ML20149L784
| ML20149L784 | |
| Person / Time | |
|---|---|
| Site: | 05000262 |
| Issue date: | 06/30/1996 |
| From: | Adams W, Condra R, Morton J OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20149L775 | List: |
| References | |
| CON-FIN-A-9093 ORISE-96-F-49, NUDOCS 9611180096 | |
| Download: ML20149L784 (40) | |
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1 I! . i j The Oak Ridge Institute for Science and Education (ORISE) was established by the U.S. Department of Energy to -
- e. undertabo national and international programs in science and engineering education, training and management systems, . l
. energy and envimament systans, and medical sciences. ORISE and its programs are operated by Oak Ridge Associated Universities (ORAU) through a management and operating contract with the U.S. Department of Energy. Establisted in 1946, ORAU is a consortium of 89 colleges anxi universities, NOTICES The opinions expressed herein do not necessarily reflect tie opinions of the sponsoring institutions of Oak Ridge .
; Associated Universities.
This report was prepared as an account of work sponsored by the United States Govenunent. : Neitler the United States Govemmert nor the U.S. Departmert of Energy, nor any of their employees, makes any warranty, expressed or implied, 'g or assumes any legal liability.or responsibility for the accuracy, completeness, or usefulness of any information, g
. apparatus, product, or process disclosed, or represents that its use would not infringe on privately owned rights.
. Reference herein to any specific commercial product, process, or service by trade name, mark, manufacturer, or
' otherwise, does not necessarily consttute or imply its erxlorsement or recommendation, or favor by the U.S. Government l
. or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of iE the U.S. Government or any agency thereof.
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RADIOLOGICAL SURVEY OFTHE BRIGHAM YOUNG UNIVERSITY L-77 RESEARCH REACTOR i PROVO, UTAH i Prepared by W. C. Adams and J. R. Morton . Environmental Survey and Site Assessment Program , Environmental and Health Sciences Division I Oak Ridge Institute for Science and Education j Oak Ridge, Tennessee 37831-0117 ! Prepared for the U.S. Nuclear Regulatory Commission Region IV Office Sponsored by the Office ofNuclear Materials Safety and Safeguards U.S. Nuclear Regulatory Commission JUNE 1996 i I FINAL REPORT 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-AC05-760R00033 with the U.S. Department ofEnergy.. Brtsham Young Universky - June 6.1996 b:\ensap\reponsWyuwyu.002
' ~ -
[ RADIOLOGICAL SURVEY OF TIIE
- BRIGHAM YOUNG UNIVERSITY L-77 RESEARCll REACTOR PROVO, UTAH Prepared by
- [> .
Date: 0 ' ' l W. C. Adams, Project Leader l Environmental Survey and Site Assessment Program Prepared by: % Date: 4 f4 J. R. MortanpealthWhysics Technician Environmental Survey and Site Assessment Program Reviewed by: hklsL % . b Date: f/6/Pt R.D. Condra, Technical Resdurce Manager Environmental Survey and Site Assessment Program 't Reviewed by: %4 A. T. Payne, Administrative Services Manager Date: bbf b ' 4 Quality Assurance / Health and Safety Manager Environmental Survey and Site Assessment Program Reviewed by: % Date: /p!6h(- T. J. Vitku/ S/rvey Projects Manager ' ' ' Environmental Survey and Site Assessment Program { Reviewed by: Date: 9b E. W. Abelquist, Assistan)4rogram Director Environmental Survey and Site Assessment Program 4 Reviewed by: Date:d/7 ' W. L. Beck, Program Director # # Environmental Survey and Site Assessment Program a Brigham Young University June 6,19% h:\essap\ reports \byu\byu.002
i P ACKNOWLEDGEMENTS i The authors would like to acknowledge the significant contributions of the following staff members: i FIELD STAFF T. J. Vitkus I t LABORATORY STAFF i R. D. Condra ; J. S. Cox : M. J. Laudeman S. T. Shipley ) CLERICAL STAFF T. S. Fox K. E. Waters l l ILLUSTRATORS - i T. D. Herrera l
'I i
I
)
l i l Brigham Young Unwenhy . June 6.1996 b:\essap\reponswyuwyu.002
t i I i r i TABLE OF CONTENTS i PAGE !
~ List of Figures . . . . . . . . . . . . . . . . . . . . . . . . .. ........... ........ .... . . . . . . . . . ii i List of Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
)
Abbreviations and Acronyms . . . . . . ......................................iv . t Introduction and Site History . . . . . . . . . . . . . . . . .... .... ... ................. 1 ! Site Description . . . . . . . . . . . . . . . . .. .... . . . . . . . . . . . .. .......... ....... 2 i Objectives . . . . . . . . . . . . . . . .. . . .. . ... ... . . . . . . . . . . . . . .. ... 3 ! Document / Data Review . . . . . . . . . . ....... ..... .. . ... . .. ... ..... .3 t
- Procedures . . . . . . . . . . . . . . . . . . . . .. .. .... .. . . .. ... ....... 3 :
Findings and Results . . . . . . . . . . . . . . . . , , .. . .... . ..... ... . .... 6 i Comparison ofResults with Guidelines . . . .. ..... ... . . .. ... ......... 6 ; Summary . . . . . . . ..... ... . . . ... .. .... ............... . ...7 . L References . . . . . . . . . . . . . . . . . . . . . . . 17 l l t Appendices: i Appendix A: Major Instrumentation
' Appendix B: . Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors.
l i l \ I i f I ! l i I i i i i a Brighum Young Universky . June 6.1996 b:\eansp\repons\byu\byu.002 i
, ,ns- ,, , - -- ,-
I 1 LIST OF FIGURES i 1 PAGE FIGURE 1: Research Reactor Facility - Floor Plan and Exposure Rate , Measurement Locations . . . . .. .. ..... , ,.. . ... .8 l FIGURE 2: Reactor Room - Measurement and Sampling Locations . . . . ..... .9 FIGURE 3: Control Room - Measurement and Sampling Locations . . ..... . .. . 10 FIGURE 4: Large Room (Accelerator Room) - Measurement and Sampling Locations . . . . I1 1 l l 1 l l l i i l l l I i I Brigham Young University . June 6.1996 h:\cssap\repons\byu'byu.002 j I I i
_ . . _ . . _ _ _ . - . _ . _ . _ _ . - _ . . . - ._-_..m. . . . _ _ _ . _ , _ . _ . . . . _ _ . . _ . _, v
- LIST OF TABLES i PAGE l f
l TABLE 1: Summary of Surface Activity Levels . . . . . . . . . . . 12 j i TABLE 2: Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 l t i i i i I l l l l
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) ABBREVIATIONS AND ACRONYMS a alpha i p beta 4 Y gamma j - pR/h microroentgen per hour
- - prem/h microrem per hour
- ASME American Society of Mechanical Engineers BKG background j- BYU Brigham Young University cm centimeter cm.2 square centimeter
- cpm counts per minute Cs cesium DOE . Department ofEnergy dpm/100 cm2 disintegrations per minute per 100 square centimeters j- EML Environmental Measurements Laboratory
- EPA . EnvironmentalProtection Agency i ESSAP Environmental Survey and Site Assessment Program
- . m meter 2
- m square meter MDC minimum detectable concentration NaI sodium iodide NIST National Institute of Standards and Technology NRC Nuclear Regulatory Commission ORISE Oak Ridge Institute for Science and Education Sr strontium j Tc technetium Th thorium U uranium 3 7.nS zine sulfide i
i i i i f 4
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h l Brighame Young Universky. June 6,1996 Y h:\essapacpons\byu\byu.002 5 m + + g- +s v, m - - - - - - , -
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l
; RADIOLOGICAL SURVEY l l OFTHE i BRIGHAM YOUNG UNIVERSITY L-77 RESEARCH REACTOR PROVO, UTAH l
1 INTRODUCTION AND SITE HISTORY h i The Brigham Young University (BYU) L-77 Research Reactor was a small, solution-type nuclear i reactor designed for laboratory use. It was manufactured by the Atomics International Division of j i North American Aviation, a predecessor of the Rocketdyne Division of Rockwell International and l delivered to the university in August of 1967. Prior to that, this reactor had been used by Atomics International at several expositions in foreign countries where it had been loaded ar d operated for demonstration purposes. The reactor consisted of a core tank-which contained the liquid fuel solution-within an inner shield tank. ! The reactor was fueled in September 1967 with an uranyl-sulfate solution with a final mass of 1447
- grams of uranium-235 (U-235). During the next fifteen years, the reactor _was used for 355 separate operations-for a total of 1799 watt-hours-in conjunction with physics classes taught at the university under U.S. Nuclear Regulatory Commission (NRC) License R-109, for research and 4
training purposes. On May 12,1982, all operations of the reactor terminated and the reactor was I defueled on May 5,1987. 1 There have been no recorded incidents, while loading or unloading fuel to the reactor, to indicate that there were any spills or other incidents that would have led to contamination of the reactor facility.
- Operation of the reactor produced fission products, however only two-St-90 and Cs-137-are likely to remain following the ten years ofinactivity. The liquid fuel is the only radioactive material associated with the reactor that presents a possibility of contaminating areas away from the reactor.
Thus, any unreported spill involving the liquid fuel contaminated with fission products which may have occurred would have primarily included uranium contamination, along with Sr-90 and Cs-137 contamination. The licensee did not find any evidence of neutron activation of the stainless steel Brigham Young University - June 6.1996 hnessap\reportstyubyu.002
shell that divided the shielding water from the inner shielding components and the radiological analyses performed on samples of the shielding water indicated that there was no contamination in the sample. According to the licensee, the current radiological status of the facility, outside the reactor shield, is at background levels; therefore, because of the sealed and contained nature of the L-77 reactor, no contamination is expected outside of the core vessel. s BYU has decommissioned the facility and plans to release the site for unrestricted use. The final ; radiological survey of the reactor and its components and the reactor facility was performed by BYU f and the results were provided to the NRC in April 1994 (BYU 1994). The reactor was disasembled ! and the contaminated materials identified during the decommissioning activities were shipped to Richland, Washington for disposal as low-level radioactive waste. The reactor fuel was shipped to EG&G in Idaho and the plutonium-beryllium neutron source has been transferred to the Physics department under a separate license. The remaining rooms in the facility-the Reactor Room, the I Control Room, and the Accelerator Room-were surveyed and await release to unrestricted use pending the NRC's termination of License R-109. ; i The U.S. Nuclear Regulatory Commission, Region IV Office, requested that the Environmental ) l Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) perform an independent radiological survey of the remaining reactor components and the ! reactor facility that are to be released for unrestricted use. ; SITE DESCRIPTION The research reactor facility is located on a hilljust south of the Joseph Smith Building on the south edge of the Brigham Young University campus. Prior to the building being converted to a research reactor facility, the building had housed the main heating systems for the university. As the university expanded, the campus central heating systems were upgraded and moved to a new location--therefore, the Old Central Heating Facility was converted to the Nuclear Laboratory Building. The building is a steel, concrete, brick, and concrete block stmeture. Some of the interior Brigham Young University June 6,1996 hdesssp\reportswyuWyu.002 l I l
1 a 6 l j walls consist of wood and sheetrock with tile or concrete flooring. There are three main rooms in ! the facility-the Reactor Room, the Control Room, and the Accelerator (Large) Room-which have a combined floor space of approximately 170 m2 (Figure 1). k, OBJECTIVES i L l
)
The objectives of this radiological survey were to provide independent document reviews and ;
.. . F
- radiological data for use by the NRC in evaluating the adequacy and accuracy of the licensee's final
- radiological survey report relative to established guidelines.
l i . DOCUMENT / DATA REVIEW i i i [ ESSAP reviewed the licensee's radiological survey data (BYU 1994,1995a,1995b, and 1995c). ' j Procedures and methods utilized by the licensee were reviewed for adequacy and appropriateness. l The data were reviewed for accuracy, completeness, and compliance with guidelines. Comment l I letters, documenting these reviews, were submitted to the NRC on July 25,1994; August 7,1995; I i and, November 17,1995(ORISE 1994,1995a, and 1995b). ) PROCEDURES l i i i During the period of April 10 and 11,1996 ESSAP performed a radiological survey of the Nuclear ; i Laboratory and remaining L-7'7 Research Reactor components. The survey was performed in
. acco:Jance with a site-specific survey plan which was submitted to and approved by the NRC i
) Region IV Office (ORISE 1996). Survey activities included a visual inspection and independent f
- measurements and sampling of the reactor facility and the reactor components to be released for l l unrestricted use. This report summarizes the procedures and.results of the survey. ;
i 1 L Brigham Young University I June 6,1996 h:\essap\ reports \byu\byu.002 l
SURVEY PROCEDURES The following procedures apply to the reactor components and to the Reactor Room, Control Room, and Accelerator Room within the L-77 Research Reactor facility. REFERENCE GRID The reference grid system established by the licensee was used where possible. Measurement locations on ungridded surfaces were referenced to prominent building features or the existing grid. Measurement and sampling locations on reactor components were referenced to reactor component nomenclature and identification numbers provided by the licensee. SURFACE SCANS Surface scans for alpha, beta, and gamma activity were performed on the floor and lower wall surfaces in the Reactor Room, the Accelerator Room, and the Control Room, using large-area and hand-held gas proportional, ZnS scintillation, and NaI scintillation detectors coupled to ratemeters or ratemeter-scalers with audible indicators. A 75% scan of the floor and 50% scan of the lower wall surfaces was performed in each surveyed room. Scans were also performed over 50 to 100% of the reactor component surfaces. SURFACE ACTIVITY MEASUREMENTS Background measurements of surface activity on poured concrete, concrete blocks, and sheet metal were performed at building locations not having a history of radioactive materials use. Direct measurements for total alpha and total beta activity were performed at 47 locations on the floor, walls, and overhead surfaces. An additional 14 direct measurements were performed on reactor components. One smear for the detection of removable activity was collected at each direct measurement location, with the exception oflead shot used for shielding, where no smear was taken. l Brighara Young University June 6.1996 h:\essap\ reports \byusbyu.002
1 l i These measurements were performed using gas proportional and ZnS scintillation detectors coupled to ratemeter-scalers. Measurement locations are shown on Figures 2 through 4. BYU stored the lead shot used for shielding in 55 gallon dmms labeled as LDP 1&2, LDP 3&4, etc. A lead shot sample from three of the dmms was collected from the surface at the top of the drum and the total beta activity determined. From drum LDP 1&2, a second sample, at 8 inches from the top surface, was also collected and the total beta activity was determined.
~ EXPOSURE RATE MEASUREMENTS Background exposure rates were determined for the building interior at 3 locations of similar construction but without a history of radioactive materials use. Facility exposure rates were measured at 5 locations (Figure 1). Measurements were performed at 1 m above the surface using a microrem meter.
SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data were returned to ORISE's ESSAP laboratory in Oak Ridge, Tennessee for analysis and interpretation. Smears were analyzed for gross alpha and gross beta activity using a low-background gas proportional counter. Smcar results and direct measurements for surface activity 2 were converted to units of disintegrations per minute per 100 cm (dpm/100 cm2 ). Exposure rates were reported in units of microroentgens per hour (pR/h). The data generated were compared to NRC guidelines established for unrestricted use (NRC 1974). Brigham Yong Universh . June 6.1996 h:\essap\reportsibyu\byu.002 l
FINDINGS AND RESULTS SURFACE SCANS Surface scans did not identify any areas of elevated direct radiation throughout the L-77 Research Reactor facility or on any of the reactor component surfaces. SURFACE ACTIVITY LEVELS Survey activity measurements and smear results are presented in Table 1. All total alpha activity measurements were below the minimum detectable concentration (MDC) of 59 dpm/100 cm2 . Total 2 beta activity ranged from less than 290 to 510 dpm/100 cm . Removable activity at all locations was less than the minimum detectable concentrations of 10 and 14 dpm/100 cm2 for gross alpha and gross beta, respectively. EXPOSURE RATES Exposure rates performed at one meter above the surface at the three background and five site locations are presented in Table 2. Background exposure rates ranged from 8 to 9 pR/h with an average of 9 pR/h. Exposure rates inside the reactor facility ranged from 8 to 10 pR/h. COMPARISON OF RESULTS WITH GUIDELINES Due to historical records which indicate the presence of beta-gamma emitters, strontium-90, and uranium, several NRC guidelines are applicable. The most restrictive surface activity guidelines are those for Sr-90. The applicable NRC guidelines for Sr-90 surface activity levels are (NRC 1974): Total Activity 1,000 -y dpm/100 cm 2, averaged over a 1 m2 area 3,000 p-y dpm/100 cm 2, maximum in a 100 cm2 area Brigharn Young University June 6,1996 hNisap\reporuWyuWyu.002 i I . . . - .
l I Removable Activity 200 p-y dpm/100 cm2 . The applicable NRC guidelines for uranium surface activity levels are (NRC 1974): Total Activity 5,000 a dpm/100 cm 2, averaged over a 1 m2area 15,000 a dpm/100 cm2 , maximum in a 100 cm2 area Removable Activity 1,000 a dpm/100 cm2 Surface activity levels at all direct measurement locations were below the guidelines for both average and maximum alpha and beta activity. All removable activity levels were below the applicable guidelines, as well. The exposure rate guideline is 5 R/hr above background (NRC 1991). All site exposure rate measurements were within the guideline limit.
SUMMARY
On April 10 and 11,1996, ESSAP performed a radiological survey of the L-77 Research Reactor components and the Nuclear Laboratory Building at Brigham Young University in Provo, Utah. Survey activities included surface scans, direct measurements of total and removable activity, and exposure rate measurements. Survey results indicated that fixed and removable activity levels were less than the applicable guidelines. All exposure rates were below guidelines and were consistent with background levels. ESSAP's survey findings are consistent with the licensee's measurements and support the licensee's conclusion that the radiological conditions of the surveyed areas and reactor components satisfy the NRC guidelines for release to unrestricted use. i Brigham Young thtversity June 6,1996 h:\essap\ reports \byu\byu.002
262-002 (2) ' LARGE ROOM (ACCELERATOR ROOM) g5
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h JL MEASUREMENT / SAMPLING T LOCATIONS
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0 6 M METERS FIGURE 1: Research Reactor Facility - Exposure Rate Measurement Locations Brigham Young Universky - June 6,1996 h:\ensap\reportsibyu\byu.002 I
262-004(1) RRF1 RRF2AA RRF3 RRF4 A RRF5 16 21 15 2 6 5
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,4 9 10 19 11 12 Y Y YY Y MEASUREMENT / SAMPLING N LOCATIONS j i 9 SINGLE-POINT FLOOR JL T
A SINGLE-POINT WALLS SINGLE-POINT CEIUNG N NOT TO SCALE FIGURE 2: Reactor Room - Measurement and Sampling Locations ( Brigbasa Young Universky - June 6,1996 b:kssap\ reports \byu\byu.002
262-003(1) CRF1 6 CRF2 A 10 47 CRF3 5 *4 O CRF5 (Console)
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NOT A PART OF RESEARCH REACTOR FACIUTY CRF4 8 g3 gl V N h MEASUREMENT /SAMPUNG JL LOCAll0NS T e SINGLE-POINT FLOOR h A SINGLE-POINT WALLS FEET 0 6 M 0 METERS FIGURE 3: Control Room - Measurement and Sampling Locations Brigham Young Universky - June 6,1996 b:WaapWponsWyuWyu.002
262-005(1) 3 e
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N MEASUREMENT / SAMPLING H LOCATIONS x T e SINGLE-POINT FLOOR A SINGLE-POINT WALLS j( NOT TO SCALE FIGURE 4: Large Room (Accelerator Room) - Meesurement and Sampling Locations Brigbarn Young University June 6,1996 h:\cusap\reponsWyu%yu.002
TABLE 1
SUMMARY
OF SURFACE ACTIVITY LEVELS FOR THE L-77 RESEARCH REACTOR BRIGHAM YOUNG UNIVERSITY PROVO, UTAH Location / Total Activity (dpm/100 cm2) Removable Activity (dpm/100 cm2 ) Grid Block' Alpha Beta Alpha Beta Reactor Room 6 1/RRF 13 <59 350 <10 <14 2/RRF 5 <59 <290 <10 <14 3/RRF 6 <59 <290 <10 <14 4/RRF 22 <59 <290 <10 <14 5/RRF 4 <59 <290 <10 <14 6/RRF 1 <59 <290 <10 <14 7/RRF 11 <59 <290 <10 <14 8/RRF 20 <59 <290 <10 <14 9/RRSW 3 <59 <230 <10 <14 10/RRSW 5 <59 <290 <10 <14 11/RRSW 8 <59 <290 <10 <14 10/RRSW 14 <59 <290 <10 <14 13/RREW 10 <59 <290 <10 <14 14/RREW 2 <59 <290 <10 <l4 15/RRNW 9 <59 <290 <10 <14 15/RRNW 2 <59 <290 <10 <14 17/RRWW l <59 <230 <10 <14 18/RRWW 5 <59 <230 <10 <14 19/RRSW 7 <59 240 <10 <14 Brigham Young University - June 6,1996 12 n;s,,,,ps,,po,, sdyusdyo.co2
TABLE 1 (Continued)
SUMMARY
OF SURFACE ACTIVITY LEVELS FOR THE L-77 RESEARCH REACTOR BRIGHAM YOUNG UNIVERSITY PROVO, UTAH Location / Total Activity (dpm/100 cm2) Removable Activity (dpm/100 cm2 ) Grid Block Alpha Beta Alpha Beta 20/RRC 18 <59 <270 <10 <14 21/RRNW 2 <59 450 <10 <l4 22/RRC 8 <59 <270 <10 <14 Control Room
- 1/CRF 4 <59 <290 <10 <14 2/CRF 5 <59 <290 <10 <14 3/CRF 4 <59 <290 <10 <14 4/CRF 3 <59 <290 <10 <14 5/CRF 2 <59 <290 <10 <14 6/CRNW 3 <59 <290 <10 <14 7/lWB 7 <59 <290 <10 <14 8/CRSW 10 <59 <290 <10 <14 9/ CREW 2 <59 420 <10 <14 10/CRNW 2 <59 510 <10 <14 Accelerator (Large) Room d 1/LRF 8 <28 <290 <10 <14 2/LRF 7 <28 <290 <10 <14 3/LRF <28 <290 <10 <14 4/LRF 5 <28 <290 <10 <14 5/LRF 4 <28 <290 <10 <14 6/LRF 4 <28 <290 <10 <14 l
Brigham Young Uruversity - June 6, IM 13 n:1,,,,,s,,po,1.sbyusdru.oo2
TABLE 1 (Continued)
SUMMARY
OF SURFACE ACTIVITY LEVELS FOR THE L-77 RESEARCH REACTOR BRIGHAM YOUNG UNIVERSITY PROVO, UTAH Location / Total Activity (dpm/100 cm2) Removable Activity (dpm/100 cm 2) Grid Block Alpha Beta Alpha Beta 7/LRF 3 <28 <290 <10 <14 8/LRF 3 <28 <290 <10 <14 9/LRF 2 <28 <290 <10 <14 10/LRF 1 <28 <290 <10 <14 11/LRW 4 <28 <290 <10 <14 10/LRW 5 <28 <290 <10 <14 13/LRW 9 <28 <290 <10 <14 14/LRT 10 <28 <290 <10 <14 15/LRW l1 <28 <290 <10 <14 Reactor Components' MB41 <28 <270 <10 <14 SS 17 <28 <270 <10 <14 A2 <28 <270 <10 <14 MB 18 <28 <270 <10 <14 M 26 <28 <270 <10 <14 MB 46 <28 <270 <10 <14 A 10 <28 <270 <10 <14 MB 50 <21, <270 <10 <14 SS 5 <28 <270 <10 <14 HEPA Filter NA <270 NA NA i Brigham Young University June 6,1996 4 h:kssap\reportsWyuWyu.(X12
TABLE 1 (Continued)
SUMMARY
OF SURFACE ACTIVITY LEVELS FOR THE L-77 RESEARCH REACTOR BRIGHAM YOUNG UNIVERSITY PROVO, UTAH Location / Total Activity (dpm/100 cm 2) Removable Activity (dpm/100 cm2) Grid Block Alpha Beta
, Alpha Beta Lead Diphenyl Shot' LDP1&2 NA8 <270 NA NA LDP 1&2,8" NA <270 NA NA LDP3&4 NA <270 NA NA LDPS&6 NA <270 NA NA
-s
'The grid identification numbers, as identified by BYU are as follows: RR = Reactor Room; CR =
Control Room; LR = Large Room; F = floor; SW = south wall; EW = east wall; NW = north wall; WW = west wall; C = ceiling; MP = metal peices from reactor that were scanned only; SS = stainless steel pieces from reactor; A = aluminum pieces from reactor; MB = mild steel beta measurement; IWB = internal wall bathroom; and LRT = Large Room Top. bRefer to Figure 2.
' Refer to Figure 3.
d Refer to Figure 4.
'The L-77 research reactor was dismantled by BYU; the identification numbers are those placed on each individual piece by BYU.
LDP is the lead diphenyl shot used as shielding in the reactor. BYU had the lead shot stored in 55 gallon drums labeled as LDP 1&2, LDP 3&4, etc. A lead shot sample from the drum was collected from the surface of the drum and the total beta activity determined. From drum LDP 1&2, a second sample, at 8 inches was also collected and the total beta activity was determined. 8NA = Not Applicable. I Brigham Young University - June 6. tu 15 nu,,,,s,,po,tisbyuseyu.co2
TABLE 2 EXPOSURE RATES FOR THE L-77 RESEARCH REACTOR BRIGHAM YOUNG UNIVERSITY PROVO, UTAH Location Exposure Rate at I m (pR/h) Reactor Facility
- 1 - Reactor Room 10 2 - Reactor Room 10 3 - Control Room 9 4 - Accelerator Room 9 5 - Accelerator Room 8 Background
- CluffBuilding, Southwest Foyer 9 CluffBuilding, Southeast Foyer 8 CluffBuilding, Northeast Foyer 9 Refer to Figure 1.
' Figure not provided.
t Brigham Young Univenky . June 6. IM Ib b:\essap\repons\byu\byu.002
REFERENCES Brigham Young University (BYU). Decommissioning Survey for the L-77 Research Reactor, Brigham Young University. Provo, Utah; April 15,1994. Brigham Young University. RE: Reply to the critique of the survey of the L-77 reactor facility at Brigham Young University. Provo, Utah; May 30,1995a. l Brigham Young University. RE: Reply to Letter dated August 22,1995. Provo, Utah; October 9, i 1995b. Brigham Young University. RE: Additional Direct Beta Measurements from the Brigham Young University Reactor Facility. Provo, Utah; December 7,1995c. Oak Ridge Institute for Science and Education (ORISE). D_ocument Review - Decommissioning Survey for the L-77 Research Reactor, Brigham Young University (Docket No. 050-262). Oak Ridge, TN; July 25,1994. Oak Ridge Institute for Science and Education. Document Review - Decommissioning Survey for the L-77 Research Reactor, Brigham Young University (Docket No. 050-262). Oak Ridge, TN; August 7,1995a. Oak Ridge Institute for Science and Education. Document Review - Decommissioning Survey for l the L-77 Research Reactor, Brigham Your.g University (Docket No. 050-262). Oak Ridge, TN; November 17,1995b.
' Oak Ridge Institute for Science and Education. Proposed Radiological Survey for the Brigham Young University L-77 Research Reactor, Provo, Utah (Docket No. 050-262, RFTA No. 94-332).
Oak Nge, TN; March 26,1996. U.S. Nuclear Regulatory Commission (NRC). Termination of Operating Licenses for Nuclear Reactors, Regulatory Guide 1.86, Washington, D.C., June 1974. U.S. Nuclear Regulatory Commission. Office of Nuclear Safety and Safeguards, Review Plan: Evaluating Decommissioning Plans for Licensees Under 10 CFR Pans 30,40, and 70, Washington, D.C.,1991. I Bngham Your.g University - June 6. IN 7 h:\ css @\reporuibyu\byu 002
l , APPENDIX A MAJOR INSTRUMENTATION Brighara Ymag Undversey June 6,1996 h:\essap\reporu\byu\hyu 002
\
6 o
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 Bicron Micro-Rem Meter (Bicron Corporation, Newburg, OH) Eberline Pulse Ratemeter Model PRM-6 l (Eberline, Santa Fe, NM) Ludlum Floor Monitor Model 239-1 (Ludlum Measurements, Inc., Sweetwater, TX) Ludlum Ratemeter-Scaler l Model 2221 (Ludlum Measurements, Inc., Sweetwater, TX) Detectors Eberline ZnS Scintillation Detector t Model AC-3-7 Effective Area,74 cm2 (Eberline, Santa Fe, NM) Ludlum Gas Proportional Detector Model 43-37 Effective Area,550 cm2 (Ludlum Measurements, Inc., Sweetwater, TX) i Brigham Young Universny . June 6,1996 b \essap\repons\byu\byu.002
Ludlum Gas Proportional Detector Model 43-68 { Effective Area,126 cm 2
- (Ludlum Measurements, Inc.,
Sweetwater, TX) Victorcen NaI Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal (Victorcen, Cleveland, OH) LABORATORY ANALYTICAL INSTRUMENTATION low Background Gas Proportional Counter i Model LB-5100-W
'(Oxford, Oak Ridge, TN) t Brtsham Young Universky . June 6.1996 h:Wenap\rtpons\byu\byu 002
i APPENDIX B SURVEY AND ANALYTICAL PROCEDURES i l I l I l 1 i i Brigham Young UWversuy June 6,1996 A-3 ,,g,,,,,,, i l l
APPENDIX B SURVEY AND ANALYTICAL PROCEDURES I 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 2 areas. Other surfaces were scanned using small area (74 cm or 126 cm2) hand-held detectors. Identification of elevated levels was based on increases in the audible signal from the recording and/or indicating inctrument. Combinations of detectors and instruments used for the scans were: Alpha - gas proportional detector with ratemeter-scaler ZnS scintillation detector with ratemeter-scaler Beta - gas proportional detector with ratemeter-scaler Gamma - NaI scintillation detector with ratemeter Surface Activity Measurements Measurements of total alpha and total beta activity levels were performed using ZnS scintillation and gas proportional detectors with ratemeters-scalers. Count rates (cpm), which were integrated over 1 minute in a static position, were converted to 2 activity levels (dpm/100 cm ) by dividing the net rate by the 4r efficiency and correcting for the active area of the detector. Because different materials (poured concrete and concrete block, brick, metal, and wood) may have very different background levels, average background counts usw v veme .w 6. ms B-1 % % ,,,, ,
i [ ( were determined for each material encountered during the survey activities--on a similar material having no known radiological history at a location separate from the surveyed area. The alpha activity background countrate for the ZnS scintillation detector and the gas proportional detector averaged I cpm for each detector. The alpha efficiency factor was 0.17 for the ZnS scintillation detector and 0.21 for the gas proportional detector. Both detector types were calibrated with a ( Th-230 calibration source. The beta activity background count rate for the gas proportional detector averaged 337 cpm for concrete and brick surfaces, 287 cpm for metal surfaces, and 206 cpm for wood surfaces. The beta efficiency factor was 0.24 for the gas proportional detector which was calibrated with a Tc-99 calibration source. The alpha minimum detectable [ 2 concentration (MDC) was 59 dpm/100 cm for the ZnS scintillation detector and 28 dpm/100 cm2 [ for the gas proportional detector. The beta activity MDCs for the gas proportional detector were 2 290 dpm/100 cm for concrete and brick surfaces, 270 dpm/100 cm2 for metal surfaces, and 230 2 dpm/100 cm for wood surfaces. The effective windows for the ZnS scintillation and gas proportional detectors were 74 cm2 and 126 cm2, respectively.
) Remavable Activity Me m-;- -uM Removable activity levels were determined using numbered filter paper disks,47 mm in diameter.
Moderate pressure was applied to the smear and approximately 100 cm2 of the surface was wiped. Smears were placed in labeled envelopes with the location and other pertinent information recorded. Frnacure Data Mencuramanic Measurements of dose equivalent rates (prem/h) were performed at 1 m above the surface using a Bicron microrem meter. Although the instmment displays data in rem /h, the rem /h to R/h conversion is essentially unity, i Bngham Young Univerniry - June 6, tu B-2 n s,,,,ps,epon wyuwyo.oo2
I ANALYTICAL PROCEDURES Removable Activity Smears were counted on a low background gas proportional system for gross alpha and gross beta activity. UNCERTAINTIES AND DETECTION LIMITS Detection limits, referred to as minimum detectable concentrations (MDC), were based on 2.71 plus 4.65 times the standard deviation ofthe background count [2.71 + (4.65/BKG)]. When the activity was determined to be less than the MDC of the measurement procedure, the result was reported as less than MDC. Because of variations in background levels and measurement efficiencies, the detection limits differ from instrument to instrument. CALIBRATION AND QUALITY ASSURANCE.
. Calibration of all field and laboratory instrumentation was based on standards / sources, traceable to NIST, when such standards / sources were available. In cases where they were not available, standards of an industry recognized organization was used. Calibration of the Bicron micrcrem meter was performed by an in-house instrument specialist.
Analytical and field survey activities were conducted in accordance with procedures from the following ESSAP documents: Survey Procedures Manual, Revision 9 (April 1995) Laboratory Procedures Manual, Revision 9 (January 1995) Quality Assurance Manual, Revision 7 (January 1995) Brighamn Young Universky . June 6,1996 h:kssap\reponseyuwyu 002
' 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.
. Quality control procedures include:
*. Daily instrument background and check-source measurements-at the beginning, during the middle, and at the end of each day-are perfo.v.ed 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.
% y u.ev.,.isy.s s.i,96 'M m:s psrepon.sny sby..oo2
I I f { ( APPENDIX C REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS
/
Brigham Young Universky . June 6,1996 h:\casap\reponsibyu\byu 002
U.S. ATOMIC ENERGY COMMISSION Juna 1974 I 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 f retain, with the Part 50 license, authorization for ( Section 50.51, " Duration of license, 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 fmm the facility. Appropriate administrative controls I 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 specifkations 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 Committw 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. DI5CUSSION 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 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 redu .x! surveillance Commission to determine whether unrestricted access requirements in that periodic surveillance of 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, 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 GUlDES c,,,,,,,u,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, negui.., ouiese ne mouse to seect be end mee ee.nen. u sne puba, metness acceptabse to the AEC regulatory etsff of emplemensme speestic paeto
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or r .e,,e,. - s r g;";c = ,,ut,^ = ,g ;e ~ ~ *- + semphones wkh them . not reewired. Memede end eelut. ens different from those set out en the giedes win be acceptones if they preeade e bee. for the g , fmdmos ,eeeerte to the neuence er continuerne of a perrmt er hcor.e by the Cenm n , , ,
- 2. Reegersh and test Reectors 7. Tiensportation
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Radioactive components may be either shipped 2. ALTERNATIVES FOR REACTOR off-site for burid ct an euthorized burid ground or RETIREMENT secured on the site. Those radioactive raaterials remaining on the site must be isolated from the public Four alternatives for retirement of nuclear reactor by physical barriers or other means to prevent public facilities are considered acceptable by the access to hazardous levels of radiation. Surveillance is Regulatory staff. These are: necessary to assure the long term integrity of the barrien. He amount of surveillance required depends a. Mothballing. Mothballing of a nuclear reactor upon (1) the potential hazard to the health and safety of facility consists of putting the facility in a state of the public from radioactive material remaining on the protective storage. In general, the facility may be site and (2) the integrity of the physical barriers. left intact except that all fuel assemblies and the Before areas may be released for unrestricted use, they radioactive fluids and waste should be removed must have been decontaminated or the radioactivity from the site. Adequate radiation monitoring, must have decayed to less than prescribed limits environmental surveillance, and appropriate security (Table 1). procedures should be established under a possession-only license to ensure that the health and The hazard associated with the returned facility is safety of the public is not endangered. evaluated by considering the amount and type of remaining contamination, the degree of confinement of b. In-Place Entombment. In-place entombment the remaining radioactive materials, the physical consists of sealing all the remaining highly security provided by the confmement, the susceptibility radioactive or contaminated components (e.g., the to release of radiation as a result of natural phenomena, pressure vessel and reactor intemals) within a and the duration of required surveillance. structure integral with the biological shield after having all fuel assemblies, radioactive fluids and C. REGULATORY POSITION wastes, and certain selected components shipped offsite. The structure should provide integrity over
- 1. APPLICATION FOR A LICENSE TO POSSESS the period of time in which significant quantities BUT NOT OPERATE (POSSESSION-ONLY (greater than Table I levels) of radioactivity remain LICENSE) with the material in the entombment. An appropriate and continuing surveillance program A request to amend an operating license to a should be established under a possession-only possessionenly license should be made to the Director license.
of Licensing, U.S. Atomic Energy Commission. Washington, D.C. 20545. "Ile request should include c. Removal of Radioactive. Components and the following information: Dismantling. All fuel assemblies, radioactive fluids and waste, and e ier materials having activities
- a. A description of the current status of the facility. above accepted tr - .tricted activity levels (Table 1) should be retre al from the site. The facility
- b. A description of measures that will be taken to owner may ' ne have unrestricted use of the site prevent criticality or reactivity changes and to with no requ. . ment for a license. If the facility minimize releases of radioactivity from the facility. owner so desires, the remainder of the reactor facility may be dismantled an:1 all vestiges removed
- c. Any proposed changes to the technical and disposed of.
specifications that reflect the possession-only facility status and the necessary disassembly / retirement d. Conversion to a New Nuclear System or a activities to be performed. Fossil Fuel System. This attemative, which applies only to nuclear power plants, utilizes the existing
- d. A safety analysis of both the activities to be turbine system with a new steam supply system.
accomplished and the proposed changes to the The original nuclear steam supply system should be technical specifications. separated from the electric generating system and disposed ofin accordance with one of the previous
- e. An inventory of activated materials and their three retirement alternatives.
location in the facility. Note: Section electronically reproduced from photocopy. C-2
. - - . ~ _ - - _ _ ~ . - . _ - . . - - - . . - - _ ~ . - _ . . . - - -
l
- 3. SURVEILLANCE AND SECURITY FOR THE f. Administrative procedures should be established RETIREMENT ALTERNATIVES . WHOSE for the notificction an6 reporting of tbnormal FINAL STATUS. REQUIRES A occurrences such as (1) the ertrance of an unauthorized POSSESSION-ONLY LICENSE person or persons into th: facJity and (2) a significant
. change in the radiation or contadnation levels in the A facility which has been licensed under a facility or the offsite environment.
possession-only- license may contain a significant amount of radioactivity in the form of activated and g. The following reports should be made: contaminated hardware and structural materials. Surveillance and commensurate security should be (1) An annual report to the Director of provided to assure that the public health and safety are Licensing, U.S. Atomic Energy Commission, not endangered. Washington, D.C. 20545, describing the results of the
- a. Physical security to prevent inadvertent exposure environmental and facility radiation surveys, the status of personnel should be provided by multiple locked of the facility, and an evaluation of the performance of barriers. The presence of these barriers should make security and surveillance measures.
it extremely difficult for an unauthorized person to gain access to areas where radiation or contamination levels (2) An abnormal occurrence report to the exceed those specified in Regulatory Position C.4. To Regulatory Operations Regional Office by telephone prevent inadvertent exposure, radiation areas above within 24 hours of discovery of an abnormal 5 mR/hr, such as near the activated primary system of occurrence. The abnormal occurrence will also be a power plant, should be appropriately marked and reported in the annual report described in the preceding should not be accessible except by cutting of welded item. closures or the disassembly and removal of substantial structures and/or shielding material Means such as a h. Records or logs relative to the following items remote-readout . intrusion alarm system should be should be kept and retained until the license is provided to indicate to designated personnel when a terminated, after which they must be stored with other physical barrier is penetrated. Security personnel that plant records: provide access control to the facility may be used instead of the physical barriers and the intrusion alarm (1) Environmental surveys, systems. (2) Facility radiation surveys,
- b. The physical barriers to unauthorized entrance into the facility, e.g., fences, buildings, welded doors, (3) Inspections of the physical barriers, and and access. openings, should be inspected at least quarterly to assure that these barriers have not (4) Abnormal occurrences.
deteriorated and that locks and locking apparatus are intact.
- 4. DECONTAMINATION FOR RELEASE FOR
- c. A facility radiation survey should be performed UNRESTRICTED USE
- at least quarterly to verify that no radioactive material is escaping or being transported through the If it is desired to terminate a license and to j containment barriers in the facility. Sampling should eliminate any further surveillance requirements, the !
be done along the most probable path by which facility should be sufficiently decontaminated to prevent radioactive material such as that stored in the inner
)
risk to the public health and safety. After the l containment regions could be transported to the outer decontamination is satisfactorily accomplished and the i regions of the facility and ultimately to the environs. site inspected by the Commission, the Commission may I authorize the license to be terminated and the facility
- d. An environmental radiation survey should be abandoned or released for unrestricted use. The performed at least semiannually to verify that no licensee should perform the decontamination using the significant arnounts of radiation have been released to following guidelines:
the environment from the facility. Samples such as ! soil, vegetation, and water should be taken at locations a. The licensee should make a reasonable effort to for which statistical data has been established during eliminate residual contamination. reactor operations. I
- b. No covering should be applied to radioactive
- e. A site representative should be designated to be surf.ces of equipment of structures by paint, plating, or responsible for controlling authorized access into and other covering material until it is known that movement within the facility. contamination levels (determined by a survey and documented) are below the limits specified in Table 1.
Note: Section electronically reproduced from photocopy. C-3
-. - - - _-.-.. - - - - - - - - . - - - _ -.- -_. ~ In addition, a reasonable fiort should be made (r.nd (4) State the fmding of the survey in units specified documented) to further minimize contamination prior to in Table 1. any such covering. After review of the report, the Commission may
- c. The radioactivity of the interior surfaces of inspect the facilities to confirm the survey prior to pipes, drain lines, or ductwork should be determined granting approval for abandonment.
by making measurements at all traps and other appropriate access points, provided contamination at 5. REACTOR RETIREMENT PROCEDURES these locations is likely to be representative cf contamination on the interior of the pipes, drain lines, As indicated in Regulatory Position C.2, several or ductwork. Surfaces of premises, equipment, or alternatives are acceptable for reactor facility scrap which are likely to be contaminated but are of retirement. If minor disassembly or "mothballing" is ! such size, construction, or location as to make the planned, this could be done by the existing operating ! surface inaccessible for purposes of measurement and maintenance pmcedures under the license in effect. l should be assumed to be contaminated in excess of the Any planned actions involving an unreviewed safety ;
- permissible radiation limits. question or a change in the technical specifications {
should be reviewed and approved in accordance with
- d. Upon request, the Commission may authorize a the requirements of 10 CFR i 50.59.
licensee to relinquish possession or contml of premises, equipment, or scrap having surfaces contaminated in If major structural changes to radioactive excess of the limits specified. This may include, but is components of the facility are planned, such as removal not limited to, special circumstances such as the of the pressure vessel or major components of the transfer of premises to another licensed organization primary system, a dismantlement plan including the ! that will continue to work with radioactive materials. infonnation required by i 50.82 should be submitted to Requests for such authorization should provide: the Commission. A dismantlement plan should be submitted for all the alternatives of Regulatory Position (1) Detailed, specific information describing the C.2 except mothballing. However, minor disassembly premises, equipment, scrap, and radioactive activities may still be performed in the absence of such contaminants and the nature, extent, and degree of a plan, provided they are permitted by existing residual surface contamination, operating and maintenance procedures. A i dismantlement plan should include the following: , (2) A detailed health and safety analysis indicating l that the residual amounts of materials on surface areas, a. A description of the ultimate status of the facility together with other considerations such as the , pmspective use of the premises, equipment, or scrap, b. A description of the dismantling activities and l are unlikely to result in an unreasonable risk to the the precautions to be taken. health and safety of the public. l
- c. A safety analysis of the dismantling activities
- c. Prior to release of the premises for unrestricted including any effluents which may be released.
use, the licensee should make a comprehensive radiation survey establishing that contamination is d. A safety analysis of the facility in its ultimate within the limits specified in Table 1. A survey report status. should be filed with the Director of Licensing, U.S. Atomic Energy Commission, Washington, D.C. 20545, Upon satisfactory review and approval of the with a copy to the Director of the Regulatory dismantling plan, a dismantling order is issued by the Operations regional Office having jurisdiction. The Commission in accordance with i 50.82. When report should be filed at least 30 days prior to the dismantling is completed and the Commission has been planned date of abandonment. The survey report notified by letter, the appropriate Regulatory should: Operations Regional Office inspects the facility and verifies completion in accordance with the (1) Identify the premises;- dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the Commission may (2) Show that reasonable effort has been made to terminate the license. If possession-only license under reduce residual contamination to as low as practicable which the dismantling activities have been conducted levels; or, as an alternative, may make application to the State (if an Agreement State) for a byproduct materials (3) Describe the scope of ;he survey and the general license. procedures followed; and Note: Section electronically reproduced from photocopy. C-4
i i TABLE 1 , ACCEPTABLE SURFACE CONTAMINATION LEVELS l Nuclide* Average6 ' Maximum6d Removable6 ' U-nat, U-235, U-238, and ; associated decay products 5,000 dpm a/100 cm 2 15,000 dpm a/100 cm 2 1,000 dpm a/100 cm 2 ! Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, Ac-227,1-125,1-129 100 dpm/100 cm 2 300 dpm/100 cm2 20 dpm/100 cm2 Th-nat, Th-232, Sr-90, Ra-223, Ra-224, U-232, I-126, I-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 Sr-90 and . others noted above. 5,000 dpm Sy/100 cm 2 15,000 dpm py/100 cm 2 1,000 dpm Sy/100 cm 2 ' '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. 6As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation. ' Measurements of average contaminant should not be averaged over more than I square meter. For objects ofless surface area, the average should be derived for each such object. dThe maximum contamination level applies to an area of not more than 100 cm 2, 2 %e 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 amouns of radioactive material on the wipe with an appropriate instrument of known efficiency, When removable contamination on objects ofless surface area is determir.ed, the pertinent levels should be reduced proportionally and the entire surface should be wiped. l r Note: Section electronically reproduced from photocopy. C-5
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