Technical Note
ORP/LV-77-3
RADIOLOGICAL SURVEYS OF IDAHO PHOSPHATE
ORE PROCESSING—THE THERMAL PROCESS PLANT
NOVEMBER 1977
^ec sr%
U.S. ENVIRONMENTAL PROTECTION AGENCY
OFFICE OF RADIATION PROGRAMS
LAS VEGAS FACILITY
LAS VEGAS, NEVADA 89114
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Technical Note
ORP/LV-77-3
RADIOLOGICAL SURVEYS OF IDAHO PHOSPHATE
ORE PROCESS ING--THE THERMAL PROCESS PLANT
NOVEMBER 1977
OFFICE OF RADIATION PROGRAMS - LAS VEGAS FACILITY
U.S. ENVIRONMENTAL PROTECTION AGENCY
LAS VEGAS, NEVADA 89114
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DISCLAIMER
This report has been reviewed by the Office of Radiation
Programs - Las Vegas Facility, U. S. Environmental Protection
Agency, and approved for publication. Mention of trade names or
commercial products does not constitute endorsement or recommen-
dation for their use.
11
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ACKNOWLEDGMENTS
The field conduct of this study was directed by Joseph A.
Cochran, whose present address is:
Chief - Sanitation Branch
Directorate of Facilities Engineering
U.S. Department of the Army
Fort Ord, California 93941
This report has been prepared by:
Gregory G. Eadie and David E. Bernhardt
Office of Radiation Programs-Las Vegas Facility
U.S. Environmental Protection Agency
Las Vegas, Nevada 89114
This study involved the cooperation and coordination of many
people. In particular, the following contributed substantially
to the project.
FMC CORPORATION
Michael Schmidt, Technical Superintendent, Pocatello
C. Dee Holmes, Environmental Coordinator, Pocatello
Dr. David C. Drown, Project Process Engineer, Pocatello
Dr. William R. King, Environmental Manager, Philadelphia
The stack samples could not have been obtained without the
extensive efforts of Dr. Drown's staff and we thank them very
much.
IDAHO DEPARTMENT OF HEALTH S WELFARE
ENVIRONMENTAL SERVICES DIVISION
Michael Christie, Health Physicist, Boise
Gary F. Boothe, Health Physicist, Boise
Charles L. Freshman, Environmental Specialist, Pocatello
The review efforts of the FMC Corporation, the State of
Idaho, and the U.S. Geological Survey were especially helpful in
the preparation of this report. Other reviewers included the
U.S. Environmental Protection Agency-Region X, and the staff of
the Office of Radiation Programs-Headquarters and Eastern Environ-
mental Radiation Facility.
111
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PREFACE
The Office of Radiation Programs of the U.S. Environmental
Protection Agency carries out a national program designed to
evaluate population exposure to ionizing and nonionizing radia-
tion, and to promote development of controls necessary to protect
the public health and safety. This report describes various
surveys which were conducted at the FMC Corporation's Thermal
Process Plant near Pocatello, Idaho to provide basic data to be
used for the evaluation of the radiological impact associated
with the phosphate industry. Readers of this report are encour-
aged to inform the Office of Radiation Programs of any omissions
or errors. Comments or requests for further information are also
invited.
Donald W. Hendricks
Director, Office of
Radiation Programs, LVF
IV
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CONTENTS
Page
ACKNOWLEDGMENTS iii
PREFACE iv
LIST OF FIGURES vi
LIST OF TABLES vii
ABSTRACT 1
INTRODUCTION 2
THE THERMAL PROCESS 3
THE RADIOLOGICAL ASSESSMENT OF THE THERMAL PROCESS PLANT 6
Gamma Radiation Surveys 6
External Dose Assessment 9
Background Radiation 9
In-Plant Radiation Exposures 9
Ambient Radon-222 Concentrations 11
Sampling System and Analytical Methods 11
Radon Measurement Results 12
Radon Dosiroetry 14
Plant Discharge of Radon 15
Thermal Process Samples 15
Environmental Samples 20
Effluent Samples 23
Airborne Particulate Radioactivity 28
Background Airborne Radioactivity Concentrations 28
In-Plant Air Sampler 28
Gross Versus Net Results 29
In-Plant Air Sampling Net Results 30
Specific Activity of Airborne Particulate Sampling 36
v
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CONTENTS (Continued)
Particle Size Characterization
Cascade Impactor Sampling
Gross Results of Impactor Sampling
Comparison of Air Sampling Results
Particle Size Analysis
Stack Sampling
Mass and Activity Balance
SUMMARY
REFERENCES
APPENDIX A
APPENDIX B
APPENDIX C
APPENDIX D
APPENDIX E
- RADIOCHEMICAL ANALYTICAL METHODS
- AIRBORNE PARTICULATE SAMPLING - GROSS
RADIONUCLIDE CONCENTRATION RESULTS
- ADDITIONAL SAMPLING OF INPUT AND SLAG PRODUCTS
- MASS AND ACTIVITY BALANCE
- UNCERTAINTIES IN LEAD-210 AND POLONIUM-210
AIR SAMPLING DATA
Page
37
37
38'
38
47
52
54
58
60
63
71
77
85
89
LIST OF FIGURES
Number
1 Flow Diagram of the Thermal Process
2 Particle Size Distribution in the Calciner
3 Particle Size Distribution in the Proportioning Bldg
4 Particle Size Distribution in the Furnace and
Pelletizer Buildings
5 Annual Mass and Activity Balance of Major Streams -
The Thermal Process Plant
Page
5
48
49
50
56
VI
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LIST OF TABLES
Number Page
1 Gamma Radiation Surveys 7-8
2 Ambient Radon-222 Concentrations 13
3 The Thermal Process-Input to Electric Furnace 18
4 The Thermal Process-Products and By-Products 19
5 Portneuf River Samples 21
6 Slag Pile Rainwater Runoff 22
7 No. 4 Furnace Cooling Water 22
8 Pond Water-Settling and Evaporation 24
9 Calciner Scrubber Effluents 25
10 Scrubber Effluents 26
11 Miscellaneous Samples 27
12 Radioactivity Content of Blank Glass Fiber Filters 30
13 Airborne Particulate Sampling in the Calciner Area 31
14 Airborne Particulate Sampling in the Proportioning Bldg. 32
15 Airborne Particulate Sampling in the Pelletizer Bldg. 33
16 Airborne Particulate Sampling in the Furnace Bldg. 34
17 Airborne Particulate Sampling 55
18 Cascade Impactor Sampling in the Calciner Area 39
19 Cascade Impactor Sampling in the Proportioning Bldg. 40
20 Cascade Impactor Sampling in the Pelletizer Bldg. 41
21 Cascade Impactor Sampling in the Furnace Bldg. 42
22 Radioactivity Content of Blank Impactor Whatman #41
Paper Filter (Slotted) 43
23 Radioactivity Content of Blank Glass Fiber Filters
(8 x 10-inch) 44
24 Summary of Gross Results of the Cascade Impactor
Sampling 45
25 Radioactive Content of Blank Glass Fiber Filter
C2.5-inch diameter) 53
26 Stack Sampling Results 55
VII
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ABSTRACT
Radiological surveys conducted at the FMC Corporation's
Thermal Process Plant in Pocatello, Idaho indicate slightly
elevated ambient levels of natural radioactivity within the
plant. Compared to an estimated natural background annual dose
equivalent rate of about 79 mrem, net gamma exposure rates ranged
from 42 mrem in general plant areas to 182 mrem per work year on
the slag pile. Ambient radon-222 concentrations, ranging from
0.17 to 1.4 pCi/1, were measured in several indoor locations, but
11 pCi/1 was measured in the Control Room of the Condenser and
Fluid Bed Building.
Elevated airborne radioactivity concentrations, orders of
magnitude greater than measured background concentrations, were
also measured in several work areas, with polonium-210 and
radium-226 being the most predominant radionuclides of the natural
uranium decay series. Particle size characterization indicates
roughly 50 percent of the arithmetic total radioactivity is
associated with the particle size fraction less than one micro-
meter equivalent aerodynamic diameter. Stack sampling results
also show that appreciable concentrations of the naturally-
occurring radionuclides, particularly polonium-210 and uranium,
are being discharged into the local environs.
A general radioactivity balance indicates that the ore is
the source of essentially all of the inlet radioactivity. The
slag accounts for essentially all of the outgoing uranium and
radium-226 and up to about 50 percent of the lead-210 and
polonium-210. The stack effluents account for about 20 percent
of the total polonium-210 activity.
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INTRODUCTION
Recent reports (Windham et al., 1976; Guimond and Windham,
1975) indicate that mining and processing of phosphate ores
technologically enhance the quantities of naturally-occurring
radioactive materials such as radium and uranium available to man
in the environment. Such radionuclides have relatively long
half-lives and will therefore persist in man's habitat for
thousands of years.
Current estimates show that the phosphate industry has
generated about twenty times the volume of waste materials as
compared to that from the uranium mining and milling industry.
The total radioactive waste inventory generated by the uranium
and the phosphate mining and milling industries are comparable to
each other although the per gram radioactivity content of the
phosphate waste is much less than that of uranium waste material
(EPA, February 1977).
In general, the phosphate industry is neither regulated nor
monitored for the possession, use, or discharge of radioactive
materials associated with phosphate rock and its products and by-
products. Recently, the State of Idaho (June 1, 1977) has
prohibited the use of phosphate slag material in the construction
of habitable structures, but has permitted the continued use of
slag for road construction, railroad ballast, and other general
purposes.
Of immediate concern is the accumulation of a data base
which will ultimately lead to an assessment of the impact on
public health due to the phosphate industry's activities. This
report discusses various radiological surveys conducted in the
FMC Corporation's Thermal Process Plant in Pocatello, Idaho. A
similar study, conducted in a wet-process plant, will be dis-
cussed in another report.
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THE THERMAL PROCESS
Phosphate rock is obtained from the Gay Mine located 35
miles northeast of the Thermal Process Plant in Pocatello,
Idaho. At the plant, phosphate rock is crushed, screened, and
briquetted. The briquettes are then fed into calciners where they
are heated to 2400° F. This calcining serves two purposes:
1) to burn out organic material and 2) to heat-harden agglomer-
ates so they will withstand further processing steps without
disaggregating. The calcined nodules pass through a proportion-
ing building where sized coke and silica are blended in. This
mixture is called the "burden" and flows by conveyor belt into
the electric furnaces. In the furnace, the high temperature
reaction (2500° to 8000° F) drives off two gases - phosphorus and
carbon monoxide. It leaves two molten residues - slag and
ferrophosphorus (a mixture of iron, vanadium, chromium, etc.). A
simplified chemical equation for the electric furnace reaction
is :
6Si02 + IOC = ?4 + 10CO + 6CaSi03
Off-gases are treated in electrostatic precipitators for dust
removal, then in a waterspray cooler where the gaseous elemental
phosphorus is condensed, collected in a sump, and pumped to
storage. Most of the carbon monoxide is used in the calciner but
some is burned to become carbon dioxide, which is released to the
atmosphere via the flare stack.
The waste product slag, which is predominately calcium
silicate, is crushed and used as an aggregate in highway surfac-
ing material and as railroad track ballast. Some other uses of
slag have been for concrete aggregate material and for use in
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insulation products. The ferrophosphorus residue is usually
custom crushed and processed to extract vanadium at another
plant. The finished product, elemental phosphorus, is shipped to
other plants for use in detergents, cleaning products, and in
numerous other applications.
In brief, the FMC Corporation's Thermal Process Plant
handles over 2,000,000 tons of raw materials (1,800,000 tons of
phosphate rock; 190,000 tons of coke; and 130,000 tons of silica)
and produces 125,000 tons of elemental phosphorus per year. From
a material balance viewpoint, about 2,000,000 tons of waste by-
products (i.e., slag and ferrophosphorus) are generated each
year. A simplified flow diagram of this thermal process is shown
in Figure 1.
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INPUT
PROCESS
PRODUCTS
& BY-PRODUCTS
'PHOSPHATE\
ROCK
COKE
\
SILICA
\
CARBON
MONOXIDE
Recycled
CALCINER
CALCINED
BRIQUETTE
ELECTRIC
FURNACE
PRECIPITATOR
(gaseous)
CONDENSERS
KSTACK VENT EXHAUST
FERROPHOSPHORUS
SALES
ELEMENTAL
PHOSPHORUS SALES
CARBON MONOXIDE
FLARE STACK
Figure 1. Flow diagram of the thermal process
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THE RADIOLOGICAL ASSESSMENT OF THE THERMAL PROCESS PLANT
Radiological surveys were completed at the FMC Corporation's
Thermal Process Plant in Pocatello, Idaho during the week of May
5 to 8, 1975. Included among these surveys were: gamma radiation
measurements; ambient radon-222 concentrations; radiological
analyses of selected input, process and product samples; airborne
particulate sampling; and particle-size characterization. Stack
sampling of representative effluent discharge vents was also
completed in September 1975. The following sections of this
report discuss the results of these surveys.
GAMMA RADIATION SURVEYS
In order to assess the gamma radiation exposure rates in the
various working areas of the Thermal Process Plant, a survey was
conducted using a portable gamma scintillator survey meter.*
This instrument was calibrated with a radium-226 standard, and
measured the relative gamma radiation exposure rate in units of
microroentgens per hour (vR/h)• Measurements were made at a
height of three feet above an area surface and the results shown
in Table 1 represent average values for each location. At the
time of these surveys, a pressurized ionization chamber (PIC) was
not available to complete radiation surveys for comparison to the
portable scintillator measurements. Although the scintillator
survey meter's response is dependent on the energy of the gamma
photon (Eadie et al., 1976), no instrument response correction
factor has been applied to the reported scintillator measurements.
Therefore, for the following dose estimates, the reported scintil-
lator measurements (Table 1) are assumed to be equivalent to the
"true" gamma exposure rate due to both the cosmic and the terres-
trial components.
* Baird-Atomic, Type NE148A-Gamma Scintillator Ratemeter.
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LOCATION DESCRIPTION
TABLE 1. GAMMA RADIATION SURVEYS
GENERAL WORK AREAS
GROSS GAMMA EXPOSURE RATE NET GAMMA EXPOSURE RATE*
yR/h
ESTIMATED
WORK YEAR DOSE EQUIVALENT**
mrem/year
Shale Area: lunch room
Ore Dump Unloader Area:
Outside
Inside Control Room
Coke Dryer: Outside
Inside Control Room
Pelletizer Building:
General Floor Area
Calciner #1:
Loading Area
Inside Control Room
Proportioning Building:
General Floor Area
Inside Lower Control
Room
Furnace Building:
West End, Lower Level -
General Floor Area
Control Room #3-4
(inside & outside)
Foreman's Office
Fluid Bed Control Rm.
General Area Near
Furnace #4
West End, Near
Slag Run
25
70
20
60
28
15
20
12
15
40
10
8
20
30
30
35
16
61
11
51
19
6
11
3
6
31
1
BKG.
11
21
21
26
32
122
22
102
38
12
22
6
12
62
2
BKG.
22
42
42
52
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00
LOCATION DESCRIPTION
TABLE 1. GAMMA RADIATION SURVEYS
GENERAL WORK AREAS(Cont'd)
GROSS GAMMA EXPOSURE RATE
uR/h
NET GAMMA EXPOSURE RATE*
uR/h
ESTIMATED
WORK YEAR DOSE EQUIVALENT**
mrem/year
Ferrophosphorus
Cooling Area #2
Cooling Area #3
Slag Pit behind
Building
Slag Area #1
Slag Area:
General Outside Area
Technical Building:
Inside
Outside
Old Kiln Building:
Inside
General Plant Areas:
Outside roadway between
Old Kiln Bldg. and
Proportioning Bldg.
Outside roadway between
Furnace Bldg. and
Pelletizer Bldg.
Background Radiation
General Pocatello,
Idaho Area
20
16
30
25
TOO
10
20 to 40
15
60
60
9
n
7
21
16
91
1
n to 31
6
51
51
-
22
14
42
32
182
2
22 to 62
12
102
102
18 mrem per Work
79 mrem per Year
Year
for
continuous exposure
+These surveys were conducted on May 8, 1975 using the Baird-Atomic ME148A-Gamma Scintillator Ratemeter.
All measurements are average instrument response, in units of yR/h, at three-foot height above surface area.
*An estimated background gamma exposure rate of 9 pR/h (using the scintillator survey meter) has been sub-
tracted from the gross value to obtain the net result. (This instrument does not respond fully to cosmic
radiation. Thus, neither the estimated background nor the gross readings reflect the true contribution of
the cosmic component.)
**Work Year has been defined as 40 h/wk for 50 wk/y (i.e., 2000 hours). The Work Year Dose Equivalent is the
net value multiplied by the Work Year time equivalent;!'.e., the dose estimate due to continuous work year
occupancy per specific area. This may be an overestimate, depending on the actual occupancy factors.
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EXTERNAL DOSE ASSESSMENT
Background Radiation
Gamma exposure rate data for non-coastal plain regions of
the United States have been summarized and indicate a mean value
of 5.2 yrem/h (46 mrem/year) absorbed dose rate equivalent in air
from terrestrial sources (NCRP, November 1975). The absorbed
dose rates (mrad/year) have been converted to dose equivalents
(mrem/year) using a quality factor of one for gamma photons,
which is consistent with the recommendations of the National
Council on Radiation Protection (NCRP, January 1971). Specific
surveys around the National Reactor Testing Station near Idaho
Falls, Idaho showed an average terrestrial dose equivalent rate
of about 6 urem/h (Oakley, 1972). Therefore, the natural ter-
restrial background radiation annual dose equivalent for the
Pocatello, Idaho area should be approximately 53 mrem per year
(i.e., continuous exposure at 6 prem/h x 24 h/day x 365 day/year).
The average dose equivalent rate due to cosmic radiation in the
Pocatello area has been estimated to be roughly 3 u^em/h, for a
scintillator survey meter measurement. This value corresponds to
an annual average cosmic dose equivalent rate of 28 mrem per year
as summarized by NCRP (November, 1975). Therefore, a total
natural background radiation exposure rate of 9 yrem/h has been
used to obtain the "net" exposure rates for the survey results as
reported in Table 1.
In-Plant Radiation Exposures
the lowest exposure rate (8 yR/h) in the Thermal Process
Plant was essentially the background exposure rate and was
measured in the Control Room #3-4 in the Furnace Building (Table
1). Therefore, for workers who spend the majority of their
working periods inside this control room, the maximum annual dose
equivalent would be about the average background rate of 18 mrem
per year (i.e., 9 PR/h x 40 h/wk x 50 wk/year). General working
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areas within the Furnace Building had net exposure rates ranging
from 1 yR/h in the West End, Lower Level to 26 yR/h around the
slag run. For general working areas of the Furnace Building, the
average net exposure rate was 21 yR/h, which corresponds to a
work year dose equivalent of 42 mrem/year.
For the general working areas of the Calciner, the net
exposure rate was 11 yR/h. Working areas within the Pelletizer,
Old Kiln, and the Proportioning Buildings had average net expo-
sure rates of 6 yR/h. The average net exposure rate was 1 yR/h
inside the Technical Building which also is essentially the
background exposure rate.
The highest net exposure rate for any of the indoor working
areas within the Thermal Plant was 31 yR/h in the Lower Control
Room of the Proportioning Building. Therefore, the maximum work
year dose equivalent for a worker in this area would be about 62
mrem per year.
In-plant radiation exposures are summarized in Table 1. By
assuming continuous exposure during the entire work period (40
h/wk x 50 wk/year) at the net gamma exposure rate levels, the
reported dose equivalents are probably overestimated. A more
precise dose estimate could be obtained by fractioning the time
spent in various work locations and using that particular exposure
rate; however, such information has not been obtained in this
study. It should also be emphasized that these in-plant external
gamma dose estimates for the work year are in addition to the
dose received by an individual due to natural background radiation
exposure. Furthermore, these estimates are based on free air
measurements. It is estimated that the dose to internal body
organs varies from 0.5 to about 0.7 of the free air measurement
due to the shielding effect of the body CO'Brien and Sanna,
1976).
10
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Due to the use of waste-product slag for on-site roadway
construction purposes, high net gamma exposure rates (averaging
about 51 uR/h) were obtained out-of-doors on the roadways.
Surveys around the ore dump unloader area indicate net exposure
rates of about 61 yR/h. The highest net gamma exposure rate
(91 yR/h) for any area in the Thermal Process Plant was out-of-
doors in the slag dump area. Although work year dose equivalent
rates have been shown in Table 1 for the out-of-doors areas, it
is obvious that these values are overestimates since no worker
spends his entire work shift directly on the slag pile or in the
ore-storage area.
Surveys conducted at a Thermal Process Plant in Florida
(Windham et al., 1976) showed external gamma dose rates slightly
higher than those measured in this study. Total measured dose
equivalent rates (including background) in the Florida plant
ranged from 100 mrem per work year in general plant areas to a
maximum of 300 mrem per work year in the ferrophosphorus and slag
storage area. For the Idaho plant, the general work areas
averaged about 42 mrem per work year, and the maximum net dose
equivalent rate (on the slag pile) was 182 mrem per work year,
above a background level of about 79 mrem per year.
AMBIENT RADON-222 CONCENTRATIONS
Sampling System and Analytical Methods
A continuous, low-volume sampling system was used to obtain
samples of ambient air for analysis of radon content (U.S. Public
Health Service, 1969). This sampling technique consists of
drawing filtered air through a small, low-volume air pump (less
than 10 ml/min sampling rate) into a 30-liter Mylar bag. The air
intake was about one meter above the ground surface. Various
sampling time periods were used for this study.
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Radon analysis was completed at the Environmental Monitoring
and Support Laboratory in Las Vegas, Nevada (EMSL-LV) using a
radon concentration apparatus for sample preparation (Johns,
1975) . This apparatus permits the isolation of radon from about
a 5-liter ambient air sample by transferring the ambient air from
the Mylar bag into a container of known volume, followed by
circulation through water and carbon dioxide traps. Radtm is
collected on two charcoal traps maintained at a temperature of
about minus 80° C using dry ice and acetone. The radon sample is
then de-emanated into a Lucas scintillation cell (125-ml volume)
using helium gas at 400° C. The Lucas cell is held for four and
one-half hours to allow for the ingrowth of the radon daughters
and then counted in a photomultiplier tube/sealer unit. The
radon activity is calculated to the mid-point of the sample
collection period to account for the radioactive decay
(T^ = 3.82 days) of radon from the time of sample collection to
time of analysis. The ambient air volume sampled is determined
by correcting for air density differences based on the nearest
1000-foot elevation increment, and the average ambient temperature
during the sampling period. The radon concentrations (in pCi/1)
reported in Table 2 therefore represent the average ambient radon
concentration for the sampling period for each specific sampling
site location.
Radon Measurement Results
Ambient radon-222 concentrations measured in the various
indoor working areas ranged from a low of 0.17 pCi/1 in the
Pelletizer Building to a high of 11 pCi/1 in the Control Room of
the Condenser and Fluid Bed Building. Harley (1975) estimated
the average background concentration of radon in surface air of
about 0.10 pCi/1 in the northern hemisphere. Other authors
(Pearson, 1967; United Nations, 1972} have reported average radon
concentrations in the general environment ranging from 0.03 to
0.4 pCi/1. Ambient outdoor radon concentrations measured in
areas associated with inactive uranium mill tailings piles
12
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indicate average levels of 0.34 pCi/1 (Monticello, Utah), 0.38
pCi/1 (Salt Lake .City, Utah), 0.51 pCi/1 (Durango, Colorado), and
0.83 pCi/1 (Grand Junction, Colorado), (U.S. Public Health
Service, 1969).
Indoors, the radon levels are typically from three to four
times the outdoor levels due to the entrapment within the structure
of radon emanation from soils beneath the structure and also from
radon emanated from the building materials, supplemented by radon
infiltration from outside air. In general, radon levels inside
structures are usually 0.6 to 1.2 pCi/1 for ventilation rates
below four air changes per hour (Johnson et al., 1973). As shown
in Table 2, the ambient radon concentrations measured indoors in
TABLE 2. AMBIENT RADON-222 CONCENTRATIONS
Date
Time
On Off
5/5/75
0958
5/5/75
1027
5/6/75
0902
5/6/75
1257
5/7/75
1223
5/9/75
1200
1330
1245
1429
1400
Grab
Sample
Location
Description
Pelletizer Building
Control Rm -Calciner
Control Room -
Condenser 5 Fluid Bed
Control Room (lower) -
Proportioning Bldg.
Control Room #3-4
Furnace Building
Carbon Monoxide
Gas Sample
Radon ^
Concentration
(PCi/1)
0.17 ± 0.036
0.63 ± 0.07
11 ± 0.28
1.4 ± 0.11
0.28 ± 0.047
3.9 ± 1.4
* Radon-222 concentration ± two-sigma counting error term.
To convert to pCi/m3, multiply the pCi/1 value in the table
by 1000.
13
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the control rooms of the Condenser and Fluid Bed, and the Propor-
tioning Buildings are higher than would normally be expected from
natural terrestrial background sources alone. It is recommended
that additional sampling for radon or radon progeny (expressed
as Working Levels) be performed to better define both
the area background and the in-plant levels,
Radon Dosimetry
The report by Swift et al. (1976) discusses various aspects
of radon dosimetry including the degree of equilibrium between
radon and its daughters, the fraction of unattached daughter ions
and certain assumptions regarding exposure modes to tissues
within the lung. For dose estimates in this study, a dose
conversion factor of 4 mrem per year, for continuous exposure to
1 pCi/m3, was used. This factor represents radon-222:polonium-
218:lead-214:bismuth-214 concentrations of 1.0:0.90:0.51:0.35
pCi/m 3, respectively. This is equivalent to an average daughter
product equilibrium ratio of 50 percent and an average ventilation
rate of one air change per hour, which is typical in living
accommodations with adequate ventilation. This dose is delivered
to the basal cell nuclei of segmental bronchi which are estimated
to lie 60 ym below the surface where the activity is deposited.
A quality factor of 10 for alpha particles has been used in
the dose calculations. Therefore, continuous exposure to an
assumed average background radon concentration of 100 pCi/m3
(0.10 pCi/1 as estimated by Harley, 1975) would result in a dose
equivalent of 400 mrem per year. For a worker exposed for an
entire work year (40 h/wk x 50 wk/year) to the maximum ambient
measured radon concentration of 11,000 pCi/m 3, the dose equivalent
would be about 10 rem per year.
* 4 mrem/year/(pCi/m3) x 2000 h j^k Year x 11,000 pCi/m3 x 10"? rem/mrem
8760h
Exposure
10 ran/year
14
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Plant Discharge of Radon
One grab sample of the carbon monoxide gas, which is recycled
through the system (Figure 1) but is ultimately discharged to the
atmosphere, was obtained. The measured radon concentration in
this sample was 3.9 pCi/1 (Table 2).
THERMAL PROCESS SAMPLES
Appendix A discusses the radiochemical analytical techniques
used for this study. Since the completion of the analyses of the
following samples, a question has arisen as to the validity of
the lead-210 radiochemical procedure. Although the procedure as
outlined in Appendix A for the lead-210 analysis was fully tested
and proved satisfactory for water sample analysis, there now
appears to be some difficulty in obtaining consistent lead-210
recovery rates for solid samples such as the ore, briquettes,
slag, and air filter samples. The inaccuracies of the lead-210
results relate to analysis of samples without the benefit of a
tracer or carrier yield. Thus the assumed yield, which was
assumed to be the same as that for test samples (i.e., 85 percent)
was in error by up to a factor of five. Comparison of the lead-
210 results as reported in Appendix C for the two EPA laboratories
indicates that the EMSL lead analyses average about a factor of
five lower than the actual lead-210 content (see Appendix E for
data which limits the uncertainty for individual results). This
error has also been noted in recent cross-check air sample results
Since the majority of samples have been consumed to obtain the
results reported here, it is impossible to re-analyze these
samples to resolve this question. Therefore, the lead-210 results
which have been obtained using the EMSL procedure (Appendix A)
have been reported here, but for solid samples, these results
probably underestimate the real sample lead-210 activity by up to
about a factor of five. Future lead-210 analyses will be con-
ducted with a revised technique.
15
-------
The problem of underestimating the lead-210 content is also
reflected in the estimate of a sample's polonium-210 content, as
discussed in Appendix A. The duplicate results reported for
calcined nodules and slag (Appendix C, Table C-2 and C-3) indicate
that, for instance, where the EMSL lead-210 values were used, the
polonium-210 values may overestimate the actual polonium-210
content by up to a factor of ten or more (ranges from 1 up to
about 20, but could be higher) i.e., if the reported EMSL lead-
210 results are actually a factor of five too low. This error
does not apply to the ore, calcined nodules, and slag samples; or
the specific stack samples where the polonium was over an order
of magnitude greater than the lead. Recognizing these difficul-
ties encountered in obtaining valid lead-210 and polonium-210
results, the following sections describe the radiological evalua-
tions of the Thermal Process.
The two sigma (95 percent) statistical counting error is
reported for all of the radiochemistry data presented in this
report. The techniques for calculating the counting errors are
illustrated in Johns (1975). The error terms for data averages,
the decay and ingrowth corrected polonium-210 results, and for
the net air sample results represent error terms obtained using
techniques for propagating error estimates as given in Appendix A
and by Eadie and Bernhardt (1976).
It should be recognized that the statistical counting error
only includes the uncertainty from the number of counts measured
from the sample and instrument background. It is not an estimate
of the total analytical uncertainty nor uncertainties in sample
collection, preparation and/or aliquoting procedures. Evaluation
of soil sampling and analytical techniques for plutonium by
Bernhardt (1976) indicate that sampling and analytical variations
of up to 50 percent and more (95 percent confidence level) are
observable in many groups of data, especially at levels within
about an order of magnitude of the analytical detection level.
Results of aliquots for ore, nodule, and slag samples (Appendix
C) indicate a general two-sigma aliquoting and analytical
uncertainty from less than 10 percent to over 50 percent. Thus,
16
-------
it should be recognized that the uncertainty of the results is
greater than the counting error.
Table 3 presents the results of analyses of samples con-
sidered to be typical of the input to the thermal process.
(Additional analyses for these samples are given in Appendix C.)
The average radium-226 content of the Idaho phosphate rock (ore)
was 26 pCi/g, and the calcined ore contained 25 pCi/g [Table 3).
Radium-226 analysis of Florida ore average about 60 pCi/g (Guimond
and Windham, 1975). Samples of silica and the two coke samples
used in the Idaho plant had radium-226 contents of 1.7, 0.70, and
0.78 pCi/g, whereas samples from the Florida plants were 0.36 and
0.28 pCi/g, respectively. The Idaho silica and coke samples are
at typical background concentrations of the naturally-occurring
radionuclides (NCRP, 1975; Eisenbud, 1963), and are at least an
order of magnitude lower in radioactivity content than the input
ore.
Results of analyses of the products and by-products of the
thermal process are shown in Table 4. The slag waste product
samples had an average radium-226 content of 32 pCi/g (Appendix
C, Table C-3) but had appreciably less polonium-210 and lead-210
as compared to the input ore (Table 3). (Specific gravity deter-
minations on samples of Idaho phosphate ore and slag indicated
average values of 2.7 and 2.8 g/cm3, respectively.) Florida
phosphate slag had an average radium-226 content of 56 pCi/g
(Guimond and Windham, 1975).
17
-------
TABLE 3. THE THERMAL PROCESS - INPUT TO ELECTRIC FURNACE
RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM, IN pCi/g
00
Radionuclide
Ra-226
Po-210+
U-234
U-235*
U-238
Th-230
Th-232
Pb-210+
Ra-228
Phosphate
Rock**
26
22
22
0.95
22
22
0.43
27
0.89
+
±
±
±
±
±
±
+
±
19
3.0
2.0
0.40
3.2
4.1
0.12
12
0.28
Calcined
Rock +
25
23
1.0
24
23
0.47
18
0.97
± 8.6
<2.65
± 2.2
± 0.24
± 1.8
± 2.0
± 0.10
± 8.6
± 0.48
Silica
1.7
2.6
1.5
1.6
0.69
0.67
± 0.
± 0.
<0.
(0.
± 1.
± 0.
± 0.
± 0.
<0.
24
90
86
07)
4
53
37
55
89
Coke
0.78 ± 0.
<1.
<0.
(<0.
•
-------
TABLE 4. THE THERMAL PROCESS-PRODUCTS AND BY-PRODUCTS
RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM, IN pCi/g
Phosphorus Product
Radionuclide (Solid)
Ra-226 0.021 ± 0.020
Po-210
U-234
U-235*
U-238
Th-230
Th-232
Pb-210 0.21 ± 0.17
Ra-228 <0.25
Ferrophosphorus Fluid Bed Prills
0.27
21
19
0.42
0.26
1.1
0.99
± 0.11 13 ± 0.65
<0.57 440 ± 27
± 5.2 <91
<0.65 (3.3)
± 4.9 <71
± 0.27 <24
± 0.20 <17
± 0.58 52 ± 1.8
± 0.94 1.9± 1.0
**
Slag
32 ±
<
25 ±
1.1 ±
25 ±
26 ±
0.59 ±
11 ±
0.96 ±
13
:16
6.7
0.58
7.0
n
0.29
7.9
0.46
*U-235 results reported in parenthesis are calculated based on the U-235 to U-238 natural activity
ratio of 1 to 21.45 (0.0466). Blanks indicate no data.
**Average result of sample analyses completed at EMSL except for lead-210 and polonium-210
(corrected) are average of EERF values only; taken from Appendix C, Table C-3.
-------
ENVIRONMENTAL SAMPLES
Grab water samples were obtained from the Portneuf River at
locations both upstream and downstream from the liquid waste
effluent discharge from both the Thermal Process and the Wet
Process Plants. Prior to analysis, the suspended solids were
allowed to "settle-out" and hence, both a liquid (soluble com-
ponent) and a suspended (insoluble component) fraction of each
water sample was processed. The water samples collected during
this study are non-potable, and human ingestion of such water is
improbable.
The radiochemical results for the Portneuf River samples are
shown in Table 5. For the liquid fraction, the results of the
upstream and downstream samples are essentially identical since
they are within the two-sigma counting error term for each
respective analysis. Similar analytical agreement was not
obtained from the results of the suspended fraction analyses.
Although the measured radium-226 content was lower in the down-
stream sample, the uranium and thorium contents were about an
order of magnitude greater in the downstream versus the upstream
sample. The higher radioactivity content in the suspended
fraction may indicate the insolubility of waste-product effluents,
It is emphasized that these results are based on analysis of only
one grab sample from both the upstream and the downstream flow.
Also, the measured radioactivity concentrations of both the
liquid and suspended fractions of this river water sample are
well within the typical background range for naturally-occurring
radioactivity (Holtzman, 1964). Any future efforts to resolve
whether there is any contribution of radioactivity to the stream
should include sediment sampling.
Table 6 presents the results of a grab sample of rainwater
runoff from the slag pile. As with the river water sample, both
a liquid and a suspended fraction of the rainwater runoff sample
20
-------
tx>
TABLE 5. PORTNEUF RIVER SAMPLES*
RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM. LIQUID FRACTION
IN UNITS OF pC1/l, AND SUSPENDED FRACTION IN UNITS OF pCi/9
Radionuclide
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
Ra-228
Upstream,
East Bank - Liquid
0.70 ±0.35
ND
1.2 ± 0.22
<0.031
0.55 ± 0.15
<0.062
<0.031
2.7 ± 1.3
<3.6
Downstream,
East Bank - Liquid
0.54 ± 0.33
ND
1.2 ± 0.20
<0.029
0.51 ± 0.11
<0.029
<0.029
<1.4
<4.0
Upstream,
East Bank-Suspended
3.6
0.067
<0
0.067
0.17
0.17
± 1.7
-
± 0.019
,0095
± 0.019
± 0.048
± 0.048
-
Downstream,
East Bank-Suspended
1.8 ± 1.3
<2.4
0.70 ± 0.20
<0.10
0.70 ± 0.20
1.6 ± 0.40
1.7 ± 0.40
6.1 ± 1.6
<20
*Grab sampled 5/8/75.
ND indicates non-detectable; blanks indicate no data.
-------
TABLE 6. SLAG PILE RAINWATER RUNOFF*
RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM
Radionuclide
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
Ra-228
Rainwater Runoff
Liquid (pCi/D
0.70 ± 0.36
0.83 ± 0.75
1.9 ± 0.25
0.062 ± 0.031
1.8 ± 0.25
0.12 ± 0.062
-------
were analyzed, with the liquid fraction having background concen-
trations of radioactivity. However, the suspended fraction had
elevated levels of radioactivity, perhaps also reflecting the
water insolubility of these radionuclides in the slag waste
material. That elevated levels of radioactivity were observed in
the suspended fraction of the rainwater runoff does indicate that
natural weathering factors, such as a normal rainstorm, may
create a pathway for increasing the contamination of the local
environs.
EFFLUENT SAMPLES
Cooling water samples from the No. 4 Furnace are shown in
Table 7. All results are within the normal fluctuation of back-
ground concentrations of the naturally-occurring radionuclides in
water.
Radiological analyses of water samples obtained from selected
settling and evaporation ponds are shown in Table 8. These
results are within normal fluctuations of background concentra-
tions as measured in the Portneuf River samples (Table 5).
Results of analyses of effluents from various facility
scrubber units are presented in Tables 9 and 10. Prior to
analysis, solids in the sample were allowed to settle-out and
hence, a liquid and a suspended fraction were analyzed. In every
case, the suspended fraction contained appreciably higher
quantities of radioactivity than did the liquid fraction,
indicating the insolubility of these components. The highest
radioactivity concentrations were measured in the samples from
the Calciner Scrubber effluents (Table 9). Medusa scrubber
samples also have slightly elevated radioactivity content compared
to normal background levels Cas reported for the Portneuf River
samples, Table 5). Table 11 presents the results for a few
miscellaneous samples.
23
-------
TABLE 8. POND WATER - SETTLING AND EVAPORATION*
RADIONUCLIDE CONCENTRATION * TWO-SIGMA COUNTING ERROR TERM, IN pCi/1
Radionucllde
Ra-226
Po-210
U-234
U-235
*" U-238
Th-230
Th-232
Pb-210
Ra-228
3-S Settling Pond
0.90 ± 0.37
NO
3.4 ± 1.8
<0.62
2.2 ± 1.2
1.0 ± 0.28
0.12 t 0.08
<3,6
4-S Settling Pond
0.33 t 0.30
66 ± 16
0.68 i 0.16
<0.04
0.40 ± 0.12
0.12 i 0.04
<0.04
110 ± 3.7
<3.5
7-S Settling Pond
1.3 ± 0.46
ND
0.65 ± 0.20
<0.05
0.45 ± 0.15
0.25 ± 0.10
<0.05
100 ± 5.8
<3.5
3-E Evaporation Pond
2.6 ± 0.62
160 i 12
1.2 ± 0.28
<0.04
0.80 ± 0.20
0.28 ± 0.16
<0.04
140 ± 3.4
<3.4
5-E Evaporation Pond**
0.60 i 0.23
170 ± 47
0.90 ± 0.25
<0.05
0.50 i 0.15
<0.05
<0.05
100 ± 14
iLS
6-E Evaporation Pond
<0.29
32 ± 10
0.68 ± 0.20
<0.04
0.48 ± 0.16
<0.04
<0.04
49 ± 2.3
<3.4
* Settling ponds grab sampled 5/7/7S; evaporation ponds trab sampled 5/S/75.
NO Indicates non-detectable; blanks ImMrate TO -tota.
** Grab sampled 8/13/75.
-------
IN)
tn
TABLE 9. CALCINER SCRUBBER EFFLUENTS*
RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM. LIQUID FRACTION
IN pCi/1, AND SUSPENDED FRACTION IN pCi/g
Radionuclide
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
Ra-228
Calciner Scrubber Calciner Scrubber
Pond Discharge Pond Discharge
(Liquid) (Suspended)
7.8 ±
ND
26 ±
1.0 ±
25 ±
21 ±
0.17 ±
13 ±
.3.1
1.0 31 ± 1.2
-
1.6
0.14
1.6
0.89
0.086
1.2
1.6 ± 1.1
No. 2 Calciner
Scrubber
Overflow (Liquid)
4.6
21
13
0.43
12
0.23
<
2.4
± 0.81
± 2.2
± 0.95
± 0.10
± 0.88
± 0.10
0.025
± 0.34
c3.5
No. 2 Calciner
Scrubber
Overflow (Suspended)
13 ± 0.93
-
4.9 ± 0.40
0.18 ± 0.044
5.0 ± 0.41
14 ± 0.56
0.28 ± 0.078
-
.1.8
*Grab sampled on 5/7/75.
ND indicates non-detectable; blanks indicate no data.
-------
TABLE 10. SCRUBBER EFFLUENTS (Sampled 5/7/751
RADIONUCLIDE CONCENTRATION t TWO-SIGMA COUNTING ERROR TERM.* LIQUID FRACTION IN pC1/l, and SUSPENDED FRACTION IN pC1/g
Radionucllde
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
Ra-228
Medusa Medusa
Fluid Bed- Fluid Bed- Scrubber Liquor- Scrubber Liquor-
Scrubber Scrubber No. 4-E, tapping No. 4-E, tapping
Liquor (Liquid) Liquor (Suspended) (Liquid) (Suspended)
2.3 ± 0.56 8.3 ± 0.53 2.5 ± 0.59 9.2 ± 7.7
NO NO ND
0.72 t 0.20 0.87 ± 0.32 2.2 ± 0.23
<0.04 <0.1l 0.057 ± 0.029
0.44 t 0.12 0.83 t 0.35 1.4 ± 0.17
<0.04 2.4 t 0.18 0.54 ± 0.11
<0.04 0.11 t 0.04 <0.029
46 ± 1.9 1900 ± 19 210 t 3.2
<3.7 <0.88 6.5 ± 3.8 <67
Medusa
Scrubber Liquor-
No. 4-W, not tapping
(Liquid)
1.
11
1.
0.
0.
1.
4.
1 ± 0.44
± 1.4
1 ± 0.14
<0.029
51 ± 0.086
23 t 0.086
<0.029
6 t 0.34
1 t 3.5
Medusa
Scrubber Liquor-
No. 4-W, not tapping
(Suspended)
24 ± 10
35 ± 7.4
0.60 ± 0.20
<0.10
0.50 * 0.20
0.60 ± 0.30
•cO.10
31 ± 3.3
<170
*NO Indicates non-detectable; blanks Indicate no data.
-------
TABLE 11. MISCELLANEOUS SAMPLES (Sampled 5/8/75)
RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM, IN pCi/1
Fluid Bed Recycle - 10,000 Gallon Sump
Furnace Slurry PHOS Water from
Radionuclide (Liquid) Furnace (Liquid)
Ra-226 9.8 ± 0.82 1.3 ± 0.30
Ra-228 2.1 ± 1.1 <1.8
Pb-210 <2.1 43 ± 2.5
27
-------
AIRBORNE PARTICULATE RADIOACTIVITY
Background Airborne Radioactivity Concentrations
Background airborne radioactivity concentrations were not
specifically obtained for the Pocatello area. However, as will
be discussed shortly, the airborne concentrations measured in the
Technical Building Library may be considered as representative of
background conditions for the purposes of this report. Other
authors have reported background airborne levels. Poet et al.
(1972) reported the average polonium-210 and lead-210 concentra-
tions in ground level air to be 0.001 and 0.01 pCi/m 3, respec-
tively. NCRP (November, 1975) summarizes reports by geographic
areas of the normal polonium-210 and lead-210 air concentrations.
In Colorado, the reported polonium-210 concentrations ranged from
0.00005 to 0.003 pCi/m3, whereas the lead-210 ranged from 0.001
to 0.021 pCi/m3. The highest reported lead-210 concentrations
were 0.026 pCi/m3 for several locations in Illinois and Utah.
NCRP (November, 1975) also discusses the background level of
airborne concentrations of the other naturally-occurring radio-
nuclides. Natural uranium (uranium-234 and -238) concentrations
18 — fi
ranged from 120 aCi/m3 (one attocurie is 10" curie or 10 pCi)
near Chicago, Illinois, to a reported high of about 400 aCi/m3
for several sites in New York State. Typical radium-226 concen-
trations are usually less than 100 aCi/m3. Sedlet et al. (1973)
reported thorium-232 and thorium-230 concentrations of 30 and 45
aCi/m3, respectively, near Chicago, Illinois.
In-Plant Air Sampler
Air samples were obtained in the in-plant areas using a
portable air sampling unit. This unit has a carbon vane vacuum
pump with a regulator which permits constant air flow sampling.
Usually, sample collections were made at two cubic feet per
*—Regulated Air Sampler, Model RAS-1, Eberline Instrument
Corporation, Santa Fe, New Mexico
28
-------
minute (CFM) using a 47-millimeter diameter, Type E glass fiber
filter. This corresponds to a linear flow rate of about 106 feet
per minute. The air sample dust load was determined by measuring
the mass of material collected on each filter.
Airborne particulate sampling was conducted in several
working areas within the Thermal Process Plant using the portable
air sampler. The radioactivity concentrations of the naturally-
occurring radionuclides are usually expressed as picocuries per
cubic meter of sampled air (pCi/m3). Dust loading of the air
filter samples was also determined and the specific activity of
the dust, expressed as pCi/g, is also given. The solubility of
airborne particulate matter was not determined in this study.
Gross Versus Net Results
Eadie and Bernhardt (1976) have reported that radiochemical
analyses of blank glass fiber filters (4-inch diameter) indicate
appreciable quantities of naturally-occurring radioactivity.
Comparison of these results with those of other facilities
indicates that the reported activity is not only due to the
composition of the filter media itself, but also due to contami-
nants in reagents, glassware, the general background level of
contaminants, etc.
Table 12 presents data on the radioactivity content of blank
glass fiber filters (4-inch diameter) and the extrapolated
activity content for the 47-mm diameter filters which were used
in this study. In order to account for this radioactivity
content associated with blank filter analyses, the appropriate
blank filter activity (Table 12) has been subtracted from the
measured gross analytical result to obtain a "net" result. No
blank subtractions have been made for the three radionuclides
(radium-228, polonium-210, and lead-210) which are at the analyt-
ical minimum detectable activity (MDA) levels. Gross analytical
results for all of the in-plant air samples are shown in Appendix
B.
29
-------
TABLE 12. RADIOACTIVITY CONTENT OF BLANK GLASS FIBER FILTERS
(pCi per filter)*
* **
Radionuclide 4-inch Diameter 47-mm Diameter
Ra-2260.35 ± 0.09 0.07 ±0.019
Po-210 <0.17 <0.036
U-234 0.10 ± 0.03 0.021 ±0.0064
U-235*** (0.0035 ± 0.0010) (0.00075 ±0.00021)
U-238 0.08 ± 0.02 0.017 ±0.0043
Th-230 0.20 ± 0.08 0.043 ±0.017
Th-232 0.13 ± 0.02 0.028 ±0.0043
Pb-210 <0.32 <0.068
Ra-228 <1.6 <0.34
+ Average pCi per filter with standard error term about this mean based on
the t-distribution at the 95 percent confidence level.
* Taken from Eadie and Bernhardt, 1976.
** Extrapolated value based on the area ratio between 47-mm and 4-inch
diameter filters of 0.214 times the 4-inch diameter activity content.
Average 47-mm filter mass ± two standard deviations was 0.1249 ± 0.0013
grams.
***Calculated U-235 content based on U-235 to U-238 natural activity ratio of
1:21.45 (0.0466).
In-Plant Air Sampling Net Results
Tables 13 to 17 present the net radionuclide concentrations
obtained for the various in-plant sampling locations. As previ-
ously noted, the lead-210 results are underestimated by up to a
factor of five, which results in a potential overestimate of
about a factor of ten or more for the polonium-210 results. Data
which narrows the uncertainty for specific samples are given in
Appendix E.
The highest airborne radioactivity was measured in the area
of the south side discharge of the Calciner (Table 13). There,
the measured airborne concentrations are orders of magnitude
greater than results obtained in the Technical Building Library
(Table 17). For example, because of the higher airborne dust
levels, the radium-226 and total uranium airborne concentrations
(i.e. ,'pCi/m3) measured in the Calciner were 3800 and 1600 times
the respective levels measured in the Technical Building Library.
Non-detectable levels of polonium-210 were measured in the Technical
Building Library (Table 17) compared to the highest value of 66
pCi/m3 measured in the area of the Calciner (Table 13).
30
-------
TABLE 13. AIRBORNE PARTICULATE SAMPLING IN THE CALCINERAREA
NET RADIONUCLIDE CONCENTRATION ± TWO-SIGMA ERROR TERM
Radionuclide
Ra-226
Po-210+
U-234
U-235***
U-238
Th-230
Th-232
Pb-210+
Ra-228+
South Side Discharge,*
Calciner Belt #1 Dump Area -
1218 to 1500 hrs. (5/5/75)
pCi/m3 pCi/g
2.7 ±
66 ±
1.5 ±
0.044
1.5 ±
2.8 ±
0.051
0.40 ±
0.18
4.9
0.12
± 0.014
0.12
0.15
± 0.022
0.28
23
38 ± 2.4
560 ± 70
20 ± 1.7
0.62 ± 0.20
21 ± 1.7
39 ± 2.1
0.72 ± 0.30
120 ± 9.5
<3.2
No. 1 Control Room**
1027 to 1520 hrs. (5/5/75)
pCi/m3 pCi/g
0.015 ± 0.010
<0.0078
0.0079 ± 0.0023
(<0. 00028)
0.0058 ± 0.0023
0.0067 ± 0.0046
0.0029 ± 0.0023
0.17 ± 0.066
<0.085
19 ± 13
<88
10 ± 3.1
(<0.37)
7.9 ± 3.
8.7 ± 6.
4.0 ± 3.
210 ± 83
<120
1
2
1
*Sampled volume of 7.9m , with a dust particulate load of 0.5663g.
*Sauipled volume of 17.6m3, with a dust particulate load of 0.0131g.
*-
**
***U-235 results reported in parenthesis are calculated based*on U-235 to U-238 natural activity ratio
of 1 to 21.45 (0.0466).
-------
TABLE 14. AIRBORNE PARTICULATE SAMPLING IN THE PROPORTIONING BUILDING
NET RADIONUCLIDE CONCENTRATION ± TVO-SIGMA ERROR TERM
to
Radionuclide
Ra-226
Po-210e
U-234
U-235
U-238
Th-230
Th-232
Pb-2100
Ra-228"
Lower Level, 1st Pillar*
West of Shack,
0947 to,1309 hrs. (5/6/75)
PCIV DCi/g
0.37
5.3
0.30
i 0.045
15 ± 1.8
i 0.89 210 i 36
t 0.036
0.014 ± 0.0069
0.32
0.72
± 0.038
± 0.062
0.035 t 0.014
0.63
<0.
'Sampled volume of 12.4m'.
"Sampled volume of 12.5m'.
^Sampled volume of 4.44m'
^Sampled volume of I4.7n',
'Gross analytical result.
t 0.11
15
12 i 1.5
0.54 ± 0.27
13 ± 1.5
29 t 2.5
1.4 t 0.58
25 + 4.5
<6.0
with a dust partlculate
with a dust partlculate
with a dust partlculate
with a dust partlculate
Lower Level Control Room**
0944 to 1316 hrs. (5/6/75)
pC1/m' Dd/a
0.12 ± 0.027
<2.7
0.095 ± 0.017
0.0044 ± 0.0039
0.
0.
<0.
0.
11 t 0.019
21 i 0.034
00056
24 ± 0.097
<0.12
16 ± 3.5
350 i 71
13 ± 2.2
0.58 ± 0.51
15 t 2.4
27 ± 4.3
<0.064
30 ± 12
<15
Upper Level , Top West
1140 to 1315 hrs. (5/6/75)
pCt/m3 pC1/q
2
<3
2
0
2
4
.9 ± 0.21 16 ± 1.2
.4 <21
.8 i 0.23 16 ± 1.3
.13 ± 0.027 0.71 ± 0.15
.8 t 0.23 16 ± 1.3
.9 i 0.25 27 ± 1.4
0.11 t 0.036 0.62 ± 0.20
6.6 ± 0.48 37 ± 2.7
<0.39 <2.2
Upper Level Control Room**
0930 to 1411 hrs. (5/6/75)
pC1/m3 pCi/g
0.48 ±
3.8 ±
0.29 ±
0.012 ±
0.32"±
0.50 ±
0.013 t
0.74 ±
<0.1
0.054
0.94
0.034
0.0058
0.035
0.048
0.0083
0.10
1
30 ± 3.4
240 ± 60
18 ± 2.1
0.73 * 0.35
19 t 2.2
31 ± 3.0
0.79 ± 0.51
45 i 6.5
<7.0
load of 0.3109g.
load of 0.0978g,
load of 0.7853g.
load of 0.2387g.
-------
TABLE 15. AIRBORNE PARTICULATE SAMPLING IN THE PELLETIZER BUILDING
NET RADIONUCLIOE CONCENTRATION ± TMO-SIGMA ERROR TERM
Radtonucllde
Operator Level , Middle** Upper Level , Top East*
Lower Level. 1000 to 1537 hrs. (5/5/75) 1003 to 1530 hrs. (5/5/75)
pC1/n3
Ra-226
Po-210*
U-234
U-235
U-238
Th-230
Th-232
Pb-210*
Ra-228*
0.28
1.1
0.45
0.023
0.24
0.50
0.019
0.66
<'
± 0.033
± 0.045
i 0.042
t 0.0096
± 0.03
i 0.044
± 0.0089
t 0.10
0.11
pCVg
20 ±
82 ±
32 ±
1.6 ±
17 ±
35 i
1.2 +
46 ±
<7.7
2.3
31
2.9
0.68
2.1
3.1
0.63
7.1
pC1/m'
0.40
1.6
0.26
0.014
0.25
0.53
0.023
0.60
-------
TABLE 16. AIRBORNE PARTICULATE SAMPLING IN THE FURNACE BUILDING
NET RADIONUCLIDE CONCENTRATION t TWO-SIGMA ERROR TERM
Radionucllde
Control Roon Level *
Between Furnace 344
0906 to 1606 hrs. (5/7/75)
pCI/*3
Ra-226
PO-210++
U-234
U-235 +++
u-238
Th-230
Th-232
Pb-210"
Ra-228++
0.
044
0.63
0.
010
±
±
i
0.012
0.23
0.0041
<0. 00094
0.
0.
0.
0.
010
018
0054
27
t
i
t
±
0.0041
0.0079
0.0047
0.054
<0.069
Tapping Level**
Between Furnace 3 S 4
0910 to 1607 hrs. (5/7/75)
pCI/g pC1/n3
16 ±
229 ±
3.5 ±
4.4
84
1.5
<0.34
3.5 ±
6.4 i
2.0 ±
98 t
<25
1.5
2.8
1.7
20
0.011 ± 0.0063
<0.048
0.0086 i 0.0046
(<0. 00048)
0.010 ± 0.0052
0.0071 ± 0.0053
<0.0030
0.25 ± 0.055
<0.067
pCI/g
6.8 ± 4.0
<69
5.6 ± 3.0
(<0.32)
6.7 ± 3.4
4.2 * 3.4
<1.9
150 ± 34
<44
Control Room for +
Furnace 3 S 4, Behind Panel
0853 to 0851 hrs. (5/7 to 8/75)
PC1/H.3
0.0082 ± 0.0030
0.16 ± 0.057
0.0022 i 0.0012
(<0. 00019)
0.0040 ± 0.0016
0.0031 ± 0.0017
<0. 00056
0.027 ± 0.013
<0.019
pCi/g
11 ± 4.
0
210 ± 79
2.9 ± 1.
6
(<0.25)
5.2 t 2.
4.1 t 2.
<0.75
36 ± 18
<25
1
6
*S«npled volune of 24.3mJ. with a dust partlculate load of 0.0678g.
••Sampled volume of 23.4m3, with a dust partlculate load of 0.0370g.
+Sampled volume of 84ra3. with « dust partlculate load of 0.0623g.
+-H3ross analytical result .
•M-HJ-235 results reported In parenthesis are calculated based on U-Z35 to U-238 natural activity ratio of 1 to 21.45 (0.0466).
-------
TABLE 17. AIRBORNE PARTICUUVTE SAMPLING
NET RADIONUCLIDE CONCENTRATION ± TWO-SIGMA ERROR TERM
Radionuclide
Technical Bulldinq* Condenser and Fluid Bed**
Library- Control Room-
0935 (5/5/75) to 1421 hrs.(5/6/75) 0904 to 1444 hrs.(5/6/75)
Old Kiln Building (Maintenance)*
Open Area - East End
0912 to 1412 hrs.(5/6/75)
pCi/m;
pCi/q
pCi/m-
pCi/g
pCi/m:
pCi/g
P-4 Loading Area++
0933 to 1438 hrs.(5/8/75)
pCi/m3
tn
Ra-226 <0.00072 <12
Po-210° NO NO
U-234 0.00077 ± 0.00063 12 ± 9.9
U-23500 (<0.000050) (<0.84)
U-238 0.0011 ± 0.00063 18 ± 9.9
Th-230 0.0016 ± 0.00065 25 ±10
Th-232 <0.000095 <1.4
Pb-210" 0.019 ± 0.0086 320 ± 140
Ra-228° <0.011 <190
0.0084 ± 0.0081 6.7 ± 6.4
0.29 ± 0.25 250 ± 210
0.014 ± 0.0038 12 ± 3.1
(
-------
Although the library had the lowest airborne radioactivity
measured in this study, even these levels were slightly elevated
compared to reported background levels reported by other authors,
but for the purposes of this study the results measured in the
Technical Building Library (Table 17) shall be considered as
background levels.
Specific Activity of Airborne Particulate Matter
The specific activity of airborne particulate matter
(reported in units of pCi/g) was also determined for the portable
air sampler filters. These results also show the disequilibrium
between the individual radionuclides of the uranium-238 decay
series, the most noticeable being the polonium-210 and lead-
210 contents compared to the other decay chain members. (Thorium-
232 and radium-228 are members of a different decay series.)
The highest polonium-210 activity was 3900 ± 430 pCi/g in
the Old Kiln Building (Table 17) with several sampling locations
having essentially a non-detectable polonium-210 level (Tables 15
and 17). The highest radium-226 activity of 38 ± 2.4 pCi/g was
obtained in the Calciner (Table 13) and the lowest activity was
6.7 ± 6.4 pCi/g in the Condenser and Fluid Bed Control Room
(Table 17).
These specific activity determinations of the airborne
particulate matter may also be compared to the results of analyses
of the input products to the thermal process (Table 3). The
activity of 22 pCi/g in phosphate rock was the highest concen-
tration measured in any of the input product samples for
polonium-210, yet specific activity determinations of the airborne
particulate samples were at least a factor of three greater than
this - except for the 21 pCi/g polonium-210 measured in the Upper
Level, Top West of the Proportioning Building. The specific
activity concentrations for the other radionuclides in the input
products are generally comparable to the airborne particulate
sample results.
36
-------
PARTICLE SIZE CHARACTERIZATION
Cascade Impactor Sampling
High volume cascade impactors were used to measure the size
distribution of airborne particulate matter for both the indoor
and outdoor environments. The impactor filter stages attach to a
high volume air sampler which is electronically controlled***
to operate at a flow rate of 40 cubic feet per minute (CFM).
This corresponds to a linear air velocity of about 72 feet per
minute through the final 8 x 10-inch filter stage.
Whatman #41 paper filters were used for each impaction
stage. Glass fiber (Type E) was used for the final 8 x 10-inch
filter. Typical particle size ranges for each filter stage, as
reported by the manufacturer, are:
Equivalent Aerodynamic Diameter at 50
Stage No. Percent Collection Efficiency (micrometer)
1 Greater than 7.2 ym
2 3.0 - 7.2
3 1.5 - 3.0
4 0.95- 1.5
5 0.49- 0.95
Final Filter Less than 0.49
"Equivalent aerodynamic diameter" is defined as the size of
a spherical particle of unit density which has the same terminal
settling velocity as the sampled particle. Radioactivity analysis
*Model 252 Series, six stage, cascade impactor from Tech Ecology, Inc.
(Now produced and marketed by Sierra Instruments, Inc.)
** Model GMWL-2000, High Volume Air Sampling System from General Metal
Works,Inc.
***Model 310/310A, High Volume Constant Flow Controller from Sierra
Instruments,Inc.
37
-------
of each impactor stage provides an indication of the activity
content for the various particle size ranges. The mass of dust
particulate loading (in grams) for each filter stage was not
determined for this study.
Gross Results of Impactor Sampling
The cascade impactor sampler was used at four locations in
the Thermal Process Plant - the Calciner, Proportioning, Pelleti-
zer, and Furnace Buildings. Gross analytical results for these
samples are presented in Tables 18 to 21. Tables 22 and 23 show
the results of radiochemical analyses of blank impactor filters -
the slotted Whatman #41 paper and the 8 x 10-inch glass fiber
filters, respectively. Since most of the blank slotted paper
filters contained radioactivity concentrations at the minimum
detectable activity (MDA) levels for the analysis, a blank filter
subtraction has not been performed for all the radionuclides for
the impactor samples. As noted in subsequent discussions, tabula-
tions of net results have been used in some of the data plots.
The previously noted reservations concerning the lead-210 and
polonium-210 data also apply to these results (i.e., lead-210 may
be low by a factor of 5 or more and polonium-210 high by a factor
of 10). Data for narrowing the uncertainty on specific samples
are given in Appendix E.
Table 24 provides a summary of the gross results of the
cascade impactor sampling for the four sampling locations. The
gross results for the Calciner (Table 18) indicate that the
majority of activity (roughly 53 percent of the total arithmetic
activity collected) was of the particle size fraction less than
0.5 ym equivalent aerodynamic diameter (i.e., the activity
collected on the final filter stage). Gross results for the
other locations also indicate that a large fraction of the
activity is in the sub-micron particle size range.
Comparison of Air Sampling Results
The activity summation for all filters of the cascade
impactor may be compared to the activity concentrations determined
38
-------
to
TABLE 18. CASCADE IHPACTOR SAMPLING IN THE CALCINER AREA
GROSS RAOIOHUCL1DE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERN. IN pC1/m3.
Radlonucllde
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
Ra-228
Filter I
0.18 ±0.02
11 ± 1.1
0.18 ± 0.018
0.011 ± 0.0035
0.17 ±0.017
0.32 ± 0.0013
0.0077 ± 0.0045
0.57 ± 0.057
<0.058
Percentage of 282
Total Actlvltv
Filter 2
0.079 ± 0.015
3.0 ± Q.39
0.041 ± 0.0074
<0.0011
0.039 ± 0.0072
0.07 ± 0.013
0.007 ± 0.0042
0.28 ± 0.056
<0.058
8.056
Filter 3
0.043 ± 0.01
1.6 t Q.26
0.022 ± 0.0053
<0. 00077
0.018 ± 0.0047
0.077 ± 0.013
0.0037 ± 0.0028
0.14 ± 0.041
<0.058
4.31!
Filter 4
0.042 ± 0.0097
1.1 ±0.24
0.056 ±0.01
<0.0026
0.046 ± 0.01
0.082 ±0.015
<0.0026
0.13 t 0.038
<0.059
3.4S
Filter 5
0.051 ± 0.011
1.6 ±0.25
0.046 ± 0.0083
<0.0016
0.019 ± 0.0054
0.049 ± 0.011
<0.0032
0.065 ± 0.036
<0.06
4.2X
Activity Summation
Final Filter For All Filters
0.38 ± 0.028
21 ±1.7
0.38 ± 0.03
0.0099 ± 0.0033
0.38 ± 0.03
0.70 ± 0.038
0.019 ± 0.0064
1.1 ± 0.071
<0.059
53%
0.78
39
0.73
<0.027
0.67
1.30
<0.043
2.3
<0.35
—
*Sanple collected 5/5/75, 1347 to 1415 hours, total volume sampled of 31.3 nt3. Sampled at South Side Discharge, Calclner Belt II Dump Area.
-------
TABLE 19. CASCADE IMPACTOR SAMPLING IN THE PROPORTIONING BUILDING*
Radlonuclide Filter 1
Ra-226 , 0.24 ±0.032
Po-210 0.57 ± 0.36
U-234 0.17 ±0.02
U-235 0.005 ± 0.0025
U-238 0.15 ± 0.018
Th-230 0.25 ± 0.028
Th-232 0.01 ± 0.005
Pb-210 0.33 ±0.082
Ra-228 <0.12
Percentage Of 24%
Total Activity
Filter 2
0.036 ± 0.012
<0.46
0.055 ± 0.013
0.0039 ± 0.0037
0.033 ± 0.0098
0.087 ± 0.02
<0.0029
0.18 ± 0.076
<0.12
13%
Filter 3
0.041 ± 0.014
<0,11
0.018 ± 0.0075
<0.0025
0.018 ± 0.0075
0.015 ± 0.0075
<0.0025
0.11 ± 0.073
<0.12
5.8%
Filter 4
0.053 ± 0.016
<1.2
0.028 ± 0.0086
<0.0032
0.019 ± 0.0071
0.022 ± 0.014
0.015 ± 0.011
0.14 ± 0.075
<0.12
21%
Filter 5
0.054 ± 0.011
<0.27
0.018 ± 0.0075
<0.0025
0.015 ± 0.0075
0.023 ± 0.0075
<0.0025
0.11 ± 0.074
<0.12
8.1%
Activity Summation
Final Filter For All Filters
0.20 ± 0.03
0.65 ± 0.37
0.15 ± 0.02
<0.0044
0.16 ± 0.021
0.28 ± 0.035
0.011 ± 0.0078
0.49 ± 0.086
<0.10
27%
0.62
3.3
0.44
<0.022
0.40
0.68
<0.044
1.4
<0.70
•Sample collected 5/6/75. 1347 to 1401 hours, total volume sampled of 15.9m3. Sampled at lower level. 1st pillar west of shack.
-------
TABLE 20. CASCADE IMPACTOR SAMPLING IN THE PELLETIZER BUILDING*
GSOSS RADIONUCLIDE CONCENTRATION t THO-SIGHA COUNTING ERROR. IN pCI/m'.
Radlonucllde
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
Ra-228
Percentage Of
Total Activity
Filter 1
0.013 ± 0.005
NO
0.016 t 0.0037
<0.00091
0.019 ± 0.004
0.026 ± 0.0062
<0.0014
0.073 ± 0.027
<0.04
15%
Filter 2
0.016
0.0051
0.0098
<0
0.0076
0.012
<0
0.025
<0
9
i 0.005
± 0.00085
± 0.0032
.001
± 0.0026
± 0.0044
.0014
± 0.017
.04
.6*
Filter 3
0.013 ± 0.005
ND
0.0098 ± 0.003
<0. 00078
<0. 00055
0.020 ± 0.0054
<0.0016
0.058 ± 0.026
<0.039
12*
Filter 4
0.011 ± 0.0041
O.070
0.0067 ± 0.0028
<0.0015
0.0085 ± 0.0032
0.017 ± 0.0051
<0.002
<0.018
<0.04
141
Filter 5
0.0088 t 0.0039
ND
0.0089 i 0.0035
<0.0018
0.011 ± 0.0035
0.011 ± 0.0053
<0.0018
0.049 t 0.027
<0.04
11%
Activity Summation
Final Filter For All Filters
0.034
0.14
0.045
0.0026
0.046
0.071
0.0027
0.088
±
±
±
±
±
t
±
±
CO,
31
0.0074
0.095
0.0069
0.0016
0.007
0.010
0.00068
0.026
.041
wt
JJ9
0
0,
0
<0
0
0
<0
0
<0
.095
.22
.096
.0086
.093
.16
.011
.31
.24
•Sample collected 5/5/75, 1102 to 1141 hours, total volume sampled of 45.5 m3. Sampled at Operator Level, middle of the Peiietlzer
Lower Level. ND Indicates non-detectable.
-------
TABLE 21. CASCADE IMPACTOR SAMPLING IN THE FURNACE BUILDING*
GROSS RADIONUCLIDE CONCENTRATION * TMO-SIGMA COUNTING ERROR TERM. IN pCI/n.3.
Radlonucllde
Filter 1
Filter 2
Filter 3
Filter 4
Filter 5
Final Flltc
Activity Sunraatlon
For All Filters
Is)
Ra-226 0.0025 ± 0.0014 0.0045 ± 0.0018
Po-210 M> NO .
U-234 0.0029 ± 0.0011 0.0024 ± 0.0011
U-235* (0.000065) (0.000047)
U-238 0.0014 t 0.00081 <0.001
Th-230 0.0037 ± 0.0015 0.0036 ± 0.0016
Th-232 0.00073 t 0.00066 0.00096 ± 0.00087
Pb-210 0.03 ±0.011 0.034 ±0.011
Ra-228 <0.016 <0.016
Percentage Of 9.2X lot
0.0019 ± 0.0014 0.0058 ± 0.002 0.0022 ± 0.0013
<0.0033 «0.031 <0.048
0.0013 ± 0.00073 0.0019 ± 0.00087 0.0013 ± 0.00072
<0.00003) (0.000051) (0.000084)
0.00064 ± 0.00055 0.0011 ± 0.00065 0.0018 ± 0.00082
0.0026 ± 0.0015 0.0025 t 0.0015 <0.001
<0.00041 <0.0011 <0.00084
0.027 ± 0.011 0.039 ± 0.012 0.064 1 0.013
<0.016 <0.016 <0.016
8.5X 16X 22X
0.0073 ± 0.0023
0.1? t 0.055
0.0026 ± 0.00099
(0.00013)
0.0027 i 0.001
0.0052 ± 0.002
<0.0014
0.061 ± 0.013
<0.016
35X
0.024
<0.20
0.012
(0.00041)
<0.0086
<0.019
<0.0054
0.26
<0.096
•Swole collected 5/7/75. 1330 to 1505 hours, total
slag slough. U-235 results reported In parenthesis
MU indicates non-detectable.
volume sanoled of 108mJ. Smoled at Control Room level
are calculated based on U-235 to U-238 natural activity
. between furnace 3 and 4 above
ratio of 1 to 21.45 (0.0466).
-------
TABLE 22. RADIOACTIVITY CONTENT OF BLANK IMPACTOR WHATMAN *41 PAPER FILTER (SLOTTED)
(pCI ± THO-SIGHA COUNTING ERROR TERM PER FILTER)
Filter
1
2
3
4
5
Average of
All Results**
Ra-226 Po-210 U-234
<0.13 <5.1 <0.36
<3.1 <0.95
0.23 ± 0.17 12 ± 6.4 1.2 ±0.85
0.37 ± 0.17 7.8 ± 4.2 <0.67
0.17 ± 0.14 10 ± 5.0 0.70 ±0.62
<0.23 <7.6 <0.78
U-235* U-238 Th-230 Th-232
<0.014 <0.31
<0.023 <0.50 <0.068 <0.068
<0.023 <0.50 <0.070 <0.034
<0.030 <0.64 <0.045 <0.034
<0.033 <0.70 <0.062 <0.048
<0.025 <0.53 <0.061 <0.046
Pb-210 Ra-228
<1.3 <1.7
<1.2 <2.0
<1.2 <2.2
<1.1 <2.0
<1.1 <2.0
<1.2 <2.0
*U-235 calculated based on natural U-235 to U-238 activity ratio of 1 to 21.45 (0.0466). Blanks indicate no data.
"All average results are reported as less than valjes since most of the analytical results of the Individual filter are less than values.
-------
TABLE 23. RADIOACTIVITY CONTENT OF BLANK GLASS FIBER FILTERS (8 BY 10-INCH)
(pC1 ± THO-SIGMA COUNTING ERROR TERM PER FILTER)
Filter
1
2
3
4
5
6
Average of
All Results**
Ra-226
0.21
0.15
0.57
0.32
0.17
0.23
0.28
* 0.08
i 0.07
t 0.13
± 0.10
t 0.07
t 0.08
t 0.16
Po-210
0.64 t 0.38
<0.53
LOST
<0.54
<0.58
<0.58
<0.57
U-234
0.29
0.34
0.35
0.36
0.25
0.30
0.32
± 0.10
± 0.12
± 0.16
± 0.13
± 0.11
± 0.13
± 0.04
U-235*
0.012 ±
0.014 ±
0.018 t
0.017 ±
0.0079 ±
0.0075 ±
0.013 ±
0.0045
0.0051
0.0084
0.0061
0.0041
0.0061
0.0044
U-238
0.26
0.31
0.38
0.36
0.17
0.16
0.27
± 0.096
± 0.11
± 0.18
± 0.13
± 0.088
± 0.13
± 0.094
Th-230
0.33
0.33
0.44
0.40
0.32
0.22
0.34
± 0.11
± 0.13
t 0.18
± 0.14
± 0.13
± 0.11
± 0.076
Th-232
0.14
0.15
0.16
0.18
0.18
0.13
0.16
± 0.072
± 0.08
± 0.15
i 0.088
± 0.096
± 0.072
± 0.021
•U-235 calculated based on natural U-235 to U-238 activity ratio of 1 to 21.45 (0.0466).
••Average of all results with standard error about this mean based 1n the t-d1str1but1on at the 95 percent confidence level.
-------
TABLE 24. SUMMARY OF GROSS RESULTS OF THE CASCADE IMPACTOR SAMPLING
Percentage of Total Radioactivity per Particle Size Range
Location
Calciner
Proportioning
Building
Pelletizer
Building
Furnace Building
Arithmetic Average
Percentage
Greater than
7.2 urn
28
24
15
9.2
19
3.0 to 7.2
vim
7.9
13
9.6
10
10
1.5 to 3.0
u m
4.3
5.8
12
8.5
7.7
0.95 to 1.5
vim
3.4
21
14
16
14
0.49 to 0.95 Less than
vim
4.2
8.1
11
22
11
0.49 vim
53
27
38
35
38
-------
from the portable air sampler for the same sampling location.
Good agreement has been obtained between the results of the
radioactivity determinations using these two different sampling
systems, even though the samples were not necessarily obtained
during simultaneous time periods.
For the Calciner, impactor results (Table 13) are roughly
one-half of the activity as measured using the portable air
sampler (first column of Table 13). In both cases, the results
indicate that polonium-210, and to a lesser extent lead-210
activities are present in concentrations in excess of the normal
equilibrium concentrations of the other decay chain members. Of
all the sampling locations, the highest airborne radioactivity
measurements were obtained in the Calciner, regardless of the air
sampling system used (this conclusion recognizes the previously
denoted uncertainties for the lead-210 and polonium-210 results).
In the Proportioning Building, the cascade impactor results
(Table 19) are within a factor of two of the individual radio-
nuclide concentrations measured using the portable air sampler
(first column of Table 14). Except for the elevated polonium-210
activity, the decay chain products appear to be in equilibrium.
Cascade impactor results in the Pelletizer Building (Table 20)
are roughly one-third of the activity measured using the portable
air sampler (first column Table 15).
Airborne radioactivity measurements within the Furnace
Building showed the best agreement between the two different air
sampling system's results (Table 21 and first column of Table
16). Results from either sampling system showed fairly constant
equilibrium concentrations within the decay chain series except
for polonium-210 content, which appears to be elevated by up to
an order of magnitude compared to the other decay chain members.
This elevationis not solely relatable to the previously mentioned
uncertainty in the analyses.
In summary, both the cascade impactor and the portable air
sampler results show elevated airborne radioactivity concentra-
tions, the most notable being polonium-210 and lead-210 for
46
-------
general plant working areas as compared to the airborne concen-
trations measured in the Technical Building Library. The lowest
airborne radioactivity concentration measured in a working area
of the Thermal Process Plant was obtained in the Furnace Building,
The highest airborne activity was measured in the Calciner.
Particle Size Analysis
Figures 2, 3, and 4 present log-normal probability plots of
data from the cascade impactor sampling. Figures 2 and 3 present
the data from the Calciner and Proportioning Building. The plots
are based on the net concentrations (gross minus activity in
blank filters) of radium-226 and polonium-210. The gross concen-
trations for thorium and uranium were used in Figure 2 because
the blank concentrations were below the detection limits. The
gross concentrations were high enough that the blank concentra-
tions made a negligible contribution. Only radium-226 concen-
trations were used for Figures 3 and 4 because the contribution
from the other radionuclides in all of the stages was not suffi-
ciently above the concentrations in the blank filters to provide
meaningful results.
The plotted points represent the percent of the collected
activity (calculated in terms of pCi/m3) less than the indicated
sizes. The actual points were calculated by allocating all the
activity in the stages and final filter subsequent to a stage to
the specified size cut off for the stage (Elder et al., 1974).
The manufacturer specified size cutoffs for the impactor at 40
cfm are 7.2, 3.0, 1.5, 0.95, and 0.49 micrometers (equivalent
aerodynamic diameter). These are partially based on a study by
Willeke (1975).
Thus, in the actual calculations, the percent of the total
activity on the 3.0, 1.5, 0.95, and 0.49 and final filter stages
was designated as less than 7.2 micrometers. The percent of "the
total activity on the final filter was ascribed to the 0.49
47
-------
10-
7—
5 —
3 —
2-
I «•
UJ
N
55 1-
UJ —I
oc _
2
.2-
0.1
\
2
Ra-226(Net)
Po-210(Net)
U-234/ 238 (Gross)
Th-230 (Gross)
\
5
10
\
20
I I I T
30 40 50 60
I
70
I
80
I
95
Figure 2.
90 95 98
ACCUMULATIVE PERCENT LESS THAN INDICATED SIZE
Particle size distribution Calciner (Log-probability plot of selected
data taken from Table 18.)
48
-------
10-
7—
5 —
3 —
2—
3.
U
N
55 1
UJ
_l
o
I-
oc
2
1.5
.5 —
.2 —
0.1
Ra-226 (Net)
Xg=1.8
I
2
I
5
I
10
I
60
I
95
20 30 40 50 60 70 80 90 95 98
ACCUMULATIVE PERCENT LESS THAN INDICATED SIZE
Figure 3. Particle size distribution Proportioning Building
(Log-probability plot of selected data taken from Table 19.)
49
-------
10-
7—
5 —
2—
UJ
N
55 1
UJ
QC __
.2 —
0.1
Q FURNACE BLDG. Ra-226 (Net)
Xg=0.76
S9=(f76=3-5
£ PELLETIZER BLDG. Ra-226 (Net)
Xg=0.75
* 5,5-0
2
D
I
5
r
10
I I I I I I
20 30 40 50 60 70
r
80
Figure 4.
90 95 98
ACCUMULATIVE PERCENT LESS THAN INDICATED SIZE
Particle size distribution (Log-probability plot of selected data
taken from Tables 20 and 21.)
50
-------
micrometer stage, etc. The lines are based on hand-drawn esti-
mates and not based on regression analysis.
The plotted distributions (Figures 2, 3, and 4) have not
been statistically tested to demonstrate the appropriateness of
describing the data with a lognormal distribution. The log-
probability plots are used as a tool to summarize the data and
describe its central tendency.
The data in Figures 2, 3, and 4 indicate geometric means
fX ) of one to three micrometers which are within expected values
g
for atmospheric and industrial dusts (Lee, 1972; Elder et al.,
1974; and Willeke, 1975). The geometric standard deviations (S )
of 110 and 60, for the data in Figures 2 and 3, appear to be
excessive, while those in Figure 4 (areas of lower concentra-
tions) are within ranges seen for ambient dust (Lee, 1972).
The high geometric standard deviations for the data in
Figures 2 and 3 probably reflect the composite of several size
distributions. That is, there may be several particle size dis-
tributions from the plant in conjunction with the ambient dust.
The Calciner (Figure 2) and Proportioning Building (Figure 3)
results indicate concentrations above the ambient levels.
Knuth (1976), in his evaluation of the impactor, indicated
that it tended to underestimate the size distribution and over-
estimate the geometric standard deviation. Thus, it is possible
that this is also reflected in the data. Although, Knuth's
results indicated that this appeared to be more of a problem for
the larger size distributions (X of 7 micrometers) but was less
of a problem for distributions around 4 micrometers.
Results from Lee (1972) indicate average geometric standard
deviations of up to 20 for ambient air. Elder et al., (1974)
indicate geometric standard deviations in the thirties for
51
-------
plutonium processes. Thus, the high values noted in this study
may be realistic. Furthermore, the data are not sufficient to
break the indicated distribution into component parts which would
have lower geometric standard deviations. The data in Figure 4
are for concentrations which appear to be near ambient levels.
STACK SAMPLING
Stack sampling was conducted in late September 1975 using
the RAG Train Stacksamplr* and methods specified in the Federal
Register, Volume 36, No. 247 (December 23, 1971). Representative
samples were obtained from each process-type discharge stack;
however, every plant discharge stack was not sampled (e.g., only
one each of the Medusa stacks was sampled, but this one sample
should be representative of the other remaining Medusa discharge
stacks).
Particulate matter was collected on a glass fiber filter
(2.5-inch diameter) which was subsequently analyzed for natural
radioactivity content. To determine the blank filter natural
radioactivity content, two sets of unused filters were analyzed
and these results are shown in Table 25. Also shown is the
extrapolated filter content based on the area ratio between the
2.5-inch and the 4-inch diameter filters of 0.391 times the 4-
inch diameter activity content as reported by Eadie and Bernhardt,
(1976). These results show fairly good agreement between the
measured and the extrapolated activity contents. Both filters
are made from the same type of material. There was less analyt-
ical uncertainty for the 4-inch filters because the sample was
based on a larger amount of material. The extrapolated filter
content (last column of Table 25) has been used as a blank filter
activity content which has been subtracted from the gross ana-
lytical result to obtain the reported "net result". No blank
* Research Appliance Corporation (RAG), Gibsonia, PA.
52
-------
TABLE 25. RADIOACTIVE CONTENT OF BLANK GLASS FIBER FILTER (2.5-INCH DIAMETER)*
(pCiYfilter ± two-sigma counting error term)
Radionuclide
Ra-226
Po-210
U-234
U-235**
U-238
Th-230
Th-232
Pb-210
Ra-228
5- Filter
Composite
0.52 ± 0.14
0.29 ± 0.11
0.060 ± 0.038
<0.0012
<0.026
<0.075
<0.080
No data
<0.81
6- Filter
Composite
0.42 ±
0.058 ±
0.022 ±
0.0012±
0.025 ±
0.089 ±
0.029 ±
0.16
0.015
0.018
0.00079
0.017
0.040
0.023
<0.088
No data
Average Blank
Filter Content1"
0.47 ± 0.21
0.17 ± 0.11
<0.041
<0.0012
<0.026
<0.082
<0.055
<0.088
<0.81
Extrapolated
Filter Content1"1"
0.14 ± 0
<0.066
0.039 ± 0
0.0014+ 0
0.031 ± 0
0.078 ± 0
0.051 ± 0
<0.13
<0.62
.035
.012
.00039
.0078
.031
.0078
*Average 2.5-inch filter mass ± two standard deviations of 0.2159 ± 0.0050 grams.
**U-235 calculated based on natural U-235 to U-238 activity ratio of 1:21.45 (0 0466)
+Average of the 5-filter and 6-filter composite results.
++Extrapolated'value based on area ratio between 2.5-inch and 4-inch diameter filters
of 0.391 times the 4-inch diameter activity content (Eadie and Bernhardt, 1976).
53
-------
subtractions have been made for the three radionuclides (radium-
228, polonium-210, and lead-210) which are at the analytical
minimum detectable activity (MDA) level. The net results for the
stack effluent discharge radioactivity concentrations are reported
in Table 26. If a polonium/lead-210 blank value of 0.1 pCi/filter
is used, the blank activity for aim3 sample is 0.1 pCi/m3. It
is then evident that the measured values are significantly above
this estimated blank value, except for the Coke Bag House.
MASS AND ACTIVITY BALANCE
General mass and activity balances are presented as accumu-
lative bar graphs in Figure 5. These balances only include the
major process streams (one percent of total mass or activity) and
are limited to the solids input and output streams and stack
effluents. Additional details concerning the balances and the
parameters used in the calculations are given in Appendix D.
It is emphasized that the balances are based on a limited
amount of data (roughly one-cubic meter stack discharge samples
of selected stacks and limited grab samples of the product and
by-product streams) taken at a point in time. There are also
uncertainties in the results due to counting statistics and
analytical variances, potential sample collection and treatment
biases, and in the case of the 1^4-210 air results, actual
errors in the analysis. The polonium-210 results for the stack
samples are sufficiently greater than the lead-210 results so
that the error in the lead-210 results has essentially no impact
on the polonium-210 results (i.e., the ingrowth correction for
polonium from lead is minimal).
The balances indicate that the phosphate ore contributes 90
percent of the input mass and essentially 100 percent of the
input activity which is estimated to be 40 curies per year of
uranium-234 and -238, radium-226, lead-210, and polonium-210
54
-------
TABLE 26. STACK SAMPLING RESULTS
(NET RAOICNUCLIDE CONCENTRATION ± THO-SIGMA ERROR TERM IH pd/m'-DRY STANDARD CUBIC METERS)
Location
Sampled
Volume (date)
Ra-226 Po-210* U-234
U-235
U-238
Th-230
Th-232
Pb-210* Ra-228*
On
Reclaim No. D
Bag House
1.09m3 (9/22/75)
2-2 Caldner
Cooler
1.26m3 (9/23/75)
2-2 Caldner
Scrubber
1.43m3 (9/23/75)
Coke Bag House
0.73m3 (9/25/75)
Fluid Bed
Scrubber
0.86m3 (9/26/75)
Medusa 4-Uest
1.12m3 (9/30/75)
0.59 ± 0.22 17 ± 3.6 0.64 ± 0.15 <0.039 0.57 ± C.14 0.81 ± 0.26 <0.16 2.7 i 0.52 <1.5
1.7 ±0.31 61+6.3 15 ±1.2 0.76 ± 0.16 13 ± 1.1 1.9 ± 0.33 0.36± 0.1C 3.3 ± 0.55 <1.3
0.31 ± 0.16 1700± 140 0.30 ± 0.10 <0.022 0.32 ± 0.092 0.55 ± 0.17 0.27± 0.12 1.0 ± 0.40 <0.00095
0.38 ± 0.27 <2.8 <0.018 <0.0091 <0.026 <0.014 <0.011 3.9 t 0.79 4.4 ± 2.5
1.2 ± 0.35 530 ± 45 0.20 t 0.12 <0.070 0.14 ± 0.12 <0.019
0.60 ± 0.23 25 ± 4.1 0.23 ± 0.10 <0.028 0.21 ± 0.10 <0.070
0.20+ 0.15 9.3 t 1.6 <2.0
<0.018 8.2 ± 0.93 <1.5
•Gross analytical result.
-------
ACCUMULATIVE MASS* (10' tons) ACCUMULATIVE ACTIVITY (CURIES)*
210 10 20 30 40
Illllllllllllllll RADIUM.226
COKE & SILICA
OUTPUT:
CALCINING PROCESS
STACK EFFLUENTS
COOLER
SCRUBBER
ELECTRIC FURNACE
STACK EFFLUENT
FLUID BED SCRUBBER
PRODUCT STREAMS
FLUID BED PRILLS
~1%
FERROPHOSPHORUS «* 1%
SLAG**
LESS THAN 0.5 Ci
0.2 Ci of Po-210
Po-210
0.3 Ci of Po-210
TOTAL STACKS
7.4 Ci of Po-210
<1 Ci of Ra, U,
& Pb-210
1.3 Ci of U
0.83Ci of Pb-210
'
of
E
I 0.4 Ci of U
0.2 Ci of Pb-210 (<0.1 Ci of Po-210)
ELEMENTAL PHOSPHORUS
LESS THAN 0.1%; i.e.,<0.04 Ci
THE INPUT AND OUTPUT VALUES ARE INDICATED AS ACCUMULATIVE VALUES. THUS SUBSEQUENT
ENTRIES ARE ADVANCED ALONG THE SCALE TO INDICATE THE SUM-TOTAL FROM PREVIOUS ENTRIES
ACCUMULATIVE OUTPUT VALUES IN EXCESS OF THE INPUT VALUES REFLECT UNCERTAINTIES IN THE
ANALYTICAL AND MASS FLOW DATA.
THE SLAG Po-210 VALUE IS BASED ON A LESS THAN VALUE OF 16 pCi/g (i.e., ALTHOUGH ANALYTICAL
RESULTS INDICATE THE PRESENCE OF Po-210 IN THE SLAG, THE INDICATED MASS BALANCE VALUE HAS AN
UNCERTAINTY APPROACHING 100 PERCENT SEE APPENDICES C AND D) .
Figure 5. Annual mass and activity balance of major streams - the thermal
process plant
56
-------
(this also applies to thorium-230 which is not shown in the
balance). Essentially all of the input uranium and radium-226,
and 90 percent of the mass are accounted for in the output slag.
Only 50 percent of the input lead-210 is accounted for in
exit streams, almost all of which is in the slag. The lead-210
results for the slag are considered valid, but errors in the
stack samples and fluid bed prills could possibly account for 10
percent or more of the lead-210. The rest of the lead-210 would
appear to be associated with fluid streams, several of which
indicated high lead-210 results (Tables 8 and 10).
About 20 percent of the input polonium-210 activity is
associated with the stack effluents (primarily the Calciner
scrubbers). The analytical uncertainty for this estimate is
greater than 10 percent (i.e., counting error and error due to
correcting for polonium-210 ingrowth from lead-210).
Over 50 percent of the polonium-210 appears to be associated
with the slag, but this value is based on using a less than value
of 16 pCi/g for estimating the slag inventory. The polonium-210
results in Table C-3 in Appendix C imply the presence of polonium
in the slag, but the correction for polonium-210 ingrowth from
lead-210 in the slag (Appendix A) introduces considerable uncer-
tainty in these results. Furthermore, the assumed slag value (16
pCi/g) appears to be contradicted by the low polonium-210 value
for the calcined rock. That is, if most of the polonium-210 is
lost from the process material during calcining, it can not be in
the final by-product slag (see Appendix D for actual inventory
values). Thus, it appears that whereas the polonium-210 inventory
value for calcined rock (4 Ci or less) is probably an under/-
estimate, the estimate for polonium-210 in slag (26 Ci or less)
may also be an overestimate. If the estimate for polonium-210 in
slag is too high, then the output polonium-210 is not accounted
for in the products used for the balance (i.e., use of 26 Ci in
slag accounts for 100 percent).
57
-------
SUMMARY
External gamma radiation measurements conducted in various
working areas of the FMC Corporation's Thermal Process Plant near
Pocatello, Idaho ranged from background to 100 yR/h. General
work areas averaged about 42 mrem per work year (2000 hours), and
the maximum net dose equivalent rate (directly on the slag pile)
was estimated to be 182 mrem per work year. The dose received
from natural background radiation in the Pocatello area was
estimated to be about 79 mrem per year due to a cosmic and ter-
restrial background exposure rate of about 9 WR/h.
Ambient radon-222 concentrations in several buildings ranged
from 0.17 to 1.4 pCi/1, except for one value of 11 pCi/1, which
was measured in the Control Room of the Condenser and Fluid Bed
Building. Ambient radon measured in two process control rooms
were higher than would normally be expected from natural terres-
trial background sources alone. Appreciable exposures from radon
and its progeny may be experienced by personnel working in several
indoor areas of the plant. Therefore, additional ambient radon
or working level determinations should be conducted to better
define this exposure source term.
Extensive radiochemical analyses of the input ore, calcined
briquettes and the waste product slag, indicate natural radio-
activity contents in excess of natural rocks and soils. Except
for the lead-210 and polonium-210 activity, there appears to be
little difference in the radioactivity content of these three
materials. For the input ore, the various decay chain members
are in secular equilibrium with the parent uranium-238. The
calcined briquettes and slag showed depleted lead-210 and
polonium-210 contents. The slag indicates somewhat less lead-210
than the calcined rock or briquettes (but statistically not
different) indicating there is some additional loss of lead-210
in the electric furnace after the calcining.
58
-------
The polonium-210 results for slag and calcined rock are
contradictory in that the slag results are higher than the rock.
Although the 100 percent error terms cause the results to overlap,
it appears that the calcined rock results may be low. Also, the
assumption that the slag concentration is equal to the less than
value is probably an overestimate. In any case, these data in
conjunction with the stack results, indicate the calcining process
is the primary source for releasing airborne polonium-210.
A general mass and activity balance (Figure 5) indicates
that the phosphate ore accounts for essentially all of the input
activity, and the slag accounts for the output activity, except
for lead-210 and polonium-210. Only about 50 percent of the
input lead-210 is accounted for in the output solids streams
(mostly in the slag) and
-------
REFERENCES
BERNHARDT, David E. (May 1976), Evaluation of Sample Collection
and Analysis Techniques for Environmental Plutonium. U.S.
Environmental Protection Agency, Technical Note ORP/LV-76-5.
EADIE, Gregory G, R.F. KAUFMANN, D.J. MARKLEY, R. WILLIAMS (June
1976), Report of Ambient Outdoor Radon and Indoor Radon Progeny
Concentrations During November 1975 at Selected Locations in the
Grants Mineral Belt, New Mexico. U.S. Environmental Protection
Agency, Technical Note ORP/LV-76-4.
EADIE, Gregory G. and D. E. BERNHARDT (December 1976), Sampling
and Data Reporting Considerations for Airborne Particulate
Radioactivity. U.S. Environmental Protection Agency, Technical
Note ORP/LV-76-9.
EISENBUD, M. (1963), Environmental Radioactivity, McGraw-Hill
Book Company, New York.
ELDER, J.C., M. GONZALES, and H.J. ETTINGER (1974), Plutonium
Aerosol Size Characteristics. Health Physics, 27:45-53.
(EPA, February 1977) - U.S. ENVIRONMENTAL PROTECTION AGENCY
(February 3-5, 1977), Report of the Workshop on Issues Pertinent
to the Development of Environmental Protection Criteria for
Radioactive Wastes. Washington, D. C.
FEDERAL REGISTER, Volume 36, No. 247 (December 23, 1971),
"Standards of Performance for New Stationary Sources." Also,
Volume 41, No. Ill (June 8, 1976), "Proposed Amendments to
Reference Methods."
••»
GUIMOND, R.J. and S.T. WINDHAM (August 1975), Radioactivity
Distribution in Phosphate Products, By-Products, Effluents, and
Wastes. U.S. Environmental Protection Agency, Technical Note
ORP/CSD-75-3.
HARLEY, J.H. (1975), "Environmental Radon," in The Noble Gases,
R.E. STANLEY and A.A. MOGHISSI, eds., U.S. Government Printing
Office, Washington, D.C. pp. 109-114.
HOLTZMAN, R.B. (1964), "Lead-210 (RaD) and Polonium-210 (RaF) in
Potable Waters in Illinois," in The Natural Radiation Environment,
J. A.S. ADAMS and W.M. LOWDER, eds., The University of Chicago
Press, Chicago, pp 227-237.
60
-------
JOHNS, F.B., ed. (February 1975), Handbook of Radiochemical
Analytical Methods. U.S. Environmental Protection Agency,
EPA-680/4-75-001.
JOHNSON, Raymond H., Jr., D.E. BERNHARDT, N.S. NELSON and
H.W. GALLEY, Jr. (November 1973), Assessment of Potential Radio-
logical Health Effects for Radon in Natural Gas.
U.S. Environmental Protection Agency, EPA-520/1-73-004.
KAPLAN, Irving (1958), Nuclear Physics, Addison-Wesley Publishing
Company, Inc. Reading, Massachusetts, U.S.A.
KNUTH, R.H. (1976), Calibration of the Sierra High Volume Slotted
Cascade Impactor: HASL-Technical Memo 76-6.
LEE, R.E. JR. (1972), The Size of Suspended Particulate Matter in
Air: Science, 178:567-575.
(NCRP, January 1971) - NATIONAL COUNCIL ON RADIATION PROTECTION
AND MEASUREMENTS, NCRP Report No. 39 - Basic Radiation Protection
Criteria. Washington, D.C., p. 135.
(NCRP, November 1975) - NATIONAL COUNCIL ON RADIATION PROTECTION
AND MEASUREMENTS,NCRP Report No. 45 - Natural Background Radiation
in the United States. Washington, D.C., p. 163.
OAKLEY, D.T. (1972), Natural Radiation Exposure in the United
States. U.S. Environmental Protection Agency, ORP/SID 72-1.
O'BRIEN, K. and R. SANNA (1976), Absorbed Dose-Rates in Humans
from Exposure to Gamma Rays. Health Physics 30:71-78.
PEARSON, J.E. (May 1967), "Natural Environmental Radioactivity
from Radon-222." Environmental Health Series, U.S. Public Health
Service Publication No. 999-RH-26.
POET, S.E., H.E. MOORE and E.A. MARTELL (1972), "Lead-210,
Bismuth-210 and Polonium-210 in the Atmosphere." J. Geophys Res.
77, 6515.
SEDLET, J., N.W. GOLCHERT and T.L. DUFFY (1973), Environmental
Monitoring at Argonne National Laboratory - 1972, USAEC Report
ANL-8007, Argonne, Illinois.
STATE OF IDAHO (June 1, 1977), Technical Policy Memorandum No. 7 -
Concerning the Use of Radium - Contaminated Phosphate Slag in
Idaho.
SWIFT, J.J., J.M. HARDIN and H.W. GALLEY (January 1976), Potential
Radiological Impact of Airborne Releases and Direct Gamma Radia-
tion to Individuals Living near Inactive Uranium Mill Tailings
Piles. U.S. Environmental Protection Agency, EPA-520/1-76-001.
61
-------
UNITED NATIONS (1972), Ionizing Radiation: Levels and Effects,
Vol. 1: Levels. (A report of the United Nations Scientific
Committee on the Effects of Atomic Radiation for the General
Assembly.) New York.
U.S. PUBLIC HEALTH SERVICE (1969), Evaluation of Radon-222 Near
Uranium Tailings Piles, DER 69-1. U.S. Department of Health,
Education, and Welfare, Rockville, Maryland.
WILLEKE, K. (1975) Performance of the Slotted Impactor. Presented
at 15th American Industrial Hygiene Conference, Minneapolis,
Minn., June 1975. Mechanical Engineering Department, University
of Minn., Minneapolis, Minn.
WINDHAM, S., J. PARTRIDGE and T. HORTON (December 1976), Radiation
Dose Estimates to Phosphate Industry Personnel. U.S. Environmental
Protection Agency, EPA-520/5-76-014.
62
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APPENDIX-A
RADIOCHEMICAL ANALYTICAL METHODS
63
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APPENDIX-A
RADIOCHEMICAL ANALYTICAL METHODS
All analyses were completed at the Environmental Monitoring
and Support Laboratory in Las Vegas, Nevada (EMSL-LV). The
following sections present brief descriptions of the analytical
methods employed for this study. Specific details of the pro-
cedures are contained in the Handbook of Radiochemical Analytical
Methods, F. B. Johns, ed. (1975).
Analysis of Radium-226. Radium-228 and Lead-210
A sequential method for the determination of radium-226,
radium-228, and lead-210 in environmental samples has been
developed by the EMSL-LV. This method is initiated by the
precipitation of radium from the sample aliquot using barium
sulfate. Barium-radium-sulfate is then dissolved in a diethylene-
triaminepentaacetate disodium solution and transferred to an
emanation tube and the radon allowed to come to equilibrium,
approximately 30 days ingrowth with its parent - radium.
Radium-226 (T^ « 1602 years) decays by alpha emission to radon -
222 CT^ =3.8 days). Radon-222, a noble gas, is then collected
from the liquid by a de-emanation technique. Radon-222 is
usually counted for 30 minutes by alpha scintillation at four and
one-half hours after the de-emanation step to allow for the
build-up of the daughters.
The solution from the radium-226 determination is saved and
the total radium is reprecipitated. Radium-228 (Th * 6.1 years)
is a beta emitter and decays to actinium-228 (T% - 6.13 hours).
The actinium is allowed to ingrow for three days and is extracted
64
-------
with diethylhexylphosphoric acid and back extracted with nitric
acid. The actinium-228 is beta counted for 30 minutes in a low-
level beta counter.
Lead is also precipitated with the radium sulfate in the
original solution. Advantage is taken of the 30-day storage, for
radon-222 ingrowth, to allow the bismuth-210 to grow in. Lead-
210 (Tis = 20.4 years) decays by beta emision to bismuth-210
(Tjg = 5.01 days). The bismuth-210 is precipitated from the
supernatant liquid in the radium-228 separation step. Bismuth is
converted to an oxide, dissolved in nitric acid, mounted on a
two-inch planchet and beta counted for 30 minutes.
Analysis of Isotopic Uranium and Thorium
Samples are decomposed utilizing techniques of nitric-
hydrofluoric acid digestion, potassium fluoride fusion or igni-
tion. The residues are dissolved in dilute nitric acid and
successive sodium and ammonium hydroxide precipitations are
performed in the presence of boric acid to remove fluoride and
soluble salts. The hydroxide precipitate is dissolved, the
solution is adjusted to 9N in hydrochloric acid, and uranium is
absorbed on an anion exchange column, separating it from thorium.
Iron is removed from the column by washing with hydrochloric acid
and the uranium is eluted with dilute hydrochloric acid. The
thorium is converted to a nitrate form and absorbed on the same
anion exchange column separating it from calcium and other inter-
ferences. The thorium is then eluted with 9N hydrochloric acid.
The uranium is electrodeposited on stainless steel discs from an
ammonium sulfate solution and subsequently counted by alpha
spectrometry. Usually 1000-minute counting times are used for
analysis. Chemical yields are normally determined by the recovery
of internal tracer standards (e.g., uranium-232 and thorium-234)
added at the beginning of the analysis.
65
-------
Analysis of Polonium-210
Samples are decomposed by digestion with hydrofluoric acid
and nitric acid in the presence of lead carrier and a polonium-
208 tracer. Polonium is co-precipitated with lead sulfide from a
dilute acid solution separating it from calcium, iron, and other
interferences. The sulfide precipitate is dissolved in dilute
hydrochloric acid and polonium is spontaneously deposited on a
nickel disk. Polonium-210 and polonium-208 tracer are measured
by alpha spectrometry. Usually, 1000-minute counting times are
used for analysis.
Polonium-210 Activity Estimations
Due to the time delay between sample collection and radio-
chemical analysis for polonium-210, the following considerations
have been utilized to estimate the polonium-210 activity in a
sample.
1. Polonium-210 decays with a radiological half-life of
138 days. Therefore, a decay correction factor should
be considered for samples which are held for a rela-
tively long period between collection and analysis.
2. Concurrently, there will be some polonium-210 ingrowth
from lead-210 and bismuth-210 contents of the original
sample during the elapsed time between collection and
analysis.
The consideration of radioactive series transformations
is discussed in detail in Kaplan (1958). For the
specific case of lead-210, the decay scheme is as shown
in Figure A-l. The solution of the system of differen-
tial equations, which describe this lead-210 decay
series, was derived by Bateman and is summarized here.
66
-------
Ti£=20.4year
A =0.0000931 day1
beta
gamma
210 Ri
83 Dl
T%=5.01day
A=0.138day-<
beta
Ti/2=138day
A =0.00501 day"1
alpha
gamma
STABLE
FIGURE A-1. LEAD-210 DECAY SCHEME
-------
a. For polonium-210 ingrowth from lead-210 in the original sample:
APb =
APo
1.01 A^e-°-0000931t * 0.0377 A0^0-138' - 1.06 A^
Where
.Pb
Po = Polonium-210 activity at time of measurement due
to ingrowth from lead-210 in sample
= lead-210 activity at time of sample collection
t = Elapsed time (in days) between sample
collection and analysis
Since the bismuth term is negligible and allowing for typical
values of t, the following simplification of the above
expression results:
= A°b (1.01- 1.06 a'0'00501')
b. For polonium- 210 ingrowth from bismuth- 210 in the original
sample:
= 0.0377 Ae-'- 0.0377
Where
Ri
APo = Polonium- 210 activity at time of measurement due to
ingrowth from bismuth- 210 in sample.
ABi = Dismuth- 210 activity at time of sample collection
This term is insignificant for typical values of t and even in
the case of the complete decay of bismuth- 210, the polonium- 210
ingrowth factor would be less than four percent (i.e., decay
constant for polonium- 210 divided by the decay constant for
bismuth- 210 equals 0.036).
68
-------
c. The polonium-210 decay correction term:
4Po _ Ao -O.OOSOlt
APo APoe
Where
Po
ApQ = polonium-210 activity at time of measurement due to decay
of original polonium-210 in sample.
o
ApQ - Original polonium-210 activity in the sample at time of
collection.
Therefore, the polonium-210 activity at the time of measurement
(Ap0) is the summation of the two source terms (i.e., due to the
ingrowth from lead-210 and due to the decay of the original
polonium-210 in the sample).
A = A™ + APo
APo APo APo
or
Apo = A°b (l.oi - 1.06 e-°-00501t) * A°oe-°-00501t
d. Due to the relatively long half-life of lead-210 (T^ = 20.4
years), the lead-210 activity at the time of measurement (A™)
is essentially equivalent to the original Pb-210 activity in
the sample (Apb). Solving the above expression for the original
polonium-210 activity in the sample at the time of collection
(ApQ) yields:
A° = eo.oosoit [A _ A (I>01 _ 1 Q6 e-o.oosoit,.
All reported polonium-210 analytical results are calculated
values obtained from the above equation. Whenever the apparent
polonium-210 content from its ingrowth from lead-210 exceeds
the measured polonium-210 activity, a non-detectable (ND) value
has been reported.
69
-------
e. The estimated counting error term associated with this calcu-
lated polonium-210 activity is simply the sum of squares of the
individual counting error terms, or:
afc = (e°-005VK + 0.2 Cl - l.l e-°-005V]
APo APo APb
Where
a^o = Estimated counting error term of the calculated
polonium-210 activity in the sample at the time of
collection.
CT. = Counting error term of the measured polonium-210 activity.
Po
CTA = Counting error term of the measured lead-210 activity.
In the tables, the estimated counting error terms at the 95
percent confidence level (i.e., twice o.o ) have been included
Pn
for the calculated polonium-210 activities.
Throughout this report, the symbol for less than (<) has been
used to indicate the equivalence of the error term (at the 95
percent confidence level) to the reported result.
70
-------
APPENDIX-B
AIRBORNE PARTICULATE SAMPLING
GROSS RADIONUCLIDE CONCENTRATION RESULTS
71
-------
TABLE B-l. AIRBORNE PARTICULATE SAMPLING IN THE CALCINER AREA
GROSS RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM.
ro
Radionuclide
South Side Discharge,*
Calciner Belt #1 Dump Area -
1218 to 1500 hrs. (5/5/75)
pCi/m3
Ra-226
Po-210
U-234
U-235***
U-238
Th-230
Th-232
Pb-210
Ra-228
2.7
66
1.5
0.044
1.5
2.8
0.055
0.40
<0
± 0.18
± 4.9
± 0.12
± 0.014
± 0.12
± 0.15
± 0.022
± 0.28
.23
pCi/g
38 ±
560 ±
20 ±
0.62 ±
21 ±
39 ±
0.77 ±
120 ±
<3.2
2.4
70
1.7
0.20
1.7
2.1
0.30
9.5
No. 1 Control Room - **
1027 to 1520 hrs. (5/5/75)
pCi/m3
0.019 ± 0.010
<0.0078
0.0091 ± 0.0023
(<0. 00032)
0.0068 ± 0.0023
0.0091 ± 0.0045
0.0045 ± 0.0023
0.17 ± 0.066
<0.085
pci/g
24 ± 13
<88
12 ± 3.1
(<0.43)
9.2 ± 3.1
12 ± 6.1
6.1 ± 3.1
210 ± 83
<120
*Sampled Volume of 7.9m3, with a dust particulate load of 0.5663g.
**Sampled Volume of 17.6m3, with a dust particulate load of O.OlSlg.
***U-235 results reported in parenthesis are calculated based on U-235 to U-238 natural activity
ratio of 1 to 21.45 (0.0466).
-------
TABLE B-2. AIRBORNE PARTICUtATE SAMPLING IN THE PROPORTIONING BUILDING
GROSS RADIONUCUDE CONCENTRATION ± TWO-SIGHA COUNTING ERROR TERM.
Radlonuclfde
Lower Level, 1st Pillar*
West of Shack,
0947 to 1309 hrs.~( 5/6/75)
Lower Level Control Room**
0944 to 1316 hrs. (5/6/75)
Upper Level, Top West*
1140 to 1315 hrs. (5/6/75)
of. i An 3 nr.i/n
Upper Level Control
0930 to 1411 hrs. (5/6/75)
nf.i /m3 nr-i In
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
Ra-228
0.37
5.3
0.30
0.014
0.32
0.72
0.037
0.63
± 0.045
± 0.89
± 0.036
± 0.0069
± 0.038
t 0.062
± 0.014
± 0.11
15
210
12-
0.54
13
29
1.5
25
<0.15
± 1.8
* 36
± 1.5
± 0.27
* 1.5
l 2.5
± 0.58
± 4.5
<6.0
0.13 ± 0.027
,2.7
0.097 t 0.017
0.0045 ± 0.0039
0.11 ± 0.019
0.21 ± 0.034
<0.0028
0.24 ± 0.097
<0.12
17
350
13
0.59
15
27
<0
30
«
± 3.5
± 71
± 2.2
± 0.51
± 2.4
t 4.3
.35
t 12
15
2.9 ± 0.21
<3.4
2.8 ± 0.23
0.13 ± 0.027
2.8 ± 0.23
4.9 + 0.25
0.12 ± 0.036
6.6 ± 0.48
<0.39
16
<
16
0.71
16
27
0.66
37
<
± 1.2
21
± 1.3
± 0.15
± 1.3
± 1.4
± 0.20
± 2.7
2.2
0.49
3.8
0.29
0.012
0.32
0.50
0.015
0.74
± 0.054
± 0.94
± 0.034
i 0.0058
* 0.035
± 0.048
± 0.0083
± 0.10
-------
TABLE B-3. AIRBORNE PARTICULATE SAMPLING IN THE PELLETIZER BUILDING
GROSS RADIONUCLIDE OWCENTRATIOH ± TWO-SI6HA COUNTING ERROR TERN.*
Radlonucllde
Ra-226
Po-210
U-234
U-235
U-238
Th-230
Th-232
Pb-210
R«.22B
Operator Level. Middle** Upper Level. Top East*
Lower Level. 1000 to 1537 hrs. (5/5/75) 1003 to 1530 hrs. (5/5/75)
pd/li3 0C1/g PCI/*3 pCI/g
0.28
1.1
0.45
0.023
0.24
0.50
0.019
0.66
-------
TABLE B-4. AIRBORNE PARTICULATE SAMPLING IN THF FURHANCE BU1LB1HG
GROSS RADIONUCLIDE CONCENTRATION ± TWO-SISMA COUNTING ERROR TERM.*
Radionucllde
Control Room Level**
Between Furnace 3 ft 4
0906 to 1606 hrs45/7/75)
pCi/m3
Ra-226
Po-210
U-234
U-235*
U-238
Th-230
Th-232
Pb-210
Ra-228
0.
0.
0.
0.
0.
0.
0.
047
63
on
<0
on
02
0066
27
// to b//b)
pCi /tn 3
0.0091 ± 0.003
0.16 ± 0.057
0.0024 ± 0.0012
(0.0002)
0.0042 ± 0.0016
0.0036 ± 0.0017
<0. 00089
0.027 ± 0.013
<0.019
PCi/g
12 ±
210 ±
3.2 ±
(0
5.5 ±
4.8 ±
<1
36 ±
<25
4.0
79
1.6
.26)
2.1
2.6
.2
18
*U-235 results reported in parenthesis are calculated based on U-23b to U-238 nacurai activity ratio of 1 to 21.45 (0.0466).
**Sarapled Volume of 24.3m3, with a dust participate load of 0.0678g.
+Sanpled Volume of 23.4m3, with a dust particulate load of 0.0370g.
++Sampled Volume of 84ra3, with a dust particulate load of 0.0623g.
-------
TABLE B-5. AIRBORNE PARTICIPATE SAMPLING
GROSS RADIONUCLIDE CONCENTRATION ± TWO-SIGTIA COUNTING ERROR TERM.*
Technical Building** Condenser and Fluid Bed ***
Library- Control Room-
Radlonuclide 0935 (5/5/75) to 142J hrs. (5/6/75) 0904 to 1444 hrs. (5/6/75)
pCi/m3
Ra-226
Po-210
U-234
U-235*
U-238
Tfi-230
Th-232
Pb-210
Ra-228
0.0013
0.00093
(0
0.0012
0.0019
<0,
0.019
± 0.0012
ND
t 0.00062
.000056)
t 0.00062
i 0.00062
.0003!
i 0.0086
PCi/q
21 ± 19
ND
15 ± 9.9
(0.93)
20 ± 9.9
30 i 9.9
<4.9
320 ± 140
<190
pCi/m3
0.012
0.29
0.015
(0
0.011
0.017
0.0057
0.20
± 0.008
± O.J5
t 0.0038
.00051)
± 0.0038
i 0.0057
± 0.0038
± 0.059
,081
pCi/g
9.6 ±
2r.O ±
13 ±
(0.
9.4 ±
14 ±
4.7 ±
160 t
<66
6.4
210
3.1
44)
3.1
4.7
3.1
47
Old Kiln Building (Maintenance)*
Open Area - East End P-4 Loadinq Area +*
0912 to 1412 hrs. (5/6/75) 0933 to 1438 hrs. (5/8/75)
pCi/m 3
0.27
43
0.20
0.0079
0.21
0.33
0.024
0.58
<0
± 0.034
± 4.9
i 0.025
± 0.0044
± 0.025
t 0.037
± 0.011
± 0.089
092
PCi/q
24 ±
3900 +
18 ±
0.69 +
18 ±
29 ±
2.1 t
51 ±
<8 1
3.0
430
2.1
0.38
2.2
3.3
0.95
7.9
High Vol. Sampler
pCi/m3
0.0032
0.038
0.0028
0.0033
0.0047
0.00053
0.018
<0
± 0.0008
t 0.027
± 0.00058
<0. 00014
i 0.00065
± 0.00022
± 0.00022
± 0.0038
Op4;>
*U-235 results reported in parenthesis are calculated based or U-235 to U-23fl natural activity ratio of 1 to 21.45
NO indicates non-detectable.
"Sampled Volume of 130m3, wfth a dust paniculate load of 0.0081g.
***Sarapled Volume of 21m3. with a dust oarticulate load of 0.0256g.
^Sampled Volume of 16.4m3, with a dust particulate load of 0.186Zq.
++Sampled Volume of 346m3.
-------
APPENDIX C
ADDITIONAL SAMPLING OF INPUT AND SLAG PRODUCTS
77
-------
APPENDIX C
ADDITIONAL SAMPLING RESULTS
On December 23, 1976, additional samples of the input ore,
calcined briquettes and of the waste product slag were obtained
from the FMC Corporation. Six samples of each type of material
were analyzed for natural radioactivity content at both the
Environmental Monitoring and Support Laboratory (EMSL) in Las
Vegas, Nevada and at the Eastern Environmental Radiation Facility
in Montgomery, Alabama. These results are shown in Tables C-l,
C-2, and C-3.
Average analytical results except for lead-210, are well
within the 95 percent confidence level, indicating excellent
analytical agreement between the two laboratories. Considering
the statistical fluctuations in the analytical results, there
appears to be little difference, except for lead and polonium-
210, in the radioactivity content of the input ore, calcined
briquettes and slag. However all three materials exhibit ele-
vated levels of natural radioactivity as compared to natural
rocks and soil, as discussed in the main text.
As noted in the text and in the footnotes for Tables C-l,
C-2, and C-3 the EMSL lead-210 results are in error, Thus, these
results, although reported, are discounted in determining the
averages.
For the input ore samples, the various decay chain members
are in secular equilibrium with the parent uranium-238. The
calcined briquettes and slag showed depleted polonium-210 contents,
averaging roughly one-half of the activity of the other individual
decay chain members. Therefore, it appears that the calcining
process is the mechanism for generating the elevated airborne
concentrations of polonium-210 as reported in the main text.
78
-------
TABLE C-l. FMC ORE (PHOSPHATE ROCK) SAHPLI5 (SA'IPLED 12/23/76) RADIONUCLIDE CONOOTSATION i TWO-SIO!A COUNTING ERROR TERM, in pCi/g
SAMPLE »1 SAMPLE »2 SAMPLE »3
• 4
F.ERF* EMSL** EERF BEL EERF MSL EERF tMSL
Ra-266 24.3 t 0.243 26 i 0.93 24.0 t 0.24 27,0.94 21.3,0.213 14,0.68 19.4,0.194 30,1.0
Po-210* 21.5,3.2 24,2.4 22.8,3.3 22,1.3 25.2,3.7 21,1.7 19.4,1.5 22,1.7
U-234 22.0,3.30 21,5.2 20.6,2.99 22,2.4 21.4,2.99 23,2.3 21.7,3.15 22,4.7
U-235 1.02,0.204 <0.78 1.12,0.217 0.98,0.41 0.985,0.187 0.89,0.37 0,97810.191 <0.45
U-238 22.3,3.35 22,5.5 21.0,3.05 20,2.2 22.2,3.11 23,2.3 21.8,3.15 26,4.7
Tn-232 0.483,0.0676 0.30,0.11 0.516,0.0722 0.37,0.13 0.439,0.0636 0.49,0.16 0.441,0.0639 0.42,0.14
Th-230 21.4 , 0.641 23 , 1.1 24.1 ± 0.723 24 ± 1.0 22.2 , 0.666 19 , 0.96 22.8 , 0.648 23 t 0.99
Th-228 0.415,0.0622 0.484,0.0677 0.427,0.064 0.479,0.0646
Tn-227 0.857,0.128 0.723,0.123 0.713,0.118 0.857,0.124
Pb-210 27.7,0.9 2.7,0.69 31.8,1.9 6.4,0.83 29.4 ,1.5 4.7,0.76 27.7,1.0 5.8*0.8
"-228 "-0 0.78,0.76 <1.0 0.74,0.66 <1.0 0.98,0.68 <1.0 0.67,0.65
•EERF - Eastern Environmental Radiation Facility, Montgomery, Alabama.
"EMSL - Environmental Vtonitoring and Support Laboratory, Las Vegas, Nevada; blanks indicate no data.
'Since these values are for geological material that has not been processed (i.e., calcined), it is assumed the polonium and lead-210
r^ ISci'? ^"i?*15"1*1 " *re the other radionuclides; therefore, these polonium values have not been corrected for ingrowth from
lead. EMSL lead-210 results have been found to be in error by up to a factor of five too low, as discussed in the text
79
-------
SAMPLE 15 SAMPLE »6 AVERAGE ± TWO STANDARD DEVIATIONS
in pCi/g
EERF BiSL EERF B4SL EERF BCL
21.1 ± 0.211 29 ± 0.98 20.8 ±0.208 53 ±1.3 21.8 ±3.86 30 ± 25
21.5 ± 1.5 23 ± 1.6 22.4 ± Ot9 21 ± 2.4 22.1 + 3.82 22 ± 2.3
21.8 ± 3.05 24 ± 3.9 20.7 ± 3.00 21 t 3.0 21.4 ± 1.18 23 ± 4.6
1.17 ± 0.221 1.2 ± 0.78 0.876 ± 0.175 0.99 ± 0.56 1.02 ± 0.212 0.88 i O.S1
21.9 ± 3.07 24 ± 4.0 20.5 ± 2.97 22 ± 3.2 21.6 ± 1.43 23 ± 4.1
0.454 ± 0.0635 0.46 ± 0.15 0.479 ± 0.0646 0.36 ± 0.13 0.469 i 0.0594 0.40 ± 0.14
22.1 ± 0.662 23 ± 1.0 22.0 ± 0.66 17 + 0.87 22.4 ± 1.86 22 ± 5.6
0.435 i 0.063 0.449 ± 0.0629 0.448 ± 0.0563
0.820 ± 0.127 0.685 ± 0.1IJ 0.776 ± 0.155 --
15.4 ±1.3 6.5 ± 0.82 31.9+1.8 7.2+1.0 27.3+12.3 5.6+3.3
' 1-0 < 0.65 <1.0
-------
TABLE C-2. WC CALCINED BRIQUETTES (SAMPLED 12/23/76) RADIONUCLIDE CONCENTRATION ± TWO-SIGMA COUNTING ERROR TERM, in pCi/g
SAMPLE »1
EERF*
Ra-226
24.2 i
Po-210+ (7.1 t
U-234 23,8 ±
U-235
U-238
Th-232
Th-230
Th-2E8
TV 227
Pb-210*
Ra-228
1.16 i
23.5 t
0.479 t
21.6 t
0.435 ±
0.694 i
19.1 t
«J
0.242
0.6)
3.56
0.225
3.53
0.0574
0.648
0.0543
0.104
0.3
L.O
BEL"
28
(9.2
22
1.1
24
0.50
23
8.1
1.3
± 0.97
i 1.1)
± 2.0
t 0.48
t 2.0
i 0.15
± 1.0
± 0.90
± 0.84
SAMPLE »2
EERF
21.8 i
(6.1 t
22.3 ±
0.975 t
2S.O i
0.404 t
23.0 t
0 520 *
0 700 *
20.5 ±
<>
0.218
0.6)
3.12
0.185
3.22
0,660fr
0.689
0.3
..0
B1SL
20 t 0.82
(8.5 i 0.88)
22 ± 1.7
0.94 t 0.30
22 1 1.8
0.52 ± 0.15
23 t 0.98
6.5 t 0.85
<0.73
S/WLE
EERF
22.4 ±
(7.4 i
21.9 ±
0.861 t
22.3 i
0.575 ±
22.7 i
25.1 ;
<]
0.224
0.6)
1.18
0.168
3.23
0.0719
0.68
0.6
1.0
• 3
SAMPLE »4
fflSL
31
(11
22
1.0
24
0.44
23
8.7
1.5
± 1.0
i 1.2)
i 2.0
i 0.33
1 1.9
1 0.14
i 1.0
± 0.93
± 0.83
EERF
20.1
(8.1
2J.2
1.14
23.6
0.410
23.8
17.4
± 0.203
± 0.9)
t 3.25
± 0.216
± 3.30
± 0.0636
± 0.713
t 0.6
<1.0
ecL
25 t 0.91
(9.3 i 0.86)
23 ± 1.5
0.94 ± 0.26
23 ± 1.6
0.49 ± 0.14
23 i 0.95
5.2 ± 0.77
< 0.74
•EERF - Eastern Envircamental Radiation Facility, Montgomery, Alabama.
**06L - Environmental Monitoring and Support Laboratory, Us Vegas, Nevada
+The B6L lead-210 results hive been foind to be in error by up to a factor of five too low, as discussed in the text. The poloniun-210
results in parenthesis represent values at date of analysis; the other poloniun-210 values have been corrected for ingrowth from lead-210
as described in Appendix A.
++ND indicates non-detectable; blanks indicate no data.
81
-------
SAMPLE '5 SAMPLE 16 AVERAGE i TWO STANDARD DEVIATIONS
EERF ESML EERF EMSL EERF " **'* ESML
20.7 t 0.207 29 ± 0.98 21.6 i 0.216 32 ± 1.0 21.8 ± 2.77
28 i 8.8
C8'2 * ^O) (11 ± 1.1) (7.9 t 0.9) (9.0 ± 0.89) (7.47 t 1.58) (9.7 * 2.l)
<2.65 ND <2.65
23.1 ± J.23 26 ± 2.S 23.0 ± 3.56 23 t 2.3 22.9 ± 1.56 23 + 3.1
0.870 * 0.178 1.2 t 0.42 1.10 ± 0.237 1.1 ± 0.43 1.02 i 0.269 1.0 t 0.21
23.7 t 3.32 25 t 2.4 23.9 t 3.7 25 ± 2.5 23.3 t 1.18 24 ± 2.3
0.465 ± 0.0674 0.42 t 0.14 0.522 t 0.0705 0.42 * 0.13 0.476 t 0.131 0.47 ± 0.087
21.4 ± 0.641 24 * 1.0 20.9 ± 0.627 24 t 1.0 22.2 t 2.22 23 * 1.0
0.406 t 0.0649 0.491 t 0.0687 0.477 t 0.0929
0.877 ± 0.136 0.691 * 0.117 0.810 t 0.261
12-8 ± 1-3 5.0*0.77 15.2*1.5 9.7*0.94 18.4* 8.60 7.2*3.9
<1>0 « 0'78 <1.0 <0.63 <1.0 < 0.95
82
-------
TABLE C-3. IWC SLAG (SAMPLED 12/23/76) RAIHONUCLIDE CONCENTRATION ± TOO-SIGMA COUNTING ERROR TEIW, in pCi/g
EERF*
Ra - 226 22.8 ± 0.228
Po - 210+ (9.8 ± 1.0)
Corrected** 8.27± 2.86
U - 234 28.4 + 4.25
U - 235 1.59 ± 0.326
U - 238 29.4 ± 4.41
Th - 232 0.648 ± 0.081
Th - 230 2S.5 * 0.764
Th - 228 0.617 i 0.0771
Th - 227 0.981 t 0.142
Pb - 210* 11.6 ± 1.0
Ra - 228 <1.0
SAMPLE »1
tWSL**
37 + 1.1
(8.3 + 1.3)
27 + 4.4
< 0.63
25 ± 4.2
0.52 ± 0.19
29 ± 1.4
2.3 + 0.67
< 0.80
SfMPJE
* 2
EERF EMSL
23.2 +
(2.2 +
ND
21.2 «
1.00 +
22.2 +
0.632 +
25.8 +
0.713 +
0.993 +
11.1 +
<
0.232 40 +
0.2) (2.4 +
2.96 27 t
0.19 1.0 i
3.1 26 *
0.0789 0.20 +
0.774 11 +
0.139
1.7 0.86 +
1.0 <
1.2
0.50)
2.4
0.37
2.4
0.055
0.40
0.61
0.76
SAMPLE »3
EERF
24.8 ±
(12.6 i
23.7 i
21.6 »
1.09 +
21.6 +
0.533 +
26.3 +
0.648 +
0.974 +
6.6 +
<
: 0
.248
2.1)
5.85
2.91
0
.201
2.92
0.0693
0.789
0.0778
0.136
0.6
1.
0
EMSL
36 1 1.1
(11 + 1.1)
27 ± 2.5
1.1 + 0.41
29 + 2.6
0.47 t 0.15
28 i 1.1
2.1 + 0.68
1.6 + 0.86
SAMPLE
EERF
32.5 + 0
.325
(3.7 + 0.5)
ND
27.2 ± 3.81
1.29+0
27.7 ± 3
0.672 + 0
26.6 + 0,
0.696 ± 0,
0.762 + 0.
7.8 i 0.
c 1.
.245
.88
,084
,797
,0835
118
7
0
* 4
EMSL
37 ± 1
(3.3 i 0
29 i 2
0.95 ± 0
25 ± 2
0.66 + 0,
23 + 0,
0.69 ± 0.
< 0.
.1
.43
.4
.35
.3
.16
,95
62
76
Wt' - tastern Environmental Radiation Facility, Montgomery, Alabama.
BMSL - Environmental Monitoring and Support Laboratory, Las Vegas. Nevada
4 as described in Appendix A™**" ™ "^ " ^^ °f anal>'sis: the other Doloniim-210 values'have been corrected for ingrowth froTle'ad^lO
ND indicates non-detectable; blanks indicate no data.
83
-------
SAMPLE »S SAMPLE »6 AVERAGE ± TWO STANDARD DEVIATIONS
in pCi/g
EERF HCL EERF BtSL EERF EMSL
32.3 ± 0.323 37 j 1.1 33.3 ± 0.333 23 t 0X7 28.2 ± 10.1 35 ± 12
(8.9*0.7) (11 ±0.95) (8.2 ±1.3) (8.0*0.72) (7.S7* 7.82) (7.3± 7.4)
v ND ND <16.0
19.7 ± 2.46 24 ± 3.3 20.4 ± 2.65 27 ± 2.0 23.1 ± 7.46 27 ± 3.2
1.63 ± 0.293 0.87 ± 0.51 0.931 ± 0.168 1.2 ± 0.33 1.26 ± 0.601 0.96 ± 0.40
18.6 + 2.32 24 * 3.3 20.3 1 2.64 28 ± 2.0 23.3 * 8.57 26 * 3.9
0.627 1 0.0783 0.73 + 0.22 0.683 t 0.0854 0.70 ± 0.18 0.632 ± 0.107 0.55 ± 0.40
28.1 ± 0.843 33 * 1.5 27.3 t 0.819 31 * 1.2 26.6 ± 1.93 26 * 15
0.672 ± 0.0839 0.775 t 0.093 0.687 t 0.110
1.03 1 0.144 1.07 ± 0.144 0563 ± 0.215
1.4 t 0.66 16.7 ± 2.8 4.2 ± 0.82 10.8 » 7.88 <1.9
<1.0 <0.80 <1.0 <0.76 <1.0 <0.91
34
-------
APPENDIX D
MASS AND ACTIVITY BALANCE
85
-------
APPENDIX D
MASS AND ACTIVITY BALANCE
The data in Table D-l show the parameters and calculated
values for the mass and activity balances. The field study was
organized to try and sample the primary plant streams, with
emphasis on effluent streams. Thus, the data for a complete mass
balance (i.e., internal plant liquid streams) is not available.
Furthermore, all the data obtained has not been utilized directly
in the balance. In some cases, mass flow information was not
available and in other instances, it was apparent that a given
stream did not contribute a significant amount of activity (i.e.,
less than one percent of the total). The present mass balance is
limited to data on solids and stack effluents. That is, liquid
effluent streams have not been included in the balances.
The primary input mass and activity is due to the phosphate
ore. Since the radionuclide values are statistically the same
and there is reasonable cause to assume equilibrium, the values
have been averaged and the input is taken as 40 Ci per year for
each radionuclide.
86
-------
APPENDIX D - TABLE D-l. MASS AND ACTIVITY BALANCE
00
MATERIAL
INPUT
Phosphate Ore
U
Ra-226
Pb-210
Po-210
Coke
U
Ra-226
Pb-210
Po-210
Silica
U
Ra-226
Pb-210
Po-210
IN PROCESS
Calcined Rock
U
Ra-226
Pb-210
Po-210
OUTPUT
Calcining Process
Stack Effluents
RADIONUCLIOE CONC.
(pd/Gran or m')a
(pCi/grain)
22 ± 3
26 t 19
27 i 12
22 ± 3
(pci/g)
<0.5 ± 0.5
0.78± 0.17
2.4 ± 0.6
<1.3 + 1.3
(pCi/g)
1 ± 1
1.7 ± 0.24
0.67± 0.55
2.6 ± 0.9
(pC1/g)
24 ± 2
25 ± 8.6
18 ± 8.6
<2.7 ± 2.7
MATERIAL FLOWb*c MATERIAL FLOW PER YEAR ACTIVITY CURIES0
Gram or m'/year
1.8E6 Ton/yr 1.6E12g/yr
36 ± 14%
42 ± 73%
44 ± 44%
36 ± 14%
Average ± 1 standard deviation 40 ± 4Ci
0.19E6 Ton/yr 0.17E12 g/yr
0.086* 100%
0.13 ± 22%
0.41 ± 25%
0.22 ± \OQ%
0.125E6 Ton/yr 0.11E12 g/yr
0.1 ± 100%
0.2 t 14%
0.08 ± 82%
0.3 ± 35%
1.5E6 Ton/yr 1.4E12 g/yr
33 ± 8%
34 ± 34%
24 ± 48%
<4 ± 100%
inn ?nn ,-fm v P-517 Dry Std. cfmf , ,„,„ .._
Caldner Cooler
(pC1/m!)
stack cfm
x (4 coolers6)
U
Ra-226
Pb-210
Po-210
Calclner Scrubber
U
Ra-226
Pb-210
Po-210
H l 1
1.7 t 0.3
3.2 t 0.6
61 * 6
(pCI/m3)
0.31± 0.10
0.3U 0.16
0.86± 0.5
1700 ±140
106.900 cfm x
d- Cf" 4.05E9 H3/>r
x (4 scrubbers6)
0.04 t 7%
0.005 ± 18%
0.01 t 19%
0.2 * 10%
0.001 t 32%
0.001 t 52%
0.003 ± 58%
6.9 * 8%
-------
Electric Furnace
Stack Effluent
Fluid Bed Scrubber
U
Ra-226
Pb-210
Po-210
Products 4 By-products
Flufd Bed Prills
U
Ra-226
Pb-210
Po-210
0.17± 0.12
1.2 * 0.35
9.2 ± 1.7
530 ±45
(PCI/9)
<8] i 81
13 ± 1
52 ± 2
440 ± 27
68.700
5.94E8 mVyr
x (1 scrubber6)
16,000 Ton/yr
0.016E12 g/yr
O.OOOU7U
0.0007129*
0.005 ±18*
0.31 ± 8*
1.3 ilOOI
0.21 i 8%
0.83 i 41
7.0 i 61
00
CO
Ferrophosphorus
U
Ra-226
Pb-210
Po-210
Slag
U
Ra-226
Pb-210
Po-210
Elemental Phosphorus
Ra-226
Pb-210
(pCi/g)
20 t 5
0.27* o.lI
1.1 ± 0.6
<0.6 l 0.6
(pC1/g)
25 t 7
32 * 13
11 * 8
<16 t 16
(pCVg)
0.0211 0.020
0.21 * 0.17
22.000 Ton/yr
1.8E6 Ton/yr
0.125E6 Ton/yr
0.02E12 g/yr
1.6E12 g/yr
0.11 El 2 g/yr
0.4 ± 25)1
O.OOSi 411
0.02 ± 55%
0.01 ±100*
40 * 28*
51 * 41*
18 ± 735!
26 * 100*
0.002± 95*
0.02 i 81*
FOOTNOTES:
a. Values based on Tables 3, 4, and 26. The net lead and polonfum-210 values are based on 0.1 pCI of activity per
filter (see Table 25). The U results are an average of the uranlum-234 and -238 results. Less than values have
been assumed to be equal to the less than value. Volumes are dry standard cubic meters.
b. Mass flow data based on private correspondence with the company and data obtained with the stack samples.
c. Exponential notation U used. That Is. £12 means 10".
d. The percent error terms are based on the error terns for the concentration of activity In the material.
This factor Is for the number of similar unsanpled stacks at the plant.
e.
f.
Correction factors for the effluent at stack conditions are based on temperature, pressure, and
moisture measurements made during the stack sampling.
-------
APPENDIX E
UNCERTAINTIES IN LEAD-210 AND POLONIUM-210
AIR SAMPLE DATA
89
-------
APPENDIX E
UNCERTAINTIES IN LEAD-210 AND POLONIUM-210 AIR SAMPLE DATA
Tables E-l and E-2 present a tabulation o£ the polonium-210
concentrations, as of time of analysis, and lead-210 concentra-
tions, as of the time of collection, for air samples listed in
Tables 13 to 21 of the text. This tabulation can be used to
assess the uncertainty associated with individual lead-210
results and to estimate the correct value.
Comparison of the lead-210 interlaboratory duplicate results
reported in Appendix C and special method evaluation samples
indicates that the EMSL lead-210 results for solids and air
samples may be underestimated by up to a factor of five or more.
Due to the delay of about 280 days between sample collection and
analysis for polonium-210, the polonium-210 values are generally
indicative of the lead-210 collected in the sample. That is,
while the polonium-210 collected in the sample would have decayed
to 25 percent of its level at time of collection prior to analy-
sis, the ingrowth of polonium-210 from lead-210 collected in the
sample would have reached about 75 percent of the lead-210
value. Thus, the polonium-210 at time of analysis would be
indicative of the lead-210 (within 50 percent) under the range of
extreme conditions of a polonium-210:lead-210 ratio of one-tenth
(normal ambient, NCRP, 1975) to three.
Therefore, the true lead-210 concentrations should fall
between the reported lead-210 values and the polonium-210 values
at time of collection (Tables E-l and E-2). Furthermore, with
minor exceptions, the lead-210 values should be near the high
side of the range indicated.
90
-------
The notable exceptions are cases where the polonium-210
concentrations are excessively above the reported lead-210 values
(about an order of magnitude). Specifically, the samples from
the Calciner South Side Discharge (from Table 13) and the Old Kiln
Building (from Table 17) indicate that the polonium-210: lead-210
ratio was probably greater than three and thus, especially for
the Calciner sample, it appears the polonium-210 in these samples
at time of analysis was primarily due to polonium-210 collection,
and not due to ingrowth. For these samples the best estimate for
lead-210 is about 2 pCi/m3 or less (up to about five times the
reported lead value).
For these two samples, and for a few other samples with a
polonium-210:lead-210 ratio at time of analysis of five or more,
there is good evidence of elevated polonium-210 and that the
decay and ingrowth corrected values reported in Tables 13 to 21
of the text are correct. For the remaining air sample results,
although there is some evidence of elevated polonium-210, the
individual sampling results are insufficient to clearly denote
even the positive presence of polonium-210 (beyond that ingrown
from the lead-210). That is, a polonium-210:lead-210 ratio of up
to four at time of analysis for these data (if analysis is at 280
days after collection) can be solely due to a factor of five
error in the lead-210 analysis and 75 percent ingrowth of polonium-
210 from the lead. Although such a situation of elevated lead-
210 with no polonium-210 is not likely, much of the data are
insufficient to demonstrate that it could not occur.
In summary, the best estimates for the lead-210 values are
the polonium-210 values at time of analysis, except for the
values denoted above. The polonium-210 values can only be
approximated as discussed above.
91
-------
APPENDIX E
TABLE E-l. ESTIMATED UNCERTAINTY FOR LEAD-210 IN AIR FILTERS
Referenced Table/
Location Description
Reported Lead-210
(pCi/m3 )
Polonium-210 at
Time of Analysis
(pCi/m3)
Table 15-Calciner Area
South Side Discharge
No. 1 Control Room
Table 14-Proportioning Bldg
Lower Level
Lower Level Control Room
Upper Level
Upper Level Control Room
Table 15-Pelletizer Bldg
Operator Level
Upper Level
Upper Control Room
Table 16-Furnace Building
Control Room Level
Tapping Level
Control Room
Table 17
Library
Condenser § Fluid Bed
Control Room
Old Kiln Building
P-4 Loading Area
0.40 ± 0.28
0.17 ± 0.066
0.63 ± 0.11
0.24 ± 0.097
6.6 ± 0.48
0.74 ± 0.10
0.66 ± 0.10
0.60 ± 0.092
0.30 ± 0.07
0.27 ± 0.054
0.25 ± 0.055
0.027± 0.013
17 ± 1.2
0.13 ± 0.026
1.8 ± 0.21
0.85 ± 0.12
5.8 ± 1.4
1.5 ± 0.22
0.78 ± 0.079
0.85 ± 0.073
0.19 ± 0.028
0.36 ± 0.042
0.20 ± 0.033
0.059 ± 0.01
0.019± 0.0086 0.011 ± 0.0035
0.20 ± 0.059 0.22 ± 0.034
0.58 ± 0.089 10 ± 1.1
0.018± 0.0038 0.023 ± 0.0062
* These data are gross results prior to blank filter activity subtraction.
+ The statistical uncertainty of values as expressed by the two-sigma counting
error must be considered in comparing values. For example 5 ± 2 is not
statistically greater than 3 ±,1. Furthermore, counting error does not
include other analytical and sample collection uncertainties.
92
-------
TABLE E-2.
APPENDIX E
ESTIMATED UNCERTAINTY FOR LEAD-210 IN AIR FILTERS*
Polonium-210 at
Referenced Table/ Reported Lead- 210
Location/Description (pCi/m3)
Table 18-Calciner Area
Filter 1
2
3
4
5
Final Filter
Table 19-Proportioning Bldg
Filter 1
2
3
4
5
Final Filter
Table 20-Pelletizer Bldg
Filter 1
2
3
4
5
Final Filter
Table 21-Furnace Building
Filter 1
2
3
4
5
Final Filter
0.57
0.28
0.14
0.13
0.065
1.1
0.33
0.18
0.11
0.14
0.11
0.49
0.073
0.025
0.058
<0.018
0.049
0.088
0.03
0.034
0.027
0.039
0.064
0.061
± 0.057
± 0.056
± 0.041
± 0.038
± 0.036
± 0.071
± 0.082
± 0.076
± 0.073
± 0.075
± 0.074
± 0.086
± 0.027
± 0.017
± 0.026
± 0.027
± 0.026
± 0.011
± 0.011
± 0.011
± 0.012
± 0.013
± 0.013
Time
3.1
0.97
0.51
0.37
0.45
6.1
0.39
0.25
'O.ll
<0.41
0.15
0.53
0.045
0.020
0.019
0.031
0.035
0.10
0.021
0.021
0.021
0.037
0.06
0.076
of Analysis
(PCi/m3)
± 0.26
± 0.089
± 0.057
± 0.052
± 0.056
± 0.43
± 0.066
± 0.12
± 0.051
± 0.067
± 0.012
± 0.0073
± 0.009
± 0.012
± 0.025
± 0.014
± 0.004
± 0.0043
± 0.0042
± 0.0052
± 0.0084
± 0.0098
* These data are gross results prior to blank filter activity subtraction.
+ The statistical uncertainty of values as expressed by the two-sigma counting
error must be considered in comparing values. For example 5 ± 2 is not
statistically greater than 3 ± 1. Furthermore, counting error does not
include other analytical and sample collection uncertainties.
93
. S. GOVERNMENT PRINTING OFFICE: 1978-785-45J
-------
TECHNICAL REPORT DATA
(Please read Inunctions on the reverse before completing)
Technical Note: ORP/LV-77-3
4. TITLE AND SUBTITLE
Radiological Surveys of Idaho Phosphate Ore Processing -
The Thermal Process Plant
3. RECIPIENT'S ACCESSION'NO.
5. REPORT DATE
NOVEMBER 1977
6. PERFORMING ORGANIZATION CODE
8. PERFORMING ORGANIZATION REPORT NO,
9. PERFORMING ORGANIZATION NAME AND ADDRESS "
Office of Radiation Programs-Las Vegas Facility
U.S. Environmental Protection Agency
P. 0. Box 15027 ,
Las Vegas, NV 89114
10. PROGRAM ELEMENT NO.
11. CONTRACT/GRANT NO.
12. SPONSORING AGENCY NAME AND ADDRESS
13. TYPE OF REPORT AND PERIOD COVERED
FINAL
14. SPONSORING AGENCY CODE
15. SUPPLEMENTARY NOTES
. ABSTRACT Kadioiogicai surveys
UL die FMC Corp
Th
l Fn
**"*» * • *%^ v^w J. LSVSX CL L- J.W11 O X11C X lllo--I -T 1. t-JT!T??S?S ^^^
Plant in Pocatello, Idaho indicate slightly elevated ambient levels of natural radio-
activity within the plant. Compared to an estimated natural background an^ull dose
equivalent rate of about 79 mrem, net gamma exposure rates ranged f^fm^m in
general plant areas to 182 mrem per work year on the slag pile? Ambient radon-222
concentrations ranging from 0.17 to 1.4 pCi/1, were meafu?ed in Sve?al SdoSr
locations but 11 pCi/1 was measured in the Control Room of the CondeSer^nd Fluid
Bed Building. Elevated airborne radioactivity concentrations, orderf of ma^itude
greater than measured background concentrations, were also measured in s^SSn^v
areas, with polonium-210 and radium-226 being the most predominant radionuIlidesTof
the natural uranium decay series. Particle size characterization indicates rouehlv
50 percent of the arithmetic, total radioactivity is associated with the pStiSf size
fraction less than one micrometer equivalent aerodynamic diameter. Stack sampling
results also show that appreciable concentrations of the naturally-occurTinr?aSo-
nuclides, particularly polonium-210 and uranium, are being discharged into lie lo^al
environs A general radioactivity balance indicates that the ore is tS source of
.ssentially all of the inlet radioactivity. The slag accounts for essentially all of
the outgoing uranium and radium-226 and up to about 50 percent of the Sad-210 and
polonium-210. The stack effluents account for about 20 percent of tiL total
KEYWORDS AND DOCUMENT ANALYSIS
DESCRIPTORS
Phosphate, Natural Radioactivity
ladium, Radon
]amma Radiation
Particle Size
Airborne Radioactivity
8. DISTRIBUTION STATEMENT
Release to public
EPA Form 2220-1 (9-73)
b.IDENTIFIERS/OPEN ENDED TERMS
Phosphate Industry
Environmental Surveys
Radiation Surveys
19. SECURITY CLASS (ThisReport)
20. SECURITY CLASS (Thispage)
Unclassified
COSATI Field/Group
1806
1807
1808
22. PRICE
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