ML20204D534
| ML20204D534 | |
| Person / Time | |
|---|---|
| Site: | 05000375 |
| Issue date: | 12/31/1986 |
| From: | Murphy G OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20204B484 | List: |
| References | |
| CON-FIN-A-9400 NUDOCS 8703250516 | |
| Download: ML20204D534 (61) | |
Text
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I e,e,a,m,
Oak Ridge Associated CONFIRMATORY RADIOLOGICAL SURVEY Universities I
Prepared for OF THE U.S. Nuclear l
Sa"m'iis[!n s L-85 REACTOR FACILITY m
Region V Office ROCKETDYNE DIVISION I
Division of Radiation Safety and Safeguards, ROCKWELL INTERNATIONAL Emergency I
ila'fioiogic'a'i CORPORATION
"'*d
"d Protection Branch l
S ANTA SUSANA, CALIFORNIA l
G.L. MURPHY I
I I
I Radiological Site Assessment Program Manpower Education, Research, and Training Division FINAL REPORT DECEMBER 1986 I
hD DOCK 5
0375 0
t '
E CONFIRMATORY RADI0 LOGICAL SUR\\CY OF THE L-85 REACTOR FACILITY I
ROCKETDYNE DIVISION ROCKWELL INTERNATIONAL CORPORATION
' SANTA SUSANA, CALIFORNIA f
Prepared by 1
G.L. MURPHY Radiological Site Assessment Program Manpower Education, Research, and Training Division Oak Ridge Associated l'niversities Oak Ridge, TN 37831-0117 I
Project Staf f 9
R.D. Condra A.$. Masvidal D.A. Cibson C.F. Weaver M.J. Lauderaan t
Prepared for i
Division of Radiation Safety'and Safeguards Emergency Preparednese and Radiological Protection Branch a
U.S. Nuclear Regulatory Commission i
Region V Office t
Final Report December 1986 l}
This report is bssed on work performed under Interagency Agreement DOE No.
40-862-86 NRC Fin.
No.
A-9400 between the U.S.. Nuclear Regulatory Commission and the U.S.
Department of Energy.
Oak ' Ridge Associated I
Universities performs complement ary
' work under contract number DE-AC05-760R00033 with the U. S. Departaint of Energy.
I N
1
~
TABLE OF CONTENTS I
Page List of Figures.....
11 List of Tables iii Introduction I
I Site Description Document Review.
2 Survey Procedures.
2 g
Res.1ts s
Ccm arlson' of Results with Guidelines.
8 8
Summary References 43 3
Appendices Appendix A: Major Analytical Equipment Appendix B: Measurement and Analytical Procedures Appendix C: Regulatory Guide 1.86, Termination of Operating i
Licenses for Nuclear Reactors t'.
n t
I i
i g
F i
LIST OF FIGURES Page lw FIGUP.E 1: General Location of Rockwell International 10 FIGURE 2:
Building T-093 11 FIGURE 3: General Floor Plan of Building T-093 12 FIGURE 4:
Surface Scans of Upper Walls and Ceiling 13 FIGURE 5: Direct Measurements of Floors and Lower Valls.
14 FIGURE 6: Transferable Contamination Surveys 15 FIGURE 7:
In-situ Ga una Spectrometry Measurements.
16 FIGURE 8: Elevated Alpha Levels in the Reactor Bay Area......
17 1
l FIGURE 9: Location of 38 Rand'om Grid Blocks on the Floor of the l
Reactor Bay Area 18 WJ FIGURE 10: Location of 6 Random Grid Blocks on the Floor of the Control Room Area 19 FIGURE 11: Location of 8 Random Grid Blocks on the Lower Walls of j
the Reactor Bay Area 20 l
FIGURE 12: Location of 4 Random Grid Blocks on the Lower Walls of the Control Roou 21 FIGURE 13: Location of indoor Exposure Rate Measurements.
22 FIGURE 14: Location of 6 Random Surface Soil Samples Around Building T-093 23 i
FIGURE 15: Location of 4 Baseline Samples in the Santa Susana Area.
24 8
.I i
8 1
11 1
LIST OF TABLES Page TABLE 1: Gross Alpha Survey Results - Reactor Bay Floor..
25 TABLE 2: Gross Beta Survey Results - Reactor Bay Floor 27 TABLE 3: Gross Alpha Survey Results - Reactor Bay Lower Walls.
29 TABLE 4: Gross Beta Survey Results - Reactor Bay Lower Walls 30 TABLE 5: Gross Alpha Survey Results - Control Room Floot 31 TABLE 6: Gross Beta Survey Results - Control Room Floor.
32 TABLE 7: Gross Alpha Survey Results - Control Room Lower Walls 33 TABLE 8: Gross Beta Survey Results - Control Room Lower Walls..
34 TABLE 9: Gross Alpha Survey Results - Reactor Bay Upper Walls / Ceilings.
35 1
TABLE 10: Gross Beta Survey Results - Reactor Bay Upper Walls / Ceilings.
36
[
TABLE 11: Gross Alpha Survey Results - Control Room Upper Walls / Ceilings 37 TABLE 12: Gross Beta Survey Results - Control Room Upper Walls / Ceilings...........
38 I
TABLE 13: Radionuclide Activities in Miscellaneous Media.
39 TABLE 14: Indoor Exposure Rate Measurements 40 TABLE 15: Exposure Rates and Radionuclide Concentrations in Surface Soil Samples 41 I
TABLE 16: Exposure Rates and Radionuclide Concentrations in Surface Soil Baseline Locations 42 ll N f
iii
[
CONFIRMATORY RADIOLOGICAL SURVEY 0F THE L
L-85 REACTOR FACILITY l
ROCKETDYNE DIVISION
(
ROCKWELL INTERNATIONAL CORPORATION L
SANTA SUSANA, CALIFORNIA INTRODUCTION I
Rockwell International Corporation operated a research reactor at the Santa Susana Field Laboratories from late 1956 until February 29, 1980.
The f acility operated under an Atomic Energy Commission (AEC) contract from 1956 through 1972, followed by Nuclear Regulatory Commission (NRC) license #R-118,
('
from 1972 to present.
The reactor was operated to provide a neutron source, for suberitical experiments, neutron radiography, and training functions.
In March
- 1980, Rockwell International applied for an NRC order authorizing dismantling of the facility, disposal of the component parts, and termination of facility license #R-118.
The fuel solution was transferred to
{
the Idaho Nuclear Engineering Laboratory in September 1982, and the NRC issued a decommissioning order on February 22, 1983.
In March 1986, Rockwell submitted a radiation survey report (RI 86) of the decommissioned facility, indicating that the facility satisfied the NRC guidelines (RG 186) for release from licensing restrictions.
At the request of the Nuclear Regulatory Commission's Region V Office, the Radiological Site Assessment Program (RSAP) of Oak Ridge Associated Universities (0RAU) conducted a confirmatory survey of the L-85 reactor facility (Building T-093).
This report presents the procedures and results of that survey.
SITE DESCRIPTION
/
The L-85 reactor was operated by the Rocketdyne Division of Rockwell International Corporation, and was located at the Santa Susana Field Laboratories, slightly northwest of Canoga Park, California (Figure 1).
The L-85 reactor was housed in Building T-093, which was constructed from concrete
(
block and sheet metal on a steel girder frame.
Building T-093 (Figure 2) consists of the reactor bay area ani a small control room, which also contains a lavatory.
The facility dimensions are 16.4 m x 9.7 m, and the reactor bay has an open ceiling approximately 10 meters high with an overhead crane (Figure 3).
i
I DOCUMENT REVIEW Prior to performing the site survey, ORAU performed a review of the licensee's documentation supporting the decommissioning proj ect.
ORAU reviewed the dismantling order (NRC 83), decommissioning plan (RI 85), the final survey (RI 86), licensee /NRC correspondence, and Rockwell internal I,
documents supporting the decommissioning activity.
Rockwell developed and implemented a
- thorough, comprehensive decommissioning plan. Many items of concern addressed by the NRC, during the dismantling, were resolved by Rockwell.
ORAU's review indicates the procedures, instrumentation, and data support the final survey report.
SURVEY PROCEDURES I
On September 30 through October 2, 1986, ORAU conducted a confirmatory radiological survey of the L-85 reactor facility.
The purpose of the survey was to verify the adequacy and accuracy of the licensee's final survey, and to confirm the radiological condition of the facility relative to the decommissioning guidelines.
A.
Indoor Areas Gridding A 1 m x 1 m grid pattern was established on the floor, using the northwest corner of the building as the baseline coordinate (0,0).
Grid I
block numbers were assigned sequentially from west to east, and each row was numbered starting from the west.
A similar floor grid was established in the control room; the lavatory was not gridded due to its s
small size.
A 1 m x 1 m grid, two meters high, was established on lower wall I
surfaces.
Grid block #1 was always established in the lower left hand corner of the wall, and the grid blocks were numbered sequentially from bottom to top.
2
lI Measurements on the upper walls and ceilings were referenced to building features, or the floor grid.
I Surface Measurements I
RSAP personnel scanned the floors and lower walls wit.it an alpha and l
beta-gamma floor monitor and Nal(Tl) gamma scintillation detectors,
.5 respectively.
Locations inaccessible to the floor monitors were scanned with hand held alpha scintillation and beta-gamma " pancake" detectors.
Upper wall and overhead surface scanning (Figure 4) on ledges, beams, piping, fixtures, equipment, and ductwork was conducted using hand held alpha and beta gamma detectors.
Elevated areas were marked for additional measurements.
I Measurements of total alpha and beta-gamma contamination levels (Figure 5) on random floor and lower wall grid blocks were performed at I
the center and four points, midway between the center and grid block corners.
Smears (Figure 6) for removable alpha and beta contamination 5;
were performed at the location in each grid block where the highest total measurement was obtained. Total and removable contamination measurements were also performed on upper walls,
- ceilings, ledges, and other horizontal and vertical surfaces.
I Exposure Rate Measurements Gamma exposure rates at 1 meter above the floor were measured at random locations and areas of elevated gamma levels, identified by the surface I
scans, using a pressurized ionization chamber.
Gamma Spectroscopy Measurements An in-situ gamma spectrum was collected at each location where gamma exposure rate measurements were made (Figure 7).
The spectra was collectedusing a high purity germanium detector positioned one-half meter W'
from the surf ace.
The gamma spectra were used to identify the residual radionuclide contaminants.
!gg I
3
I Miscellaneous Sampling I
Random samples of paint were collected in the reactor bay area and the control room.
Samples of residue on horizontal surfaces, concrete dust and chips, standing water (located in electrical chase), and drain residues were collected and returned to ORAU for analyses.
B.
Outside Areas Surface Measurements I
ORAU personnel performed walkover surface
- scans, using gamma scintillation detectors, to a distance of 10 meters from the building in all directions.
Scans were extended to cover the access roads and 1
parking lots to a distance of 80 meters from the building.
Soil Sampling l
Surface soil samples were collected in east, south, and west areas adjacent to Building T-093.
The area directly north of the building was j
covered by a small building and large rock out croppings.
I C.
Baseline Measurements 1
Indoor Areas l
Building T-453 located approximately 50 meters south of Building T-093 l
l was used to establish the baseline for gamma exposure rate measurements.
Building T-453 has a similar construction history as T-093, and is l
located in a non-restricted area which has no history of non-sealed source radioactive materials use.
Exposure rate measurements and gamma spectra were collected using the pressurized ionization chamber and the gamma spectroscopy system.
Outdoor Areas Gamma exposure rate measurements and surf ace soil samples were collected at four off-site locations surrounding the Santa Susana Field Laboratory.
I t
I lI Sample Analysis and Interpretation of Data Smears were couated to determine gross alpha and beta activity.
Soil and miscellaneous residue samples were analyzed by gamma spectroscopy for cobalt-60, cesium-137, europium-152/154, uranium-238/235, and other identifiable photopeaks.
Major analytical equipment used in support of this i
1 survey is listed in Appendix A, and Appendix B describes the measurement and analytical procedures.
Results were compared with guidelines established by the Nuclear Regulatory Commission for release of facilities for unrestricted use (Appendix C).
An additional guideline, was established to limit the exposure rate to SpR/h above ambient background at one meter from the surface, or the licensee may present a worst-case analysis which establishes the dose-rate to a theoretical individual occupying the area, to less than 10 mrem / year.
RESULTS Indoor Areas Baseline Exposure Rate j
The baseline exposure rate of 12 uR/h, was established in Building T-453.
Surface Scans Alpha and beta-gamma scans identified two areas of elevated alpha I
activity in grid blocks 16 and 81 on the floor of the reactor bay (Figure 8).
These areas did not exceed the release guidelines, but Rockwell elected to perform further remedial action to ensure that residual contamination was as low as reasonably achievable.
E Surface Contamination Levels i
Tables 1 through 8 summarizes the results of surface contamination measurements performed on 56 random floor and lower wall grid blocks of I
Building T-093 (Figures 9 - 12 ).
The total activity presented in the tables I
5
I is a direct measurement which contains non-removable activity, as well as, removable activity, if present.
The total alpha activity ranged from the 2
2 minimum detectable activity of 11 dpm/100 cm to 460 dpm/100 cm, in grid 2
block 81.
The highest average grid block results (160 dpm/100 cm ) occurred 2
in block 16.
The removable alpha activity ranged from an MDA of 3 dpm/100 cm 2
to 14 dpm/100' cm.
The total beta activity ranged from a MDA of 400 dpm/100 2
2 cm to 4900 dpm/100 cm, in grid block 39.
The highest average grid block g
2 u
results (1800 dpm/100 cm ) occured in grid blocks 39 and 47.
The removable 2
2 beta activity ranged from an MDA of 7 dpm/100 cm to 12 dpm/100 cm,
Tables 9 through 12 summarize the results of 30 single point surface contamination measurements performed on upper walls and ceilings of Building 2
T-093.
The total alpha activity ranged from an MDA of 18 dpm/100 cu g,
2 160 dpm/100 cm,
and the removable activity ranged from an MDA of 2
2 3 dpm/100 cm to 14 dpm/100 cm.
The total beta activity ranged from an MDA 2
2 of 400 dpm/100 cm to 1400 dpm/100 cm, and the removable beta activity' ranged I
2 2
from an MDA of 7 dpm/100 cm to 20 dpm/100 cm,
Radionuclide' Activity in Miscellaneous Media Table 13 summarizes the radionuclide concentrations in 12 samples of miscellaneous media.
Dust residues were collected at seven locations in the reactor bay area.
Radionuclides in trace amounts were identified in several of these samples.
One sample collected along the east wall ledge indicated the presence of Co-60, Cs-137, Eu-152, Eu-154, and U-238.
Two paint samples were collected from the reactor bay and one from the control room, one reactor bay sample contained Cs-137 and U-238 in trace amounts. A smear from the sink I
trap of the control room lavatory had no detectable radionuclides.
The minimum detectable activities reported are quite high due to the small sample size and relatively small area smeared.
Concrete chips were collected from the scabbled area of the Reactor Bay.
This sample contained 11 pCi/g of Eu-152, the highest europium content of any miscellaneous sample.
Radiostrontium analysis on this sample indicated the Sr-89 was less than 0.3 pCi/g and Sr-90 was less than 0.4 pCi/g.
A very small sample of standing water (less than 2 ml) was collected from the electrical chase in the reactor I
I 6
E bay floor. No identifiable photopeaks were observed. The sample was analyzed for gross alpha and beta activity. The results were 1.4 0.8 pCi/l alpha and 5.6 i 1.2 pCi/l beta.
Exposure Rate Measurements f~L Table 14 summarizes the exposure rate measurements, taken at six random y
locations and five additional locations identified by surface scans I
(Figure 13). The exposure rate levels in Building T-093 ranged from 12 to 18 pR/h, compared to a baseline level of 12 pR/h measured in Building T-453.
L The gamma spectra collected in Building T-453 identified only the presence of natural radionuclides normally present in building materials. A review of the gamma spectra collected at each exposure rate measurement location indicated the presence of Co-60 and Eu-152 in the elevated exposure rate areas.
[
Although Co-60 and Eu-152 photopeaks are present in the random locations, L
their relative peak areas are considerably smaller and are attributed to the close proximity of the hot spots from the scabbled floor.
Outside Areas Surface Measurements No locations of elevated direct radiation levels were identified by gamma scans of the areas within 10 meters of the buildings, parking lots and access roads, or drainage ditches. The exposure rates measured at one meter from the ground at the soil sampling locations, ranged from 16 to 18 pR/h, compared to 10 to 13 pR/h at offsite locations.
Although the onsite locations are slightly elevated, this is probably related to variations in natural radionuclide concentrations in dif ferent types of soil.
Soil Samples Six random soil samples were collected around Building T-093, and the results are summarized in Table 15 (Figure 14).
The radionuclide concentrations are typical of concentrations present in four baseline samples I
l I
I collected offsite (Figure 15).
Only photopeaks associated with naturally occurring radionuclides were detected in the surface soil samples.
l COMPARISON OF RESULTS WITH GUIDELINES The survey findings indicate the total residual contamination is less 2
2 l
than the NRC guidelines of 5000 dpm/100 cm average, 15000 dpm/100 cm 2
maximum, and 1000 dpm/100 cm removable, for alpha residual contamination.
l The survey findings also indicate that the total beta residual contamination I
I is less that the NRC guidelines for Sr-90 (most restrictive category),
2 2
2 1000 dpm/100 cm average, 3000 dpm/100 cm maximum, and 200 dpm/100 cm removable except for grid blocks 5, 13, 32, 39, 41, 47, 59, and 67.
- However, I
strontium 89/90 results of randomly selected concrete samples, analyzed by ORAU and the licensee, indicate the strontium 89/90 contamination to be less than the minimum detectable activity of 16 dpm (0RAU data).
No radionuclide contamination of surface soils was detected.
The water sample results are less than the EPA drinking water guidelines of 15 pC1/1 gross alpha and 50 l
pCi/l gross beta.
The only guideline which was not met, was the exposure rate criteria.
The baseline exposure rate was established as 12 UR/h, and the highest exposure rate in the reactor bay at one meter above the intersection of grid blocks 41, 42, 50, 51 was 18 pR/h.
This result is probably slightly elevated due to the " bowl" geometry of the scabbled floor.
The licensee proposed a l
scenario (RAI 86) of an individual occupying a desk for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> /per year adjacent to the scabbled area prior to backfilling.
At the highest exposure l
rate location, this individual would receive a dose of 4.4 mrem / year.
Our data supports their scenario, although we differ slightly in predicted dose, 4 mrem per year, which would be within the 10 mrem / year guidelines.
If this area were backfilled with concrete to render the facility usable, the exposure I
rate would likely be reduced to within the guideline level of 5 $/h above background.
SUMMARY
At the request of the Nuclear Regulatory Commission, Region V Office, ORAU conduced a confirmatory radiological survey of the Rockwell International L-85 Reactor facility, located in Santa Susana, California.
The survey was
E 1
performed on September 30 through October 2, 1986.
The purpose was to verify e
the radiological condition of the facility relative to NRC guidelines for release for unrestricted use.
Radiological information collected included gamma exposure rates, surface contamination levels, gamma spectra, and radionuclide concentrations in soil and miscellaneous media.
Based on the final results of the ORAU survey, and a review of the decommissioning documentation, it is ORAU's opinion that the L-85 Reactor facility (Building T-093) has been remediated to the existing NRC guidelines, with the exception of the exposure rate criteria.
ORAU suggests that the L
At^a^ concept has been met, and that restoration of the remediated area will reduce the exposure rate to the levels established by the Dismantling Order.
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A.am e
FIGURE 5: Direct Measurements of Floors and Lower Walls. Total alpha and beta-gamma contamination I
measurements were made using portable scalers and detectors. The grid system (X's on the floor) is visible in the remediated area of the concrete.
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Y' FIGURE 6: Transferable Contamination Surveys.
Smears were taken at the location of the highest radiation measurement in each grid block.
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FIGURE 7:
In-situ Gamma Spectrometry Measurements.
Gamma spectra were collected at random locations and locations of elevated radiation levels, using a high purity germanium detector and 4096 channel pulse height analyzer.
o o
r m
r t
r a
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. ts...
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SOUm WAU.
l NORE WAM.
u 14 5
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+
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METER n.
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WEST WALL FIGURE 11: Location of 8 Random Grid Blocks on the Lower Walls of the Reactor Bay Area
EAST WAU.
.. -0 g
o s
NORM WAU.
SOUN WAU.
+
f, v
. r. -
2 0
1 ucna l
l
.s WEST WALL FIGURE 12: Location of 4 Rondom Grid Blocks on the Lower Wolls of the Control Room
m m
M M
q g
g g
g g
M M
q q
36 e
34 61 e
42 51 41 50 e
e 67 31 58 e
w 93
)
[/
V 11 g
l e
28 e
, ACTUAL MEASUREMENT
+
LOCATION O
1 1
I METER FIGURE 13: Location of Indoor Exposure Rote Measurements
M M
M 6
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4 N
5 G"
M
+
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6 3
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l eT ca g f n r i i
ud Si l
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WOOLSEY 1
2
.....~. RESERVOIR '
.O ROCKYELL CANYoy
..~... ~
' ~.'.~.'.'.'.'.'.~.~.
INTERNATIONAL
... ~
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o CAN y
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1 FIGURE 15: Location of 4 Baseline Sampics in the Santo Susana Area l
z -
m_,
- i...
TABLE 1 GR03S ALPHA SURVEY RESULTS - REACTOR BAY FLOOR ROCKWELL INTERNATIONAL
. SANTA SUSANA, CALIFORNIA w
CridBk'ocK8 Total Activity (dpa/100 cm )
Removable Actigity 2
Range Average (dpm/100 cm )
4 51 -
82 59
<3 5
<20 - 190 120
<3 10 11 -
82 61
<3 11
<20 - 100 61
<3 i
13
<20 - 120 59
<3 16
<14 - 1400b 330b
<3
!7 19 - 100 45
<3 19 19 -
37-24
<3 23
<14 -
56 33
<3
(
24 (14 -
37 17
<3 s
L
/
32
< 14 -
37 19
<3 36
<14'-
37 22
<3 37
<14 --
56 21
<3 (f
41 - 210 80
<3 39 ;
<14 -
37 26
<3 40 41
<14 -
28
<14
<3
{-
4.4
<14 -
37 28
<3 45-10 -
46 30
<3 47 37 -
74 57
<3
(
55
<14 -
46 30
<3 L
59
<14 -
37 26
<3 02 (11 -
51 39
<3 63 (11 -
61 35
<3
(
67 20 -
41 35
<3
'~
72 41 -
61 45
<3 73 <
28 -
93 45
<3
[
75 19 -
46 33
<3 79 41 -
61 51
<3 81
<20 - 5300b 930b 14 1 10e
<14 -
56 28
<3 83
[~i-35
<14 -
56 35
<3
' 87
<14 -
46 22
<3
,t 28 - 100 48
<3 i
91
<!4 -
65 30
<3
{
92 93 19 -
56 34
<3 95 37 -
93 46
<3 25 f;,
3.
'_n.
I TABLE 1 (Continued)
CROSS ALPHA SURVEY RESULTS - REACTOR BAY FLOOR ROCKWELL, INTERNATIONAL SANTA SUSANA, CALIFORNIA 2
Crid Block Total Activity (dpm/100 cm )
Removable Actigity Range Average (dpm/100 cm )
99
<14 - 110 39
<3 100 51 - 270 120
<3 16b
<14 - 390 160
<3 81b
<20 - 460 120
<3 Release Criterta 15000 5000 1000 I
8 Refer to Figure 9.
bArea recleaned and surveyed. Additional data provided at bottom of I
table.
CError is 20 based only on counting statistics.
I I
I
.g-26
l
{
TABLE 2 GROSS BETA SURVEY RESULTS - REACTOR BAY FLOOR ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA
(
2 Grid Blocita Total Activity (dpa/100 cm )
Removable Activity 2
Range Average (dpm/100 cm )
[
4 680 - 1200 950 (7
l 5
650 - 2500 1450C
<7
['
10 440 - 1000 820
<7 L
11 580 - 920 770
<7 13 850 - 1200 1000c
<7 16 750 - 1800 790
<7
(~
17
<440,1400 840 (7
19 540 - 1100 960
<7 23
<490 - 1200 640
<7
[
24
<490 - 1200 670 (7
32 820 - 2700 1600c
<7 36
<490 - 880 790
<7 r
37
<490 - 780 590
<7 L
39 510 - 4900c 1800c
<7 40
<490 - 1600 840
<7 41 1200 - 1700 1500C
<7
(
44 540 - 990 760
<7 45
<490 - 1200 730
<7 47 990 - 3600C 1800c
<7
[-
55 540 - 1000 730
<7 59 850 - 3300c 1600c
<7 62 680 - 1800 990
<7 63
<440 - 1100 780
<7
(
67 750 - 1300 1000C
<7 72 (440 - 1300 800
<7 73
<490 - 1300 710
<7
(
75
<490 - 1100 650 12 1 8b 79 680 - 1300 980
<7 81
<440 - 1300 770
<7
{.__
83
'<490 - 680
<490
<7 85
<490 - 990 550 8i6 87
<490 - 1100 570 10 1 7 91
<490 - 1500 730
<7
[
92
<490 - 950 650
<7
(
[
27
TABLE 2 (Continued)
GROSS ALPHA SURVEY RESULTS - REACTOR BAY FLOOR I
ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA I
2 Grid Block Total Activity (dpm/100 cm )
Removable Actigity Range Average (dpm/100 cm )
93
<490 - 1100
<490
<7 I
95
<490 - 510
<490
<7 99 510 - 1100 710
<7 100
<440 - 1200 890
<7 Release Criteria 3000 1000 200 aRefer to Figure 9.
b rror is 2a based only on counting statistics.
E cExceeds criteria for Sr-90.
Sr-90 analysis of concrete samples indicates minimal presence of Sr-90.
I I
I
'l i
I I
I 2e
I TABLE 3 GROSS ALPHA SURVEY RESULTS - LEACTOR BAY LOWER WALLS I
ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA 2
Crid Blocka Wall Total Activity (dpm/100 cm )
Removable Activity 2
Locations Range Average (dpm/100 cm )
5 North 31 - 100 57
<3 I
17 North
<20 - 100 65
<3 3
East 51 - 200 98
<3 19 East 51 - 170 110
<3 14 South 82 - 170 130
<3 16 South 61 - 210 130
<3 6
West
<20 - 200 94
<3 13 West
<20 - 37 22
<3 Release Criteria 15000 5000 1000 I
aRefer to Figure 11.
8 E
I I 1
I
=
E I
29
I TABLE 4 GROSS BETA SURVEY RESULTS - REACTOR BAY LOWER WALLS ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA E
2 Grid Blocka Wall Total Activity (dpm/100 cm ) Removable Activity 2
Location Range Average (dpm/100 cm )
5 North
<440 - 540
<440
<7 17 North
<440 - 540
<440
<7 3
East
<440 - 480
<440
<7 19 East (440
<440 9
7b 14 South
<440 - 610
<440 10 7
16 South
<440 - 610
<440
<7 6
West
<440
<440 10 1 7 13 West
<440
<440
<7 Release Criteria 3000 1000 200 I
aRefer to Figure 11.
b rror is 2a based only on counting statistics.
E t
i t
8 I
i se i
TABLE 5 GROSS ALPHA SURVEY RESULTS - CONTROL ROOM FLOOR ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA 2
Grid Blocka Total Activity (dpm/100 cm )
RemovableAct{vity Range Average (dpm/100 cm )
2
<18 - 73 25
<3 I
7
<18 - 82 58
<3 L
8
<18 - 46
<18
<3 21
<18 - 64 35
<3
~
22
<18 - 82 36
<3 2Lb
<18 - 270 70
<3 Release Criteria 15000 5000 1000 B
[
aRefer to Figure 10.
bL refers to area in lavatory.
1 v
I Fs I
l 31 l
y TABLE 6 GROSS BETA SURVEY RESULTS - CONTROL ROOM FLOOR ROCKWELL INTERNATIONAL f
SANTA SUSANA, CALIFORNIA L
1 2
Grid Blocka Total Activity (dpm/100 cm )
Removable Activity 2
Range Average (dpm/100 cm )
2
<400 - 530
<400
<7 7
<400
<400
<7 L
8
<400
<400
<7 21
<400 - 560
<400
<7 22
<400 - 680
<400
<7
{
L 2tb
<400 - 650
<400
<7 I
p Release Criteria 3000 1000 200 aRefer to Figure 10.
{
bL refers to area in lavatory.
I 1
F l
J L
[
L rL TABLE 7 GROSS ALPHA SURVEY RESULTS - CONTROL ROOM LOWER WALLS ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA 2
Grid Blocka Wall Total Activity (dpm/100 cm ) Removable Activity 2
Location Range Average (dpm/100 cm )
4 North
<18 - 42 25
<3 6
East
<18 - 55 37
<3 3
South
<18
<18
<3 5
West
<18 - 27
<18
<3 Release Criteria 15000 5000 1000 aRefer to Figure 12.
E W
1 e
33
fI l
TABLE 8 GROSS BETA SURVEY RESULTS - CONTROL ROOM LOWER WALLS ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA l
2 Grid Blocka Wall Total Activity (dpm/100 cm ) Removable Activity 2
Location Range Average (dpm/100 cm )
F 4
North (400
<400
<7 6
East
<400
<400
<7 3
South
<400 - 650
<400
<7
[
5 West
<400 - 470
<400
<7 Release Criteria 3000 1000 200 g
.R.i.r t.....r.
1,.
ul u
T%J L
F a
I l
I 34
[
TABLE 9 F
}
GROSS ALPHA SURVEY RESULTS - REACTOR BAY UPPER WALLS / CEILINGS ROCKWELL INTERNATIONAL r
SANTA SUSANA, CALIFORNIA b
a Location TotalActivigy Removable Activity 2
(dpm/100 c:a )
(dpm/100 cm )
E West Wall-4m high @ 9m N 37
<3 West Wall-6m high @ 6m N 37
<3 Ceiling (Grid Block 38 Floor) 27
<3 g
L West Wall-7m high @ lim N
<18
<3 Traveling Crane Support NW Corner
<18
<3 North Wall-Sm high @ 2m E 37
<3
(
North Wall-6m high @ 6m E 55
<3 North Wall-7m high @ Sm E 27
<3 East Wall-8m high @ 9.5m N 64
<3
(
East Wall-4m high @ 10m N 46
<3 L
East Wall-4m high @ 8m N 37
<3 Ceiling Beam-7m high @ 6.5m N x 4.5m E 160 10 1 8b Ceiling (Grid Block 34 Floor)
<18
<3 East Wall-Sm high @ Sm N 64
<3 Ceiling Beam-8m high @ 6m N x 4m E 27
<3 Ceiling (Grid Block 81 Floor) 110
<3
{
East Wall-4m high @ 3m N 37
<3 South Wall-3m high @ 9m E
<18
<3 South Wall-5m high @ 8m E 110
<3
(
Ceiling (Grid Block 95 Floor) 120
<3 L
South Wall-Sm high @ 3m E 120
<3 Ceiling (Grid Block 91 Floor)
<18
<3 f
West Wall-6m high @ 3m N 82
<3 West Wall-7m high @ 6m N 130
<3 West Wall-5m high @ Sm N 120 14 1 10
{
Release Criteria 15000 1000 r
aSingle point measurements performed at random locations.
b rror is 2a based only on counting statistics.
I E
L r
35
I l
TABLE 10 l
GROSS BETA SURVEY RESULTS - REACTOR BAY UPPER WALLS / CEILINGS 8
ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA I
l Location Total Activip)a Removable Activity 2
(dpm/100 cm (dpm/100 cm )
l West Wall-4m high @ 9m N 450 14 i 7b West Wall-6m high @ 6m N
<440
<7 g
Ceiling (Grid Block 38 Floor)
(440
<7 i
l West Wall-7m high @ lim N
<440
<7
.l Traveling Crane Support NW Corner
<440 (7
W North Wall-Sm high @ 2m E
<440
<7 li North Wall-6m high @ 6m E 1400
<7 North Wall-7m high @ Sm E
<440
<7 East Wall-8m high @ 9.5m N
<440
<7 l
East Wall-4m high @ lOm N 820 20 9
East Wall-4m high @ 8m N 1100
<7 L
Ceiling Beam-7m high @ 6.5m N x 4.5m E 1400 13 7
Ceiling (Grid Block 34 Floor)
(440
<7 l
East Wall-5m high @ Sm N
<440
<7 I
Ceiling Beam-8m high @ 6m N x 4m E
<440
<7 Ceiling (Grid Block 81 Floor) 450
<7 l
East Wall-4m high @ 3m N 990
<7
'g South Wall-3m high @ 9m E
<440 9
6 E
South Wall-Sm high @ 8m E 1300 12 7
L Ceiling (Crid Block 95 Floor)
<440
<7 I
South Wall-Sm high @ 3m E 1200
<7
.I Ceiling (Grid Block 91 Floor) 650
<7 West Wall-6m high @ 3m N 650 13 7
l West Wall-7m high @ 6m N 710
<7 West Wall-Sm high @ Sm N
<440 12 7
Release Criteria 3000 200 t
aSingle point measuremento performed at random locations.
l l
bError is 2o based only on counting statistics.
I I
1 36
i I TABLE 11 GROSS ALPHA SURVEY RESULTS - CONTROL ROOM UPPER WALLS / CEILING 1
ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA Total Activip)a Removable Activity Location 2
(dpm/100 cm (dpm/100 cm )
Ceiling (Grid Block 20 Floor) 27
<3 South Wall-3m high @ 1.5m E
<18
<3 Top of Ventilation Duct 0 5m E of West Wall 27
<3 I
Ceiling (Grid Block 2 Floor)
<18
<3 Ceiling (Grid Block 3Lb Floor
<18
<3 Release Criteria 15000 1000 aSingle point measurements performed at random locations.
bL refers to area in lavatory.
I I
I I
I I
I
'l n
I 37
(
c' k
TABLE 12 GROSS BETA SURVEY RESULTS - CONTROL ROOM UPPER WALLS / CEILING ROCKWELL INTERNATIONAL i
SANTA SUSANA, CALIFORNIA Total Activip)a
[
Location Removable Activity 2
(dpm/100 cm (dpm/100 cm )
I Ceiling (Grid Block 20 Floor)
<400
<7 L
South Wall-3m high @ 1.5m E
<400
<7 Top of Ventilation Duct @ Sm E of West Wall
<400
<7 Ceiling (Grid Block 2 Floor)
<400
<7 b Floor)
<400
<7 Ceiling (Grid Block 3L Release Criteria 3000 200
/
L aSingle point measurements performed at random locations.
bL refers to area in lavatory.
L e
t
(
38
1 M
I~'L._ _f W fDT R
NWW WW P
TABLE 13 RADIONUCLIDE ACTIVITIES IN MISCELLANEOUS MEDIA ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA 2
Radionuclide Activity (dpm/100 cm )a Sample No.
Location Co-60 Cs-137 Eu-152 Eu-154 U-238 U-235 1-Residue Reactor Bay North Wall Beam
<1.5 17 1 6b
<3.6
<l.8
<0.9
<7.1 2-Residue Reactor Bay East Wall Beam
<0.7
<l.0
<1.7
<0.8 15 17
<2.0 3-Residue Reactor Bay East Wall Beam
<0.7 2.1 i 1.5
<1.5
<0.7
<3.1 4-Residue Reactor Bay East Wall g
Ledge 1.0 1 0.7 6.3 i 1.6 1.8 i 1.6 0.9 1 0.8 22 i 8
<l.7 5-Residue Reactor Bay South Wall Ledge
<2.9 2.6 1 0.9
<0.9
<0.5 1.7 10.7
<l.9 6-Residue Reactor Bay West Wall Ledge
<0.8
<2.1
<l.6
<0.9 23 i 10
<5.0 7-Residue Reactor Bay Floor
<0.2 0.7 1 0.2 0.3 1 0.2 0.2 0.1 1.6 1 0.7
<0.3 8-Paint Reactor Bay North Wall
<12
<7.9
<13
<7.0
<160
<33 9-Paint Reactor Bay South Wall
<10 35 i 18
<15
<7.3 303 i 173
<44 10-Paint Control Room North Wall
<6.6
<6.6
<ll
<6.6
<56
<35 11-Smear Lavatory Sink Trap
<ll
<ll
<l7
<8.2
<70
<41 12-Conrete Chips Reactor Bay Floor 0.9 i 0.7
<0.4 11 1 2
<0.9
<4.1
<l.4 e
2 2
aSample was collected from an area exceeding 100 cm, but the data has been normalized to 100 cm,
bError is 20 based only on counting statistics.
cConcrete chips and dust were collected from scabbled area of the Reactor Bay Floor.
Results are reported in pCi/g.
TABLE 14 INDOOR EXPOSURE RATE MEASUREMENTS 8
ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA I
l Location Grid Blocka Exposure Rate (pR/h) at 1 m Above Surface
,I
=_
I l
Reactor Bay 11 13 Reactor Bay 28 13 g'
Reactor Bay 36 14 Reactor Bay 56 13 I
Reactor Bay 67 14 Reactor Bay 93 12 l
Reactor Bay 34 14 7
'g Reactor Bay 61 17 g
Reactor Bay 58 15 Reactor Bay 31 14 Reactor Bay Corner 41/42 and 50/51 18
- I Bldg. T-453 Center of Bldg.
12 (Background Location) aRefer to Figure 13.
.I lI I
1I I
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40
h v_
o n_
o fL F1 n_
o.
o n-F L_J L M F
TABLE 15 EXPOSURE RATES AND RADIONUCLIDE CONCENTRATIONS IN SURFACE SOIL SAMPLES ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA Exposure Rate Radionuclide Concentrations (pCi/g)
Locationa At I m From Surface Co-60 Cs-137 Eu-152 Eu-154 U-238 U-235 (pR/h) 47.5m West of Reactor Bldg.
18
<0.1 0.4 1 0.lb
<0.1
<0.1 1.4 1 0.9
<0.2 l
42.5m West of Reactor Bldg.
16
<0.1
<0.1
<0.1
<0.1 2.5 1 0.7
<0.2 i
Sm East of Reactor Bldg.
16
<0.1
<0.1
<0.1
<0.1 1.0 1 0.5
<0.2 i
10m East of Reactor Bldg.
16
<0.1 0.4 1 0.1
<0.1
<0.1 1.3 1 0.8
<0.2 6m South of Reactor Bldg.
16
<0.1 0.1 1 0.1
<0.1
<0.1 2.1 10.6
<0.2 10m South of Reactor Bldg.
16
<0.1
<0.1
<0.1
<0.1 2.1 10.6
<0.2 aRefer to Figure 14.
b rror is 2a based only on counting statistics.
E I
w
_m rn cy3
_r w rmm-TABLE 16 EXPOSURE RATES AND RADIONUCLIDE CONCENTRATIONS IN SURFACE SOIL BASELINE LOCATIONS ROCKWELL INTERNATIONAL SANTA SUSANA, CALIFORNIA Exposure Rate Radionuclide Concentrations (pCi/g)
Locationa At I m From Surface Co-60 Cs-137 Eu-152 Eu-154 U-238 U-235 Woolsey Canyon Rd.
1.6 km from gate 13
<0.1
<0.1
<0.1
<0.1
<0.7
<0.2 Woolsey Canyon Rd. @
Valley Circle 13
<0.1
<0.1
<0.1
<0.1 0.7 1 0.4b
<0.2 Valley Circle @
Schumann Drive 13
<0.1
<0.1
<0.1
<0.1
<l.5
<0.3 Valley Circle @
<0.1
<0.1
<0.1
<0.1
<l.5
<0.3 Roscoe Dr.
10 Bldg. T-453 (Indoor) 12 c
aRefer to Figure 15.
bError is 2a based only on counting statistics.
cExposure rate data only.
rn
I REFERENCES NRC 83 Rockwell International, Docket No.
50-375, Order Authorizing Dismantling of Facility and Disposition of Component Parts, USNRC, Division of Licensing, February 22, 1983.
RAI 86 Letter from M.E.
Remley, Rockwell A.I.
to H.
- Denton, USNRC, dated I
03/06/86.
Attachment:
Revised Radiation Survey Report for L-85 Nuclear Examination Reactor (7).
RG 186 Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors", USNRC.
RI 85 Procedure for Dismantling and Decontaminating the L-85 Reactor Facility, Rocketdyne Division, Rockwell International Corporation, July 9, 1985.
5 RI 86 Radiation Survey Report of the L-85 Reactor Facility Following I
Dismantlement and Decontaminaton of the Facility, Docket No. 50-375, Rocketdyne
- Division, Rockwell International Corporation, March 6, 1986.
I I
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43
E t
[
d E
rL APPENDIX A l
MAJOR ANALYTICAL EQUIPMENT
[
[
E
lI I
APPENDIX A l
l f
Major Analytical Equipment The display or description of a specific product is not to be construed as an endorsement of that product or its manufacturer by the authors or their employer.
.g
- W A.
Direct Radiation Measurements Eberline " RASCAL" Portable Ratemeter-Scaler I
Model PRS-1 (Eberline, Sante Fe, NM)
I j
Eberline PRM-6 i
Portable Ratemeter (Eberline, Sante Fe, NM)
Ludlum Alpha Beta Floor Monitor Model 239-1 l
(Ludlum, Sweetwater, TX)
Ludlum Model 2220 Portable Scaler-Ratemeter (Ludlum, Sweetwater, TX) j Eberline Alpha Scintillation Probe W
Models AC-3-7 (Eberline, Sante Fe, NM)
Eberline Beta-Gamma " Pancake" Probe Model HP-260 (Eberline, Sante Fe, NM) fl l
Victoreen Beta-Gamma " Pancake" Probe l
Model 489-110 (Victoreen, Inc., Cleveland, OH)
I Reuter-Stokes Pressurized Ionization Chamber Model RSS-111 I
(Reuter-Stokes, Cleveland, OH)
Victoreen NaI Gamma Scintilltion Probe-I Model 489-55 (Victoreen, Inc., Cleveland, OH)
I High Purity Germanium Detector Model GEM-13180-S, 13% efficiency (EG&G ORTEC, Oak Ridge, TN)
I I
A-1
I I
i Multichannel Analyzer l
Canberra Series 30 (Canberra Instruments, Meridian, CT)
B.
Laboratory Analyses l
Low Background Alpha-Beta Counter W
Model LB5110-2080
(
(Tennelec, Inc., Oak Ridge, TN)
Source Serial #
Calibration Efficiency I'a Am-241 1777-3-84
.276 5
ca-'37 2069-2-15
.411 Ge(Li) Detectors Model LGCC2220SD, 23% efficiency (Princeton Gamma-Tech, Princeton, NJ) f I
Used in conjunction with:
Lead Shield, SPG-16 i
f (Applied Physical Technology, Smyrna, GA)
High-Purity Germanium Detector j
Model GMX-23195-S, 23% efficiency (EG&G ORTEC, Oak Ridge, TN)
Used in conjunction with:
f Lead Shield, G-16 (Gamma Products Inc., Palos Hills, IL)
Multichannel Analyzer ND-66/ND 680 System (Nuclear Data, Inc., Schaumburg, IL)
High Purity Germanium Coaxial Well Detecter Model GWL 110210-PWS-S, 23% efficiency (EG&G ORTEC, Oak Ridge, TN)
C.
Site Specific Equipment List Scaler /Ratemeter Detector Calibration Date PRM-6
- 5 NaI
- 11 Daily Onsite PRM-6
- 6 Na1
- 8 Daily Onsite PRM-6
- 8 NaI
- 14 Daily Onsite
'I PRM-6
- 9 NaI
- 13 Daily Onsite RASCAL
- 11 AC-3
- 3 09/24/86 RASCAL
- 13 AC-3
- 12 09/23/86 RASCAL
- 14 AC-3
- 14 09/23/86 I
- 13 HP-260
- 13 09/24/86 RASCAL
- 14 HP-260
- 7 09/24/86 RASCAL
- 14 HP-260
- 11 09/24/86 I
A-2
( -- - - - - -
l r
b C
FL J
A I
L F
APPENDIX B MEASUREMENT AND ANALYTICAL PROCEDURES l
B
,b I
L m
I 1
L APPENDIX B IL Analytical Procedures H
L Alpha and Beta-gamma Measurements Measurements of total alpha radiation levels were performed using Eberline Model PRS-1 portable scaler /ratemeters with Model AC-3-7 alpha scintillation probes.
Measurements of total beta gamma radiation levels were performed using Eberline Model PRS-1 portable scaler /ratemeters with Model
{
HP-260 thin-window " pancake" G-M probes.
Count rates (cpm) were converted to 2
disintegration rates (dpm/100 cm ) by dividing the net rate by the 4x efficiency and correcting for the active area of the detector. Although other factors (i.e. backscatter) can affect the calibration, they are considered insignificant for the measurements performed.
Effective windme areas were 59
~
2 2
cm for the ZnS detectors and 15 cm for the G-M detectors. Background count rates for ZnS alpha probes averaged approximately I cpm; the average background count rate was approximately 40 cpm for the G-M detectors.
Surface Scan Surface scans of grid blocks in the facility 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.
Identification of elevated levels was based on increases in the audible signal from the recording or indicating instrument.
Alpha and beta gamma scans of large surface areas on the floor of the facility were accomplished by use of a gas 2
proportional floor monitor, with a 600 cm sensitive area.
The instrument is
{
slowly moved in a systematic pattern to cover 100% of the accessible area.
Combinations of detectors and instrument for the scans were:
Beta-Gamma - G-M probe with PRM-6 ratemeter.
Beta-Gamma - G-M probe with " RASCAL" scaler /ratemeter..
Gamma
- NaI scintillation detector (3.2 cm x 3.8 cm crystal) with E
PRM-6 ratemeter.
Alpha
- ZnS probe with " RASCAL" scaler /ratemeter.
Alpha / Beta - Gas proportional floor monitor with PRM-6 ratemeter or
{
Ludlum Model 2220 scaler /ratemeter.
B-1
I Gamma Exposure Rate Measurements Measurements of gamma exposure rates were performed using a Reuter-Stokes pressurized ionization chamber. The average of a minimum of five readings was determined at a distance of 1 meter from the surface to the center of the chamber.
Gamma spectra was collected at each location where exposure rate measurements were made, using a portable high purity germanium detector and I
MCA system.
Removable Contamination Measurements Smears for determination of removable contamination levels were collected on numbered filter paper disks 47 mm in diameter, then placed in individually labeled envelopes with the location and other pertinent information recorded.
The smears were counted on a low background alpha-beta counter.
I Soil Sample Analysis Soil samples were dried, mixed, and a portion sealed in 0.5-liter Marinelli beaker.
The quantity placed in each beaker was chosen to reproduce the calibrated counting geometry and ranged from 400 to 900 g of soil.
Net soils weights were determined and the samples counted using Ge(L1) and intrinsic germanium 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 I
computer capabilities inherent in the analyzer system. The individual spectra were reviewed for identifiable photopeaks related to nuclear
- fuels, activation, or fission products.
Miscellaneous Media Miscellaneous media of very small sample size (dust, residue, standing water, concrete chips, etc.) were analyzed in a high purity Germanium Coaxial I
well detector coupled to a Nuclear Data Model ND-680 pulse height analyzer.
t The individual spectra were reviewed for identifiable photopeaks related to fuel, activation, or fission products.
I B
B-2
I I
i Errors and Detection Limits l
The errors associated with the analytical data presented in the tables of this report, represent the 95% (20) confidence levels for that data.
These errors were calculated, based on both the gross sample count levels and the l
associated background count levels.
When the net sample count was less than the 20 statistical deviation of the background count, the sample concentration I
was reported as less than the minimum detectable activity
(<MDA) or concentration (<MDC).
Because of variation in Compton contribution from other radionuclides in the samples, the MDA's/MDC's for specific radionuclides differ from sample to sample.
I l
Calibration and Quality Assurance I
l Laboratory and field survey procedures are documented in manuals I
developed specifically for the Oak Ridge Associated Universities' Radiological Site Assessment Program.
t With the exception of the measurements conducted with portable gamma l
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.
Gamma spectrometry systems are calibrated as needed using NBS l
traceable standards; quarterly NBS traceable standards are counted to verify ef ficiency factors, and daily backgrounds and standards are counted to check calibration drift or loss of resolution.
Quality assurance results are maintained on file.
Quality control procedures on all instruments 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 I
I B-3
I I
lI l
'I lI lI lI AerENoIx C I
aEcutiT0er cu oE 1.ee TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS I
I I
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I il I
I I
\\
W
- 0' '
June 1974 U.S. ATO',dtC ENERGY COMMISSION r REGULATO'RY GUIDE 4
[
DIRECTORATE OF REGULATORY STANDARDS ns o',
q
[
REGULATONY GUIDE 1.36 L
TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS
~
IL A. INTRODUCTION A licensee havmg a possessioti only beense must b
reta;n, with the Part 50 license, authonzation for specid Section 50.51, " Duration oflicense, renewal," of 10 nudsar rr.stenal (10 CFR Part 70, "Special Nudear CFR Part 50, "Licensmg of Produr tion and Utilization ' Matenal"), byproduct matenal (10 CFR Port 30,"RtJer p
Facilities," requires that each heen* te operate a of General Applicability to Licensing of Byproduct production and utilization facihty be issued for a Matenal") and source rr.atenal (10 CFR Part 40, specified duration. Upon expiration pf the speified "1.icereir4 of Source Matsnal"), until the fuel, radio period, the license may be either renewed or termmt.ted acts compenents, ar.d sources ara removed from the WE by the Commission. Section 50.82, "Applicatiota fot. facility. Appropriate administrative controls and facivy termmation oflicenses," spacifies the requirements that requirements are imposed 'i/ the Part10 license and the must be satisfied to termnate an operating license, technical specifications to assure that proper surveillance including the requirement that the dismantlement of the is performed and that the reactor facility is maintained l
facility and disposal of the component parts not be in a safe condition and not operated.
inimical to the common defense and secunty or to the health and safety of the public. This guide describes A possession.only ID:ense permits various options s'td methods and procedures considered mediptable by the procedures for decomt'tissiontng, such as mothbaE,
4 Regulatory staff for the termmation. of operstmg entombment, or dismanthng. The requirements imputed licenses for nuclear reacton. The Advisory Committee cepend on the opbon selected.
on Reactor Safeguards has been consuusd conceming this guide and has concurred in the regulatory posinen.
Sects 50.82 provides that the h:msee may dis,
rnantle md dispose of the component pans of a nucMr B. DigCUSSION r: actor in accordance with existing repalations. For research reactors and entical facilities, thu has ususMy When a heensee decides to terrntnate he nuclear meant the disassembly of a reactor and its slupment reactor operstmg license, he may, as a first step m tne offute, somenmes to another appropnately bcensed process, request that hs optratmg beense be amended to organizauon for further use. The ste from wluch a restnet him to possess but not operate the-facihty. The reactor has been removed mun be decontaminated, as advantage to the bcensee of convertmg to such a necessary, and inspected by the Commission to deter.
possession only license u reduced surveillance require-mme wnether unrestncted access can bc approved. In menu in that periodic surveillance of equipment im-tne case of nuclear power reactors, dismantlmg has portant to the safety of reactor operation u no longer usually been accomphshed by shippmg fuel offsite, required. Once thn possesmon only license is issued.
making the reactor inoperable, and dnposms of some of reactor operation is not permitted. Other acuvities the radioacuve componenu.
related to cessation of operations such as unloadmg fuel from the reactor and placing it in storage (either onsite Radioactive components may be either shipped off-of offsite) may be contmued.
site for bunal at an authonzed bunal ground or secured s,
g.-
USAlc RECULATORY CUIDES c.o
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on the ste.Those radwactive materials remaimng on the fluids and waste should be removed from the site.
site must be isolated from the pubhc by physical bamers Adequate radiation morutonng, environmental surveil-E or other means to prmnt public access to hazardous lance, and appropnate secunty procedures should be levels of radiation. Surveinan:e u necessary to assure the estabhshed under a possession only beense to ensure that j
long ' term mtegnty of the bamers. The amount of the health and safety of the pubhc is not endangered.
surveillance required depends upon (1) the potential I
hazard to the healtl*.and sakty of the public from
- b. In. Place Entombment. In place entorri.unt con-radioacuve matensi vemammg on the site and (2) the nsts of acahng all the remammg highly radioactive or mtegnty of the phyucal barners. Before areas may be contammated components (e.g., the presure vessel and 1
releard for unnstncted use, they must have been reactor mternals) withm a structure integral with the decontammated or the radioactmty must have decayed biological shield after havmg all fuel anembbes, radio-l to less than presenbed hauts (Table !).
active fluids and wastes, and cenam selected com-I ponents shipped offsste. The structure should provide I
The hszard asociated mth the retired facihty u mtegnty over the penod of time m whien significant evaluated by conndenng,the amount and type of quantities (greater than Table I levels) of radioactmty remamir4 contannnation, the degree of confinement of remam with the material in the entombment. An g
tne remt.imng radioactive traterials. the physical secunty appropnate and contmumg surveillance program should g
povided by tne confinement, the susceptibility to be estabbshed under a possession-only license.
release of radiation as a reault of natural phenomena, l
and the duration of requhol surveillance.
- c. Removal of. Radioactive Components and Dis.
I mantling, All fuel anemblies, radioactive fluids and C. MGULATORY POSITION waste, and other materials having activities above ac-l cepted unrestncted activity levels (Table I) should be I. APPUCATICN FOR A UCENSE TO POSSESS BUT removed irom the site. The facility owner may then have NOT OrUt ATE (POSSESSION-ONLY UCENSE) unrestricted use of the site with no requirement for a license. If the facility owner so desires, the remainder of A requut to amend an operating license to a the reactor facility may be dismantled and all vestges I
possenion only beense should be made to the Director removed and disposed of.
of Ucensing. U.S. Atomic ?.rmg/ Commurion, Washmg-l tons D.C. 20545. The sequest should include the
- d. Conversion to a New Nuclear System or a Fossil followmg information:
Fuel System. This alternative, which applies only to I
nuclear power plants, utihzes the existing turbme system
- a. A descripuon of the current status of the facihty.
with a new steam supply system. The origmal nuclear l
steam supply system should be separated from the
- b. A dese.iption of measu es that will be taken to electric generstmg system and disposed of m accordance
-I prevent enucality or ren:tmiy changes and to mirumize with one of the previous three retirement attemauves.
releases of radioactmty hom the facility.
I
- 3. SURVEILLANCE AND SECURITY FOR THE RE.
I
- c. Any proposed. changes to the tec!mical specifica.
TIREMENT ALTERNATIVES WHOSE FINAL tions that icGect the possession-only faculty status and STATUS REQUIRES A POSSESSION ONLY j
the necessary duanembly/retuement activiues to be UCENSE performed.
'I
~
A facihty which has been licensed under a posses-
- d. A safety'.malyns of both the actmties to be non only beense may contam a significant amount of
(
accomphshed rud the proposed changes to the techmcal radioacuvity m the form of acuvated and contaminated
'm specificauons. '
hardware and structural matenals. Surveillance and g
commensurate security sht,'Jd be provided to anure that
- e. fn inventory of acusated materials and their the pubhc health and safety are not endangered.
locanon in thi facihty.
(I
- a. Physical security to prevent inadvertent exposure
- 2. ALTERNAHVE3 FOR REACf0R RETIREMEhT of personnel should be 'provided by mult ple locked bamers. The presence of these bamers should make it Four altsmatives for retirement of nuclear reactor extremely difficult for an unauthonzed person to gain (j
facihties e.e censidered acceptable by the Regulatory access to areas where radiation or contamination levels 3
sitri. Thee are:
exceed those specified in Regulatory Position C.4. To f
prevent inadvertent exposure, radiation areas above S
- a. Mothba!1mg, Mothballing of a nuclear reactor mR/hr, such as near the activated pnmary system of a i
isen:y consists of puttmg the facility in a state of power plant, should be appropnately marked and should
{
pronive storage, in general, the facihty may be left not be accesible except by cuttmg of welded closures or intz.ct except that til fuel asemblies and the radioactive the disassembly and removal of substantial structures I
C-2
'I and/or shieldmg matenal. Means such as a remote.
(1) Environmental surveys.
i readout mtruson alarm system should be provided to Ie indicate to designated personnel when a physical barrier (2) Facihty radiation surveys.
is penetrated. Secunty personnel that provide access control to the facihty may be used mstead of the (3) Inspecuons of the physical barriers, and I
physacal bamars and the mtrusion alarm systems.
(4) Abnormaloccurrences.
- b. The physical bamers to unauthorized entrance into the facihty, e.g., fences, buildmss, welded doors, and access openmgs, should be inspected at least
- 4. DECONTAMINATION FOR RELEASE FOR UN.
quarterly to assure that these bamers have not detenor-RESTRICTED USE ated and that locks an f lockmg apparatus are mtact.
I If it is desired to terminate a license and to ehmmate
- c. A facihty radiation survey should be performed at any further surveihance requirements, the facihty should least quarterly to verify that no radioactive matenal is be sufficiently decontammated to prevent risk to the escapmg or bems transported through the containment public health and safety. After the decontammation is I
bamers in the facihty. Samphng should be done along satisfactorily accomplished and the site inspected by the most probable path by which radioactive matenal the Commission, the Commission may authonze the such as that stored in the irmer contamment regions hcense to be terminated and the facility abandoned or could be transported to the outer regions of the facility released for unrestricted use. The licensee should per.
.I and ultimately to the environs.
form the decontamination usmg the 'following guide.
lines:
- d. An environmental radiation survey should be performed at least semiannually to verify that no
- a. The licensee should make a reaconable effort to I
siimficant amounts of radiation have been :ticassd to the eliminate residual conta:nmation.
l environment from the facility. Samples such a.s soil, vegetation, and water should be taken at locations for '
- b. No covenng should be applied ta radioactive which statistical data has been established durmg teactor surfaces of equipment or structures by pamt,platmg, or i
l operations.
other covenng material untilit is known that contamina-tion levels (detennmed by a survey and documented) are
- e. A site representative should be designated to be below the limits specified in Table L In addition, a responsible for controlling authonzed access into and reasonable effort should be made (and documented) to movement within the facihty.
further muumize contamination prior to any such covenng.
- f. Administrative procedares should be established I
for the notificauon and reporting of abnormal occur.
- c. The radioactmty of the interior surfaces of pipes, rences such as (1) the emrance of an unauthonzed dram hnes, or ductwork should be determined by person or persons into the facility and (2) a significant mak2ng measurements at all traps and other appropnate I
change in the radiation or contammation levels m the access pomts, provided contammanon at these locations facibty or the offute envronment.
is likely to be wpresentanve of cont =nnnstion on the mtens f the pipes, drain hnes, or ductwork. Surfaces
- g. The followmg reports should be made:
of Pemio. equmment, or scrap which are likely to be I
c.; e d be are of such size, construcuen, or (1) An annual report to the Director of Licensmg,
).e
,o to maxe the surface maccessible for purposes U.S. Atomic Energy Corumsmon, Washmgton, D.C.
of measurernent should be assumed to be contammated 20545, describing the results cf the environmental and in excess of the permissable radiation Imuts,
,I facihty radiation surveys, the status of the facility, and an evaluation of the performance of security and
- d. Upon request, the Commission may authorize a surveinance measures, licensee to rehnquish possession or control of premis:s, I
equipment, or scrap having surfaces contaminated in (2) An abn~:aai rcurrence report to the Regula-excess of the hrmu spc.:ified. This may include, but is tory Operations R nal Office by telephone within 24 not hmited to, speial circumstances such as the transfer hours of discovery of an abnormal occurrence. The of premises to another licensed organization that will I
abnormal occurrence will aim be reported in the annual continue to work with radioactive materials. Requests report described in the preceihng item.
for such authonzation should provide:
- h. Records or logs relat:ve to the followmg items (1) Detailed, specific information describing the I
should be kept and retamed until the beense is tenni-gemises, equipment, scrap, and radioactive contami.
f nated, after which they may be stored with other plant nants and the nature, extent, and degree of residual records:
surface contammanon.
I C-3
I (2) A detailed health and safety analysis indi.
or a change in the technical specifications should be catmg that the residual amounts of matenals on surface reviewed and approved m accordance with the require.
areas, together with other considerations such as the ments of 10 CFR 950.59.
prospecuve use of the premises, equipment,or scrap.are unlikely to result m an unreasonable risk to the health if major structural changes to radioactive components and safety of the pubhc.
of the facihty are planned, such as removal of the pressure vessel or major components of the primary
- e. Pnor to release of the premises for unrestncted system, a dismantlement plan meiudmg the information use, the hcewee should make a comprehensive radiauon required by 650.82 should be subnutted to the Commis-survey estabhslung that contamination is withm the sion. A dismantlement plan should be subnutted for all hmits specifed in Table 1. A survey report should be the alternatives of Regulatory Position C.2 except filed with the Director of1.icennng. U.S. Atomic Energy mothbalhng. However, mmor disassembly activities may Commission, Washmgton, D.C. 20545, with a copy to still be performed in the absence of such a plan.
the Director of the Regulatory Operations Regional provided they are permitted by exsstmg operstmg and Office havmg junsdiction. The report should be filed at maintenance procedures. A dismantlement plan should least 30 days pnor to the planned date of abandonrrant.
include the followmg:
The survey report should:
- I
- a. A description of the ultimate status of the facility (1) Identify the premises;
- b. A description of the dismantling activities and the (2) Show that reasonable effort has been made to precautions to be taken.
I reduce residual contamination to as low as practicable levels;
- c. A safety analysis of the dismantling activities including any effluents which may be released.
I (3) Describe the scope of the survey and the general procedures followed;and
- d. A safety analyns of the facility in its ultimate status.
(4) State the fmding of the survey in units I
specified in Table 1.
Upon satisfactory review and approval of the dis.
mantling plan, a dismantling order is issued by the After review of the report, the Commission may Commission m accordance with 650.82. When dis.
.g inspect the facilities to confirm the survey prior to mantling is completed and the Commission has been g
granung approval for abandonment.
notified by letter, the appropriate Regulatory Opera.
tions Regional Office inspects the facility and verifies
- 5. REACTOR RETIREMENT PROCEDURES completion in accordance with the dismantlement plan.
If residual radiation levels do not exceed the values m I
As indicated in Regulatory Position C.2, several Table I, the Commission may termmate the beense. If alternauves are acceptable for reactor facility retirement.
these levels are exceeded, the beensee retams the If mmor disassembly or mothbalhng" is planned, this possession.only license under which the dismantling could be done by the existing operating and mamte.
activities have been conducted or, as an alternative, may nance procedures under the beense m effect. Any make appheauon to the State (if an Agreement State) planned actions involymg an unreviewed safety question for a byproduct matenals bcense.
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I C-4
I TABLEI ACCEPTABLE SURFACE CONTAMIN ATION LEVEL.3 l
REMOVABLE e NUCllDEa AVERAGE c MAXIMUMbd b
b U-nat, U 235, U 238, and 5,000 dpm al100 cm 15 000 dpm al100 cm 1,000 dpm c/l00 cm2 2
2 associated decay products 2
20 dpm/100 cm2 Transuranics, Ra 226, Ra 228, 100 dpm/100 cm2 300 dpm/100 cm Th-230.Th 228,Pa 231, Ac 227,1125,1 129 2
2 2
200 dpm/100 cm I
Th-nat,Th-232 St 90, 1000 dpm/100 cm 3000 dpm/100 cm Ra 223, Ra 224 U-232, 1 126,1-131,1 133 2
2 2
Beta gamma emitters (nuclides,
5000 dpm $9/100 cm 15,000 dpm $9/100 cm 1000 dpm Eq/100 cm wtth decay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above.
aWhen surface conummation by both alpha-and beta-gamma emittmg nuchdes exists, the hmits established for alpha-and
.I beta-gamma-emittmg nLclides should apply sndependently.
bas used in this table, dpm (dtantepations per rrunute) means the rate of emisson by radioacuve matenal as determined by correcting the counts per minute observed by an appropnate detector for backpound, efficuney, and geometne factors assocuted with the mstrumentauon-cMeasurements of average contaminant should not be averaged over more than 1 square meter. For objects ofless surface area, the I
average should be denved for each such object.
2 dThe maxunum contammation level appbes to an area of not more than 100 cm.
- The amount of removable radioactive matenal per 100'em2 of surface area should be detemuced by wipmg that area with dry filter or soft absorbent pape:, applymg moderate pressure, and assesang the amount of radioactive material on the m1pe with an appropnate instrument of known efficiency. When removable contamination on objects of less surface vea is deternuned. the perunent leveh should be reduced proporuonally and the entste surface should be m1 ped-lI
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I C-5