA Reactor v
l CONFIRMATORY SURVEY OF THE UNIVERSITY OF TEXAS TRIGA REACTOR AUSTIN, TEXAS INTRODUCTION AND SITE HISTORY The University of Texas TRIGA reactor was originally licensed by the Atomic Energy Commission (AEC), predecessor to the U.S. Nuclear Regulatory Commission (NRC), for operation at 10 kW. The license was amended in 1968 for operation at 250 kW In May 1985, The University of Texas requested authorization to dismantle the TRIGA reactor facility and to dispose of the component parts, in accordance with the plan submitted as part of the application to terminate the NRC license R-92 (Docket 50-192). In March 1987, the NRC issued the order to dismantle the reactor following shipment of the fuel offsite. At the time that the NRC order was issued, the reactor was still operating and remained in intermittent operation until May 1988, when it was permanently shut down. The reactor fuel was transferred by the Nuclear Engineering Teaching Laboratory (NETL) staff to a new TRIGA research reactor facility (NRC License R-192), constructed approximately 11 kilometers from the main University campus in Austin. Following transfer of the fuel, the dismantling and decommissioning program was initiated. A preliminary radiological survey was conducted by both the NETL and Quadrex Corporation, a University of Texas contractor. The neutron-activated parts of the aluminum reactor pool liner, a portion of the concrete reactor shield structure, and other radioactive waste materials were shipped to Barnwell, South Carolina, a low-level disposal site. The licensee and Quadrex Corporation, performed a final radiological survey, and the data were provided to the NRC in December 1992.' 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. At the request of the U.S. Nuclear Regulatory Commission, Region IV Office, the Environmental Survey and Site Assessment Program (ESSAP) of Oak Ridge Institute for Science r Univerniy of Texas - My 8,1993 TRIGA Reactor
l and Education (ORISE) conducted confirmatory activities and additional radiological surveys at this facility. This report describes the procedures and results of those activities. SITE DESCRIPTION I The reactor room (Room 131) is located on the first floor of Taylor Hall, on the campus of the University of Texas at Austin, Travis County, Texas (Figures 1 and 2). The reactor room has 2 approximately 150 m of floor space. The reactor pool is approximately 10 m deep and is located near the center of Room 131 (Figure 3). There is also a crawlspace underneath the reactor room with access to the bottom of the outer layer of the reactor pool and to the area where fuel storage pits were formerly located. The areas immediately adjacent to the reactor room were not included in the decommissioning plan for this reactor facility; however, at the request of the NRC, limited survey activities were performed in these areas. I At the time of the initial ESSAP survey of this facility, the reactor pool liner was still in place. l The aluminum pool liner was subsequently removed, and the exposed concrete walls of the former reactor pool were surveyed by the licensee, prior to the ESSAP follow-up survey. I OBJECTIVES I The objectives of the confirmatory process are 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 report, relative to established guidelines. DOCUMENT REVIEW As part of the confirmatory activities ESSAP reviewed the licensee's radiological survey data.' 23 Analytical procedures and methods utilized by the licensee were reviewed for adequacy and appropriateness. The data were reviewed for accuracy, completeness, and compliance with guidelines. J University of Texas - July 8.1993 TRIGA Reacto, 2
PROCEDURES On April 5 and 6,1993, ESSAP performed a confirmatory survey of the University of Texas TRIGA reactor located on the campus of the University of Texas at Austin. Subsequent to further remediations and removal of the reactor pool liner by the licensee, ESSAP performed a follow-up survey of the facility on June 1,1993. The surveys were conducted in accordance with survey plans which were submitted to and approved by the NRC Region IV Office." REFERENCE GRID The 1 m X 1 m alphanumeric reference grid system, established on the floor of the reactor room by the licensee, was used by ESSAP to reference measurement and sampling locations (Figure 3). Measurement locations on the wall and ceiling surfaces were referenced to the floor grid. Sampling locations from the crawlspace were also referenced to the reactor room floor grid. The reactor pool surfaces were not gridded. SURFACE SCANS Surface scans for alpha, beta, and gamma activity were performed on floors and lower walls (up f to 2 m) in the reactor room, using large area gas proportional and NaI scintillation detectors, coupled to ratemeter-scalers and ratemeters with audible indicators. Beta and gamma surface scans were performed in the reactor pool area. Gamma scans were performed in accessible areas of the crawlspace underneath the reactor room. Cursory beta and gamma surface scans were performed in areas adjacent to the reactor room. Areas of elevated direct radiation, identified by scans, were marked for further investigation. SURFACE ACTIVITY MEASUREMENTS Direct measurements to determine total alpha and total beta surface activity were performed on 57 randomly selected grid blocks on the floor and lower walls in the reactor room. Measurements were performed at the center of each grid block. Six measurements were University of Texas . My 8.1993 TRIGA Reactor 3 l
performed on upper wall and ceiling surfaces. Thirty measurements were performed on the top edge, floor and walls of the reactor pool. Sixteen measurements were performed in areas adjacent to the reactor room. A smear sample for determining removable activity was obtained from each measurement location. Smear samples for determining H-3 and C-14 activity were collected from the center point of 14 randomly selected grid blocks in the reactor room. Measurement and sampling locations for total and removable activity are illustrated in Figures 3-6. I EXPOSURE RATE MEASUREMENT I Background exposure rate measurements were performed at 3 locations within Taylor Hall, having similar construction as the reactor room but without a history of radioactive materials use. Exposure rate measurements were performed at I m above the surface at 5 locations in the reactor room and 2 locations in the adjacent areas, using a pressurized ionization chamber (PIC). Measurement locations for Room 131 are shown in Figure 3. I SOIL SAMPLING I Thrcc Avil muyics were collected from the crawlspace underneath the reactor room. The approximate locations of these samples relative to the floor grid in the reactor room are indicated in Figure 3. Oae sample was obtained near the northwest corner of the outer layer of the reactor pool. Two additional samples were obtained from former locations of fuel storage pits. SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and survey data were retumed to the ESSAP Oak Ridge laboratory for analyses and interpretation. Smears were analyzed for gross alpha and gross beta or H'-3/C-14 activity. Direct measurement and smear data were converted to units of disintegrations per minute per 2 2 100 cm (dpm/100 cm ), and exposure rate measurements were reported in microroentegens per hour ( R/h). 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. Soil and I University of Texas -July 8.1993 TRIGA Reactor 4
concrete sample results were reported in units of picoeuries per gram (pCi/g). Additionr.1 information concerning major instrumentation, sampling equipment, and analytical procedutes is provided in Appendices A and B. Results were compared to NRC guidelines which are provided in Appendix C. FINDINGS AND RESULTS I DOCUMENT REVIEW ESSAP reviewed the licensee's radiological survey data and comments were provided to the NRC.* In ESSAP's opinion, the licensee's measurements of removable alpha and beta activity, l exposure rate measurements, and the survey performed subsequent to removal of the reactor pool liner provide an adequate description of the radiological condition of the facility. However, because of the methodology used in determining total surface activity in the reactor room, the licensee's data are not directly comparable to the NRC surface contamination guidelines. I Confirmatory activities are, by definition, limited to those types of measurements, performed l by the licensee, which are adequate to demonstrate compliance with applicable NRC guidelines. ESSAP's measurements of removable surface activity and exposure rate in the reactor room and l the survey of the reactor pool can be considered confirmatory activities. Although ESSAP's measurements of total surface activity in the reactor room and the surveys performed in the adjacent areas are not " confirmatory", they will be of use to the NRC in the overall evaluation of the radiological status of the facility. SURFACE SCANS Alpha, beta, and gamma surface scans of the reactor room floor and lower walls identified several locations of elevated beta radiation near or at the top edge of the reactor pool. These locations were marked for further investigation. Cursory scans in areas adjacent to the reactor room identified one location in room 125 with elevated beta activity. This location was also marked for further investigation. University of Texas - My B.1993 TRIGA Reumr 5 ]
SURFACE ACTIVITY LEVELS . i In the reactor room, results of total activity measurements, performed on randomly selected grid blocks on the floor and lower walls, were all less than the detection limits of the procedure which were 69 dpm/100 cm2 and 1700 dpm/100 cm2 for alpha and beta, respectively. Removable activity levels at all but two of these locations were less than the minimum detectable 2 activity of the procedure which were 12 dpm/100 cm for alpha,17 dpm/100 cm2 for beta, 2 6 dpm/100 cm for H-3, and 5 dpm/100 cm2 for C-14. Removable beta activity at grid blocks L,1 and C,2 were 18 and 19 dpm/100 cm 2, respectively. Surface scans had identified several locations of elevated beta radiation near the reactor pool. l The results of surface activity measurements, performed at those locations prior to and after remediation by the licensee, are summarized in Table 1. Prior to remediations, the total surface l activity measurements were <66 dpm/100 cm for alpha and ranged from 11,000 to 2 2 46,000 dpm/100 cm for beta. After the remediations, the total surface activity measurements l ranged from <68 to 120 dpm/100 cm2 for alpha and < 1,300 to 5,200 dpm/100 cm2 for beta. The removable activity levels, before and after the remediation, were <12 dpm/100 cm 2 for l alpha and ranged from <17 to 22 dpm/100 cm2 for beta. The size of the area with 5,200 dpm/100 cm2 (#9, Figure 4) was less than 100 cm 2. I Results of total activity measurements on the floor and walls of the reactor pool were all less than the detection limits of the procedure which were 68 dpm/100 cm2 and 1,300 dpm/100 cm 2 for alpha and beta, respectively. Removable activity levels were < 12 dpm/100 cm2 for alpha and ranged from <20 to 22 dpm/100 cm2 for beta. Results of surface activity levels in areas adjacent to the reactor room are summarized in Table 2. At one location in Room 125 (W4,S4), identified by surface scans, the beta activity 2 was 47,000 dpm/100 cm . A portion of the wood floor was removed by NETL staff while ESSAP was on site. The beta activity from the follow-up measurement taken at this location 2 was < 1500 dpm/100 cm . Results of total activity measurements at other locations were less than the detection limits of the procedure which were 69 dpm/100 cm2 and 1500 dpm/100 cm 2 J University of Texas - July 8,1993 TRIGA Reactor 6
I fur alpha and beta, respectively. Removable activity levels were less than the minimum 2 detectable activity of the procedure which were 12 dpm/100 cm and 17 dpm/100 cm2 for alpha and beta, respectively. I EXPOSURE RATES I Background exposure rate was 11 R/h. Exposure rate measurements at 5 locations in the reactor room and 2 locations in the adjacent areas ranged from 9 to 13 R/h (Table 3). RADIONUCLIDE CONCENTRATIONS IN SOIL SAMPLES Radionuclide concentrations in crawlspace soil samples are 1 :sented in Table 4. In all three samp:.es, the concentrations of Co-60 and Eu-152 were <0.1 and <0.2 pCi/g, respectively. l The concentrations of U-235 ranged from 0.1 to 0.2 pCi/g. The concentrations of U-238 ranged from 1.2 to 1.7 pCi/g. l l COMPARISON OF RESULTS WITII GUIDELINES l 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. The major contaminants identified by the licensee, prior to remediation, 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 /100 cm2 , averaged over a 1 m2 area l 15,000 dpm /100 cm2 , maximum in a 100 cm2 area l Removable Activity l l 1000 dpm /100 cm2 Universey of Texas . My 8.1993 TRIGA Reactor 7
l 1 The surface contamination guidelines for uranium are: Total Activity 5,000 dpm a/100 cm 2, 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. The NRC guideline for exposure rate at I m above surface is 5 R/h above background." All exposure rates measured in this survey were within this limit. [ The soil concentration guideline for enriched uranium is 30 pCi/g.' Based on a U-234:U-235 ratio of 30 to 1, the highest total uranium concentration in the samples collected ( (crawlspace, underneath grid block H4) is 7.9 pCi/g which is well below the 30 pCi/g limit. There are no specific concentration guidelines for Co-60 and Eu-152 in soil.
SUMMARY
On April 5 and 6,1993, ESSAP conducted confirmatory activities and additional radiological surveys of the University of Texas TRIGA reactor on the campus of the University of Texas at Austin, Texas. Confirmatory survey activities included document reviews, surface scans, measurements of removable surface activity, and exposure rate measurements in the reactor room. Additional survey activities included cursory scans of the areas immediately adjacent to the reactor room, measurements of total surfne activity in the reactor room and adjacent areas, and soil sampling in the crawlspace. I University of Texas July 8,1993 TRIGA Reactor 8
[ Several locations were identified in the reactor room with elevated beta activity. These locations were near or at the edge of the reactor pool. Subsequently, the reactor pool liner was removed by the licensee and the locations of elevated activity were remediated and surveyed. At the request of the NRC, Region IV Office, ESSAP surveyed the remediated areas and the reactor pool floor and walls on June 1,1993. The follow-up confirmatory measurements support the licensee's conclusion that the radiological condition of the reactor room satisfies NRC's guidelines for release to unrestricted use. The survey activities in areas immediately adjacent to the reactor room identified one { location in room 125 with 47,000 beta dpm/100 cm2 This area (approximately 50 cm) 2 was effectively remediated while ESSAP was on site. The results of total and removable activity at other measurement locations in the adjacent areas were within applicable NRC guidelines. These adjacent areas were not included in the decommissioning activities, and it is not clear whether the activity found was due to reactor-related operations. [ [ [ [ ( [ [ [ [ c Unsversity of Teams July 8.1993 TRIGA Reuw 9 I
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l 0 e . eiT' e [ A [ [ MEASUREMENT / SAMPLING LOCATIONS h SURFACE ACTMTY JL 0 TOWER WALLS AND FLOOR T g SURFACE ACTMTY; H-3/C-14 ( LOWER WALLS AND FLOOR h g SURFACE ACTMTY FEET UPPER WALLS AND COUNG r 0 6
& EXPOSURE RATE
( 0 CRAWLSPACE SOIL METERS f L FIGURE 3: Room 131, Reactor Room - Measurement and Sampling Locations e I l L. University of Texas . July 8,1993 TRIGA Reactor 12
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I I I I I " n MEASUREMENT /SAMPUNG h LOCATIONS I hSCABBLEDAREA I S SINGLE-POINT O b 6 METERS L FIGURE 4: Top Edge of the Reactor Pool - Measurement and Sompting Locations I University .f Texas - My 8,1993 TRIGA Reaciw 13
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6 METERS FIGURE 6: Floor of the Reactor Pool - Measurement and Sampling Locations r L caivmay v Tex = . Jay 3, im mca a=== 15
TABLE 1 SU3131ARY OF SURFACE ACTIVITY AIEASURE51ENTS LOCATIONS OF ELEVATED DIRECT RADIATION UNIVERSITY OF TEXAS TRIGA REACTOR AUSTIN, TEXAS Total Activity Removable Activity Location p (dpm/100 cm') (dpm/100 cm') Alpha Beta Alpha Beta Prior to Removal of Reactor Pool Liner 1 < 66 46,000 < 12 < 17 2 < 66 29,000 < 12 < 17 3 < 66 21,000 < 12 22 4 < 66 19,000 < 12 < 17 5 < 66 28,000 < 12 < 17 6 < 66 11,000 < 12 < 17 After Removal of Reactor Pool Liner 7 120 2500 < 12 < 20
. 8 < 68 < 1300 < 12 < 20 9 < 68 5200 < 12 < 20 10 < 68 2500 < 12 < 20
- Refer to Figure 4.
I I 1 i I l Univenity of Tczas - July 8.1993 TRIGA Reactor 16
I TABLE 2 SUT131ARY OF SURFACE ACTIVITY 31EASURE51ENTS AREAS ADJACENT TO THE REACTOR ROO51 I UNIVERSITY OF TEXAS TRIGA REACTOR AUSTIN, TEXAS Total Activit Removable Activity (dpm/100 cmb' (dpm/100 cm 2) Location. Alpha Alpha I Room 125 Beta Beta W2.S2 < 69 < 1500 < 12 < 17 W4.S4 < 69 < 1500* < 12 < 17 W7, S1 < 69 < 1500 < 12 < 17 ; I Doorway to Room 125 A < 69 < 1500 < 12 < 17 l Doorway to Room 131 < 69 < 1500 < 12 < 17 Room 125 A , l Center of Room < 69 < 1500 < 12 < 17 i Hallway ; W2.S0 < 69 < 1500 < 12 < 17 Doorway to Room 131 < 69 < 1500 < 12 < 17 Room 131 A Center of Room ! < 69 <1500 < 12 < 17 Room 133 W6. S1 < 69 < 1500 < 12 < 17 WO, S0 < 69 < 1500 < 12 < 17 l W2, S2 Room 135
< 69 < 1500 < 12 < 17 l Center of Room Doorway to the Han
< 69
< 69
< 1500
< 1500
< 12
< 12
< 17
< 17 Room 135 A _
Center of Room < 69 < 1500 < 12 < 17 Doorway to Room 133 < 69 < 1500 < 12 < 17 f
'These areas were not gridded. A description of the measurement location is provided.
I When coordinates are given, they designate the distance in meters from the northeast corner of the room. For examale, "W., S2" in Room 125 indicates a measurement location at 2 m west and 2 m south of tae northeast corner of Room 125. I %is measurement was taken after removal of a portion of the wood floor by NETL while ESSAP was on site. Prior to this remediation the beta activity was 47,000 dpm/100 cm2. unim.iw or rm. .uy . im roar me== 17
TABLE 3 EXPOSURE RATE MEASUREMENTS f UNIVERSITY OF TEXAS TRIGA REACTOR - AUSTIN, TEXAS Location Exposure Rate (gR/h) Facility Exposure Rates' Room 131, Grid Block 15 10 Room 131, Grid Block Ill 11 Room 131, Grid Block L2 10 Room 131, Grid Block C2 10 Room 131, Grid Block D10 11 Office, S.E. Corner of Room 131 13 Room 125, Center of Room 9 Background Exposure Rates Hallway in front of Room 133 11 Taylor Hall Entrance Hallway 11 Hallway between Room 131 and Main Hallway 11
- Refer to Figures 2 and 3.
as [ Univereirf of Texas-July 8.1993 TRIGA Reactor 18
TABLE 4 RADIONUCLIDE CONCENTRATIONS ( IN CRAWLSPACE SOIL SAMPLES L UNIVERSITY OF TEXAS TRIGA REACTOR AUSTIN, TEXAS Location in Radionuclide Concentrations (pCi/g)* Pace
- Co-60 Eu-152 U-235 U-238 Grid Block H4 < 0.1 <0.2 0.2 0.1 1.7 i 1.1 Cdd Block H5 < 0.1 < 0.2 0.1 0.1 1.2 1.0 Grid Block J11 < 0.1 < 0.2 0.1 0.1 1.3 1.1
' Uncertainties represent the 95% confidence level, based only on counting statistics. ' Locations listed here designate the grid block on the reactor room floor corresponding to the approximate locations of soil sample collected from crawlspace underneath the floor. Refer to Figure 4.
univmi, w Tem ur s. im rucumu,, 19
REFERENCES
- 1. University of Texas, Austin, Texas, " Final Termination Survey," December 1992.
f
- 2. Quadrex Corporation, " Project Report for Decontamination and Survey for Unconditional Release of the Reactor Room in Taylor Hall at the University of Texas,"
January 15, 1993.
) 3. Letter from T. Bauer (University of Texas) to Document Control Desk (NRC),
Reference:
" Inspection Report 50-192/93-01, Confirmation Survey, Removal of Pool Liner," May 3,1993.
- 4. Oak Ridge Institute for Science and Education, " Radiological Survey Plan for the University of Texas TRIGA Reactor, .\ustin, Texas," March 30,1993.
- 5. Letter from A. Ansari (ORISE) to B. Murray (NRC, Region IV),
Reference:
" Confirmatory Survey of the Reactor Pool Area, University of Texas," May 25,1993.
- 6. Letter from A. Jaberabonansari (ORISE) to B. Murray (NRC, Region IV),
Reference:
" Final Status Survey of the University of Texas TRIGA Reactor," March 19, 1993.
- 7. Letter from A. Jaberabonansari (ORISE) to B. Murray (NRC, Region IV),
Reference:
" Comments on Quadrex Corporation Project Report for Decontamination and Survey of the Reactor Room in Taylor Hall at the University of Texas," March 24, 1993.
- 8. U.S. Nuclear Regulatory Commission, " Termination of Operating Licenses for Nuclear Reactors," Regulatory Guide 1.86, Washington D.c., June 1974.
- 9. U.S. Nuclear Regulatory Commission, " Disposal of Onsite Storage of Thorium and Uranium Wastes from Past Operations," 46 FR 52061, Washington, D.C., October 23, 1981.
- 10. U.S. Nuclear Regulatory Commission, Office of Nuclear Safety and Safeguards,
( " Review Plan: Evaluating Decommissioning Plans for Licenses Under 10 CFR Parts 30,40, and 70," Washington, D.C.1991. University of Texa, July 8.1993 TRIGA Reactor 20
l
)
APPENDIX A MAJOR INSTRUMENTATION ( University of Texas - July 8,1993 TRIGA Rescur
APPENDLX 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 Effecdve Area,15.5 cm' (Eoerline, Santa Fe, NM) 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) University of Texas My E.1993 TRIGA Reaciar A-1
Ludlum Gas Propcational Detector Model 43-68 Effective Area,100 cm2 (Ludlum Measurements, Inc., Sweetwater, TX) Reuter-Stokes Pressurized Ionization Chamber Model RSS-111 (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 Model No: ERVDS30-25195 (Tennelec, Oak Ridge, Th9 Used in conjunction with: Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, TN) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT) High-Purity Germanium Detector Model GMX-23195-S,23% Eff. (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) Iow Background Gas Proportional Counter Model LB-5100-W i (Oxford, Oak Ridge, TN) f' Tri-Carb Liquid Scintillation Analyzer Model 1900CA I (Packard Instrument Co., Meriden, CT) University of Texas - July 8,1993 TRIGA Rc== A-2
APPENDIX B SURVEY AND ANALYTICAL PROCEDURES 1 l i Wmsh of Texas Ju!y 8,1993 TRIGA Reactor
APPENDLX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES Surface Scans 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. A large surface area, gas proportional floor monitor was used to scan the floors of the surveyed areas. Other surfaces were scanned using smaller area (100 cm)2 hand-held gas proportional detectors. Identification of elevated levels 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: Alpha-Beta - gas proportional detectors with ratemeter-scalers f Gamma - NaI scintillation detectors with ratemeters Surface Activity Measurements f Measurements of total alpha and beta activity levels were performed using ZnS scintillation and GM detectors with ratemeters-scalers. Count rates (cpm), which were integrated over 1 minute in a static position, were converted - 2 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 to 0.19 for the ZnS scintillation detectors. The beta activity background count rates for the GM detectors ranged from 57 to 61 epm. Beta efficiency factors ranged from uwm.ny aTm. .uy . ms moue.c= B-1 l
0.15 to 0.16 for the GM detector. The effective windows for the ZnS scintillation and GM 2 detectors were 59 cm and 15.5 cm 2, 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, obtained from each measurement location, were placed in labeled envelopes with the location and other peninent information recorded. In addition, at 14 measurement locations, smear samples for H-3/C-14 analysis were obtained. These smears were placed in labeled glass vials and capped. Exnosure Rate Measurements Measurements of gamma exposure rates were performed at 1 m above surface using a pressurized ionization chamber (PIC). Miscellaneous Samples Soil Sampling Approximately 1 kg of soil was collected at each sample location. Collected samples were placed in a plastic bag, sealed, and labeled in accordance with ESSAP survey procedures. Univerniry of Texas - July 8.1993 TRIGA Re.n= B-2 l
ANALYTICAL PROCEDURES Removable Activity Gross 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. 9 Miscellaneous Samples Gamma Spectrometry ' Solid Samoles Samples of solid material (soil) 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 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. Energy peaks used for determination of radionuclides of concern were: Co-60 1.173 MeV Eu-152 0.344 MeV University of Texas- July 8,1993 TRIGA Reactor B-3
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 data 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 report.- 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. CALIBRATION AND QUALITY ASSURANCE 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) University of Texas Juh 5.1993 TRIGA Reacw B-4 l
.x 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. 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 or pressurized ionization chambers was performed by the manufacturer. Quality control procedures include:
- Daily instrument backgroer.d 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. Univenay of Texas - July 8,1993 TRIGA Rc== B-5
( ( [ [ APPENTIX C REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS AhT GUIDELINES FDR RESIDUAL CONCENTRATIONS OF THORIUM AhT URANIUM WASTES IN SOIL i [ [ r u % <rm.-u x . im rum %
U.S. ATOMIC ENERGY COMMISSION Juns 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 Material *), byproduct material (10 CFR Part Facilities," requires that each license to operate a 30,
- Rules of General Applicability to Licensing of production and utilization facility be issued for a Byproduct Material"), and source material (10 CFR specified duration. Upon expiration 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 by the Commission. Section 50.82, " Applications for from the facility. Appropriate 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 mothballing, 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. Section 50.82 provides that the licensee may B. DISCUSSION dismantle and dispose of the component parts of a nuclear reactor in accordance with existing regulations. When a licensee decides to terminate his nuclear For research reactors and critical facilities, this has reactor operating license, he may, as a first step in the usually meant the disassembly of a reacter 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 ( requirements in that periodic surveillance of equipment Commission to determine whether unrestricted access 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, making 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 GUIDES c,,,,,, ,,,,, , , , besetwy c.oes we me d te swenbe w make maabe to it. pake '""8 "* u.s. Atenue Er rgy ce nnua a we.*wnstes D.c. 2ms. methode acceptease to the AEC regJetory staff of emplemenung specifes parte Anem Duner of bouletery Stowents. comnants and suggeers for of the commesion's regJetiers, to engmete techrwouse seed by the etaff m p8em"*me e those gaces we omweged eM enound be um to N evaiwetsrg specafic swontems se postJeted acciaents, or te provide swoonce to Sewn of t% comnussen, M. Atenu Enew amnussen. Wesugin, apphcents. RegJetory Owoes are r et eestatmas for regJetene and C. 2WE. Auntem Chef, Pdhc Proceedenge staff. compience with them o not eenered. Wetheos and solutione differers from those est aus in the genes ein be acceptable if they provide a base for the , , fmd eqweete te the mouence or cente%ence of a permet er kcense by the
- 1. Power Reactre 6. Products
- 2. Reeweh and cut Reectors 7. Transportation
~ ...hed g - b. re. d .e..d . approp_e. t. emme..,e t - -1~e -~ . o-~
.or_.ms .,o to r.t. r.. mf.m.t.m ., omr. e.
l t g,-~,my g,, =i ~
- Note: Section electronically reproduced from photocopy. C-l
Rrdioactive components may be either shipped Four alternatives for retirement of nuclear reactor off-site for burial at an authorized burial ground or facilities are considered acceptable by the secured on the site. Those radioactive materials Regulatory staff. These are: remammg 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. Suncillance 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 and safety of radioactive fluids and waste should be removed the public from radioactive material remaining on the from the site. Adequate radiation monitoring, site and (2) the integrity of the physical barriers. environmental suncillance, and appropriate security Before areas may be released for unrestricted use, they procedures should be established under a must have been decontammated or the radioactivity possession-only license to ensure that the health and I must have decayed to less than prescribed limits (Table 1). safety of the public is not endangered.
- b. In-Place Entombment. In-place entombment i The hazard associated with the returned facility is evaluated by considering the amount and type of consists of sealing all the remaining highly radioactive or contaminated 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 I security provided by the confinement, the susceptibility to release of radiation as a result of natural phenomena, having all fuel assemblies, radioactive fluids and wastes, and certain selected enmponents shipped and the duration of required surveillance. offsite. 'Ibe structure should provide integrity over I C. REGULATORY POSITION the period of time in which significant quantities (greater than Table I levels) of radioactivity remain with the material in the entombment. An i 1. APPLICATION FOR A LICENSE TO POSSESS BUT NOT OPERATE (POSSESSION-ONLY appropriate and continuing surveillance program 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 Dismantling. 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 accepted unrestricted activity levels (Table 1) the following information: should be removed 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 I owner so desires, the remainder of the reactor
- b. A d:scription of measures that will be taken to facility may be dismantled and all vestiges removed prevent criticality or reactivity changes and to and disposed of.
I minimize releases of radioactivity from the facility.
- d. Conversion to a New Nuclear System or a
- c. Any proposed changes to the technical Fossil Fuel System. This attemative, which applies specifications that reflect the posswion-only facility only to nuclear power plants, utilizes the existing I status and the necessary disassembly / retirement turbine system with a new steam supply system.
i activities to be performed. The original nuclear steam supply system should be separated from the electric generating system and I 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 altematives. I technical specifications.
- 3. SURVEILLANCE AND SECURITY FOR TIIE
[ e. An inventory of activated materials and their RETIRGIENT ALTERNATIVES WIIOSE l 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 Note: Section electronicany reproduced from photocopy. C-2
contammeted hardware and structural materials. g. The following reports should be made: Survcillance 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 of personnel should be provided by multiple locked Washington, D.C. 20545, describing the results cf the environmental and facility radiation surveys, the status barriers. He 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 a power plant, should be appropriately marked and within 24 hours of discovery of an abnormal occurrence. The 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 substantial item. [ structures and/or shielding material. Means such as a remote-readout intrusion alarm system should be h. Records or logs relative 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 persormel that terminated, after 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 intrusion alarm systems. (1) Environmental surveys,
- b. He physical barriers to unauthorized entrance (2) Facility radiation surveys, into the facility, e.g., fences, buildings, welded doors, and access openings, should be mspected at least (3) Inspections of the physical barriers, and quarterly to assure that these barriers have not deteriorated and that locks and locking apparatus are (4) Abnormal occurrences.
intact.
- c. A facility radiation survey should be performed 4. DECONTAMINATION FOR RELEASE FOR at least quarterly to verify that no radioactive material UNRESTRICTED USE
[ 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 survei!!ance requirements, the radioactive material such as that stored in the inner facility should be sufficiently decontammated to prevent containment regions could be transported to the outer risk to the public health and safety. After the regions of the facility and ultimately to the environs. decontamination is satisfactorily accomplished and the site inspected by the Commission, the Commission may
- d. An environmental radiation survey should be 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 for which 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.
for the notification and reporting of abnormal In addition, a reasonable effort should be raade (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 determined r I Note: Section electronically reproduced from photocopy. C-3
by makmg measurements at cll traps and other After review of the report, the Commission may appropriate access points, provided contammation at inspect the facilities to confirm the survey prior to these locations is likely to be representative of granting approval for abandonment. I 2 contamination on the interior of the pipes, drain lines, or ductwork. Surfaces of premises, equipment, or scrap which are likely to be contaminated but are of 5. REACTOR RETIREMENT PROCEDURES such size, construction, or location as to make the I 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 I planned, this could be done by the existing operating
- d. Upon request, the Commission may authorize a and maintenance procedures under the license in effect.
licensee to relinquish possession or control of premises, Any planned actions involving an unreviewed safety I equipment, or scrap having surfaces contammated in excess of the limits specified. This may include, but is question or a change in the technical specifications should be reviewed and approved in accordance with not limited to, special circumstances such as the the requirements of 10 CFR $ 50.59. transfer of premises to another licensed organization I that will continue to work with radioactive materials. If major structural changes to radioactive Requests for such authorization should provide: components of the facility are planned, such as removal of the pressure vessel or major components of the I (1) Detailed, specific information describing the premises, equipment, scrap, and radioactive primary system, a dismantlement plan including the information required by i 50.82 should be submitted to contammants and the nature, extent, and degree of the Commission. A dismantlement plan should be I 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 I together with other considerations such as the operating and maintenance procedures. A prospective use of the premises, equipment, or scrap, dismantlement plan should include the following: g are unlikely to result in an unreasonable risk to the l 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 report 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. I Atomic Energy Commission, Washington, D.C. 20545, with a copy to the Director of the Regulatory d. A safety analysis of the facility in its ultimate Operations regional Office having jurisdiction. The status. i report should be filed at least 30 days prior to the planned date of abandonment. The survey report Upon satisfactory review and approval of the should: dismantling plan, a dismantling order is issued by the I (1) Identify the premises; Commission in accordance with i 50.82. dismantling is completed and the Commission has been When notified by letter, the appropriate Regulatory (2) Show that reasonable effort has been made to Operations Regional Office inspects the facility and I reduce residual contamination to as low as practicable levels; verifies completion in accordance with the dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the Commission may I (3) Describe the scope of the survey and the general procedures followed; and 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 (4) State the finding of the survey in units specified (if an Agreement State) for a byproduct materials I in Table 1. license. I l Note: Section electronically reproduced from photocopy. C-4
l l TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS I 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' I,000 dpm a/100 cm' Transuranics, Ra-226, Ra-228, I Th-230 Th-228, Pa-231, Ac-227, I-125, I-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th-nat, Th-232, Sr-90, Ra-223, 1 Ra-224, U-232, I-126, I-131, I-133 1,000 dpm/100 cm2 3,000 dpm/100 cm' 200 dpm/100 cm' Beta-gamms emitters (nuclides with decay modes other than alpha emission or I spontaneous fission) except Sr-90 and others noted above. 5,000 dpm 07/100 cm2 15,000 dpm By/100 cm2 1,000 dpm #7/100 cm2 aWhere 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 I correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation.
Measurements of average contammant should not be averaged over more than I square meter. For objects ofless surface area, the average should be derived for each such object. I 'I'he maximum contamination level applies to an area of not more than 100 cm 2. 2 "I'be amount of removable radioactive material per 100 cm 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 I 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. I ! l l l l 11 L F Note: Section electronically reproduced from photocopy. C-5
( 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 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 (pCilg) Material f r vari us options ( l' 2* 3' 4' 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 oBased 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 pR/h above background from direct external exposure.
- Based on limiting individual dose to 170 mrem /yr.
- 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.
[ , c t Note: Section electronically reproduced from photocopy. C-6}}