ML20196D522
| ML20196D522 | |
| Person / Time | |
|---|---|
| Site: | 05000124 |
| Issue date: | 01/31/1988 |
| From: | Murphy G OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20196D519 | List: |
| References | |
| CON-FIN-A-9076-3 ORAU-88-A-74, NUDOCS 8802170192 | |
| Download: ML20196D522 (61) | |
Text
.
ORAU 88/A-74 m
Prepared by Oak Ridge Associated CONFIRMATORY RADIOLOGICAL SURVEY Universities Prepared for OF THE U.S. Nuclear So*'n"I'is%n s ARGONAUT REACTOR FACILITY Region ll Office Supported by VIRGINIA POLYTECHNIC INSTITUTE Division of Industrial and AND STATE UNIVERSITY Medical Nuclear Safety BLACKSBURG, VIRGINIA G.L. MURPHY Radiological Site Assessment Program Manpower Education, Research, and Training Division FINAL REPORT JANUARY 1988 hbk A O 24 P
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ORAU 88/A 74 6
CONFIRMATORY RADIOLOGICAL SURVEY
^
o0F THE ARGONAUT REACTOR FACILITY.
VIRGINIA ' POLYTECHNIC INSTITUTE AND. STATE UNIVERSITY.
BLACKSBURG, VIRGINIA Prepared by LG. L. Murphy Radiological Site Assessment Program.
Manpower Educatica,.Research,'and Training' Division Oak Ridge Associated Universities Oak Ridge, Tennessee 37831-0117 Project Staff J.D. Berger R.C. Rookard R.D. Condra D.S. Styers*
F. A. Lange C.F. Weaver Prepared for U.S. Nuclear Regulatory Commission Region II Office FINAL REPORT January 1988 i'
This' re port is based on work performed under Interagency Agreement DOE No.
40-816-83 NRC Fin. No. A-9076-3 between the U.S.
Nuclear Regulatory Commission and the U.S.
Depa rtment of Energy.
Oak Ridge Associated Universitics performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy.
- Currently with:
ASG, Inc., Oak Ridge, TN.
u
G-rr TABLE OF CONTENTS l
Page
's 11
-List of Figures-
-List of Tables iii Introduction and Site = History 1
1 Site _ Description -
-2
' Survey Procedures Results 6
Comparison of Results With Guidelines 10
-Summary
.... eL......~........-............
11 References.
38 l
. Appendices L
Appendix A: Major Analytical Equipment i
Appendit B: Measurement and Analytical Procedures Appendix C: Regulatory Guide 1.86 - Termination of Opersting Licenses For Nuclear Reactors
(
Appendix D:
Proposed Confirmatory Survey Plan for the Argonaut l
Reactor Facility Virginia Polytechnic Institute and State University f
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t 1
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I LISr 0F FIGURES Page FIGURE 1: Map of Virginia, Showing Approximate Location of Virginia Polytechnic Institute and State University, Blacksburg, 13 Virgi.11a.
FIGURE 2 Map of Area Surroundirg VPT & SU Showing Locations of Background Measurements and Sampling Locations.
14 FIGURE 3:
Floor Plan of Robeton Hall............
15 FIGURE 4:
Room 6 Floor Plan Shosf.ag Locations of Measurements and Grid System Installed.
15 FIGURE 5:
Room SA Floor Plan Showing Locations of Measurements and Grid 17 System Installed...
FIGURE 6:
Room 8 Floor Plan Shosit:g Locations of Measurements and Grid System Installed.
18 FIGURE 7:
Room 8A Floor Plan S'.iowing Locations of Meaourebents and Grid System Installed......
.19
- ! G!>RE 8: Room 103 Floor Plan Shouing Locations of Measurements 20 FIGURE 9: Room 10 Floor Plan Showing Crid Blocl.s Surveyed and Grid 21 System Installed FIGURE 10:
Room 10 Floor Plan Showing Locations of Upper Wall 22 Measuremett ?.s.
FIGUPE 11: Room 10 Sump hhewing Lne.aiions of Contamination Measurecents 23 FIGURE 12: Graphic Representstion of Fuel Rod Storage Cells.
24 FLUURE 13 Room 10 Showi'ig Sampling Locations of Soil and Miscelinneous 25 Media
... c.......................
FIGURE 14:
Roou 10 Showica Lceations of PIC vs NaI Cross-Calibration Measurements..
26 FIGUPE 15: Map of Area Surroundtag VPI & SU Shouing Locations of VPI 27 Landfill...
i FIGURE 16: Map of VPI landfill Showitig locations of Sur f:.ce Soil and Concrete 28 Sampling.
~
7 f
11
I t
l LIST OF TABLES Pace 1ABLE 1: Direct Radiation Levels and Radionuclide Concentrations Measured at Baseline Sampling Locations.
29 TABLE 2t Direct Radiation Levels Measured at Locations in the 30 VTAR Facility...
31 TABLE 3:
Summary of Surface Contamination Measurements..
Tt.BLE 4:
Radionuclide Concentrations in Miscellaneous Media 34 TABLE 5: Radionuclide Concentrations in Soil Samples Collected From the Reactor Pit Area 35 TABLE 6:
Direct Radiation Levels and Radionuclide Concentrations in 36 Surface Soil Samples, Landfill...
TABLE 7:
Direct Radiation Levels and Radionuclide Concentrations in Miscellaneous Concrete and Surface Soil Samples, Fron Locations of Elevated Direct Radiation in the Landfill...
37 fii
- s s
s
-;I Y
hW CONFIRMATORY RADIOLOGICAL SURVEY' 0F THE ARGONAUT REACTOR FACILITY VIRGINIA POLYTECHNIC INSTITUTE ~AND STATE UNIVERSITY BLACKSBURG, VIRGINIA INTRODUCTION AND SITE HISTORY The Virginia Tech Argonaut'_ Reactor ' (VTAR) Facility is located in Robeson Hall on the - northwest - corner of _. the main -- campus of Virginia Polytechnic
-Institute and State University (VPI), -'between the Appalachian and Blu ' Ridge.
Mountains in Blacksburg, Virginia (Figure 1).
The VTAR is an Argor.
type research and training reactor, originally designed. and installed by American Standard - Nuclear Division.
The reactor was used as a part of the ' Nuclea r
._ Enginee ring curriculum for basic research in neutron physics, neutron radiography, neutron' activation analysis, technical training and Reactor Operator training.
The reactor began operation in June 1959, with a maximum power-level of 10 kW(th); the reactor was modified, and Nuclear Regulatory Commission (NRC) license number R-62 was amended to allow a' maximum power of-100 kW(th) in 1966. The reactor was shut down on July 14, 1983, and the license was amended for possession only in April 1985. -Reactor fuel was shipped to the Department of Energy in late 1985 and early 1986.
Dismantlement and decommissioning operations began in September 1985, and were completed in January - !?87 by Chem-Nuclear Systems, Inc., Columbia, South Carolina.
The Decommissioning Final Report was issued by VPI in April'1987.I At the request of the NRC, Region II, the Radiological Site Assessment Program of Oak Ridge Associated Universities (ORAU) conducted a radiological survey to evaluate the VPI reactor facilities radiological status relative to the NRC guidelines for unrestricted use.
f.
SITE DESCRIPTION o
The VTAR was located in Room 10, on the ground floor of Robeson Hall (Figure 2).
Room 10 has a high bay, which extends through the first floor elevation.
Rooms 6 (Anteroom), 6A (Electronics Shop), 8 (Sample Prep Room) 1 I
and 8A (Storeroom) are located adjacent to Room 10, and access doors connects Rooms 6 and 8 with Room 10.
On the first floor, Room 108 (Control Room) was accessible to Room 6 via a spiral staircase (Figure 3).
Room 108 has a
{
viewing window which permits direct observation into Room 10.
The VTAR core was heterogenous in design, using 93% enriched MTR type uranium-aluminum matrix fuel elements.
Thermal power output was limited to 100 kW, with light water used as a coolant and part of the moderator.
The remainder of the moderator consisted of graphite blocks which surrounded the fuel and water moderator.
Room 10 contained two fuel storaga racks.
The east rack was used only for unieradiated fuel, and consisted of 16 tubes imbedded in the concrete floor. The west rack was used for spent fuel, and was constructed similar to the east rack.
SURVEY PROCEDURES Document Review
/
ORAU reviewed the VPI final survey report and supporting documentation for the VTAR facility.
Approximately 10% of the raw data on a room by roon basis were to be compared to the results in the final report.
Facility Survey Gridding Confirmatory measurements were referenced to an ORAU grid system (2 m x 2 m), established on the floors and lower walls of rooms 6, 6A, 8, and 8A.
The licensee grid (6' x 6') was subdivided into 3'
x 3'
squares in room 10.
The upper walls, ceilings, and Room 108 were not gridded.
Mea sure me n t s and samples from the ungridded surfaces were referenced to the floor and lower wall grids, or to pertinent building features.
2
e 4
Exposure Rate Measurements Gamma exposure rates at 1 meter above the floor were measured at 12
-locations within the VTAR area, using NaI(TI) gamma - scintillation detectors cross calibrated onsite with a. pressurized ionization chamber.
Surface Scans and Measurement of Total and Removable Contamination Thorough, systematic alpha, beta-gamma, and gamma scans were performed on floors and lower walls (up to 2 m) using a gas proportional alpha / beta floor monitor, zine sulfide alpha detectors, "pancake" GM detectors, and NaI(TI) scintillation detectors coupled to scalers /ratemeters with audible indicators.
Representative areas on overhead surfaces (higher than 2 m) such as ledges, beams, pipes, ductwork and miscellaneous equipment were also scanned.
Thirty grid blocks on the floors and lower walls in the reactor high bay were randomly selected for surface contamination measurements (Figures 9-11).
Total measurements of alpha and beta-gamma contamination levels were systematically performed at the center and four points, midway between the center and block corners.
Smears for removable alpha and beta contamination were performed at the location in each grid block where the highest direct reading was obtained.
Total and removable contamination levels were also measured at 18 locations on the upper walls, ceilings and miscellaneous overhead objects.
Twenty-three single point measurements for total alpha and beta-gamma contamination were performed on the floors and lower walls of rooms 6, 6A, 8, 8A, and 108 (Figures 4-8).
Smears for removable alpha and beta contamination were also performed at each location.
Total and removable contamination
('
measurements were also performed at four random locations on the upper walls and ceilings.
Paint Samples Although historical data indicated the release of U-235 materials and/or fission products, beta-gamma surface scans would have detected residual 3
I contamination indicative of contamination beneath a painted surf ace.
No elevated beta-gamma levels associated with painted surf aces was detected and no paint samples were collec;ed, in accordance with the proposed survey plan l
(Appendix D).
Fliscellaneous Samples Four samples of concrete were collected from the reactor pit area, and a swipe was collected f rom a drainline located in the southeast corner of the reactor pit. A residue sample was collected f rom the f loor in Room 108 where a contaminated metal slug was f ound. The metal slug and contaminated wax were removed for dis pos al by Chen-Nucicar.
The origin of the item was not identif ied during the survey, and its apparent use was as a "door stop."
Room 10 Soil Samples Nine surf ace soil samples (0-15 cm) were collected from the reactor pit area.
These samples were collected f rom the exposed surf ace after remedial action uas completed.
Roof Beta-gamma and gamma scans were conducted on the roof of Robeson Hall.
Three single point direct meas u rements and s mears were collected from two exhaus ts on the wes t end of the building.
Landfill Survey tiis ce llaneous concrete, rubble and soil f rom the reactor pit area were disposed in a landf ill operated by VPI.
The materials were surveyed by Chen-Suclear prior to disposal.
The NRC requested that a limited survey of the landf ill be included as a part of the f acility survey.
4
e i
e
'?
h Cridding
' The area was ' not gridded due to time constraints.
Measurements were 2
. referenced-to pertinent land features.
Surface Scans Limited random walkover gamma scans were performed using NaI(TI)-
= scintill'ation detectors coupled to.ratemeters with audible-indicators.- Areas of elevated radivition levels were identified for later characterization.
Exposure Rate Measurements
' Gamma exposure rates were determined ~ at 4 random locations and at 4 locations of elevated radiation levels, identified by the walkover survey. -
Exposure re.tes were raeasured at the soil surface and at 1 m above the surface, using a NaI(T1) scintillation detector, cross calibrated against a pressurized ionization chamber.
Soil Sampling Surface soil samples were obtained at the location of each random gamma exposure rate measurement, and from areas of elevated surface radiation levels identified by walkover scans.
Miscellaneous Sampling Three concrete samples were collected from rubble piles where elevated radiation levels were detected.
Background Samples and Measurements Samples of surface soil were collected and background exposure rates were measured at 8 offsite locations (Figure 2) in the area around the VTAR facility to establish baseline radionuclide soil concentrations and direct 5
l radiation levels.
Als o, background gamma exposure rates at I meter above the f loor were meapured at 3 locations in Robeson Hall and 2 locations in Vallace Hall.
The areas in Robeson Hall did not have a prior his tory of us e of radioactive materials.
Sample Analysis and Interpretation of Results Soil and paint samples were analyzed by gamma spectrometry, and the spectra were reviewed f or identif iable photopeaks, with particular attention to U-238, U-235, Th-232, Co-60, Cs-137, Eu-152, Eu-154 and Ba-133.
Smea rs obtained for the determination of removable contamination were analyzed for gross alpha and beta activity.
Additional inf ormation concerning najor instrumentation, sampling equipment, and analytical procedures is provided in Appendices A and B.
Results were compared with NRC guidelines, es tablis hed f or telease of f acilities f or unrestricted use (Appendix C).
RESULTS Document Review In general, the decontamination plan appears to have been adequately developed and implemented to ensure the NRC guidelines were met.
The A
inf ormation contained in the final survey report did not adequately summarize the radiological s tatus of the site.
Dr. September 3,
1987, the licensee submitted additional data to the NRC in the form of Amendment 1.'
The amendment did clarif y the radiological status of the site, the survey procedures, and methodology. However, ORAU believes that the medium potential contamination areas should have been surveyed at a higher percentage coverage than the low potential contamination areas.
Als o, percentages of grid blocks surveyed (direct measurements) should only be used af ter the total area has been 100% scanned for all radionuclides involved.
The licensee's decommissioning fint.1 report indicated that a lead curtain was activated and could not be shipped to an active burial site.
The current s tatus of the lead curtain is :
6
l o currently stored at the VPI Waste Storage Building, which is in an isolated area of the campus.
o the item has been transferred to the VPI Broad license #45-09475-30 (See page 10 of Reference 1).
o the item will be encapsulated and shipped for disposal at an approved burial uite at a later date, upon approval of the encapsulation procedure.
No other mixed hazardous waste items were identified by VPI.
Background Levels and Baseline Concentrations Background exposure rates and baseline radionuclide concentrations in soil from the vicinity of the VPI site are presented in Table 1.
Exposure rates ranged from 9 to 15 LR/h, at one meter f rom the surface.
Uranium-238 concentrations ranged from
<0.7 pCi/g to 4.7 pCi/g.
Cobalt 60 and C s-137conce nt ra t ions were (0.1 pCi/g and from
<0.1 to 0.7 pCi/g; Th-232 concentrations ranged from 0.6 pCi/g to 1.8 pCi/g.
Interior exposure rate measurements from three background locations in Robesor. Hall ranged from 4 to 11 LR/h with an average of 8 $/h.
Two background neasurements f rom Wallace Hall averaged 8 LR/h, consistent with the average from Robeson Hall.
Faci'.ity Surver Exp>sure Rate Measurements One area of contamination was identified in Room 108 (low probability area) during the beta-gamma surface scans.
The area was pointed out to the licensee, and remediated during the ORAU survey.
No other contaminated areas were identified.
7
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I
' Gamma scans inside the VTAR facility indicated no elevated exposure rate levels above the background range of 4 to 11 wR/h (Table 2).
Contamination Measurements Results -of total and removable contamination measurements are summarized in Table 3.
Alpha and beta-gamma levels were = below the = release guidelines except f or one location in Room 108 (Figure 8).
The area was cleaned by the licensee, and' resurveyed.
The maximum alpha measurement (af ter remediation) 2 was
.150 dpm/100cm and the maximum beta-gamma measurement was 2
2 1400 dpm/100cm.
The highest alpha grid block average was 72 dpm/100cm 2
and the highest beta-gamma grid block average was 920 dpm/100cm.
The 2 and maximum removable alpha and beta contamination levcis were 7 dpm/100cm 2
14 dpe/100cm, respectively.
Miscellaneous Samples The radionuclide concentrations in f our concrete samples collected f rom the reactor pit area are presented in Table 4.
Two samples had ceasurable concentrations of Co-60 (1.0 pCi/g) and Eu-152 (3.7 pCi/g), while other radionuclide concenttations were less than the minimum detectable concentrations, or comparable to background levels.
A residue sample collected f rom exterior ductwork had measurable Co-60 2
2 2
(2200 dpm/100cm ),
Cs-137 (90 dpm/100cm ),
Eu-152- (2700 dpm/100cm ) and 2
Eu-154 (1600 dpm/100cm ).
The ductwork was removed by the licensee for proper dis pos al.
A residue sample collected from room 108 indicated the presence of U-238 (1500 dpm/100 cm ), while otter radionuclides of concern were nondetectable.
This area was cleaned by the licensee and resurveyed by ORAU.
The contamination levels following remediation were within the guideline levels.
A wipe pulled through a drain in the reactor pit area contained very low 2
2 levels of Co-60 (9
dpm/100cm ),
Eu-152 (13 dpm/100cm ),
and U-238 (30 dpm/100 cm#).
8
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) '
Soil Samples
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E The radionuclide concentrations in nine soil samples collected f rom the s
reactor pit are presented in Table 5.
The soil concentrations are generally
[
in the baseline range, and all concentrations except U-238 are less than 1 pCi/g.
Uranium 238 ranged f rom <l.0 to 4.5 pCi/g, which is consistent with e
background concentrations collected from the greater Blacksburg area.
6 I
r Roof Random beta-gamma and gamms scans of the roof and the interior of exhaust vents of Robeson Hall did not indicate the presence of any elevated radiation levels or measurable contamination.
Total alpha contamination was less than 2
21 dpm/100cm,
and the removable contamination ranged from
<2 to 2
5 dpm/100cm.
The total beta-gamma contamination ranged from 730 to 2
1600 dpm/100cm, and the removable beta contamination ranged from 7 to 2
8 dpm/100cm,
f
?
1.andfill Survev The VPI landfill (Figure 15) is located at the intersection of Price's Fork Road and Highway 460, slightly west of Blacksburg.
The landfill is f
located on VP1 property, and was originally used as a sanitary landfill.
Its
[
use is currently restricted to construction naterials and rubble. There is no T
historical data that indicates radioactive materials were disposed of in the g
landfill prior to the reactor decommissioning project.
4 Surface Scans h'
i E
Walkover surface scans conducted in the landfill identified four areas of 1
elevated radiation levels, (Figure 16), that were associated with large r
concrete block rubble piles.
The elevated areas ranged up to six times background.
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Exposure Rate Measurements Exposure _ rate _ measurements at 1 m above the surf ace f rom the random soil sampling locations averaged 11 to 15 4/h (Table. 6).-
Exposure rate-measurements at one meter _ f rom the elevated _ locations in the rubble piles could not be determined due to the shielding provided by the concrete blocks.
Contact exposure rate ' measurements were perf ormed in areas' where measurements at one meter were -not representative of the surf ace radiation levels.
The contact exposure rate measurements in the elevated areas renged f rom 27 to 78 wR/h.
Radionuclide Concentrations in Soil Radionuclide concentrations in soil are presented in Table 6 and 7.
Six of -eight samples have concentrations typical of background levels.
The two samples with elevated radionuclide concentrations were collected f rom areas of elevated radiation levels identified by the walkover surf ace scans.
Rubble Pile'-2 had concentrations as f ollows:
Co-60, 10.6 pCi/g; Eu-152, 30.1 pCi/g; and Eu-154, 23.3 pCi/g.
Rubble Pile-3 had concentrations of Co-60, 3.0 pCi/gt Eu-152, 1.4-pCi/g; and U-238, 7.1 pCi/g.
Miscellaneous Sample Concentrations Radionuclide concentrations in mis cellaneous concrete samples from the landfill are presented in Table 7.
Elevated concentrations of Co-60,-
38.3 pCi/g; Eu-152, 150 pci/g; and Eu-154, 110 pCi/g were measured in Sample Rubble Pile-5.
COMPARISON OF RESULTS WITH GUIDELINES NRC surf ace contaminat ton guidelines for release of facilities for unrestricted use are presented in Appendix C.
The principal radionuclides of interest are U-238, Co-60, Cs-137, and Eu-152.
The criteria for residual alpha contamination ist s
10 m.
Total Contamination' 15,000 dpm'100 cm2 (maximum in a-100 cm2
/
area)
-5,000.dpm/100 cm2 (averaged over 1 m ).
2 Renovable Contamination 2
~
't,000 dpm/100 cm For residual beta-gamma contamination, the NRC guidelines aret Total Contamination 2
15,000 dpm/100 cm2 (maximum in a 100 cm area).
2 5,000 dpm/100 cm2 (averaged over 1 m )
Removable Contanination 2
1,000 dpm/100 cm All total and removable alpha and beta-gamma levels as well as removable contamination measurements were within these guidelines.
Although site specific soil concentration guidelines were not developed for the VTAR facility, the radionuclide concentrations in the soils f rom the reactor pit area were within the baseline levels for soils in the southwest Virginia area.
Radionuclide concentrations in concrete, ventilation duct residue and drain residues were detectable.
Following removal of the exhaust duet and blower assembly from Room 10, the direct radiation levels were below guideline
' levels for surface contamination.
SUMMARY
On July 28-30, 1987, Oak Ridge Associated Universities performed a confirmatory radiological survey of the VTAR facility located in Robeson Hall 11
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Blacksburg, Virginia.
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r The sutvey included surface
- alpha, gamma and beta-gamma
- scans, measurement of direct and removable contamination levels, and the measurement l:(,;, -'
g.. -.,t of radionuclide concentrations in soil, concrete rubble, and tesiduo samples.
g C
The findings support the close-out survey performed by the licensee, and 5,
confirm that the radiological condi ions nf the VTAR f acility satisfy the NRC Pc. ?., - -
,.l' gridelines establishe.1 for release 'or unrestricted use.
The information from
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the landfill survey is inconclusi-e.
One soil sample contains radionuclide
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sites.
Exposu re rate measurement l ac'. 2 it not sufficient to determine if the g
concrete rubble meets the guidaMnes of 5 kR/h above background at one neter.
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Oak Ridge Associated l'riversities recommends that additional survey work be
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performed in the landfill area.
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@ CONCRETE SAMPLE N
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)
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TC PRICE'S FORK ROAD
'4 FIGURE 16:
Map of VPl Loadfill Showing Locations of Surface Soil and Concrete Sampling 28
--. w
.s Y
~
TABLE 1 DIRECT RADIATION LEVELS AND RADIONUCLIDE CONCENTRATIONS MEASURED AT BASELINE SAMPLING' LOCATIONS ARCONAUT REACTOR FACILITY l
VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY l
I Camma Exposure Rates at I m above the Surface Radionuclide Concentrations Locationa (PR/h)
Co-60 Cs-137-Eu-152 Eu-154 U-238 Th-232 1
13
<0.1 0.7 1 0.2b
<0.2
<0.1-
<0.7 1.8 1 0.5
- I
~
<l.2 1.0 1 0.4 2
11
<0.1 0.5 i 0.2
<0.1
<0.1
. 4.7~t.1.2 1.5 1 0.5 3
13
<0.1 0.4 2 0.1
<0.1
<0.1 4
12
<0.1 0.6 1 0.2
<0.2
<0.1
~<l.2
<0.6 5
12
<0.1 0.5 1 0.2
<0.2
<0.1 2.2 10.8 1.3 1 0.4 ~
6 9
<0.1 0.2 1 0.1
<0.1
<0.1
<0.9.
1.0 1 0.4 w
7 14
<0.1 0.7 1 0.2
<0.1
<0.1 1.6 1 1.3 1.5 1 0.8 8
15
<0.1
<0.1
<0.2
<0.1 l '+ 1 0. 6 1.5 2 0.3 aRefer to Figure 2.
bUncertainties are 20 based only on counting statistics.
/
\\
TABLE 2
[.-
DIRECT RADIATION LEVELS HEASURED AT LOCATIONS IN THE VTAR FACILITY l
VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY BLACKSBURG, VIRGINIA Room
. Figure Grid Block Gamma Exposure Races @ lm above the surf ace ( LR/h) 6' 4
F5 11 6A 5
F1 10 8
6 F5 8
108 8
9 i
-10 14 F34 5
I 10 F31 6
10 F22 5
f 10 F20 5
10 RS31 6
10 RS2/3 6
10 Center of Hole 11 I
k I
30
m
,- n
-m
--.- w
~.g
~,
i I
TABLE 3 i
l
SUMMARY
OF SURFACE CONTAMINATION K ASUREMENTS j
VTAR FACILITY VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY BLACKSBtRG, VIRGINIA l
Location Number of TOTAL CONTAMINATION REMOVABLE CONTAMINATION Number of Crid Blocks Alpha (dpm/100 cm )
Beta-Gamma (dpm/100 cm2)
At ha Range Beta Range Grid Blocks 2
2 2
Surveyeda Highest Grid Range of Highest Grid Range of (dpm/100 cm ) (dpm/100 cm )
Exceeding Block Avg.
Measurements Block Avg.
Measurements
' Criterla 6
l Floors 8 Lower 3
<21
<21
<390 5390
<2-
<5 - 14 0
Walls Upper Walls a 1
<21
<21
<390
<390 3
<5 0
Celling
~
6A Floors & Lower 2
31
<21 - 31
<390
<390
<2
<5 0
Walls Upper Walls, Docts 1
52 52
. <390
<390
<2 9
0 Cellings S
Floors & Lower 6
31
<21 - 31 1400
<390 - 1400 2-3
<5 - 12 0
Walls Upper Walls, Ducts 1
<21
<21
<390.
<390
<2
<5 0
Cellings
r TABLE 3 (Continued)
SUMMARY
OF SURFACE CONTAMINATION KASUREMENTS VTAR FACILITY VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY BLACKSBURG,' VIRGINIA Location Number of TOTAL CONTAMINATION REMOVABLE CONTAMINATION Number-of.
i 2
2 Grid Blocks Alpha (dpm/100 cm )
Beta-Ganuna (dpm/100 cm )
Alpha Range.
Beta Range Grid Blocks 2
2 Surveyed Highest Grid Range of Highest Grid Range of (dpm/100 cm ) (dpe/100 cm 3.
Exceeding' Block Avg.
Measurements Block Avg.
Measurements Criteria 84 Floors & Lo.c:-
1
<21
<21
<390
<390
<2
<5 0,
.i g
N Walls Upper Walls &
1
<21
<21 480 480
<2
<5 0
Celling
-l i
j 108 i
Floors & Lower 15
,200b D
b D
D -
' 5' 110D 1
1200 3000
<390-29000
<2-41 l
Walls Upper Walls, Ducts c
Ceilings 4
u l
.a
. TABLE 3 (Continued)
Supe 4ARY OF SURFACE CONTAMINATION MEASUREMENTS VTAR FACILITY VIRGlNIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY BLACKSBURG, VIRGINIA ah Location Number of TOTAL CONTAMINATION REMOVABLE CONTAMINATION
.Numbe. of-Grid Blocks Aloha (dpm/100 cm )
Beta-Gamme (dpm/100 cm )
Alpha Range
' Beta Range Grid Blocks l
Surveyed Highest Grid Range of Highest Grid.
Range of (dpm/100 ce$ (dpm/IOO cm$ ~
Exceeding Block Avg.
Measurements Block Avg. - Measurements Criteria to Floors & Lower 30 72
<21-72 920
<380-920
<2-7
<5-12 0
l Watisd Upper Wat is, 13 91
<14-41 440
<380-440
<2-7
<5 0 Cellings La Sump 5
31
<21-31 600
<390-600
<2
<5 -
0 East Fuel Cells 4
e c
<380
<380
<2
<5 0
West Fuel Cells 5
c c
1000
<380-1000-
<2
<5 0
108 Floors & Lower Walls After 15 150
<21-150 500
<390-500 e
c 0
RemedIaI Action aSingle point measurement unless otherwise Indicated.
Nemedial action performed; area resurveyed.
cNo measurement taken, dFive point measurements.
4
-- t w,
s e
TABLE 4 RADIONUCLIDE CONCENTRATIONS IN MISCELLANEOUS MEDIA' VTAR FACILITY VIRGINIA POLYTECl!NIC INSTITUTE AND STATE UNIVERSITY BLACKSBURG, VIRGINIA
.3
]
=_.
Radionuclide Concentrations (pCi/g)
Location Figure Media Co-60 Cs-137 Eu-152 Eu-154 U-238 Cl 13a Concrete 1.1 2 0.2
<0.1 3.7_ 1 0.5 2.6 20.3
-2.6 1 1.5 C2 13 Concrete
<0.2
<0.1 0.6 1 0.3
-<0.1
<l.5-C3 13 Concrete 0.4 1 0.2
<0.1 0.7 1 0.4'
.<0.2 3.4.2 2.1 C4 13 Concrete 0.3 1 0.1
<0.1 0.7 20.3
. <0. 2 -
1.3 1 1.2 Radionuclide Concentrations (dpm/100 cm2)
Co-60 Cs-137 Eu-152 E9-154 U-238 b
vent Residue 2200 90 2700 1600
<l40 8
Residue
<11
<ll
<31
<ll.
1500 D
13 Drain Wipe 9
1.8 13
<2 30 auncertainties are 20 based only on counting statistics.
broom 10 exhaust vent on Roof - no figure provided.
s.JF
._..,_m,
0 TABLE 5 RADIONUCLIDE CONCENTRATIONS.IN SOIL SAMPL'3S COLLECTED FROM THE REACTOR PIT AREA h
VTAR FACILITY I
VIRGINIA POLYTECHNIC INSTITUTE AND STATE UN*.VERSITY BLACKSBURG, VIRGINIA Radionuclide Concentrations pCi/g
-Locationa Co-60 Cs-137 Eu-152 Eu-154 U-238 S1 0.4 1 0.2b
<0.1
<0.2
<0.1 3.4 ! 0.7 S2 0.5 1 0.3
<0.1 0.8 t 0.2
<0.1 2.7 2.1.3 53
<0.1
<0.1
<0.2
<0.1 2.8 1 2.8 S4
<0.1
<0.1
<0.2 (0.1
<l.1 SS
<0.1
<0.1
<0.2
<0.1 4.5 t 1.1 S6
<0.1
<0.1
<0.2
<0.1 4.5 t 1.1 i
S7
<0.1
<0.1
<0.2
<0.1 1.7 1 1.4 S8
<0.2
<0.1
<0.2
<0.1 3.3 1 0.8 S9
<0.1
<0.1
<0.1
<0.1
<l.0 8 Refer to Figure 13.
bUncertainties are 20 based only on counting statistics.
i 4
35
1 1
TABLE 6 DIRECT RADIATION LEVELS AND RADIONUCLIDE' CONCENTRATIONS
-IN SURFACE SOIL SAMPLES LANDFILL VTAR FACILITY VIRGINIA -POLYTECHNIC INSTITUTE AND STATE UNIVERSITY.
.i
-BLACKSBURG, VIRGINIA 1
Exposure Rate Radionuclide Concentrations pCi/g a
(vR/h @ l m)
Co-60 Cs-137 Eu-152 Eu-154 U-238 Th-232 Location 25 m N of Rubble '
14
<0.1 0.2 2 0.lb
<0.2
<0.1 4.1 21.0' l.4 20.8
~
60 m W of Rubble 14 (0.1
<0.1
<0.2
<0.1 3.5.2 2.2 2.7 20.8 60 m S of Rubble 15
<0.1
<0.1
<0.2
<0.1
<l.2 1.6'20.6 40 m E of Rubble 15
<0.1 0.8 2 0.2
<0.2
<0.1
<l.3' l.4 '2 0.6
$ aRef er to Figure 16.
bUncertainties are 20 based only on counting statistics.
C0ther radionuclide(s) detected, but concentrations are less than 0.5 pCi/g.
e
.. - ~
g TABLE 7 DIRECT RADIATION LEVELS AND RADIONUCLIDE CONCENTRATIONS IN MISCELLANEOUS CONCRETE AND SURFACE SOIL. SAMPLES FROM LOCATION OF ELEVATED DIRECT RADIATION IN THE LANDFILL VTAR FACILITY.
VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY-BLACKSBURG, VIRGINIA.
.l Exposure Rate Radionuclide Concentrations pCi/g Location (pR/h @ l m)
Co-60 Cs-137 Eu-152 Eu-154 U-238 Th-232 a
l l
l l
Soil w
b Rubble Pile - 1 13
<0.1
<0.1
<0.1
<0.1
<l.1 1.4.20.6 Rubble Pile - 2 11 10.6 1 0.7
<0.2 30.1 2 1.6 23.3 2-0.8
<2.5
<0.8.
Rubble Pile - 3 14 3.9 2 0.4
<0.1 1.42 0.3
<0.1 7.1 2 1.4 1.5 2 0.7-Rubble Pile - 4 14
<0.1 0.2 20.1
<0.2
<0.1
<l.6 1.7 2 0.4 U
Concrete Rubble Pile - 5 78C 38.3 2 2.1
<0.5 150-25 110 23
<7.7
<2.0 Rubble Pile - 6 36' l.0 2 0.3
<0.1
<0.2
<0.2
<l.0
<l.2 Rubble Pile - 7 27 2.920.7
<0.2 13.3 2 1.7 1.0 2 0.9
<2.1
<0.9 C
aRef et Figure 16.
_s bUncertainties are 2u based only on counting statistics.-
cExpos ure rates at one meter were not determined, because the conf iguration of the rubble piles provided shiciding.
between the elevated areas and a reading location one meter away.
j>
I f
g p-REFERENCES l.
)
- 1.
. Virginia' Tech. Argonaut Reactor Facility Decommissioning Final Report,
-Dated' April 1987,' prepared..by Chem-Nuclear Systems, Inc.,;Barnwell, SC.
2.-
Amendment l.to Final Report, Dated' September 3, 1987, prepared by Department.of Health and~ Safety, VPI~& SU,.Blacksburg, VA.
T 4-
)
P 38
L APPENDIX'A MAJOR ANALYTICAL EQUIPMENT
(
l}
l:
APPENDIX.A l
MAJOR ANALYTICAL EQUIPMENT.
h-
.The' display or description of a specific product is not to be construed asian endorsement of that, product or;its manufacturer by the authors orj their U
employer.
-A.
Direct Radiation Measurements Eberline'"RASCAL" Portable Ratemeter-Scaler i
Model' PRS-1 (Eberline, Sante Fe, NM)
Eberline PRM-6 Portable Ratemeter (Eberline, Sante Fe, NM)
Ludium Alpha / Beta Floor Monitor Model'239-1 (Ludium, Sweetwater, TX)
Eberline Alpha Scintillation Probe Model AC-3-7 (Eberline, Sante Fe, NM)
Eberline Beta-Gatma "Pancake" Probe Model HP-260 (Eberline, Sante Fe, NM)
Victoreen Beta-Gamma "Pancake" Probe Model 489-110 (Victoreen, Inc., Cleveland, OH)
Victoreen Na1 Gamma Scintillation Probe t
Model 489-55 (Victoreen, Inc., Cleveland, OH)
Reuter-Stokes Pressurized Ionization Chamber Model RSS-Ill (Reuter-Stokes, Cleveland, OH)
B.,
Laboratory Analyses Low Background Alpha-Beta Counter Model LB5110-2080 (Tennelec Inc., Oak Ridge, TN) g A-1
3___________.__
ic.
y
/>
c t
Ge(Li) Detector-Model LGCC2220SD, 23% efficiency
. (Princeton Gamma-Tech, Princeton, NJ)
[
Used in' conjunction with:
Lead Shield, SPG-16 (Applied Physical Technology, Atlanta, GA)'
High Purity Germanium Detector Model GMX-23195-S, 23% efficiency (EG&G ORTEC, Oak Ridge, TN) u Used in conjunction with:
Lead Shield, G-16 (Gamma Products, Inc., Palos Hills, IL)
High Purity Germanium Coaxial Well Detector Model'GWL-110210-PWS-S, 23% efficiency
- (EG&G ORTEC, Oak Ridge, TN)
Used _ in conjunction with:
Lead Shield, G-16 (Applied Physical Technology, Atlanta, GA)
Multichannel analyzer ND-66/ND-680 System (Nuclear Data, Inc., Schaumburg, IL) 1 L
A-2 l
A
/
l i
I APPENDIX B k
MEASUREMENT AND ANALYTICAL PROCEDURES i
(
4
/;
APPENDIX B L'
Measurement and Anclytical Procedures l'
Gamma Scintillation Measurement Surf ace scans.and measurements of gamma exposure rates were performed using Eberline Model PRM-6 portable ratemeters with Victoreen Model 489-55 gamma scintillation probes containing 3.2 cm x 3.8 cm NaI(T1) scintillation probes. Count rates were ' converted to expc- -re rates ( pR/h) by cross-calibrating with a Reuter-Stokes Model RSb-lli pressurized ionization chamber at eleven representative onsite locations.
Alpha and Beta-Gamma Measurements i
Floors were scanned for elevated alpha / beta levels by passing slowly over f
the surf ace with a Ludium Model 239-1 Gas Alpha Proportional Floor Monitor with a 600 cm sensitive area. Other surfaces were scanned using Eberifne Model PRS-1 portable scaler /ratemeters coupled to alpha scintillation probes.
Measurements of total alpha radiation levels were performed using Eberline Model PRS-1 portable scaler /ratemeters with Model AC3-7 alpha scintillation probes. Measurement of direct beta-gamma radiation levels were i
performed using Eberline Model PRS-1 portable Scaler /ratemeters with Model RP-260 thin-window pancake GM probes. Count rates (cpm) were converted 2
to disintegration rates (dpm/100 cm ) by dividing the net rate by the 4x efficiency and correcting for active area of the detector. The effective 2
2 window area was 59 cm for the alpha detectors and 15 cm for the GM detectors. The average background count rate was approximately 2 cpm for the alpha probes and 44 cpm for the GM probes.
Removable Contamination Measurements Gross Alpha and Gross Beta Smears for determination of removable contamination levels were collected on numbered filter paper disks 47 mm in diameter, then placed in individually fl-1 i
I
+
l l,
labeled envelopes with the location and other pertinent information recorded.
I The smears were counted on a low background gas proportional alpha-beta counter.
Soil Sample Analysis Gamma Spectrometry Soil samples were mixed, and a portion placed in a 0.5 liter Marinelli beaker. The quantity placed in each beaker was chosen to reproduce the calibrated counting geometry and ranged f rom 600 to 800 g of soil. Net soil weights were determined, and the samples counted using intrinsic germanium ~and--
Ge(Li) detectors coupled to a Nuclear Data Model ND-680 pulse height analyzer system.
Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system. Each spectra was scanned for identifiable photopeaks which could be attributed to the VTAR operations.
f Miscellaneous Sample Analysis Concrete and residue samples were analyzed by determining the sample weight, and performing gamma spectrometry by counting the sample on an intrinsic germanium detector coupled to a Nuclear Data Model ND-680 pulse height analyzer system. The drain swipe was analyzad by gamma spectroscopy and related to area based on pipe length and diameter.
t Uncertainties and Detection Limits The uncertainties associated with the analytical data presented in the tables of this report, represent the 95% (20) confidence levels for that data. These uncertainties were calculated based on both the gross sample count levels and the associated background count levels. When the net sample count was less than the 20 statistical deviation of the background count, the sample concentration was reported as less than the minimum detectable concentration (<MDC).
Because of variation in background levels and the effects of the Compton continuum caused by other constituents in the samples, the MDC's for specific radionuclides dif fer f rom sample to sample.
B-2
Calibration an'd Quality Assurance
/
Laboratory and field survey procedures are documented. in manuals
(:
developed specifically for the Oak Ridge Associated Universities' Radiological-Site Assessment Program.
With the exception of the measurements conducted with portable -gamma scintillation survey meters, instruments were calibrated with NBS-traceable
^
standards. The calibration procedures for the portable gamma instruments are performed by comparison with an NBS calibrated pressurized ionization chamber.
- -Quarlity--control-procedureno-all-inst ruments-included-daily-background-- --
---=
and check-source measurements to confirm equipment operation within acceptable statistical' fluctuations. The ORAU laboratory participates in the EPA and EML
~ Quality Assurance Programs.
(
I B-3
}
DRAFT s
I l
p...._..
Appendix C Regulatory Guide 1.86 Terminaton of Operating Licenses for Nuclear Ractors
.O
]
n.
U.S. ATOMIC ENERGY COMMISSION I
p rREGULATORY GUIDE DIRECTORATE OF REGUL*4 TORY STANDARDS
('
c,%i ed 1
REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUOTION A licensee having a possession.only beense must tetun, mth the Part 50 licera.,authonzation for spectil Se: tion 50.51, "Duration of license, renewal," of 10 nucleu matenal (10 CFR Pan 70, "Special Nudeu CFR Pan 50, "1.ieensmg of Produ: tion and Uthi*ation Materia!"), byprodu:t material (10 CFR Pan 30,"Rules Facibties," requires that each b:ense to operate a of General Applicabihty to Licensing of Byproduct produ: tion and utilization fa:ihty be issued for a Material"), and source matenal (10 CFR Part 40, spe:ified duration. Upon expiration of the specified "Li:ensmg of Source Matenal"), until the fuel, radio-penod, the license may be either renewed or terminated
- tive componenu, and sour:es are removed from the by the Commission. Section 50.S2, "Applications for facility. Appropriate administrative controls and facility termmation of li:enses," spe:!!ies the requirements that requirements are imposed by the Part 501icense and the must be satisfied to terminate an operating license, tedni:al spe:ifi:stions to assure that proper surveillance -
in:luding the requirement that the dismantlement of the is performed and that the rea: tor facility is maintained fa:ility and 6sposal of the component parts not be in a safe condtion and not eperated.
inimi:a] to the common defense and secunty or to the heahh and safety of the publi:. This guide describes A ponenion only li:ense permits nrious options and methods and procedures considered acceptable by the procedures for decommissioning, such as mothballing.
Regulatory staff for the termination of operatmg entombment, or dsmanthng. The requirements imposed licenses for nudeu rea: tors. The Advisory Committee cepend on the opton seit:ted.
on Rea: tor Safeguards hu been consulted concerning this guide and hu con:urred in the regulatory position.
Se: tion 50.82 proddes that the licensee may dis.
mantle and dspose of the component puts of a nuclear B. DISCUSSION reactor in accordance v.ith existing reFulations. For rescuch res: tors and critical facilities, this has usually When a li:ensee decides to terminate his nuclear meant the disassembly of a rea: tor and its shipment reactor operating li:ense, he may, as a first step in the offeite, sometimes to another appropriately b:ensed proccu, request that his opentmg li:ense be amended to organi:ation for further use. The site from which a restn:t him to posseu but not operate the facility. The rea: tor has been removed must be deconta ninated, as advantage to the li:ensee of converting to such a necessary, and inspe:ted by the Corr =ussion to deter.
poneuion.only li:ense is reduced surveillance require.
mme whether unrestri:ted ::ess can be approved. In menu in that periodi: surveillance of equipment im-the case of nu: lear power rea: tors, dismantling has portant to the safety of rea: tor operation is no longer usually been :complished by shipping fuel offsite, required. Once this ponession only license is issued, making the reactor inoperable, and esposing of some of rea:ter operation is not permitted. Other :tytties the radion:tive components.
related to cessation of operations such as unloading fuel from the rea: tor and placing it in storage (either onsite Radioa:tive components may be either shipped off.
site for bunal at an authented burial ground or se:ured of effsite) may be contmued.
c c.
w m ~.- ~
usAte ntovuTony cuiots ca. 20! l.
t,
-y, c,es==.es.
%v.a.
,_.~..o -,.a,..
m.,
.v
- e..ve.i e
.. ~. c v a...
- i..,. c. -,. -.,
~~..
s,... c -,, -
.c......,,.,n,.
.. ~.i,e.
~...,.,
. ~ s.
c.
. e. m i.
c
.o
=~,J.M ';" ' ' "". ".L'""'."4 ", ';'.,*."*',', ',1". :'".",, ".,
v.-...,,~,~-
~ C" ".'" " "".","" ". ':*.~' ", ;"" '.";" LL.'*2'. "~ "
i...e-e.~,
- 2. 3,.
t.~,,...,
1 S g W h s
. S..
S. N "=
- 8..
89 4 ad
.ct l
iC (
a,
,g ege
.u
.s e.m.,
1 C-1
on the site. Those radioactive materials remaining on the.
Cuids and waste should be removed from the site.
L site must be isolated from the public by physical barriers Adequate radiation monitoring. environmental surveD.
or, other mean: to prevent public access to hazardous lance, and appropriate security procedures should be levels of radiation. Surveillan:e is necessary to usure the established under a possession only license to ensure that long term integnty of the bunen. The amount of the health and safety of the public is not endangered.
surveillance required depends upon (1) the potential hanrd to the health and safety of the public from
- b. In-Place Entombment. In place entombment con.
rad oacuve material remaining on the site and (2) the sists of scaling au the remaining highly radioactive or integnty of the physical bamers. Before areas rnay be contaminated components (e.g., the pressure vessel and released for unrestricted use, they must have been reactor intemals) within a structure integral with the decontaminated or the radioactivity must have decayed biological shield after having au fuel anembbes, radio-to less than pres:ribed hmits (Table 1).
active fluids and wastes, and cenatn selected com.
ponents shipped offute. The structure should provide The hazard ano:iated with the retired fa:ility is intepnty over the period of time in whl=h significant evaluated by considering the amount and type of quantities (greater than Table I levels) of radica:tivity remaining contamination, the degree of confinement of remam with the material in the entombment. An the remammg radioa:tive matenals, the physical secunty appropriate and continuing surveitance prograin should provided by the confinement, the susceptibihty to be established under a possession only license, fflfw of radifilon as a result of natural phenomena, and ti.e d#Ation of requited surveillance.
- c. Removal of. Radioactive Components and Dis.
mantling. All fuel assemblies, radioactive fluids and C. REGULATORY PO:ITION waste, and other materials having a:tiuties above ac.
cepted unrestricted a:tivity levels (Table 1) should be
- 1. APPUCATION FOR A LICENSE TO POSSESS BLT removed from the site. The fa:Ility owner may then have NOT OPERATE (POSSESSION.ONLY LICENSE) unrestn:ted use of the site with no requirerrent for a license. If the facility owner so desires, the rentinder of A request to amend an operating license to a the rea: tor fa:ihty may be dismant]cd and all vestiges possenion only license should be made to the Dire: tor terroved and 6sposed of.
f.
of 1.jeensing, U.S. Atomi: Energy Commission, Washing-ton, D.C. 20545. The request should indude the
- d. Conversion to a New Nudear Synem or a Fossu fotowmg informatient Fuel System. This a]temative, whi:h applies orJy to nudiar power plants, utH:zes the ex: sting turbine system
- a. A des:ription of the current status of the fa:uity.
with a new steam supply system. The original nudear steam supply system should be sepunted frem the
- b. A des:ription of measures that wiD be taken to electric generating system and disposed ofin :cordance prevent enticality or reactivity chtnges and to minimize with one of the previous three retirement alternatives.
releases of radioactivity from the fa:ility.
- 3. SURVEILIANCE AND SECURITY FOR THE RE.
- c. Any proposed changes to the technical speciSca-TIREMENT ALTERNATIVES WHOSE FINAL tions that reDect the ponession only facility status and STATUS REQUIRES A POSSESSION.ONLY I
the necessary disauembly/ retirement a:tiviues to be LICENSE performed.
~ A facility whi:h has been licensed under a posses.
- d. A safety analysis of both the a:tmties to be sion only license may contain a signincant' amount of
- comphshed and the proposed changes to tne technical radca:tivity in the form of activated and contaminated speciScations.
hardware and stru:tural materials. Surveillance and commensurate secunty should be provided to assure that
- e. An inventory of activated materials and their the pubbe health and safety ut not endangered.
location in the fa:ihty.
j
- a. Physietl se:unty to prevent inadvertent exposure
- 2. ALTERNATIVES FOR REACTOR RETIREMENT of personnel should be provided by multiple locked bamers. Tne presence of these bamers should make it Four altamatives for retirement of nucleu rea: tor extremely diffi: ult for an unauthonzed person to Fain facilities are considered a::eptable by the Regulatory acceu te aren where radiation or contamination levels j
staff.These are:
exceed those spe:ified in Regulatory Position C.4. To i
prevent inadvertent exposure, radiation areas above 5
- a. Mothballing. Mothballing of a nudear rea:ter mR/hr, such as near the activated pnmary system of a fa:ility connsis of puttmg the facility in a state of power piant, should be appropriately marked and should protective storage, in general, the facihty may be left not be a::essible except by cuttmg of welded closures or mtact except that all fuel assemblies and the radion:tive the d:sassembly and removal of substantial stru:tures 1.66 2 C-2
l and/or shielIing material. Means such as a remote.
(1) Environmental surveys, readout intrusion alann system should be provided to indicate to designated personnel when a physical barrier (2) Facility radiation surveys, io penetrated. Secunty penonnel that provide access control to the facility may be used instead of the (3) Inspections of the physical barriers, and physical barriers and the intrusion alarm systems.
(4) Abnormal e+curtences.
- b. The physical barriers to unauthorized entrance into the facihty, e.g., fences, buDdings, welded doors, and access openings, should be inspected at least
- 4. DECONTAMINATION FOR RELEASE FOR UN-quanerly to assure that these bamers have not detenor-RESTRICTED USE ated and that locks and locking appantus are intact.
If it is desired to terminate a license and to eliminate
- c. A facility radiation survey should be performed at any funher surveillance requirements, the facility should least quarterly to verify that no radioactive matenal is be sufficiently decontaminated to prevent risk to the es: aping or being transported through the containment public health and safety. After the decontamination is bamers in the facihty. Sampling should be done along satisfactorily accomplished and the site inspe:ted by the most probable path by which radioactive material the Commission, the Commission may authonze the such as that stored in the inner containment regions license to be terminated and the facility abandoned or could be transponed to the outer regions of the fa:llity released for unrestricted use. The licensee should per-and ultimately to.tht; environs.
form the decontamination usmg the 'following guide-lines:
- d. An emironmental radiation survey should be performed at leas semiannually to verify that no
- a. The licensee should rnake a reasonable effort to signficant amounts of radiation have been released to the eliminate residual centammation.
endronment from the facility. Samples such as so!!,
vegetation, and water should be taken at locations for
- b. No covering should be applied. to radioactive whi:h statistical data has been established dunns rea: tor surfaces of equipment or stru:tures by paint, plating. or operations.
other covenng material untilit is known that contamina-tion levels (detenmned by a survey and documented) are
- e. A site representative should be designated to be below the limits specified in Table L in ad6 tion, a responsible for controlling authonzed a: cess into and reasonable effen should be made (and documented) to movement within the fa:Eity, funher minimize contamination prior to any su:h cove:ing.
f.' Administrative procedures should be established for the notification and reponing of abnormal o::ur-
- c. The radion:tivity of the interior surfaces of pipes,.
rences such as (1) the entrance of an unauthorized drain 1mes, or ductwork should be determined by
. person or penons into the fa:ility and (2) a significant making measurements at all traps and other appropriate J
change in the radiation or contamination levels in the access points, provided contamination at these locations facility or the offsite emironment.
is likely to be representative of contamination on the interior of the pipes, drain lines, or ductwork. Surfa:es
- g. The following repons should be made:
of premises, equipment, or s: rap whi:h are IJkely to be contaminated but are of such size, constru: tion, or
,(1) An annual report to the Director of Licensing, location as to make the surfa:e inaccessible for purposes U.S. Atomic Energy Commision, Washington, D.C.
of measurement should be assumed to be contammated.
20545, describing the results of the environmental and in excess of the perrdssable radiationlimits.
fa:11ity radiation surveys, the status of the is:ility, and an evaluation of the performance of security and
- d. Upon request, the Commission-may authorize a surveillance measures.
li:ensee to relinquish posseuien or control of premises, equipment, or s: tap haung surfaces contardnated in (2) An abnormal occurrence report to the Regula-excess of the limits specified. This may include, but is tory Operations Regional Office by telephone within 24 not hmited to, spe:ial circumstances such as the transfer houn of 6s:overy of an abnormal occurrence. The of premises to another licensed organization that wEl abnormal occurrence will also be reponed in the annual cortinue to work with radios:tive materials. Requests report des:ribed in the preceding item, for su:h authonzation should providet
- h. Re:ords or logs relative to the foDowing items (1) Detailed, specific information des:rfbing the should be kept and retained until the license is termi-premises, equ:pment, : tap, and radioactive contami-nated, after which they may be stored with other plant nants and the nature, extent, and degree of residual records:
surface contammation.
1.86 3 C-3
t f
(2) A detailed health and safety analysis indi-or a change in the technical specifications should be cating that the residual amounts of materials on surface reviewed and approved in accordance with the require-areas, together with other considerations such as the ments of 10 CFR 650.59, prospe:tive use of the premises, equipment, or scrap, are unlikely to result in an unreasonable rssk to the health if major structural changes to radioactive components and safety of the pubhc.
of the facility are planned, such as removal of the pressure vessel or maor components of the primary
- e. Prior to release of the premises for unrestncted system, a 6smintlement plan including the information use, the li:ensee should make a comprehensive radiation required by $50.82 should be subrrutted to the Commis-survey estabbshing that contamination is within the sion. A dismantlement plan should be submitted for a!!
limits specified in Table 1. A survey report should be the alternatives of Regulatory Position C.2 except filed with the Director of Licensing. U.S. Atomic Energy mothballing. However, minor disassembly activities may Comm:ssion, Washinfton, D.C. 20545, with a copy to still be performed in the absence of such a plan, the D: rector of the Regulatory Operations Regional provided they are permitted by existmg operating and Office havingjurisd: tion.The report should be filed at maintenance pro:edures. A dismantlement plan should least 30 days prior to the planned date of abandonment.
include the foDowing:
The survey report should:
- a. A description of the ultimate status of the facility
(!) Identify the premises;
- b. A des:ription of the dismantling activities and the (2) Show that reasonable effort has been made to precautions to be taken.
redu:e residual contamination to as low as practicable
- c. A safety analysis of the dismantling activities levels; in:luding any effluenu which may be released.
(3) Describe the s: ope of the survey and the general procedures foDowed; and
- d. A safety analysis of the fa:Ility in its ultimate status.
(4) State the fm6ng of the survey in units j
spe:ified in Table 1.
Upon satisft: tory review and approval of the dis-mantling plan, a 6smantling order is issued by the i
After review of the report, the Commission may Commission in accordan:e with (50.82. Wnen dis-inspe:t the fa:ilities to confirm the survey prior to rnantling is completed and the Commissica has been grannng approval for abandenment.
notified by letter, the appropriate Regulatory Opera-tions Reponal Offi:e inspe:ts the fa:ility and verifies
- 5. REACTOR PJ.TIREMEhT PROCEDURES completion in accordance with the dismantlement plan.
If residual r:6ation levels do not ex:eed the values in As in6:sted in Regulatory Position C.2, several Table 1, the Commission may terminate the licenw. If altematives are a :eptable for reactor fa:ility retirement.
these levels are exceeded, the licensee retains the if ::unor disassembly or "mothballing" is planned, this possession enly license under which the dismantling
- culd be done by the existing operating and ma:nte-activities have been condu:ted or, as an alternatin,may nan:e procedures under the license in effect. Any make application to the Statt (if an Agreement State) planned actions involving an unreviewed safety question for a byproduct materials license.
1.86-4 C-4
TABLEI ACCEPTABLE SURFACE CONTAMINATION LEVELS NUCLIDEa AVERAGE c l
MAXIMUMbd REMOVABLE e f-b b
2 2
1,000 dpm c/l00 cm 2
15.000 dpm c/100 cm U.nat, U 235, U 238, and 5,000 dpm a/100 cm associated decay products 2
2 20 dpm/100 cm 2
300 dpm/100 cm Transuranies, Ra 226, Ra 228, 100 dpm/100 cm Th 230,Th 228,Pa 231.
Ac 227,1125,1129 2
2 200 dpm/100 cm Th nat,Th 232, Sr 90, 1000 <!pm/100 cm2 3000 dpm/100 cm Ra 223, Ra 224, U 232, 1126,1131,1133 2
2 2
Beta gamma emitters (nuclides.
5000 dpm $q/100 cm 15,000 dpm $9/100 cm 1000 dpm $q/100 cm with decay modes other than alpha emission or spontaneous fission) i except St 90 and others noted above.
- Where surface contamination by both alpha and beta pmma-emi: ting nuclides exists, the hmits esublished for alpha sr.d beu-pmma-emitting tuchdes should apply inaependently.
bas used in this table, dpm fdiuntepauens per minute) means the rate of emission by radioactive material as determined by ccrrectirig the counts per minuts observed by an apprepnate detector for backpound, efDesency, and geometric factors associated ulth the instrumenuuon-Measuremenu of average conumirunt should not be averaged over more than I square meter. For objects of less surface area 2e C
aver:Fe should be derrved for each such object, 2
dThe maxtmum contaminationlevel appbes to an aret of not more than 100 cm,
2 of surface area should et determined by siping that area sith dry fUter or
'ne amount of removable radioactrve materM per 100 cm soft abserbent pape:, a; plying moderate preuute, and anesang the amount of radsoactive materal on the sipe sith an apprognate instrument of knesn efDeiency. Vnen removable conurn:r.auon on obytets of less surface area is dettmuned, the perunent leveh should N reduced proportionaDy and the entre surface should be wipd-1.66 5 C-5
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Appendix D Proposed Confirmatory Survey Plan For the Argonaut Reactor Facility Virginia Polytechnic Institute & State University
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PROPOSED CONFIRMATORY. SURVEY PLAN
~FOR THE 5
A?COSAUT. REACTOR FACILITY -
d VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIVERSITY
' ~
BLACKSBURG, VIRGINIA Q
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Site History and Descr ption 4.
i. /'
i The Virginia Tech Argonaut Reactor (VTAR) Facility is located in Robeson i
Hall on the ' northwest corner of the main campus of Virginia Polytechnic Institute and State University (VPI), between the Appalachian and Blue L
Ridge Mountains in Blacksburg, Virginia.
The VTAR is an Argonaut type research and. training reactor, originally. designed and installed by
'American Standard Nuclear Division.
The reactor was used as a part of k
the Nuclear Engineering curriculum for basic research in neutron physics, neutron radiography, neutron activation analysis, technical training and Reactor Operator training.
y
~
The' VTAR core was heterogenous in design, using 93% enriched MTR. type uranium-aluminum matrix-fuel elements.
Thermal power output was limited to 100 kW(th), with light water used as coolant and part of the moderator.
The remainder of the moderator consisted of graphite blocks.
I which surrounded the fuel and water noderator.
The VTAR was located.in Room 10, on the ground floor of Robeson Hall, with a high bay area which extends through the first floor le.ve l.
Rooms 6.(Anteroom), 6A (Electronics Shop), 8 (Storeroom) and 8A (Sample Prep Room) are located adjacent to Room 10, and an access door connects Rooms 6 and 10.
On the first floor level, Room 108 is accessible to Room 6 via a spiral staircase.
Robeson Hall is also occupied by the Physics faculty, staff and graduate student offices, other research laboratories, shops, and classrooms.
Prepared under Interagency Agreement DOE No. 40-816-83 NRC Fin. No. A-9076-3 between the U.S.
Nuclear Regulatory Commission and the U.S.
Department of Energy.
July 16, 1987 D-1
W The u reector began -operation in June 1959, with a maximum power level of a
C T
10. kW( th); the reactor was modified, and Nuclear Regulatory Commission lit.ense number R-62 was amended to allow a maximum power of 100 kW(th) in.
1966.
The reactor was shut down on July - 14,-1983, and the license was amended for possession only in April 1985.
Reactor fuel was shipped to
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the Department of Energy in late 1985 and early 1986.
Dismantlement and decommissioning operations began in September 1985, and were completed in January 1987 by Chem-Nuclear Systems, Inc., Columbia, South ; Carolina.
The Decommissioning Final Report was issued by VPI in April 1987.
7 I
Region Il of the Nuclear Regulatory Commission, has requested that Oak s
Ridge Associated Universities perform a confirmatory survey of the site.
II.
Purpose f.
The purpose of the survey is to confirm that the radiological data presented by Virginia Polytechnic Institute & State University, relative to release of the facility for unrestricted use.
III.
Responsibility Work described in this sr.tvey plan will be performed under the
(
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supervision of Mr. J.D.
Berger, Manager and Mr. G.L. Murphy, Assistant I
1 Mar.ager with the Radiological Site Assessment Program of Oak Ridge l
Associated Universities (0RAU).
IV.
Procedures 1.
Oak Ridge Associated Universities will review VPI's close-out survey report and supporting documentation concerning site decommissioning activities.
2.
A one meter grid will be established on the floor and lower wall (up to 2 m) areas of Rooms 6, 6A, 8, 8A, 10, and 108.
The upper walls
{,-
and ceilings will not be gridded.
Measurements and samples from the ungridded surf aces will be referenced to the floor and lower wall grid, or to pertinent building features.
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3.
The floor and lower valls will' be surf ace ' scanned using d.aIU/D ga ama tseintillation detecUrs (maximum diatance f rom o siirf ace' 5 cm), gas -
proportional and ZnS alpha detectors (marimum fdistane.c 'f rom Stirf ace l' cm), 'and gas priiportional and "pancake" GM bsta-J uma detectors (maximum 91 stance frof surf ace 1 cm).- Locations of elevated readings
~
will be not ed for. further investigation.
_.y 4..
Exposure rate measurements, using ' gamma scintillation detectors will be nale at one meter f rom. the floer, lower walls, and areas _of elevated gamIV radiation identified by surface scans.
The gamma.
scint iMatton detect,s ra ' ' will be cross-calibrated onsite with a pressurleed' ion -chamber. - Exposure rate measurements will be made with a. pressurit.ed iopizatQn chatabbr 'at a minimum of six locations, as suggested by.the measurements with the igamma scintillation
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ca}ibra tion' locat b,ug and loculon8 > ide ntified by detectors.
At r,_
.2 obtained.,using a solid gamma surface scany," gamma spenya 'will ' be.
to varif y the (jplic.abilith of 'the calibration curve state detector, to measurements throughout the "facility, The background exposure rate will be es ta';11shed with the~ predsurlied ionization chamber.
The location for determination of the background exposure rate will be an yrea that 13 not radiolog$cally contaminated, outside the
~
restricted area and of similar construction material.
5.
Measurements of total and removable alpha and beta-gamma contamination will be performed on a minimum of 30 of the floor and lower. wall grid blocks, selected at random.
One set of five direct measurements will be obtained for each surveyed grid block, and one smear wi.11 be taken for each set of five measurements.
Additional measurements will be performed at locations identified by the surface C
i;cin.
6.
, Direct measurements and smnars will be obtained on the upper walls
~
' and ceilingr.
Particular attentica will be given to cracks, beams, piping., ledges, duct % and other surf aces where material might settle or accumulate. T%se surveys will include the inside surfaces of any d rains, 4 e x*aau s t air docts, and storage wells.
The number of survey locations will be ' determined by results as the survey progresses; hovever, a minimum of 33 teasurements will be performed.
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Direct'. measurements and smears will be obtained at locations of L
. elevated. contact radiation levels identified by the surface scans.
f.
8.
Samples of. residues, including ' water, will be collected from floor cracks or joints,. beams, inside and outside of piping, ledges, air du.:t s: :.Lorage wells, and other surfaces as appropriate.
9.
Samples of paint will' be removed f rom. a minimum of 10. locations on the floor or other surfaces, if direct measurements suggest possible residual contamination.
. 10.
Other sample media and locations will be added to the survey, based on findings as the survey progresses.
V.'
Data and Sample Analysis Direct measurements, expocure rate data, and gamma spectra will be evaluated onsite to determine the need for additional decontamination.-
Smears and other samples will be returned to ORAU laboratories in Oak Ridge, Tennessee, for analysis.
Findings of the independent measurements will be compared to the Regulatory Guide 1.86 release criteria as referenced in the f acility Dismantlement and Decommissioning Plan (refer to attachment).
In addition, the exposure rates within the -
facility shall not exceed 5 WR/h above ba.ckground at I meter from accessible surfaces.
VI.
Tentative Schedule Site Visit July 22, 1987 Site Survey July 27 - August 1, 1987 Draft Report September 15, 1987 Final Report October 15, 1987 D-4
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