ORNL/EPA-2
Potential Radiological Impacts of Recovery
of Uranium from Wet-Process Phosphoric Acid
Final Report to the Environmental Protection Agency
W. Davis, Jr.
F. F. Haywood
J. L. Danek
R. E. Moore
E. B. Wagner
E. M. Rupp
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ORNL/EPA-2
Dist. Category UC-11
HEALTH AND SAFETY RESEARCH DIVISION
POTENTIAL RADIOLOGICAL IMPACTS OF RECOVERY
OF URANIUM FROM WET-PROCESS PHOSPHORIC ACID
Final Report to the
Environmental Protection Agency
W. Davis, Jr.*
F. F. Haywood
J. L. Danek
,R. E. Moore
E. B. Wagner
E. M. Rupp
P. J. Walsh, Project Coordinator
J. E. Fitzgerald, Project Officer
Chemical Technology Division.
Date Published: January 1979
Research sponsored by the Environmental
Protection Agency under Interagency Agree-
ment EPA-IAG-D5-E681AG under Union Carbide
Corporation contract W-7405-eng-26 with the
U.S. Department of Energy.
OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
operated by
UNION CARBIDE CORPORATION
for the
DEPARTMENT OF ENERGY
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iii
CONTENTS
Page
Abstract vii
1. Summary and Conclusions 1
2. Introduction 8
2.1 References for Section 2 10
3. Objectives and Assumptions 11
3.1 Objectives 11
3.2 Selection of Model Plants 12
3.2.1 Model 1: Reductive stripping process 12
3.2.2 Model 2: Oxidative stripping process 14
3.2.3 Model 3: Alkylpyrophosphoric acid process 14
3.3 Management of Radioactive Effluents and Wastes 18
3.4 Cost Parameters 19
3.5 Equipment Operation 19
3.6 Plant Siting 20
3.7 Radiological Impact 20
3.8 References for Section 3 21
4. Source Terms for Release of Radioactive Materials 23
4.1 Description of Model 1 23
4.1.1 Potential for accidental releases 27
4.2 Description of Model 2 28
4.2.1 Potential for accidental releases 28
4.3 Description of Model 3 29
4.3.1 Potential for accidental releases. 29
4.4 Composition and Amount of Radioactive Material Processed. . 30
4.5 Description of Treatment Methods for Airborne
Radioactivity 33
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IV
Page
4.5.1 Bag filter 37
4.5.2 Venturi scrubbers 38
4.5.3 Other wet scrubbers 39
4.5.4 High-efficiency particulate air (HEPA) filters. . . 39
4.6 Description of Case Studies and Source Terms 41
4.6.1 Models 1 and 2 42
4.6.2 Model 3 43
4.7. References for Section 4 43
5. Miscellaneous Wastes 50
5.1 References for Section 5 50
6. Costs for Radwaste Treatment 51
6.1 Capital Costs 51
6.1.1 Direct costs 51
6.1.2 Indirect costs 52
6.1.3 Capital cost 52
6.1.4 Annual fixed charge 52
6.1.5 Annual operating and maintenance cost 52
6.1.6 Total annual cost increment for Case 2 53
6.2 References for Section 6 53
7. Onsite and Environmental Monitoring 54
7.1 Description of Site and Specific Release Points 54
7.1.1 Survey plan 56
7.2 Radiological Survey Techniques 61
7.2.1 Isokinetic stack monitoring 61
7.2.2 Particle size measurements 66
7.2.3 Atmospheric spot sampling 70
7.2.4 Soil sampling and analysis 72
7.2.5 Environmental gamma-ray measurements using an
in-situ measuring technique 72
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Page
7.2.6 Liquid and sediment sampling and analysis 73
7.2.7 Gamma-ray exposure rate measurements 73
7.3 Survey Results 74
7.3.1 Release through the stack 74
7.3.2 Determination of size distribution 76
7.3.3 Determination of uranium concentration in air ... 77
7.3.4 Radionuclide concentration in soil 79
7.3.5 Radonuclide concentrations in water and sediments . 85
7.3.6 Background measurement 88
7.4 Conclusion 92
7.5 References for Section 7 93
'v
8. Environmental Impact 94
8.1 Radiological Impact of Airborne Effluents During
Operations 94
8.1.1 Models and assumptions 95
8.1.2 Site specific meteorological, population, and
agricultural data 96
8.1.3 Radiation dose commitments from airborne
effluents 106
8.1.4 Post-operational source terms 112
8.1.5 Post-operational pathways of exposure 113
8.1.6 Estimates of post-operational doses 114
8.2 Positive Radiological Impacts 118
8.2.1 Source terms 120
8.2.2 External radiation dose estimates 124
8.2.3 Internal dose estimates 126
8.3 References for Section 8 130
9. Overview and Recommendations 133
9.1 Summary of Net Impact 133
9.2 Information Gaps and Research Needs 134
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vi
Page
10. Appendixes 137
Appendix A. Description of Sampling Train 139
Appendix B. Pertinent Operating Procedures at URC for Changing
Product Drums (with Respect to Source Sampling) . . 141
Appendix C. Description of Ge(Li) Detector System 143
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POTENTIAL RADIOLOGICAL IMPACTS OF RECOVERY
OF URANIUM FROM WET-PROCESS PHOSPHORIC ACID
W. Davis, Jr. J. L. Danek
F. F. Haywood E. B. Wagner
R. E. Moore E. M. Rupp
P. J. Walsh, Project Coordinator
1. SUMMARY AND CONCLUSIONS
1.1 Summary
A study was made to determine the radiological impacts associated
with recovery of uranium from wet-process (WP) phosphoric acid in central
Florida. Releases of radioactive materials from uranium recovery plants
result in a negative impact (increased dose commitment) on the populations
surrounding the plants. On the other hand, removal of uranium and other
radionuclides from phosphoric acid prevents their distribution on farm
lands, and urban gardens and grasses via fertilizers; this results in a
positive impact (decreased dose commitment) on the associated populations.
This study considers the potential negative impacts of current and pro-
jected recovery processes in a site-specific manner using detailed state-
of-the-art methodologies. Positive impacts are treated in a generic
sense using U.S. average values for important variables such as average
and maximum fertilizer application rates and quantities of radionuclides
in fertilizer.
For purposes of this study, uranium recovery plants are assumed to
recover uranium from WP phosphoric acid. Commercial-scale experience
with such plants is limited. Thus, three model plants were selected based
on current and projected plans as determined by a literature search,
discussions with industry representatives, and discussions with the devel-
opers of processes for removal of uranium from phosphoric acid (Sect. 3).
Each of the models employs solvent extraction of uranium from "black" or
"green" phosphoric acid produced at a WP phosphoric acid plant; this acid
is assumed to contain 30% P2°s and 0.17 g of uranium per liter (Sect. 3.2).
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The major differences in the model plants, for purposes of estimating
radiological impacts, relates to physical location of the two cycles
(Models 1 and 2) , the nature of the product (Model 3 differs from
Models 1 and 2) , and treatment methods for control of airborne particu-
lates. The product of both Models 1 and 2 is assumed to be I^Os- How-
ever, for Model 1, the first cycle of the uranium recovery process is
located at the WP phosphoric acid plant; the second cycle is at a
different location. The nominal output of each first-cycle module is
150 metric tons of uranium per year (MTU/yr) as l^Og. The second cycle
of Model 1 will process product solution from six first-cycle units,
corresponding to a nominal 900 MTU/yr. Thus, Model 1 involves shipping
of first-cycle product solution from WP plants to the central uranium
recovery plant and return of phosphoric acid from the central facility to
the appropriate WP plant. Both cycles of Model 2 are located at the
phosphoric acid plant. Model 2 is assumed to have a capacity of 200 MTU/yr
as UsOs- Model 3 is located at the phosphoric acid plant, but involves
precipitation of uranium as crude UF^ (green salt). Alternatively, the
crude green salt could be processed further and the uranium precipitated
as ammonium diuranate (ADU) . The ADU would be calcined to the product
11303 in a manner similar to the calcination of ammonium uranyl tricarbonate
(AUT) in Models 1 and 2 (Sect. 4.3). Model 3 also is assumed to have a
capacity of 200 MTU/yr U308.
The radionuclides of concern in recovery of uranium from WP phosphoric
acid are members of the natural uranium and thorium decay series . The
following assumptions, summarized in Table 4.3, are made concerning the
radionuclide content of WP phosphoric acid and the products of uranium
recovery (V^OQ or UF4) :
1. The uranium concentration is 0.17 g /liter;
2. The quantity of radium is only 1% of its equilibrium value,
most of the 22^Ra being precipitated along with calcium
during production of WP phosphoric acid;
3. The thorium/uranium ratio is the same as in the marketable
rock;
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4. Freshly precipitated AUT from Models 1 and 2 is assumed
to be free of thorium, radium, and other radioactive
decay products;
5. Assumption 4 is used for Model 3 if the crude UF^. is
subsequently dissolved, extracted, and precipitated as
ADU. However, if the product of Model 3 is crude UF^,
this is assumed to contain all of the thorium, but not
other decay products that are initially present in the
WP phosphoric acid.
Source terms for release of radionuclides are developed (Sect. 4) for
all three model plants and for two treatment methods for airborne particu-
late control, Case 1 and Case 2. For Models 1 and 2, Case 1 control is
based on the use of a bag filter and Case 2 on the use of a bag filter
followed by a HEPA filter. For Model 3, Case 1 control is based on a
venturi scrubber; in Case 2, the venturi scrubber is followed by a HEPA
filter. Essentially all of the radioactive material discharged from any
of the model plants is due to particulate matter that becomes airborne
in the drying, calcination, and product packaging areas (Sect. 4.4). No
significant releases of solid or liquid wastes, radioactive (Sect. 3.3)
or otherwise (Sect. 5), are expected.
The releases of 234U, 235U, and 238U isotopes from Models 1 and 2
are about 3.3, 0.15, and 3.3 mCi/yr, respectively, for Case 1 (bag
filter) and 1.7, 0.08, and 1.7 yCi/yr, respectively, for Case 2 (bag
filter plus HEPA filter). These releases are calculated for a product
rate of 1000 MTU/yr (Table 4.5). For Model 3, releases of the uranium
isotopes are about an order of magnitude higher for Case 1 (venturi
scrubber) if the product is U308. However, if the product is crude UF^,
releases of thorium isotopes (227Th, 228Th, 230Th, 231Th, 232Th, and
234Th) are about 70 mCi/yr for Case 1; about 94% of this is due to 234Th
and 230Th. Addition of HEPA filters (Case 2) reduces the releases by a
factor of 2000 (Table 4.6).
For all three models, Case 1 treatment methods (bag filters for
Models 1 and 2, venturi scrubber for Model 3) are adopted as base cases
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for estimation of radiological impact and for estimation of costs of
additional control. The addition of HEPA filters (Case 2) is the only
measure necessary to meet lowest achievable emission control technology.
The total annual cost increment of Case 2 compared to Case 1 for any
production rate of ^1000 MTU/yr is about $9000.
Field measurements were conducted (Sect. 7) at the only commercial
uranium recovery plant in operation. The plant was in a "shakedown"
period following initial start-up and was still experiencing operational
difficulties that required shutdowns during the field work. However, it
was possible to collect samples and make measurements directed toward
validation of model calculations. Sampling points were chosen to allow
calculations of efficiency of radwaste treatment systems for this facility
as well as quantities of radioactivity in the air, water, and soil in the
vicinity. Analysis of data from field measurements indicates that slight
contamination of the on-site property due to 238U has already occurred.
However, concentrations of this nuclide are below maximum permissible
concentrations (in water) for the general public. In most cases, measured
concentrations of 238U ancj 232Th are typical for reclaimed lands of this
region; decay products of these nuclides appear to be in near-equilibrium
concentrations.
The major contribution to atmospheric releases of radioactivity occurs
in the drumming building, shown in Fig. 7.1. It has been determined that
this release is uranium with natural isotopic composition. Two points
are available for atmospheric releases from this building. The first is
the off-gas treatment system, which consists of a bag filter followed in-
line by a HEPA filter on the roof of the building. This represents the
last control point of release into the atmosphere. The average concentra-
tion of natural uranium released from this stack was 3.42 x 10~12 yCi/ml.
This corresponds to an annual mass emission rate of approximately 2 mg/yr
and an activity emission rate of 1.34 x 10~3 yCi/yr (at a production
rate of about 50 MTU/yr). This concentration is slightly below the maxi-
mum permissible concentration in air for an individual of the general
public. It does not take into account dilution in the atmosphere, which
will greatly reduce the specific activity (Sect. 7).
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The second release occurs from the two room-ventilation fans located
on the roof of the drumming building. Concentrations of natural uranium
being released were variable, probably as a result of disruptions in
plant operations during the field measurements. Of the data accumulated,
the worst case (average of two values) corresponded to 47 x 10~12 uCi of
natural uranium being released to the atmosphere per milliliter. This is
equivalent to an annual mass emission rate of 29 kg/yr. Once again,
dilution in the atmosphere has not been applied.
Efficiency of the off-gas treatment system was determined. Bay and
in-line HEPA filters gave an overall decontamination factor of 2 x 10 .
This is much lower than calculated and much lower than industrial experi-
ence would indicate. However, control methods at the facility are still
being developed, and with proper installation and maintenance, performance
should eventually be in accord with industrial experience. Even as oper-
ated presently, however, radionuclide levels in airborne effluents did not
exceed maximum permissible levels for an individual of the general popula-
tion (Sect. 7).
Aerodynamic particle size range of uranium-bearing dust in the drum-
ming building air was characterized. This building consists of a single
room on each of two floors; freshly calcined powder (assumed to be U308,
see Sect. 3) is packaged in drums on the lower floor. Particle size
measurements on the first and second floors revealed that approximately
70% of the particles on the first floor had diameters larger than 9.2 y,
but approximately 75% of the total airborne uranium on the second floor
consisted of particles with diameters between 1.0 to 5.5 y.
The source terms (Sect. 4) for release of radionuclides from all the
model plants and airborne effluents treatment methods provide input to
models and computer codes (Sect. 8) that are used to estimate doses to
the population surrounding release points during plant operation. As
additional input to the models, population, land use, and meteorological
data were collected for two reference sites in Florida where uranium
recovery plants will be located. The maximum individual dose (individual
located 0.5 mile downwind of the source, in the open, and all food pro-
duced in that area) is from Model 3, Case 1, crude UF^ product; the dose
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is about 250 mrem/yr to bone (Tables 8.. 15 and 8.16) for a 1000-MTU/yr
product rate. Inhalation of 230Th is the primary contributor to this
dose. Addition of Case 2 treatment methods (HEPA filter) reduces this
dose to about 0.1 mrem/yr. The maximum individual doses are much lower
for all other models and treatment cases. The annual dose to the popula-
tion surrounding model uranium recovery plants in Florida could be as
high as 426 person-rem/yr to bone for a 1000-MTU/yr product rate [again
for Model 3, Case 1, crude UF4 (Table 8.18)] with the plant located
about 10 miles south of Tampa. All other dose estimates were lower, and
Case 2 treatment methods would reduce this maximum population dose to
about 0.2 person-rem/yr.
As stated previously, the annual cost of Case 2 compared to Case 1
for Model 3, crude UFi^, is estimated to be $9000. Thus the dollar costs
per person-rem for a 426 person-rem/yr reduction in bone dose would be
only $21/person-rem for a 1000 MTU/yr product rate. The cost/benefit
ratio for all other models would be higher. For example, the cost per
person-rem for Case 2 over Case 1 for Model 3, U308 product would be
about $950/person-rem for a 9 person-rem/yr reduction in lung dose
(Sect. 9), and would be higher for plants with product rates <1000 MTU/yr.
The long-lived uranium and thorium isotopes released during operation
of uranium recovery plants will persist in the environment for thousands
or even millions of years after the plants cease operation. However, the
long-term doses to populations surrounding the plant sites are calculated
to be very low (Sect. 8.1.6) and would be insignificant compared to back-
ground doses.
The potential reduction in doses (positive impact) associated with
reduction of uranium and thorium in fertilizer could easily exceed the
negative local impacts of plant operation (Sect. 8.2). Lack of informa-
tion precluded a detailed assessment of potential positive impacts.
However, even order of magnitude estimates clearly show that the overall
radiological impacts of recovering uranium from WP phosphoric acid could
be positive, especially if best-achievable control technologies are
utilized.
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1.2 Conclusions
Assessment of the negative and positive impacts of recovering uranium
from WP phosphoric acid involves many uncertainties, and thus assumptions,
especially regarding long-term behavior of the natural uranium and thorium
isotopes of concern. Because conservative assumptions are generally made,
calculated impacts would be greater than any actually observed. However,
such conservatism would lead to overestimation of positive impacts as
well as negative impacts; thus, more credits would be taken for positive
impacts than is realistically achievable. If the positive and negative
impacts could be compared on a common basis, then the conservatism would
factor out. Unfortunately, the impacts of uranium recovery cannot be
compared on a point-by-point basis. The major negative impacts are essen-
tially short term in nature and involve populations surroundins the plants
during operation. The major positive impacts are long term and involve
different populations in different locations.
Within the scope of this study, positive impacts are addressed in a
crude generic sense using simplified methodologies to obtain order-of-
magnitude estimates. The maximum potential positive impacts were not
calculated; hence the problem of overestimation of positive impacts due to
conservatism is tempered to some degree. The estimations indicate that
positive impacts of removal of uranium from WP phosphoric acid should
substantially exceed the negative impacts. More detailed study will be
necessary to quantify the net impact.
This study considered only the process of removal of uranium from WP
phosphoric acid and not the phosphate industry in general. Occupational
exposures to personnel in the phosphate industry, uses of reclaimed mining
lands, and the composition of wastes from the phosphate industry in general
may be influenced by the uranium recovery activities. These possible
relationships should be investigated more closely.
The major conclusions concerning the impact of recovery of uranium
from WP phosphoric acid are:
1. The best-achievable control methods involve the addition
of HEPA filters to base-case plants at an annual cost of
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8
about $9000 and will reduce negative impacts to less than
0.2 person-rem/yr for the "worst-case" release. Such an
impact is considered insignificant in comparison to the
impact of doses due to background radiation for the same
population.
2. Addition of HEPA filters is recommended for Model 3,
Case 1, crude UF^, because the maximum individual dose
could approach 250 mrem/yr, and the cost of reducing this
exposure to less than 0.1 mrem/yr is only $36/mrem. The
cost of reducing population dose from a maximum of 426
person-rem/yr to about 0.2 person-rem/yr is only $21/
person-rem. Both of these estimates are based on a
1000 MTU/yr product rate. For a 200 MTU/yr product rate,
those costs would be $180/mrem and $105/person-rem,
respectively> since the doses would be about five times
lower, but the cost of addition of the HEPA filters
would remain essentially constant (Sect. 9).
3. Radiological assessment of the environment of the only
commercial uranium recovery operation [Uranium Recovery
Corp. (URC)] shows that some slight contamination of the
property due to 238U releases has occurred. However, the
levels of this contamination are below maximum permissible
values for the unrestricted public as defined in Title 10,
Code of Federal Regulations, Part 20, 1977. Reclaimed
lands in the immediate vicinity of URC appear to contain
uranium and thorium decay products that are essentially
in equilibrium with the parent nuclides at concentrations
typical for reclaimed land in the area of central Florida
east of Tampa.
2. INTRODUCTION
This study was performed through an Interagency Agreement between the
Environmental Protection Agency and the Energy Research and Development
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Administration (now the Department of Energy) to evaluate the potential
radiological impacts of uranium recovery of phosphates in Florida. The
scope of the work includes a literature search, field radiation measure-
ments, and radiological assessments.
The literature search, discussions with representatives of industry,
and discussions with scientists at Oak Ridge National Laboratory (ORNL),
who developed processes for removal of uranium from phosphoric acid,
indicate that uranium can now be recovered economically from phosphatic
materials only as a by-product of wet-process (WP) phosphoric acid produc-
tion. Thus model processes, which are representative of current or
planned commercial operations, were developed on the basis of using WP
acid as a feed stock to assess the release of radioactive materials to
the environment (source terms). However, the model processes (or base
cases) do not represent the design of any particular facility. The
effectiveness of existing or potential radioactive waste treatment systems
was estimated by comparing source terms for the various systems and the
resulting environmental impacts. Costs for radwaste treatment were esti-
mated on the basis of experience with similar systems used in the nuclear
industry.
The only commercial uranium recovery operation in Florida is that of
the Uranium Recovery Corporation (URC) near Mulberry. Thus, field measure-
ments were limited to one site. Sampling points were chosen to allow
calculations of efficiency of the radwaste treatment systems for this
facility as well as quantities of radioactivity in the air, water, and
soil in the vicinity of the facility. The model processes include one
similar to this facility so that, at least for one case, model calculations
could be supported by measurements. Measurements were also made in criti-
cal areas of the facility (dusting in calcination and packaging operations)
on particle size distribution and to obtain data on the fraction of the
product which becomes airborne. Particle size distribution and amount of
airborne material are key parameters in estimating source terms for any
radwaste treatment system.
The radiological assessment considers both the local impacts of the
uranium recovery facilities and the positive impact. Models 1 and 2
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10
correspond to removing uranium from phosphoric acid and, thus, greatly
reducing the deposition of uranium and its decay products on agricultural
lands and home gardens via fertilizers. Model 3 corresponds to removing
both uranium and thorium from phosphoric acid and greatly reduces deposition
of uranium, thorium, and their decay products on agricultural lands and
home gardens via fertilizers. A simple, generalized methodology was used
to assess positive long-term impacts. Radiological assessment methodologies
2
developed previously for application to various parts of the nuclear fuel
cycle were used to estimate local negative impacts.
A previous study on milling of uranium ores is most closely related
to the present study. However, a major difference (and advantage in terms
of impact) is that in processes for recovery of uranium from phosphoric
acid the uranium is already dissolved, and no additional mining, leaching,
or tailings disposal are required. The major problem in the uranium
recovery plant is from dusting in calcination and packaging operations.
These operations are very similar to the corresponding operations in uranium
ore milling.
2.1 References for Section 2
1. F. J. Hurst, W. D. Arnold, and A. D. Ryon, "Progress and Problems of
Recovering Uranium From Wet-Process Phosphoric Acid," presented at
the 26th Annual Meeting of the Fertilizer Industry Roundtable,
Atlanta, Ga. (October 1976).
2. R. E. Moore, AIRDOS-II Computer Code for Estimating Radiation Dose
to Man From Airborne Radionuclides in Areas Surrounding Nuclear
Facilities, ORNL-5245 (April 1977).
3. M. B. Sears, R. E. Blanco, R. C. Dahlman, G. S. Hill, A. D. Ryon,
and J. P. Witherspoon, Correlation of Radioactive Waste Treatment
Costs and the Environmental Impact of Waste Effluents in the
Nuclear Fuel Cycle for Use in Establishing "As-Low-As-Practicable"
Guides - Milling of Uranium Ores, ORNL/TM-4903, vol. 1 (May 1975).
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11
3. OBJECTIVES AND ASSUMPTIONS
3.1 Objectives
The objective of this study is to evaluate the radiological impact
of the recovery of uranium from WP phosphoric acid using currently avail-
able or conceived treatment systems. There are two opposing aspects of
this evaluation. First, there will be an increase in the radiological
impact in the area surrounding the uranium recovery plant; second, there
will be a decrease in the radiological impact of using phosphate fertilizer
throughout the rest of the United States. This decrease is due to the
removal of uranium or both uranium and thorium from phosphoric acid that
will subsequently be used primarily in the manufacture of fertilizer.
The radiological impact in the vicinity of the uranium recovery plant will
depend on plant capacity. Generally, the amount of waste effluent to be
treated increases with the plant size; that is, larger treatment systems
are required for larger plants. However, the fraction released is assumed
to be essentially the same for large and small systems. A larger total
amount of radioactive material is thus released from the larger unit when
it is operating on the same type (but larger volume) of radioactive
effluent. The calculated total amounts of radioactive materials released
are defined, but these are expected to vary with the plant size. Values
derived in this study for a single size of conceptual plant can be
extrapolated to larger or smaller plants. The quantities of radioactive
wastes were selected on the assumption that a careful internal waste manage-
ment program will be followed (see Sect. 3.5).
Estimates are made of the average radioactive and some nonradioactive
releases and the cost of radioactive waste treatment operations. In
general, the plant will operate continuously; however, continuous operation
is not emphasized since those portions of the plant producing radioactive
discharges (or the whole plant) could be shut down in the event of failure
of a radioactive waste treatment unit. Only potential releases from
normal operations, including anticipated abnormal conditions,* have been
considered in this study.
*An abnormal condition is a transient-process state, or a state resulting
from an unusual incident in which operation parameters affecting control
of radioactive materials (in the gasborne and liquid effluents) move out
of the normal operating range.
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3.2 Selection of the Model Plants
Various processes have been suggested for recovering uranium from
1 2
phosphate rock; most have been summarized by Ross or Ring. Three model
plants were selected for the present study on the basis of plants already
existing or being designed for construction in central Florida. Each of
the models employs solvent extraction of uranium from "black" or "green"
3 4
phosphoric acid produced at a WP phosphoric acid plant; this acid is
5
assumed to contain 30% ?2®5 and 0-17 g of uranium per liter.'
3.2.1 Model 1: Reductive stripping process
CO
The reductive stripping process of Hurst and Grouse is based on
contacting the uranium-bearing phosphoric acid with a solution of 0.5 M
di(2-ethylhexyl) phosphoric acid (DEPA) and 0.125 M trioctylphosphine oxide
(TOPO) in a kerosene-like or aliphatic diluent (hereafter simply called
diluent). This contacting transfers most of the uranium, present as U(VI),
to the organic phase. Uranium is stripped back into an aqueous phase that
contains iron primarily in the form Fe(II), thereby reducing uranium to the
U(IV) state. Product solution from this first cycle (which is less volumi-
nous and more concentrated in uranium than the 30% acid by a factor of about
50) is oxidized, extracted into 0.3 M DEPA—0.075 M TOPO in a diluent,
scrubbed, and stripped with ammonium carbonate. The last steps of the
uranium recovery operation are precipitation as ammonium uranyl tri-
carbonate (AUT), filtration, washing, calcination to 1)303, and packaging
for shipment to a uranium hexafluoride production plant.
The first cycle of the uranium recovery process of Model 1 (Fig. 3.1)
is located at the WP phosphoric acid plant; the second cycle is at a dif-
ferent location. The nominal output of each first-cycle module is 150
MTU/yr as t^Og.* The second cycle of Model 1 will process product solution
from six first-cycle units, corresponding to a nominal 900 MTU/yr. Thus,
Model 1 involves shipping of first-cycle product solution from WP plants
to the central uranium recovery plant, and return of phosphoric acid "from
the central facility to the appropriate WP plant. The predominant radio-
logical impact for Model 1 will be associated with the second cycle.
*For the purpose of this report, it is assumed that the product is rela-
tively pure UsOg. as stated in Sect. 4.1. This assumption is not critical
to the assessment of radiological impact, as noted in Sect. 8.
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ORNL DWG 77-848 Rl
PLANT
VENT
54% P,0, ACID
W P PHOSPHORIC
ACID PLANT
Fig. 3.1. Flowsheet for Model 1. This uses a first cycle of the reductive stripping
process with di(2-ethylhexyl) phosphoric acid (DEPA) and trioctylphosphine oxide (TOPO) at
the WP phosphoric acid plant and a second cycle of DEPA/TOPO extraction at a central
processing plant.
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14
Each first-cycle module corresponds to 95% recovery of uranium from
a 5 to 6 M H3POi,. solution containing 0.17 g of uranium per liter. The
model WP acid plant has a capacity of 352,000 metric tons (MT) or
388,000 short tons (ST) of P20s per year; this nearly equals the average
9-11
capacity of the 12 WP plants in the vicinity of Tampa, Florida
(Table 3.1). The capacity of the first cycle of Model 1 is larger than
the smallest-sized plant considered economically operable for uranium
12
recovery [i.e., 100 short tons of uranium per year (STU/yr)].
3.2.2 Model 2; Oxidative stripping process
An oxidative stripping process has been described ' ' that differs
from the reductive stripping process in the first cycle, but which is the
same in the second cycle. Thus, in the first cycle of Model 2, the 30%
phosphoric acid from the WP plant is contacted with a solution of a
commercial mixture of mono- and dioctylphenylphosphoric acids (OPAP) in a
14
diluent. The OPAP solution, typically 0.32 M, is reported to extract
uranium more efficiently than does the DEPA-TOPO solution of Model 1.
Uranium is stripped back into an aqueous phase of about 10 M t^POi^. contain-
ing an oxidizing agent. This aqueous phase is then processed in a second
cycle of solvent extraction based on DEPA-TOPO, as shown in Fig. 3.2.
All of the Model 2 plant is located at the WP phosphoric acid plant
and is assumed to have a capacity of 200 MTU/yr.
3.2.3 Model 3; Alkylpyrophosphoric acid process
During the 1950s, the U.S. Atomic Energy Commission provided financial
support for the development of several methods for recovering uranium from
Florida phosphates. One of these is based on extraction of uranium from
WP HoPOtf, by a mixture of, primarily, caprylphosphoric acids. This process
involves a single extraction cycle and, with minor modifications, was
developed at both the Bonnie Chemical Plant of International Minerals
and Chemical Corp. (IMC) at Bartow, Florida, and the U.S. Phosphoric
1 fl — 1 ft
Products (USPP) Division of Tennessee Corp. at, Tampa, Florida.
U.S. Phosphoric Products is now owned by Gardinier Co. This process,
Fig. 3.3, involves extraction of uranium in the U(IV) valence state and
-------
Table 3-1 Production capacities of wet-process phosphoric acid plants in central Florida
PI » n •)- lo/^st. i rm
Company
Agrico chem-Williams
Borden chem Co
CF Industries
Cities Service
Kngflhurd M&C-Conserve
Farmla.nd Industries
Gardinier0
Grace ?* USS Agri. Chem.
W. K. Grace
International Minerals
Hoyster Co.
USS />.r;ri. Chem.
County City
Polk
Manatee
Polk
Hillsborough
Hillsborough
Polk
Polk
Hillsborough
Polk
Polk
Polk
Polk
Polk
Polk
Central Florida
U.S. Total
Piercf
Piney Point
Bonnie
Plant City
Tampa .
Nichols
Pierce
Tampa
Bartow
Bartow
Mulberry
Mulberry
Bartow
Ft. Me ado
Total
Pi tv
•> a
population
50
75
Not listed
15,'t51
"79,000
300
"50
:.79,ooo
i:.,89i
1!:,891
Not listed
.".,701
1- , 89D
li , 37U
Actual Production,
U.S. Total d
Thousand MT P00c/yr
From ref. 11.
From ref. 9 .
MT I
J/yre
Annual capacity (Thousand short tons P'.0cj)°
1973
270
175
6Uo
•.50
5)4 li
150
1)55
-
-
315
-
135
90
176
3200
6)438
5919
2900
1170
197)i
270
175
6140
•50
-
150
'•55
51. 1;
-
330
-
135
90
17(".
3260
6693 .
6186
2960
1190
1975
300
175
6Uo
6: '5
-
150
1,55
5)4)4
-
330
750
135
90
176
It 370
8638
6889
Central
3960
1600
19Y6
300
375
690
G'.''j
-
150
''55
5)1)1
-
'••.30
750
] '-15
90
170
It It 20
88). S
Florida
ItOlO
1615
1077
300
175
690
C/.5
-
150
1.55
514)1
380
' 330
C50
135
_
176
>t710
9.05)4
1978
300
175
690
(7:5
-
150
'155
5)4)1
380
3-;o
750
135
_
176
It7l0
905)4
1979
300
175
690
(V.5
-
150
)i55
5)4),
380
330
750
135
-
17')
1)710
205))
J'j80
300
175
690
U Og/ST
l-n
-------
ORNL DWG 77-849 R2
3 *{ MANUFACTURE/
94X P,O, ACI
P PHOSPHORIC
ACID PLANT
SECOND CYCLE
Fig. 3.2. Flowsheet for Model 2. This uses a first cycle of oxidative stripping with
octylphenylphosphoric acids (OPAP) and a second cycle of DEPA/TOPO both at the WP phosphoric
acid plant.
-------
ORNL DWG 77-090
ORGANIC
PHOSPHATE
IN KEROSENE
H2S04
34% P20SACID
Fig. 3.3. Flowsheet for Model 3. This uses a single-cycle of alkylpyrophosphoric acid
(APPA) extraction prior to UF^ precipitation; alternatively, the UF^ is dissolved for further
purification.
-------
18
its subsequent precipitation as ul\ (green salt). The IMC process has
been described as being subject to many difficulties including:
(1) degradation of as much as 1/3 of the inventory of pyrophosphoric acids
to orthophosphoric acids in a day; (2) loss of kerosene equal to 1% of the
volume of acid treated under the best of conditions; (3) severe emulsion
problems that slowed separation of organic and aqueous phases; and
(4) fouling of the iron used to reduce uranium to the tetravalent step.
Some of these problems were alleviated by use of centrifugal phase separa-
tors, keeping reducing-iron surfaces relatively clean by performing the
reduction in a ball-mill type of operation, and cooling and clarifying
the WP phosphoric acid before extraction; however, the process was quite
difficult to control.
Many of the technical difficulties reportedly have been solved since
the late 1950s and the alkylpyrophosphoric acid process is expected to
contribute to recovery of uranium from WP phosphoric acid in a few years.
Model 3 is located at the WP phosphoric acid plant and is assumed to have
a capacity of 200 MTU/yr.
3.3 Management of Radioactive Effluents and Wastes
The flowsheets in this study illustrate very low, but not "zero"
release of radionuclides.
Airborne effluents. Airborne effluents consist primarily of process
product dusts that become airborne in dust collector ducts during the final
drumming operation in the plant. In Models 1 and 2, this dust will be
11303; in Model 3, the dust will be uT<\ plus some thorium fluorides. There
will also be very small quantities of uranium (and thorium in the case of
Model 3) decay products in these dusts, because the solvent extraction
product occasionally will not be calcined or dried immediately. Ventilation
air is assumed to be discharged to the environment through a HEPA filter.
In the calculation of source terms (Sect. 4), it is assumed that radio-
activity in the ventilation exhaust is negligible in comparison with that
in air discharged from the drumming equipment.
-------
19
Liquid effluents. There will be no intentional release of any
radioactive liquid wastes to the environment. Nonradioactive liquids,
including process cooling water, will be released (Sect. 5). It should
be noted that phosphoric acid depleted in uranium is the primary raffinate
from solvent extraction operations. This acid is returned to the WP
phosphoric acid plant.
Solid wastes. Solid wastes will be generated in the second cycle of
solvent extraction and in the calcination and u^Os-packaging operations.
Radioactive contamination will be due primarily to natural uranium (and
thorium in Model 3) and its decay products.
3.4 Cost Parameters
The base cases are representative of concepts described in the litera-
ture, scaled to a size similar to that described by Ross and compatible
with analyses of effluent samples taken during the course of this study.
Capital and annual costs are estimated for only one type of waste effluent
treatment beyond base cases. The calculations of incremental annual costs
(Sect. 6) and the incremental changes in the environmental impact (Sect. 8)
are correlated in terms of cost/benefit ratios in Sect. 9. The estimated
costs are based on a new plant; backfitting costs could be readily calcu-
lated also.
The capital costs of the model plants are expected to be significantly
different; however, product cost, in terms of dollars per metric ton of
uranium, may be almost the same for all model plants.
3.5 Equipment Operation
For the purpose of this study, it is assumed that all chemical and
radioactive waste streams will be passed through all equipment installed
for their treatment, even though chemical or radioactive impurity concentra-
tions are below "permissible" licensing levels. In particular, it is
assumed that no treatment system will be bypassed. The equipment is ade-
quately sized to ensure a high level of operating flexibility and
efficiency.
-------
20
3.6 Plant Siting
The model plants are located east and south of Tampa Bay in Florida.
The two largest cities in this area are Tampa and St. Petersburg, with
populations of about 279,000 and 236,000 respectively. Other cities
(and their populations) within approximately 50 miles of the site of the
model plants are Clearwater (77,500), Lakeland (47,500), Sarasota (46,000),
Bradenton (26,000), and Winterhaven (17,600). Rivers in the area include
the Alfia, Hillsborough, Manatee, Peace, Little Manatee, and Withlacoochee.
Also within 50 miles of the site are many lakes and the wetlands north of
Lakeland and Winterhaven and those east of Sarasota. Population distribu-
tion is based on model plants being located near Mulberry (population, 2700)
in Polk County and south of Tampa. Site selection is described in detail
in Sect. 8.
3.7 Radiological Impact
Radiation doses to the populations surrounding the model plant are
estimated using procedures developed by the staff of the Health and Safety
Research Division at ORNL. Pathways are considered for external radiation
dose from sources outside the body and for internal dose from sources in
the body. Immersion in the airborne effluents as they are diluted and
dispersed leads to external exposure, and inhalation causes internal
exposure. The deposition of radioactive particulates on the land surface
leads to direct external exposure and to internal exposure by the ingestion
of food products through various food chains. Similarly, swimming in
waters containing radionuclides can lead to external exposure, whereas the
harvest of fish or drinking from the waters can lead to internal exposure.
The estimated radiation doses to individuals and to the human popula-
tion are calculated for annular distances out to 55 miles in 22.5° sectors
using the site parameters listed in Sect. 8. Population annual average
doses (person-rem/yr), the sum of the doses to all individuals in the
population considered, are calculated for the total body and for individual
organs. Similar calculations were made of the adult maximum annual total-
body and organ doses (mrem/yr). Details of dose models, assumptions, and
methods are given in Sect. 8.
-------
21
3.8 References for Section 3
1. R. C. Ross, "Uranium Recovery from Phosphoric Acid Nears Reality as
a Commercial Uranium Source," Eng. Min. J. 176, 80 (December 1975).
2. R. J. Ring, Manufacture of Phosphatic Fertilizers and Recovery of
Byproduct Uranium - A Review, AAEC/E355 (November 1975).
3. A. V. Slack, ed., Phosphoric Acid, Fertilizer Science and Technology
Series, Marcel Dekker, N.Y., 1968.
4. T. P. Hignett, Characteristics of the World Fertilizer Industry —
Phosphatic Fertilizers, Tennessee Valley Authority, Report No. S-422,
prepared for use at United Nations International Symposium on
Industrial Development, Athens, Greece (December 1967).
5. F. J. Hurst, D. J. Grouse, and K. B. Brown, Solvent Extraction of
Uranium from Wet-Process Phosphoric Acid, ORNL/TM-2522 (April 1969).
6. F. J. Hurst, D. J. Grouse, and K. B. Brown, "Recovery of Uranium
from Wet-Process Phosphoric Acid," Ind. Eng. Chem., Process Des.
Dev. 11, 122 (1972).
7. F. J. Hurst and D. J. Grouse, "Reductive Stripping Process for the
Recovery of Uranium from Wet-Process Phosphoric Acid," U.S.
Patent 3,711,591 (Jan. 16, 1973).
8. F. J. Hurst, W. D. Arnold, and A. D. Ryon, "Recovering Uranium from
Wet-Process Phosphoric Acid," Chem. Eng. 84_56, (Jan. 3, 1977).
9. E. A. Harre, M. N. Goodson, and J. D. Bridges, "Fertilizer Trends —
1976," Tennessee Valley Authority, Bulletin Y-lll (March 1977).
10. J. R. Douglas, Jr., "World Phosphate Fertilizer Industry at Crossroads,"
Tennessee Valley Authority, Report No. Z-67, presented at Phosphate-
Sulfur Symposium, John's Island, Fla., Jan. 15-16, 1976.
11. 1976 Commercial Atlas and Marketing Guide, 107th ed., Rand McNally,
Chicago, 1976.
12. C. L. Bieniewski, F. H. Persse, and E. F. Brauch, Availability of
Uranium at Various Prices from Resources in the United States, U.S.
Dept. of the Interior, 1C 8501 (1971).
-------
22
13. F. J. Hurst and D. J. Grouse, "Oxidative Stripping Process for the
Recovery of Uranium from Wet-Process Phosphoric Acid," U.S.
Patent 3,835,214 (Sept. 10, 1974).
14. F. J. Hurst and D. J. Grouse, "Recovery of Uranium from Wet-Process
Phosphoric Acid with Octylphenylphosphoric Acid," Ind. Eng. Chem.,
Proc. Des. Dev. 13_, 286 (1974).
15. B. F. Greek, 0. W. Allen, and D. E. Tynan, "Uranium Recovery from
Wet-Process Phosphoric Acid," Ind. Eng. Chem. 49, 628 (1957).
16. P. D. V. Manning, I. M. LeBaron, and F. Crampton, "Recovery from
Phosphate Rock," pp. 375-86 in Uranium Ore Processing, J. W. Clegg
and D. D. Foley, eds., Addison-Wesley, Reading, Mass., 1958.
17. R. H. Kennedy, "Recovery of Uranium from Low-Grade Sandstone Ores
and Phosphate Rock," pp. 216-26 in Proceedings of a Panel on
Processing of Low-Grade Uranium Ores, June 27 — July 1 1966, IAEA,
Vienna.
18. C. S. Cronan, "Capryl Pyrophosphate Ester Extracts Uranium from Wet-
Process Phosphoric Acid," Chem. -Eng. 66(9), 108 (1959).
-------
23
4. SOURCE TERMS FOR RELEASE OF RADIOACTIVE MATERIALS
Models 1 and 2 are similar in that each consists of two cycles of
solvent extraction. In the absence of production plant data, it is assumed
that the reductive (Model 1) and oxidative (Model 2) stripping processes
are technically equal and yield a product from the second cycle that is
free of thorium impurities which are contained in the WP phosphoric acid.
On the other hand, the UF^ from the alkylpyrophosphoric acid process of
Model 3 is assumed to contain all of the thorium initially in the WP acid.
All model plants are assumed to be located in west-central Florida.
Source terms, which specify the annual releases of radioactive materials,
are based partly on analyses of the operation of uranium mills and partly
on additional measurements made for this study (Sect. 7).
4.1 Description of Model 1
Model 1, Fig. 3.1, of this study consists of a combination of six
first-cycle extraction units, each located at a WP phosphoric acid plant,
and a second cycle located elsewhere. The product rate from each first-
cycle unit is 150 MTU/yr as l^Og; thus, the capacity of the central unit
is about 900 MTU/yr, or about 1000 STU/yr as U308.
The function of the first cycle of solvent extraction, which is based
2-4
on the reductive-stripping process of Hurst and Grouse, is beneficiation
of the uranium in WP phosphoric acid, increasing its concentration by a
large factor. Typically, the product from this operation will contain
.7 to 10 g of uranium per liter, but the feed contains only about 0.17 g
of uranium per liter. Also in the feed solution are small to large
quantities of dissolved or suspended iron, aluminum, calcium, sulfate,
fluoride, and other chemicals, as shown in Table 4.1 (see Sect. 5 for a
discussion of disposal of miscellaneous solids). The function of the
second cycle is to convert the uranium in the first-cycle product to a
relatively pure form of ^03- This product may be shipped to a uranium
hexafluoride conversion plant for further purification as a step in pre-
paring the uranium for use in nuclear fuels. The 30% phosphoric acid
-------
Table A.I. Chemical compositions and flows of some 30% PzOs black or green
acids In wet-process phosphoric acid plants in central Florida
First-cycle feed
Concentration Relative flow mass flow ratios
Constituent
H3PO,,
P205
Al
A1203
Fe(II)
Fe(total)
Fe203
Ca
CaO
V
U
SO.,
S03
F
Atomic ratios
Al/P
Fe(II)/P
Fe/P
Ca/P
V/P
U/P
S/P
F/P
Units
M
wt %
wt %
lit
wt %
lit
lit
wt %
lit
wt 7,
lit
lit
lit
wt %
lit
wt %
g-atom/g-atom P
Range
5.0 -
(39.3 -
(28.5 -
3 -
0.3 -
6 -
2 -
0.1 -
0.14 -
19 -
21 -
0.018 -
0.0009 -
0.018 -
0.008 -
0.0003 -
0.00010 -
0.033 -
0.18 -
, rates / thousand MT \
Typical Typical0 (MT/MTU) \ 150 MTU recovered /
6.0
45.4)e 41.7
32.9)e 30.2 30.18 1860 - 2980 (2480) f 294 - 470
6 10-50 1.6 - 7.9
1.9 0.67
0.8
12 15 - 100 2.4 - 16
1.1 0.70
4 3-35 0.5 - 5.5
0.1 0.07
0.3 0.4 - 2.6 0.06 - 0.41
0.19 1.0 0.158
31 80 - 370 12 - 58
3.1 1.65
30 90 - 260 14 - 41
2.0 1.88
0.044 0.044 0.015
0.0029
0.043 0.016 0.010
0.020 0.004 0.003
0.0012
0.00016
0.065 . 0.091 0.048
0.32 0.25 0.23
From Hurst and Crouse.
Dihydrate process data from Hignett.
CHemihydrate and anhydrite process data from Yasuda and Miyamoto, Chap. 4, Table 3, p. 319 in ref. 6.
Based on 95% recovery of uranium.
eBased on densities of pure H.3PO.,/H20 solutions at 25°C: p(5 M) = 1.2472 g/ml; p(6 M) = 1.2952 g/ml.
Corresponds to average value used in this report at 0.17 g
-------
25
from the WP plant is clarified and then contacted with a solution contain-
ing 0.5 M DEPA and 0.125 M TOPO in a diluent. Based on the chemical
compositions listed in Table 4.1, the flow rate of the WP acid to the
first-cycle module will be in the range 1.3 to 2.5 m3/min (350 to 660 gpm)
(Table 4.2) for a flow of 158 MTU/yr to the first cycle. Uranium extracted
in this cycle will be stripped back into an aqueous phase by a solution
containing iron that is primarily in the Fe(II) valence state. This leads
2+
to a reduction of U02 > the predominant form of extracted uranium, to the
U(IV) valence state. The reduced form of iron in the strip solution is
obtained by reducing some of the iron impurity already in the black or
green acid (Table 4.1). This is accomplished by contacting it with iron
metal (Fig. 3.1).
Table 4.2. Stream concentrations and flow rates based on a
first-cycle product rate of 150 MTU/yr
Stream 1
From WP plant to first-
cycle module
Stream 2,
Uranium cone, g/£
H3POit cone, M
P205 cone, wt %
Uranium flow, MT/yr
P205 flow, thousand MT/yr
Solution density, g/ml (MT/m3)
Solution flow,
thousand MT/yr
thousand m3/yr
short tons/day
gal/min
Range
0.14 - 0.19
5-6
28.5 - 32.9
294 - 470
1.247 - 1.295
895 -1650
690 - 1320
2700 - 5000
350 - 660
Average
0.17
5.35
30
158
352
1.264
1170
930
3540
470
rirst
second
7-10
6.
32.9
150
6.4 -
1.295
19.4 -
15.0 -
58.6 -
7.5 -
-co-
cycle
9.1
27.8
21.4
83.7
10.8
-------
26
Product from the first cycle will contain 7 to 10 g of uranium per
liter; some volumetric and mass flow rates are given in Table 4.2.
Uranium from the first cycle will be oxidized to the U(VI) state before
being fed to the second cycle.
Oxidized uranium will be extracted with a solvent similar to that
used in the first cycle, except that the concentrations of DEPA and TOPO
will be about 0.3 M and 0.075 M, respectively, instead of the 0.5 M and
0.125 M values used in the first cycle. Raffinate from the second-cycle
extractor will be sent back to the first cycle; the loaded organic phase
will be scrubbed with a small volume of water to remove phosphoric acid
extracted or entrained in the organic phase. The organic solution from
the scrubbing operation will then be contacted with ammonium carbonate
solution under conditions that cause precipitation of uranium as a rapidly
separating ammonium uranyl tricarbonate (AUT). After removal of solution
by filtration or centrifugation, the AUT will be dried and then calcined
to U308.
This model involves trucking 58 to 84 ST of solution per day in each
direction between each first-cycle module and the central plant (Table 4.2)
This corresponds to 23 to 34 tank trucks per day to and from the central
plant, based on 15 ST of acid per truck, or 4 to 6 tank trucks per day
to and from each first-cycle unit.
It will be necessary to minimize the difference in phosphoric acid
flow to and from the central plant to ensure that the number of truckloads
in each direction is the same. To achieve this objective, an evaporator
is used in the second cycle to concentrate the 0.1 M t^PO^ from the
scrubbing operation to 6 M or higher. Water product from this evaporator
will be cooled and recycled to the scrubbing system (Fig. 3.1).
Off-site releases of radioactive materials consist of airborne dusts
and radon gas, primarily from the second cycle of the model plants, as
shown in Fig. 3.1. Depending on the geology and water table, there is
potential for underground migration of radioactive materials in seepage
of liquid accidentally released. However, radioactive liquids will not
be intentionally released.
-------
27
4.1.1 Potential for accidental releases
The only accident likely to release radioactive materials in signifi-
cant amounts involves a truck traveling from a WP acid plant to the central
plant. Accidents involving trucks have been analyzed according to overall
frequency and severity; data are continuously collected and analyzed
by the U.S. Department of Transportation and by the Florida Department
12
of Highway Safety and Motor Vehicles. The probability of an accident
occurring in transporting WP phosphoric acid from the WP plant to the
central second-cycle plant is only about one per million vehicle miles.
It is estimated that the average distance between the six first-cycle
modules and the central plant will not exceed 10 miles. Thus, the 23 to
34 shipments per day, 365 days/yr, correspond to 84,000 to 124,000 vehicle
miles per year, and to the probability of one accident every 8 to 12 yr.
Only about one accident in 100 will be severe; ' thus the probability
of spillage of a major portion of the 15 ST of phosphoric acid and of the
contained 70 to 100 kg of uranium during 30 to 40 yr of operation is less
than 0.05.
Other data pertaining to traffic accidents involving hazardous mate-
rials, including phosphoric acid, are available from the Department of
Transportation. The following table shows a list of some data concerning
the number and severity of accidents in Florida involving truck transporta-
tion of phosphoric acid. If statistics were available on the number of
shipments and total miles traveled in this form of transportation, then
the tabular data could be used to calculate specific accident severity
and frequency rates. Unfortunately, such statistics do not appear to be
available.
Year
1973
1974
1975
1976
No. of
accidents
0
1
2
1
No. of
people
killed
0
0
1
0
No. of
people
injured
0
2
5
0
Property
damage
($)
0
10,000
55,000
6,000
-------
28
4.2 Description of Model 2
Model 2, Fig. 3.2, also includes two cycles of solvent extraction.
1*5 1 /
The first cycle uses the oxidative stripping process of Hurst and Grouse '
which employs octylphenylphosphoric acid (OPAP) in a diluent; the second
cycle uses the same DEPA-TOPO combination that is used in the second cycle
of Model 1. However, Model 2 is located entirely at the WP phosphoric
acid plant and has a product rate of 200 MTU/yr.
The 30% phosphoric acid from the WP plant is clarified and then con-
tacted with a solution containing 0.3 to 0.4 M OPAP in diluent. Based on
chemical compositions listed in Table 4.1, and an assumed 95% recovery of
uranium, the flow rate of WP acid to this first cycle will be in the range
1.8 to 3.3 m3/min (470 to 880 gpm) for a flow of 210 MTU/yr. Any uranium
not in the U(VI) state will be oxidized during the stripping operation that
will be performed with 10 M HgPO^ containing an oxidant. Before being fed
into the second cycle of solvent extraction that is performed with DEPA-
TOPO in a diluent, the product from the first cycle will be diluted with
water or dilute ^PO^ to a concentration of about 6 M ^PO^ and 7 to 10 g
of uranium per liter. The final operations in the model are the same as
in Model 1, namely, precipitation of AUT from the DEPA-TOPO solution by
addition of ammonium carbonate, separation of the precipitate followed by
its drying, calcination to l^Og, and drumming of the
Off-site releases of radioactive materials consist of airborne dusts
and radon gas, again primarily from the second-cycle drumming operations.
As in Model 1, the potential for underground migration of radioactive
materials in liquids accidentally released will depend on geology and the
level of the water table. Liquids containing radiologically significant
impurities will not be released intentionally.
4.2.1 Potential for accidental releases
Accidental releases of WP phosphoric acid at the plant do not appear
to have been of any consequence in the past and should not be so during
the uranium recovery operations. Since all of these operations will be
performed at the WP plant in Model 2, there will be no potential for
tractor-trailer accidents such as might occur with Model 1.
-------
29
4.3 Description of Model 3
Model 3 is based on extraction of uranium by alkylpyrophosphates in
a process sponsored by the AEC during the 1950s. Cronin has
described the use of caprylpyrophosphate at U.S. Phosphoric Products Co.
(now a division of Gardinier Co.); Greek et al. and Manning et al.
described the decylpyrophosphate process of the International Minerals
18
and Chemical Corp.; Kennedy described some of the difficulties encoun-
tered with this process.
This model (Fig. 3.3) involves the use of a 5% solution of alkyl-
pyrophosphate in a kerosene diluent to extract U(IV), the valence state
obtained by reduction of U(VI) and Fe(III) with scrap iron. Most of the
calcium in this organic extract is precipitated as CaSOij by the addition
of sulfuric acid. After decantation, HF solution, a component of the WP
phosphoric acid, is added to the U(IV)-rich organic solution, thereby
inducing precipitation of uranium as a crude green salt, UF^. This solid
is separated from liquids by centrifugation and may then be dried, drummed,
and shipped to a UFg conversion plant where additional purification could
be obtained by distilling the UF6. Alternatively, the crude green salt
could be dissolved in nitric acid, which would result in uranium being
oxidized to the U(VI) valence state. Uranium could be extracted from this
solution with any one of several solvents. Finally, by adding ammonium
hydroxide, ammonium diuranate (ADU) would be precipitated leaving much of
the iron and other impurities in the aqueous solution. The ADU would be
calcined to the product H^OQ in a manner similar to the calcination of AUT
in Models 1 and 2.
4.3.1 Potential for accidental releases
In common with Model 2, all operations at the Model 3 plant are per-
formed at the WP plant. Thus, accidental releases of uranium-bearing
solutions are not expected to have any environmental consequence (see
Sect. 8.1).
-------
30
4.4 Composition and Amount of Radioactive
Material Processed
Model 1 represents a combination of six reductive-stripping, first-
cycle solvent extraction units, each at a WP phosphoric acid plant and
each producing 150 MTU/yr. The second cycle of this model is located at
a different site and will process 900 MTU/yr from the six first-cycle
modules. Models 2 and 3 represent different processes but have the same
capacity, 200 MTU/yr. Each of these models is located entirely at the
WP acid plant. It is assumed that the environmental impact due to
releases of radioactive materials originates solely from operations in
which the uranium product is precipitated, dried, calcined, and drummed.
In the base cases of Models 1 and 2 of this study, the AUT is transferred
to a furnace where it is dried and calcined; it is then conveyed to a
drumming station (Fig. 4.1). Most of the airborne uranium oxide will
originate in the drumming operation. Model 3 involves the precipitation
of U(IV) as crude UF^. This may either be dried and drummed or it may be
dissolved in nitric acid, extracted a second time as U(VI) , precipitated
as ammonium diuranate (ADU) , dried, calcined, and drummed. Either product
would be shipped to a UF6 conversion plant.
The following assumptions can be made concerning the radionuclide
content of the phosphoric acid solution fed to the first-cycle extraction
unit: (1) the quantity of 226Ra is only 1% of that which would be in
secular equilibrium in the 238U decay chain; and (2) the thorium/uranium
19-22
ratio is the same as in the marketable rock. As would be expected,
radium largely remains with the calcium sulfate rather than dissolving
in the phosphoric acid solution. These assumptions are summarized in
Table 4.3.
Source terms presented below are based partly on measurements of
bench-scale experiments and partly on a conservative assumption. Some
searches for radium in fresh H^OQ from the second cycle of Models 1 or 2
23
failed to detect this element. In addition, extraction coefficients do
not favor carry-over of thorium in the two-cycle processes. Thus, freshly
precipitated AUT from Models 1 and 2 are assumed to be free of thorium,
radium, and other radioactive decay products. This same assumption is
-------
31
FLEXIBLE
COUPLING
ORNL-DWG 77-1302
[VENT
BAG
FILTER
55 GAL.
DRUM
WEIGHING SCALES
Fig. 4.1. Drumming station at each of the model uranium recovery
plants.
-------
Table 4.3. Principal radioactive nuclides in phosphoric acid solution received
at first-cycle extraction system
Nuclide
mass
Element No .
U
234
235
238
Th
227
228
230
231
232
234
Ra 226
Nuclide
mass3
(g/g-mole)
238.0289
234.0409
235.0439
238.0508
227.0278
228.0287
230.0331
(231.03)
232.0382
(234.04)
226.0254
Half-
life3
(years)
2.47E+05
7.1 E+08
4.51E+09
(18.5)b
1.913
8.0 E+04
(1.06)b
1.41E+10
(24.1)b
1600
Isotopic
distrib.
(at. %)
100.0000
0.0055
0.720
99.28
100.0000
1.65E-9
1.89E-8
5.29E-2
8.85E-12
99.95
4.46E-10
100.
Quantity
(g/MTU)
1 . OOE+6
5.41E+1
7 . 11E+3
9.93E+5
3.23E+4
5.20E-7
6.00E-6
1.68E+1
2.86E-7
3.23E+4
1.44E-5
3.34E-3
present
(g-atom/MTU)
4.20E+3
2.31E-1
3.03E+2
4.71E+3
1.39E+2
2.29E-9
2.64E-8
7.36E-2
1.24E-9
1.39E+2
6.15E-8
1.48E-5
Activity
(Ci/MTU)
6.81E-1
3.33E-1
1.52E-2
3.33E-1
7.06E-1
1.61E-2
4.92E-3
3.33E-1
(1.52E-2)d
3.55E-3
(3.33E-l)d
3.33E-36
Reference
ratio
(Ci/Ci 238U)
0.049°
0.015°
1.03°
0.011°
o.oie
Nuclide masses and half-lives are taken from ref. 21, uranium isotopic abundances are taken from ref. 22.
Half-life is in days.
For marketable rock from Table 1 of ref. 19.
Assumed to be the same as its a-emitting parent.
8This is 1% of the equilibrium value.
-------
33
used for Model 3 if the crude UF^ is subsequently dissolved, extracted,
and precipitated as ADU. However, if the product from Model 3 is crude
UF^, this is assumed to contain all the thorium initially in the WP
phosphoric acid but not other decay products.
4.5 Description of Treatment Methods
for Airborne Radioactivity
Essentially all of the radioactive material discharged from any of
the model plants is associated with particulate matter that becomes air-
borne in the drying, calcination, and product packaging areas. The excep-
tion to this is a small contribution due to 220Rn and 222Rn generated
during storage of the product before shipment to a fluorination plant for
conversion to UFg. The decay chains leading to radon formation are shown
in Fig. 4.2, and quantities of some of the nuclides formed during decay
are shown as a function of time in Figs. 4.3 and 4.4 for 1 MTU.
The drumming station at each of the model plants is shown in Fig. 4.1.
The station is assumed to be the same for all plants, with a capacity of
1000 MTU/yr, and an air-flow rate (when the blower is operating) to the
primary filter of no more than a few hundred cubic feet per minute. Thus
only a single 2-ft x 2-ft x 12-in. -thick HEPA filter will be needed for
the Case 2 analysis described below. Except for scale and the type of
primary filter, the conceptual drumming station is essentially the same
as that used at the Cincinnati plant of the National Lead Company of Ohio.21*
About 1% of the D^OQ or UFtf product is assumed* to become airborne in
the drying, calcination, and product packaging areas; this corresponds to
about 71 Ib of U^OQ per day, or 80 Ib of UF^ per day, for a product rate
of 1000 MTU/yr. The efficiencies with which air-filtration equipment will
remove this material will depend on its particle size distribution.
*The average loss of U308 at uranium mills was 0.02% (Table 4.2 of ref. 1)
and at least 98% of airborne D^OQ was recovered (about 2% of airborne \]$Q
was released) before being exhausted through a vent in the roof (Sect.
9.4.2 of ref. 1). These two values imply that less than 1% of the V^OQ
product became airborne during solids-handling operations. In Sect. 7 it
will be shown that data obtained as part of this study correspond to only
0.005% of the uranium product becoming airborne for the U,RC recovery
process.
-------
URANIUM-238 DECAY SERIES
THORIUM-232 DECAY SERIES
URANIUM-235 DECAY SERIES
Fig. 4.2. Nuclides of the 238U, 232Th, and 235U decay chains.
-------
35
ORNL DWG.77-4824
10'
\
6
H
O
Q
O
cr
Q.
z
>-
H
>
h-
O
I- 2
- I0'q
o
Q
10'
,-3
10'
,-4
TOTAL
238
•235,
235U. 23lTh
0
10 20 30 40 50
TIME SINCE PURIFICATION (DAYS)
60
70
Fig. 4.3. Buildup of radioactivity after separating uranium in
30% acid from all other radioactive elements.
-------
36
10
.0° h-
i 10-'
o
cr
Q.
>-
\-
>
o 10
<
o
Q
<
o:
-2
10
-3
10
.-4
"
227.
Th
ORNL DWG. 77-1068
TOTAL
"V 23°Th, 234Th
223_ 2I9_
Ra. Rn
224
232
Th
220
Ra
Rn
I
I
10 20 30 40 50
TIME SINCE PURIFICATION (days)
60
70
Fig. 4.4. Buildup of radioactivity after separating uranium and
thorium in 30% acid from all other radioactive elements.
-------
37
Table 4.4 shows a summary of the data concerning this parameter for
25
western uranium mills. Since there are no similar data for a plant
recovering uranium from phosphoric acid as 11303 or UF4 by any of the
models of this report, except for the data obtained as part of this
study (Sect. 7), it is assumed that the dusts of each model plant have
essentially the same size distribution as that found in the western mills
and that data of Table 4.4 are applicable. To reduce the quantity of
radioactive materails (primarily dusts) discharged from the uranium
recovery plant, the following equipment is considered: bag filter,
venturi scrubber, wet scrubber, and HEPA filter. The first three of these
and many other dust collection systems have been extensively described in
.. 26~28
other reports.
Table 4.4. Average median particle size of
dust samples collected inside uranium mills
Concentrate av mass
Mill medium size (ym)
c
E
F
G
I
L
Av
2.0
2.9
2.3
2.4
3.1
2.2
2.5
•3
Data derived from ref. 25.
4.5.1 Bag filter
The bag filtc
1 ym from cooled, dry streams. Dusty gas flows through a filter made of
9 f\—"}0
The bag filter is quite efficient for removing fine dusts to
-------
38
compressed felt and deposits particles in the voids. As the voids fill,
a cake builds up on the fabric surface and the pressure drop increases
to a point where the deposited dust must be removed by a reverse jet of
air from the "clean" side. Cleaning may be either by pulsing a jet of
compressed air through valves controlled by a timer or by a reverse jet
through a blow ring which moves continuously up and down the bags. Very
high concentrations of dust can be handled because the maximum period
between cleaning cycles is only a few seconds. High concentrations of
dust are usually an advantage, since the deposited dust tends to be
dislodged in "slabs" rather than redispersed in the gas phase. The pulse-
jet type has proved to be reliable in UF6 plants, with a long bag life
31-33
and relatively little maintenance; this is in contrast to the
31
mechanical problems associated with the blow-ring mechanism. All UFg
plants use bag filters to recover uranium dusts from materials-handling
operations. Primary bag filters are designed to return material to the
process automatically; dust from secondary bag filters is collected in
drums and recycled.
Long-term plant and laboratory investigations have shown that under
typical industrial conditions the reverse-jet bag filter is 99.9% effi-
o/• 9 -t
cient. ' Losses are primarily from leaks around seals or holes in
the bag. Under optimum conditions (i.e., no leaks), the average effi-
ciency of the blow-ring type of bag filters at one uranium refinery was
99.986"%. Efficiencies remain close to 100% for particles down to
i 27
1 pm.
The primary bag filter was assumed to have an efficiency of 99.9%,
and the system of primary plus secondary bag filters an efficiency of
34
99.986%. The second unit would collect dust which had leaked through
the first unit and would ordinarily collect relatively little material.
In the present report, and in the absence of data on the use of bag filters
in processes for recovering uranium from phosphates, a conservative 99.9%
efficiency is assumed for a reverse jet or pulse bag filter.
4.5.2 Venturi scrubber
In common with other dust removal equipment, the venturi scrubber is
capable of high-collection efficiencies in the removal of dusts as small
-------
39
9 A 9 fl
as 1 ym. Its efficiency is dependent on the pressure drop; when
this is increased to the range of several tenths of inches of water gage,
the efficiency for removing 2-ym particles approaches 100%. Venturi
scrubbers are currently being used on the yellow cake dryer at one
uranium mill and on a dry-ore grinding circuit at another.
4.5.3 Other wet scrubbers
During drying and calcining AUT (Models 1 and 2), ammonia and carbon
dioxide will be liberated; similarly, ammonia will be liberated during
the drying and calcining of ADU. These off-gases will carry some small
quantities of uranium-bearing compounds which will be collected in wet
scrubbers used to recover ammonia and carbon dioxide for recycle. Various
9 fi—9 ft
types of wet scrubbers are available for ammonia recovery that will
also remove particulate matter with efficiencies in the range 90% or
greater for particles sized to 1 pm. These include spray towers, impinge-
ment scrubbers, wet dynamic scrubbers, and orifice scrubbers in addition
to the venturi scrubber mentioned above.
4.5.4 High-efficiency particulate air (HEPA) filters
HEPA filters have been used for many years in the nuclear industry to
remove radioactive particles from air streams; they represent the best
achievable control technology for collection of small particles. An
extensive description of their construction, installation, and properties
35
has recently been written. One standard HEPA filter has a cross section
of 2 by 2 ft and a depth of 1 ft, with an air capacity of about 1000 cfm
(Table 3.1 of ref. 35). The filters are installed in banks to achieve the
required system capacity. These filters are expendable (single-use) pleated
mats of Fiberglas paper. They are specified to exhibit a minimum effi-
ciency of 99.97% for 0.3-ym-diam particles and a maximum resistance (when
clean) of 1.0-in. 1^0 pressure when operated at rated airflow. Tests of
filter efficiency are conducted in special facilities which ensure that no
significant leakage occurs around the sides of the filter or through other
bypasses. It is necessary to construct an equally tight filter enclosure
in a field installation to achieve the rated filtration efficiency. The
construction of tight filter enclosures is a difficult engineering task.
Testing of the individual filter banks in place in the enclosure, both
-------
40
before and periodically during the service period by the dioctylphthalate
(DOP) smoke test is required to ensure that no significant leaks are
present in either the filter or the enclosure.
Variables that have been considered in HEPA filter performance analyses
include the particle size distribution of the various plutonium aerosols
"}fi
encountered. A literature survey has not indicated a gross variation
in the range of reported particle sizes in field operations.
Numerous tests have been performed with plutonium aerosols in small
37
laboratory and large-scale field installations. A detailed survey has
found large-scale filter systems in operation at the Rocky Flats Plant
which produced overall mass removal efficiencies of 107 or greater. One
such system showed a removal efficiency of 99.999% across the first two
banks of a system of four HEPA-filter banks in series, 94% across the .
third filter bank, and 83% across the fourth filter bank. The low-
efficiency value for the fourth bank was attributed to probable bypassing
of gases and was not a measure of filter media performance. This system,
which was about 15-years-old, does not represent those presently installed
at Rocky Flats where most of the filter plenums have been replaced or
modified within the last few years. ' The newer plenums are designed
to facilitate testing of individual filters and filter banks and to ensure
that each stage of filtration can be certified to be at least 99.95% effi-
cient with pneumatically generated DOP aerosol. Data obtained for some
four-stage systems at Rocky Flats show efficiencies of >99.99% to 99.998%
for fourth stages and 99.997% to 99.999% for first stages. These effi-
ciencies of production-scale equipment equal (or perhaps exceed) those
40-42
obtained by others in laboratory tests using plutonium aerosols in
small installations that are tightly sealed and tested periodically for
leaks with DOP. Removal efficiencies of at least 99.97% have been observed
for each of three single-filter stages in series. Nuclear Regulatory
Commission Guide 3.12 for the design of plutonium ventilation systems
indicates that removal efficiencies of >99.95% should be obtained for a
single bank of HEPA filters if the installation containing the filters
is constructed according to the recommended guidelines and is tested for
43
leaks after installation of the filters. Consequently, a value of
-------
41
99.95% has been used in this study to represent the rated efficiency of
each HEPA filter.
Performances of air cleaning systems in nuclear facilities during the
44
years 1966-1974 have been summarized. The present analysis presumes that
performances of HEPA-filter installations will not be subject to the design
and operational errors of these earlier systems. However, there is a
potential for mechanical damage of the filters during their initial installa-
tion and during replacement in the enclosures, although such damage will be
located when routine preoperative testing with DOP is employed, as at the
ERDA Rocky Flats Plant. After operations have started, filter efficiency
can be decreased through: (1) attack by corrosive chemicals, such as
hydrogen fluoride;* (2) degradation of the binder for the filter fibers
by condensed moisture;* (3) matting of the retained particles, which
decreases the resistance of the binder to moisture and causes an increase
in pressure drop; (4) degradation by high temperatures (fire);* and
(5) damage by sudden pressure surges. Thus continuous monitoring of the
pressure drop across the filter and periodic testing with DOP are required
to ensure that the filters are operating satisfactorily.
4.6 Description of Case Studies and Source Terms
Case studies described in this section are based on the use of equip-
ment already installed or planned for installation at the URC plant or
other uranium recovery facilities now in the design stage. Similar equip-
ment is already in use in uranium mills in the western United States.
The source terms do not include a contribution of radioactive material from
drumming-room (or other operating-room) air since these rooms are presumed
to be free enough of radioactivity to permit safe entry by plant personnel
without recourse to dust masks. Hence, the relatively large uranium
content in the drumming-room ventilation at URC (Sect. 7.3.3) was probably
^However, filter media that are very resistant to damage by hydrogen-
fluoride, water, and fire have already been produced,^ but not used
in commercial installations. Development of media to resist attack
by other chemicals such as nitrogen oxides is also in process.
-------
42
due to start-up difficulties and was not representative of expected condi-
tions; in addition, URC planned to add a HEPA filter on the discharge of
this ventilation.
4.6.1 Models 1 and 2
Two case studies for both Models 1 and 2 are described in this section.
Case 1 involves use of only a bag filter (reverse-jet or pulse); Case 2
corresponds to the addition of one bank of HEPA filters downstream from
the bag filter. The case studies are based on the assumption that product
from the second extraction cycle initially contains no radioactive nuclides
other than 238U, 235U, and 234U. This assumption is based on chemical
46
analyses that show the thorium content to be below the limit of detect-
ability of 20 ppm.
Case 1. This case is the base case of the uranium recovery process.
It corresponds to the transport of 1 wt % of the plant product to a bag
filter, which removes 99.9 wt % of airborne matter. Quantities of indi-
vidual nuclides and the total quantity of radioactive material leaving the
plant, based on 1000 MTU/yr,.are shown for both models in Table 4.5. This
case corresponds to an overall plant containment factor (CF) of 105 where
CF is defined as uranium production rate/rate of discharge of uranium to
the atmosphere.
Table 4.5. Source terms for uranium Models 1 and 2
immediately after purification
Nuclide
U-234
U-235
U-238
Total
Radioactivity
content
(Ci/MTU)
3.33 E-l
1.52 E-2
3.33 E-l
6.81 E-l
For product
Available
activity
(Ci/yr)
3.33 E+2
1.52 E+l
3.33 E+2
6.81 E+2
rate =
Activity
Case la
3.3 E-3
1.5 E-4
3.3 E-3
6.8 E-3
1000 MTU/yr
emitted (Ci/yr)
Case 2b
1.7 E-6
7.6 E-8
1.7 E-6
3.4 E-6
Based on air transport of 1% of the available activity and a filter bag
efficiency of 99.9 wt %.
Based on the parameters of Case 1 and 99.95 wt-, % efficiency of a HEPA
filter.
-------
43
Case 2. This case represents the addition of HEPA filters having
a dust-removal efficiency of 99.95% downstream from the bag filter. The
overall CF for this case is 2 x 108. The decontamination factor due to
the filter system (DF = rate at which uranium enters the bag filter system/
rate of discharge of uranium to the atmosphere) is 2 x 10^. This is the
case study and DF with which field data in Sect. 7 will be compared.
4.6.2 Model 3
Two case studies for Model 3 that were operated to produce crude UF^
or more-refined l^Og are described in this section. Case 1 involves only
the use of a venturi scrubber; Case 2 corresponds to the addition of one
bank of HEPA filters downstream from the venturi scrubber. The case
studies are based on the assumption that all of the thorium in the WP
acid is carried along during the precipitation of crude UF^ . However, if
this is oxidized and dissolved in nitric acid, it is assumed that ADU,
subsequently precipitated by the addition of ammonia, will initially be
free of thorium.
Case 1. This case is the base case of the uranium recovery process.
It corresponds to the transport of 1 wt % of the plant product to a
venturi scrubber, which removes 99 wt % of airborne matter. Quantities
of individual nuclides and the total quantity of radioactive material
leaving the plant, based on 1000 MTU/yr, are shown for both modes of
operation (UF^ or l^Og product) in Table 4.6. This case corresponds to
an overall plant CF [CF = (uranium + thorium) production rate/rate of
discharge of (uranium + thorium) to the atmosphere in the case of crude
product] of 104.
Case 2. This case represents the addition of HEPA filters with dust
removal efficiency of 99.95% downstream from the venturi scrubber. The
overall CF is 2 x 107.
4.7 References for Section 4
1. M. B. Sears, R. E. Blanco, R. C. Dahlman, G. S. Hill, A. D. Ryon,
and J. P. Witherspoon, Correlation of Radioactive Waste Treatment
Costs and the Environmental Impact of Waste Effluents in the Nuclear
Fuel Cycle for Use in Establishing "As-Low-As-Practicable" Guides —
Milling of Uranium Ores, ORNL/TM-4903, vol. 1 (May 1975).
-------
44
Table 4.6. Source terms for uranium recovery Model 3 immediately
after purification with UF4 or U308 product
Nuclide
U-234
U-235
U-238
Th-227
Th-228
Th-230
Th-231
Th-232
Th-234
Total
Radio-
activity
content
(Ci/MTU)
3.33 E-l
1.52 E-2
3.33 E-l
1.61 E-2
4.92 E-3
3.33 E-l
1.52 E-2
3.55 E-3
3.33 E-l
1.39 E+0
For product
Available
activity
(Ci/yr)
Product is crude UF^
3.33 E+2
1.52 E+l
3.33 E+2
1.61 E+l
4.92 E+0
3.33 E+2
1.52 E+l
3.55 E+0
3.33 E+2
1.39 E+3
rate = 1000 MTU/yr
Activity emitted
Case la
3.3 E-2
1.5 E-3
3.3 E-2
1.6 E-3
4.9 E-4
3.3 E-2
1.5 E-3
3.6 E-4
3.3 E-2
1.4 E-l
(Ci/yr)
Case 2*>
1.7 E-5
7.6 E-7
1.7 E-5
8.0 E-7
2.5 E-7
1.7 E-5
7.6 E-7
1.8 E-7
1.7 E-5
6.9 E-5
U-234
U-235
U-238
Total
Product is semi refined U308
3.33 E-l 3.33 E+2 3.3 E-2
1.52 E-2 1.52 E+l 1.5 E-3
3.33 E-l 3.33 E+2 3.3 E-2
6.81 E-l
6.81 E+2
6.8 E-2
1.7 E-5
7.6 E-7
1.7 E-5
3.4 E-5
QBased on air transport of 1% of available activity and a venturi scrubber
efficiency of 99 wt %.
Based on the parameters of Case 1 and 99.95 wt % efficiency of a HEPA
filter.
-------
45
2. F. J. Hurst, D. J. Grouse, and K. B. Brown, Solvent Extraction of
Uranium from Wet-Process Phosphoric Acid, ORNL/TM-2522 (April 1969).
3. F. J. Hurst, D. J. Grouse, and K. B. Brown, "Recovery of Uranium from
Wet-Process Phosphoric Acid," Ind. Eng. Chem., Proc. Des. Dev. 11, 122
(1972).
4. F. J. Hurst and D. J. Grouse, "Reductive Stripping Process for the
Recovery of Uranium from Wet-Process Phosphoric Acid," U.S.
Patent 3,711,591 (Jan. 16, 1973).
5. T. P. Hignett, "Characteristics of the World Fertilizer Industry —
Phosphatic Fertilizers," Tennessee Valley Authority, Report No. S-442,
prepared for use at United Nations International Symposium on
Industrial Development, Athens, Greece, December 1967.
6. A. V. Slack, ed., Phosphoric Acid, Fertilizer Science and Technology
Series, Marcell Dekker, New York, 1968.
7. Environmental Survey of Transportation of Radioactive Materials to and
from Nuclear Power Plants, WASH-1238 (December 1972).
8. Final Environmental Statement, Light-Water Breeder Reactor Program,
ERDA-1541 (June 1976).
9. Transportation Accident Risks in the Nuclear Power Industry 1975-2020,
EPA-520/3-75-023 (March 1975).
10. L. B. Shappert, W. A. Brobst, J. W. Langhaar, and J. A. Sisler,
"Probability and Consequences of Transportation Accidents Involving
Radioactive-Material Shipments in the Nuclear Fuel Cycle," Nucl.
Safety 14, 597 (1973).
11. R. E. Kidwell and B. V. Chatfield, U.S. Department of Transportation,
personal communication, May 1977.
12. "Standard Summary of Motor Vehicle Accidents in Florida," available
from Department of Highway Safety and Motor Vehicles, Tallahassee,
Florida 32304.
13. F. J. Hurst and D. J. Grouse, "Oxidative Stripping Process for the
Recovery of Uranium from Wet-Process Phosphoric Acid," U.S.
Patent 3,835,214 (Sept. 10, 1974).
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46
14. F. J. Hurst and D. J. Grouse, "Recovery of Uranium from Wet-Process
Phosphoric Acid with Octylphenylphosphoric Acid," Ind. Eng. Chem.,
Proc. Des. Dev. 13, 286 (1974).
15. C. S. Cronan, "Capryl Pyrophosphate Ester Extracts Uranium from Wet-
Process Phosphoric Acid," Chem. Eng. 66, 108 (1959).
16. B. F. Greek, 0. W. Allen, and D. E. Tynan, "Uranium Recovery from
Wet-Process Phosphoric Acid," Ind. Eng. Chem. 49, 628 (1957).
17. P. D. V. Manning, I. M. LeBaron, and F. Crampton, "Recovery from
Phosphate Rock," pp. 375-86 in Uranium Ore Processing, J. W. Clegg
and D. D. Foley, eds., Addison-Wesley, Reading, Mass., 1958.
18. R. H. Kennedy, "Recovery of Uranium from Low-Grade Sandstone Ores
and Phosphate Rock," pp. 216-26 in Proceedings of a Panel on
Processing of Low-Grade Uranium Ores, 27 June — 1 July 1966, IAEA
Vienna.
19. R. J. Guimond and S. T. Windham, Radioactivity Distribution in
Phosphate Products, By-Products, Effluents, and Wastes, U.S.
Environmental Protection Agency, ORP/CSD-75-3 (August 1975).
20. R. J. Guimond, "The Radiological Impact of the Phosphate Industry -
A Federal Perspective," pp. 254-72 in Proceedings of the 8th National
Conference on Radiation Control, May 2-7, 1976, Springfield, 111.
HEW(FDA)77-8021.
21. R. C. Weast, ed., Handbook of Chemistry and Physics, 55th ed.
CRC Press, Cleveland, 1974.
22. N. E. Holden and F. W. Walker, "Chart of the Nuclides," General
Electric Co., Knolls Atomic Power Laboratory, llth ed., April 1972.
23. F. J. Hurst, ORNL, personal communication, November 1976.
24. A. F. Pennak, Director of Engineering, National Lead Company of
Ohio, personal communication, July 1977.
25. W. B. Harris, A. J. Breslin, H. Glauberman, and M. S. Weinstein,
"Environmental Hazards Associated with the Milling of Uranium Ore,"
AMA Arch. Ind. Health 20, 366 (1959) (see Table 8 on p. 374).
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47
26. C. J. Stairmand, "Removal of Grit, Dust, and Fumes from Exhaust
Gases from Chemical Engineering Processes," The Chemical Engineer
194, 310-26 (December 1965).
27. C. J. Stairmand, "Removal of Dust from Gases," pp. 364-402 in
Processes for Air Pollution Control, 2nd ed., G. Nonhebel (ed.),
The Chemical Rubber Co., Cleveland, 1972.
28. G. D. Sargent, "Dust Collection Equipment" Chem. Eng. ^6(2), 130-50
(1969).
29. K. J. Caplan and M. G. Mason, "Efficiency of Reverse-Jet Filters on
Uranium Refining Operations," pp. 77-85 in Air Cleaning Seminar,
Ames Laboratory, Sept. 15-17, 1952, WASH-149 (March 1954).
30. D. S. Ensor, R. G. Hooper, and R. W. Scheck, Determination of the
Fractional Efficiency, Opacity Characteristics, Engineering and
Economic Aspects of a Fabric Filter Operating on a Utility Boiler,
prepared by Meteorology Research, Inc., for Electric Power Research
Institute, EPRI-297 (November 1976).
31. R. Johnson, Lucius Pitkin Co., Metropolis, 111., personal communica-
tion to M. B. Sears, ORNL, Dec. 12, 1974.
32. J. Craig, Engineering Manager, Kerr-McGee Sequoyah UF6 Production
Facility, Okla., personal communication to M. B. Sears, ORNL,
Oct. 16, 1974.
33. J. Thomas, Technical Superintendent, Allied Chemical UF6 Plant,
Metropolis, 111., personal communication to M. B. Sears, ORNL,
Nov. 13, 1974.
34. M. B. Sears, R. E. Blanco, B. C. Finney, G. S. Hill, R. E. Moore,
and J. P. Witherspoon, Correlation of Radioactive Waste Treatment
Costs and the Environmental Impact of Waste Effluents in the Nuclear
Fuel Cycle — Conversion of Yellow Cake to Uranium Hexafluoride.
Part I. The Fluorination — Fractionation Process, ORNL/NUREG/TM-7
(September 1977).
35. C. A. Burchsted, A. B. Fuller, and J. B. Kahn, Nuclear Air Cleaning
Handbook, ERDA 76-21 (1976).
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48
36. W. Davis, Jr., High-Efficiency Particulate Air Filters — State of
the Art Summary Pertaining to Plutonia Aerosols, ORNL/TM-4463
(April 1974).
37. N. Hetland and J. L. Russell, Jr., "Adequacy of Ventilation Exhaust
Filtering System for New Plutonium Facilities," p. 619 in Proceedings
of the 12th AEC Air Cleaning Conference Held in Oak Ridge, Tennessee,
August 28-31, 1972, CONF-720823 (January 1973).
38. F. J. Linck and J. A. Geer, "In-Place Testing of Multiple Stage HEPA
Filters," p. 526 in Proceedings of the 13th AEC Air Cleaning Conference
Held in San Francisco, Calif., August 10-15, 1974, CONF-740807.
39. A. J. Oliver and C. J. Amos, "Ventilation Design for New Plutonium
Recovery Facility," p. 320 in Proceedings of the 13th AEC Air Cleaning
Conference Held in San Francisco, Calif., August 10-15, 1974,
CONF-740807.
40. H. J. Ettinger, J. C. Elder, and M. Gonzales "Size Characteristics
of Plutonium Aerosols," p. 740 in Proceedings of the 12th AEC Air
Cleaning Conference, Held in Oak Ridge, Tenn., August 28-31, 1971,
CONF-720823 (January 1973).
41. H. J. Ettinger, J. C. Elder, and M. Gonzales, Performance of Multiple
HEPA Filters Against Plutonium Aerosols, Progress Report for Period
January 1 through June 30, 1973, LA-5349-PR (July 1973).
42. M. Gonzales, J. C. Elder, and H. J. Ettinger, "Performance of
Multiple HEPA Filters Against Plutonium Aerosols," in Proceedings
of the 13th AEC Air Cleaning Conference Held in San Francisco, Calif.,
August 10-15, 1973, CONF-740807.
43. Nuclear Regulatory Commission, Regulatory Guide 3.12, "General Design
Guide for Ventilation Systems of Plutonium Processing and Fuel
Fabrication Plants" (August 1973).
44. D. W. Moeller, "Performance of Air Cleaning Systems in Nuclear
Facilities," in Semiannual Progress Report, March 1, 1974 — August 31,
1974, Harvard Air Cleaning Laboratory, COO-3409-5 (December 1974).
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49
45. W. I. Belvin, M. A. Krimmel, H. G. Schwalbe, and E. N, Gleaton,
Development of New and Fluoride Resistant HEPA Filter Medium, Final
Report by the Herty Foundation, TID-26649 (Aug. 19, 1975).
46. F. J. Hurst, ORNL, personal communication, March 1977.
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50
5. MISCELLANEOUS WASTES
The operation of any of the model uranium recovery processes described
in this report or any other uranium recovery process will generate miscella-
neous wastes. These include sanitary wastes, packaging materials from
supplies, combustion products from the power plant, oils and greases from
equipment maintenance, discarded process equipment (e.g., HEPA filters,
bag filters, valves, small instruments, laboratory apparatus), protective
clothing, and chemicals (e.g., solids from clarification of WP acid prior
to the first extraction cycle) in the main process waste streams. All of
the solid wastes that are contaminated with natural uranium or thorium or
that are uncontaminated will be placed in on-site landfills such as the
gypsum piles or tailings ponds. Volumes of miscellaneous solid wastes
will be trivial in comparison with volumes of tailings or gypsum piles.
Any equipment moved off site, such as during decommissioning of the
uranium recovery plant, probably would comply with Nuclear Regulatory
Commission Regulatory Guide 1.86.
No liquids containing radioactivity would be discharged to the environ-
ment without prior treatment. Instead they would probably be treated by
the double-liming procedure, as are other liquids from phosphate mining
2
activities, to ensure a pH of 6 to 9, thereby effectively removing
radionuclides.
5.1 References for Section 5
1. Nuclear Regulatory Commission, Regulatory Guide 1.86, "Termination
of Operating Licenses for Nuclear Reactors" (June 1974).
2. R. J. Guimond, "The Radiological Impact of the Phosphate Industry —
A Federal Perspective," pp. 254-72 in Proceedings of the 8th National
Conference on Radiation Control. May 2-7, 1976, Springfield, 111..
HEW (FDA) 77-8021.
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51
6. COSTS FOR RADWASTE TREATMENT
Beyond the base case for each model plant, only one other radwaste
treatment is analyzed; it involves the addition of two HEPA filters, one
downstream from the primary filter at the product drumming station, and
the other on the ventilation air exhausted from the space in which the
drumming station is contained. Without a filter, the ventilation exhaust
would become the dose-limiting pathway after a HEPA filter is installed
downstream from the primary filter.
The present analysis is based on a production rate to 1000 MTU/yr;
air flow to the primary filter (Fig. 4.1) at the drumming station is
expected to be less than a few hundred cubic feet per minute. Since a
single 2-ft x 2-ft x 12-in. thick HEPA filter has a rated air flow of
1000 scfm when initially installed, only one filter is needed at the
drumming station. Costs are obtained by increasing those which pertain
2
to 1973 dollars by the factor 1.41. This factor corresponds to a 41%
3
inflation from 1973 to 1976. It is assumed that the filter housings
will be installed on and/or in an existing building and that no new
structure will be needed.
6.1 Capital Costs
The capital cost of the installation of the two HEPA filters is the
sum of the direct and indirect costs. A summary of the methods used for
estimating the direct and indirect costs is presented in the following
2
sections.
6.1.1 Direct costs
The major equipment components consist of HEPA filters, which include
housing, blowers, dampers and drives, and ducts. The following table lists
direct costs for the installation of two filter units.
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52
Dollars
HEPA filters 4250
Blowers (1000 cfm each) 2400
Ducts 2050
Dampers and drives 4250
Total direct cost (rounded upward) 13,000
6.1.2 Indirect costs
2
Indirect costs are calculated to equal 1.4 times direct costs,
$18,200.
6.1.3 Capital cost
The capital cost is the sum of the direct and indirect costs,
$31,200.
6.1.4 Annual fixed charge
The annual fixed charge is assumed to equal 26% of the capital
cost, $8100.
6.1.5 Annual operating and maintenance cost
The annual operating and maintenance (O&M) cost is calculated as
follows:
Dollars
3% of direct cost 390
Blower power cost 25
OOP Leak test cost (1/6 the value 140
used in ref. 2 x 1.41)
Filter replacement (1/yr) 170
Total O&M (rounded upward) 725
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53
6.1.6 Total annual cost increment for Case 2
The total annual cost increment for Case 2 over Case 1 for any
production rate £1000 MTU/yr is the sum of the annual fixed O&M charges,
or about $9000, rounded upward from $8800.
6.2 References for Section-6
1. C. A. Burchsted, A. B. Fuller, and J. E. Kahn, Nuclear Air Cleaning
Handbook, ERDA 76-21 (1976).
2. W. S. Groenier, R. E. Blanco, R. C. Dahlman, B. C. Finney, A. H. Kibbey,
and J. P. Witherspoon, Correlation of Radioactive Waste Treatment Costs
and the Environmental Impact of Waste Effluents in the Nuclear Fuel
Cycle for Use in Establishing "As-Low-As-Practicable" Guides — Fabrica-
tion of Light-Water Reactor Fuels Containing Plutonium, ORNL/TM-4904
(May 1975).
3. "Economic Indicators," Chemical Engineering, McGraw-Hill, New York,
p. 7 (Jan. 17, 1977).
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54
7. ONSITE AND ENVIRONMENTAL MONITORING
The radiological survey described in this section was conducted on
properties of the Uranium Recovery Corporation (URC), a commercial proc-
essing company, in Mulberry, Florida. This study was directed toward
determining background levels of specific radionuclides both on- and off-
site and measuring the releases of radioactivity into the surrounding
environment under normal operating conditions. The URC process extracts
uranium from WP phosphoric acid. From a radiological point of view, the
primary concern is the release of uranium to the environs. The daughter
products of the uranium decay series (Fig. 4.2) are of less concern due
to the high affinity of the solvent for uranium. After removal of uranium,
the acid solution is returned to the WP plant for the production of fertil-
izers and other by-product materials.
7.1 Description of Site and Specific Release Points
A simplified schematic diagram of the URC site is given in Fig. 7.1;
this diagram represents an area of approximately 2-1/4 acres. Of primary
concern is the drumming building, which contains the product drumming area
where uranium is precipitated as ammonium uranyl tricarbonate (AUT), dried,
and calcined.* The calcine (U30g) powder is dropped into 55-gal drums and
sealed prior to shipment to a UF6 conversion plant. The drumming station,
where calcined dust drops into the drum via a "mini-spout" system, is
shown in Fig. 4.1. During the filling of a drum, there is no airflow
through the 3-in.-diam duct from the drum to the bag filter. Air flows
through the duct (shown by the arrows in Fig. 4.1) only during actual
changing of drums. Most of the ^Og particulate matter that becomes air-
borne from the filled drum flows to the bag filter. This minimizes escape
of U30g dust into the product room during the drum-change process. The
airflow is then vented from the bag filter by a blower through a HEPA
filter, and is finally released to the atmosphere through a 6.25-in.-diam
*The calcine ranged from brown to black. In subsequent discussion it is
referred to as UsOg, but its chemical composition was not determined.
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55
ORNL-DWG 77-18279R
X X X X X X X X X * X X X
WAREHOUSE 8 SHOP
CONTROL ROOM
CHANGE HOUSE
DRUMMING
BUILDING
PROCESSING
/ X AREA
Fig. 7.1.
0 20 40 60 80 (00
SCALE (ft)
Plan view of the Uranium Recovery Corporation site.
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56
duct which extends through the roof to a height of about 8 ft. In addition
to this duct, two room-ventilation fans, rated at 13,700 cfm each, are also
located on the roof. These fans represent another potential source for
release of t^Os dust which may be suspended in the drumming room air. The
locations of both the duct and fans on the roof of the drumming building
are shown in Fig. 7.2.
The URC property and the surrounding area are shown in Fig. 7.3.
Background samples (water, sediment, and soil) were collected both on- and
off-site and were analyzed for specific radionuclides to characterize back-
ground radionuclide concentrations in the local and surrounding areas. The
methods used for this assessment are presented below.
7.1.1 Survey plan
Based on a brief preliminary survey of the site, a comprehensive sur-
vey plan for the radiological assessment of the site and the surrounding
area was established. The survey consisted of determination of the
following:
1. isotopic concentrations of uranium in the drumming station
exhaust duct upstream from the bag filter and downstream
from the HEPA filters;
2. the particle size distribution of l^Og dust in the drumming
building;
3. the concentration of 1)303 dust suspended in air in the drum-
ming building as well as on the roof;
4. concentrations of 235U and 238U in soil at 1, 3, 5, and
12-in. depths, at intervals of approximately 100 ft within
the fenced-in area of the plant, as shown in Fig. 7.4;
5. concentrations of 230Th, 210Pb, 235U, 238U, and 226Ra for
surface and 6-in. deep soil samples collected at approxi-
mately 50-ft intervals along liquid drainage and runoff
paths on the site, as shown in Fig. 7.5;
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57
ORNL-DWG 77-18277R
N
vf T * n r r * ''*
BAGHOUSE STACK
WITH HEPA FILTER
o
iCENTRIFUGEl
i I
v i ' J i 11/111
0 1
10
SCALE (ft )
Fig. 7.2. Plan view of equipment on roof of drumming building.
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OBKL-DWG 77-18262 R
Ln
OO
Fig. 7.3. Plan view of the area surrounding the Uranium Recovery
Corporation showing sampling locations. (Sample numbers are given as
FLOS1, for off-site jjoil sample 1, as FLW11, for water and sediment
Camples 11, etc.)
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59
ORNL-DWG 77-18281R
0 20 40 60 80 100
SCALE (ft)
Fig. 7.4. Locations of the 12-in.-deep core sampling points
(including sample numbers, such as FLS1) on the URC site.
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60
ORNL-DWG 77-18278R
FLCM6 FLD17 FLD18
0 20 40 60 80 tOO
SCALE (ft)
Fig. 7.5. Locations of water and drainage path sampling points
on the URC site. (Sample numbers are given as FLD1, for Drainage
soil sample 1, as FLW1, for water and sediment samples 1, etc.)
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61
6. concentrations of 230Th, 210Pb, 226Ra, 235U, and 238U in
liquid and corresponding sediment samples, collected from
drainage paths and standing water on the site, also shown
in Fig. 7.5;
7. concentrations of terrestrial radionuclides at possible
"collection points" on the site, as shown in Fig. 7.6,
using an in situ gamma spectral measurement technique;
8. concentrations of 230Th, 210Pb, 235U, 238U, and 226Ra
(gamma spectral analysis) for a series of soil, water,
and sediment samples collected in the general area sur-
rounding the URC plant, as shown in Fig. 7.3, but at a
distance from the plant sufficient to establish back-
ground concentrations in the local area. In addition,
measurements of the gamma-ray exposure rate were made 1 m
above the ground where each soil sample was collected.
7.2 Radiological Survey Techniques
The instrumentation and measurement techniques used to carry out the
radiological assessment of the URC site and surrounding area are described
in this section.
7.2.1 Isokinetic stack monitoring
To obtain accurate results for 11303 particle concentrations, sampling
in the ducts must be done in accordance with isokinetic principals. This
sampling was performed after consultation with air pollution engineers in
the Environmental Engineering Department at the University of Florida.
As mentioned previously, the duct ventilation system consists of a
bag filter followed by an in-line HEPA filter, which represents the last
control point prior to discharge of air to the atmosphere. Stack (or duct)
sampling was performed upstream from the bag filter in the 3-in.-diam
duct, and downstream from the HEPA filter in the 6.25-in.-diam duct in
order to establish a.decontamination factor (DF) for the overall filtration
system.
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62
ORNL-DWG 77-18280R
0 20 40 60 80 100
SCALE (ft)
Fig. 7.6. Location of in situ terrestrial gamma spectra measuring
points on the URC site. (Sample numbers are given as FLGSl, FLGS2, etc.)
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63
As shown in Fig. 7.7, no ideal port was available for sampling in the
6.25-in.-diam duct; therefore, a vertical sampling port was chosen in which
minimal perturbations were expected.* The cross section of the stack at
the sampling port was divided into 12 equal areas, as shown in Fig. 7.8,
with traverse points located at the centroid of each area. The velocity
profile at each sampling site was characterized by traversing the stack
with standard pitot tubes and an inclined manometer prior to actual stack
sampling. Two tubes used to perform the velocity measurements gave the
same results. Clogging of the tubes was not evident. Temperature and
static pressure within the stack were measured at only one traverse point
and were assumed to be constant throughout the cross section of the stack.
This assumption appears to be valid since ambient air is exhausted through
the duct far downstream from the sampling port where mixing of the air
would be thorough.
Actual sampling downstream from the HEPA filter was performed with a
0.5-in.-diam nozzle and the sampling train described in Appendix A. Each
run consisted of isokinetic sampling at all 12 points in the stack for the
same length of time at each point. The velocity at each sampling point
remained constant during the sampling period. Actual sampling was performed
under steady plant operating conditions. Procedures concerning the changing
of product drums, which are pertinent for off-gas sampling, are described
in Appendix B.
Three runs were made at this port to determine the concentration of
uranium downstream from the HEPA filter. Pertinent data for all runs which
have been calculated by methods similar to those presented in 40 CFR Part 60,
Appendix A, are shown in Table 7.1. Total uranium present and the isotopic
abundances were determined by mass spectrometric techniques. The sensi-
tivity of this method is in the low parts-per-billion range.
Due to the absence of an adequate sampling port in the 3-in.-diam
duct, a vertical sampling port was chosen in which minimal perturbations
*As defined in Title 40, Code of Federal Regulations, Part 60, Appendix A,
an ideal sampling site is at least 8-stack diameters downstream and
2-diameters upstream from any flow disturbance.
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64
ORNL-DWG 77-18284
16"
SAMPLING—»•
PORT |
7"
U-6.25"-
18"
I i
8"
J_
12" -
HE PA
FILTER
1
3"
XROOF
if \
N \
Fig. 7.7. Location of sampling port for 6.25-in.-ID duct.
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65
ORNL-DWG 77-18283R
TRAVERSE POINT
PITOT TUBE SENSING POINTS
FOR STACK VELOCITY TRAVERSE
AND SAMPLING LOCATIONS
12 POINTS
6 NORTH-SOUTH, 6 EAST-WEST
6.25 in.-ID STACK
DISTANCE FROM INSIDE
WALL OF STACK TO
NUMBER TRAVERSE POINT (in.)
\ 0.275
2 0.919
3 1.844
4 4.406
5 5.331
6 5.975
Fig. 7.8. Equal area layout for the 6.25-in.-ID duct.
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66
were expected. The cross section of the duct at the sampling point was
divided into eight equal areas, with location of traverse points at the
centroid of each area.
Table 7.1. Source sampling results from 6.25-in.-diam duct
Run
No.
1
2
3
Sampling
duration
(min)
120
60
60
Average
stack
volumetric
flow rate
(dscfm)a
135
136
146
Total
uranium
activity
x 106
(yci)b
8.09
9.82
2.44
Uranium
cone
x 1012
(yCi/ml)c
2.60
6.21
1.44
Mass
emission
rate x lO1*
(g/hr)d
8.82
21.2
5.29
Percent of
isokinetic
sampling
106
107
107
Stack gas flow rate, reduced to dry standard conditions of 70°F and
29.92 in. Hg.
Total uranium collected on filter in sampling train.
'Represents the average concentration of uranium in air (corrected to dry
air conditions) in the stack at the release point.
This hourly mass emission rate applies only to the time corresponding to
operation of the stack blower.
The methods of velocity traverse characterization and actual isokinetic
stack sampling were similar to those used at the 6.25-in.-diam duct with the
following modifications: (1) the length of time per sample run was shortened
to minimize the effect of dust loading on the filter paper; and (2) a 0.25-
in.-diam nozzle was used in this smaller diameter duct. Pertinent results
for this duct are shown in Table 7.2. The data were compiled using the
same methods as for the 6.25-in.-diam duct.
7.2.2 Particle size measurements
A six-stage Anderson cascade impactor was used for particle size distri-
bution measurements on both the first and second floors of the drumming
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67
building. A final seventh stage, consisting of a 0.45-y membrane filter
paper, was placed in line with the impactor to ensure entrapment of all
particles of at least 0.45-y size. The impactor was specifically calibrated
for l^Og dust at a constant flow rate of 1 cfm. The following distribution
was determined for each stage, with an effective aerodynamic cutoff diameter
between stages of 85%:
Particle diam (y)
Stage 1: <9.2
Stage 2: 5.5 to 9.2
Stage 3: 3.3 to 5.5
Stage 4: 2.0 to 3.3
Stage 5: 1.0 to 2.0
Stage 6: up to 1.0
Stage 7: <0.45
To ensure sampling at the proper flow rate, the impactor was used in con-
junction with a calibrated dry gas meter (±2% error) and a regulated
positive displacement vacuum pump.
Table 7.2. Source sampling results from 3-in.-diam duct
Run
No.
1£
2
3
Sampling
duration
(min)
8
16
16
Average
stack
volumetric
flow rate
(dscfm)a
105
108
108
Total
uranium
activity
x 103
(yci)b
2.53
1.57
1.73
Uranium
cone
x 109
(yCi/ml)c
15.14
4.58
5.04
Mass
emission
rate
(g/hr)d
4.00
1.24
1.36
Percent of
isokinetic
sampling
101
101
101
Stack gas flow rate reduced to dry standard conditions of 70°F and
29.92-in. Hg.
Total uranium collected on filter in sampling train.
"Represents the average concentration of uranium in air (corrected to dry
air conditions) in 3-in. duct.
Represents the mass discharge rate to the bag filter.
a
"Run No. 1 was taken with no cover on the product drum during the run.
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68
Two types of sample collection surfaces were prepared. The first
consisted of 3-l/4-in.-diam, 1/16-in.-thick standard (commercial) glass
collection plates. A thin layer of silicone grease was smeared on the
collection side of each plate to ensure entrapment of the proper size
particles at each stage. In addition to the actual samples being assayed
for total uranium on each stage (by mass spectrometric technique), a blank
sample smeared with a thin layer of silicone grease was also assayed to
determine the background quantity of uranium present. Net values in total
uranium could then be established (assuming the background quantity of
uranium remained constant).
The second method of particle collection consisted of standard (commer-
cial) 3-1/2-in. diam stainless-steel collection plates, each covered with
a 0.22-y pore-size membrane filter paper. The papers were cut by a die to
ensure a proper fit in the dishes and a uniform area of collection between
papers. The samples were analyzed for 235U by a neutron-induced fission
technique. The samples were irradiated in the Oak Ridge Research Reactor
and the delayed neutrons from the fission of 235U atoms were counted. The
2
sensitivity of this method is in the parts-per-billion range.
In addition to 235U analysis, some of these samples were analyzed for
238U to confirm the natural abundance ratio of the isotopes. This method
involves neutron activation of 238U atoms according to the reaction:
238U (n,Y) 239U •* 239Np +g ,
where the gamma emissions from 23%p are counted. A blank filter sample
was analyzed for both 235U and 238U to determine the net uranium per stage.
Three samples were taken under various conditions in the drumming
building, two on the first floor and one on the second floor. Duplicate
samples were collected with both collection techniques for intercomparison.
Specific sampling conditions and results are shown in Tables 7.3 to 7.5.
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69
Table 7.3. Particle size characterization of 1)309 dust in air
on the ground level of the drumming building 1 ft from closed drum
Stage
1
2
3
4
5
6
7
Total
Stage
1
2
3
4
5
6
7
Total
Particle
aerodynamic
diam
(y)
>9.2
5.5 to 9.2
3.3 to 5.5
2.0 to 3.3
1.0 to 2.0
sl.O
,0.45
Table 7.4. Particle
on the ground level of
Particle
aerodynamic
diam
(y)
>9.2
5.5 to 9.2
3.3 to 5.5
2.0 to 3.3
1.0 to 2.0
sl.O
£0.45
Uranium on
Method 1
yg % of total
8.91 66.69
0.91 6.81
1.07 8.01
1.23 9.21
1.04 7.78
0.07 0.52
0.13 0.97
13.36 99.99
size characterization of
the drumming building 1
Uranium on
Method 1
yg % of total
1.17 70.06
0.22 13.17
0.26 15.57
a a
a a
a a
0.02 1.20
1.67 100.
the stage
Method 2
yg % of total
5.60 70.18
0.43 5.39
0.22 2.76
0.24 3.01
0.44 5.51
0.07 0.88
0.98 12.23
7.98 99.96
U30s dust in air
ft from open drum
the stage
Method 2
yg % of total
0.53 0.30
0.07 0.04
0.18 0.10
4.86 2.76
168.0 95.45
0.85 0.48
1.45 0.82
176.0 99.95
Not detected.
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70
Table 7.5. Particle size characterization of l^Og dust in air on
the second floor of the drumming building directly under the calciner
Stage
1
2
3
4
5
6
7
Total
Particle
aerodynamic
diam
(y)
>9.2
5.5 to 9.2
3.3 to 5.5
2.0 to 3.3
1.0 to 2.0
<1.0
>0.45
Method
yg %
13.28
7.68
31.32
37.21
24.41
1.72
0.02
116.
Uranium on
1
of total
11.45
6.62
27.0
32.08
21.04
1.48
0.02
99.7
the stage
Method
yg %
2.77
3.38
9.53
10.11
28.41
5.75
8.52
68.5
2
of total
4.04
4.93
13.91
14.76
41.47
8.39
12.44
99.9
7.2.3 Atmospheric spot sampling
Due to the anticipated low concentrations of airborne uranium, a high-
volume air sampler* was used for sampling at randomly selected locations.
The filter media consisted of Hollingsworth and Vose Type-70 cellulose-
asbestos filter paper with a retention efficiency of <98% (at the flow
3
rates encountered) for 0.3-y-diam particles. Each paper was cut by a
template of approximately 4-in. diam to ensure uniform paper area. These
papers (a total of eight) and a blank sample to establish the quantity of
uranium in the paper itself were cut from the same roll. The samples were
then analyzed by isotopic dilution and mass spectrometric techniques.
Each air sample was calibrated (with filter paper and backing screen
in place) prior to and at the termination of sampling to determine the
actual volume of air that passed through the sampler. No appreciable dust
*Model No. TFIA, Staplex Corporation.
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71
loading occurred during actual periods of sampling. Eight spot samples
were taken in the drumming building and on the roof at specified locations,
These locations, results, and other pertinent remarks are given in
Table 7.6.
Table 7.6. Concentrations of natural uranium in air inside of
and on the roof of drumming building
Sample
No.
Date
Uranium
concentration
x 1012
(yCi/ml)
Remarks
7-7-77
7-7-77
7-8-77
7-8-77
7-14-77
7-8-77
7-14-77
2.91
31.73
63.12
8.37
2.27
11.75
1.49
Ground level, 10 ft from product drum
with "mini-spout" closed. Entrance
door was closed for several hours prior
to and during sampling.
Top level, 6 ft above floor directly
under fan 1 (Fig. 7.2). Both fans
were on continuously with entrance
door closed for several hours prior
to and during sampling.
Top level, 6 ft above floor directly
under fan 1. Both fans were on con-
tinuously with entrance door open for
several hours prior to and during
sampling.
Top level, 6 ft above floor directly
under fan 2 (Fig. 7.2). Both fans
were on continuously with entrance
door closed for several hours prior to
and during sampling.
Top level, 6 ft above floor directly
under fan 2. Both fans were on con-
tinuously with entrance door open for
several hours prior to and during
sampling.
On roof, near surface by fan 1 outlet.
On roof, near surface by fan 1 outlet.
Calciner shut down for approximately 38 hr prior to sampling.
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72
7.2.4 Soil sampling and analysis
Twelve-inch-deep soil samples were taken at intervals of approxi-
mately 100 ft on the URC site, as shown in Fig. 7.4. Brass and stainless
steel tubes, 1 in. ID, were driven into the ground to a depth of 12 in.
At ORNL, a milling machine was used to cut each tube to obtain samples at
1, 3, 5, and 12-in. depths. A correction was made for compaction of the
soil in the tubes. Each sample was extracted from the tube (1 in. wide),
completely dried, and homogenized prior to analyzing for 23^u and 23°U
by neutron-induced fission and neutron activation techniques respectively.
Approximately 10% of each sample that was extracted from the tube was
actually irradiated, thus emphasizing the importance of homogenization.
Surface and 6-in.-deep soil samples were also collected on the site
along liquid drainage and runoff paths, as shown in Fig. 7.5. A minimum
of 500 g (dry weight) per sample was packaged in plastic bags (double
bagged) and returned to ORNL for sample preparation and analysis. Once
received, each sample was dried and then crushed to a particle size no
greater than 500 y diam. Each sample was then proportioned in the follow-
ing manner for specific radioanalysis:
1. Approximately 340 cc of each sample was bottled, sealed, and stored
for about 30 days to allow 222Rn progeny to approach equilibrium
with 226Ra. The samples were then counted using a Ge(Li)-detector
system and the spectra were resolved by computer techniques. This
system and the counting technique are described in Appendix C.
2. Approximately 50 g of each sample was analyzed specifically for
230Th and 210Pb by radiochemical techniques.
3. Several grams of each sample were analyzed for 235U and 238U by
techniques mentioned earlier.
7.2.5 Environmental gamma-ray measurements using an in situ
measuring technique
A 55-cm3 lithium-drifted germanium [Ge(Li)] semiconductor detector
with a 4096 channel Nuclear Data (ND) 100 analyzer was used to accumulate
gamma-ray spectra of the on-site grounds. The specific locations are
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73
shown in Fig. 7.6. The detector, situated 1 m above the ground, detects
gamma radiation from an area of about 100 m2. The accumulated spectra
were stored on magnetic tape and resolved by use of a computer program.
A technical description of the in situ measurement technique is given in
ref. 4 with only minor modifications being applied for this specific
assessment.
7.2.6 Liquid and sediment sampling and analysis
Duplicate liquid samples were collected in 1-liter polyethelene jars
at each location shown in Fig. 7.5. After collection, 10-ml of concentrated
nitric acid was added to each sample to minimize plating out of ions on the
sides of the container. One liter of each sample was analyzed specifically
for 230Th and 210Pb by radiochemical techniques, and for 235U and 238U
by neutron-induced fission and neutron activation techniques respectively.
The duplicate of each sample was filtered to remove suspended particulates
> 0.8 y, concentrated by evaporation, and analyzed specifically for soluble
radium using a radon-emanation technique.
A sediment sample (minimum of 500 g) was also collected at each loca-
tion where a liquid sample was taken. These sediment samples were handled
in the same manner as soil samples. The procedure is discussed in Sect.
7.2.4.
7.2.7 Gamma-ray exposure rate measurements
External gamma exposure-rate levels were measured at specific locations
within the area surrounding the URC plant. These measurements were made
1 m above each point where a soil sample was collected (see Fig. 7.3).
The detection system consists of a Geiger-Muller survey meter with a
specially designed energy correction shield such that the response of the
detector is nearly energy independent. Exposure rates were estimated
from the count rate of this detector. A calibration factor was obtained
through repeated measurements using NBS-calibrated, gamma-ray sources.
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74
7.3 Survey Results
7.3.1 Release through the stack
Using the source sampling techniques presented above, the release of
uranium through the stack was determined; results are presented in Table 7.1.
The average concentration from three runs was determined to be 3.42 x
10~12 yCi/ml with a standard error* of 1.44 x 10~12. This corresponds to
a mass emission rate of 1.18 x 10~3 g/hr. Once the effluent leaves the
stack, it is dispersed rapidly as the cloud leaves the site of release,
thus reducing the uranium concentration. Through mass spectrometric tech-
niques, it was determined that 238U, 235U, and 234U occurred in natural
isotopic abundances. Thus the specific activity of natural uranium
(0.677 pCi/yg) was used to convert raw data (in micrograms of total
uranium) to activity (yCi). For comparison, the maximum permissible
concentrations allowed in air [MFC (air)] for natural uranium in restricted
and unrestricted areas are 1 x 10~10 and 5 x 10~12 yCi/ml respectively.
The concentration of uranium (natural) entrapped in the 3-in.-diam
duct prior to collection in the total filtration system (i.e., bag and
HEPA filter) is shown in Table 7.2. The average concentration from three
runs was determined to be 8.25 x 10~9 yCi/ml, with a standard error of
3.44 x 10~9. This corresponds to a mass rate of 2.2 g/hr discharged to
the bag filter.
All sampling runs were performed within isokinetic tolerances. No
explanation can be given for the relatively high concentration of uranium
in run 2 in the 6.25-in.-diam duct (Table 7.1). Run 1 in the 3-in.-diam
duct was performed with no sampling cover on the product drum during the
entire sampling period. Table 7.2 shows that the concentration of uranium
in the duct without the cover on the drum is approximately three times
greater than with the cover in place (normal operating procedure). The
air flow downstream from the bag and HEPA filters was approximately 25%
greater than upstream. A plausible explanation for this difference is
*Standard error is defined as S = S/y^T, where S is the standard
deviation and n = the number or samples. This is also known as
the standard deviation of the mean.
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75
leakage of air into the duct around the blower housing and joints in the
duct. During the period of sampling, the blower was located between the
bag filter and the HEPA filter.
An overall DF for the total filtration system can be calculated using
the equation
_ mass rate discharged to the bag filter
mass-emission rate from stack downstream from the filter system.
The reciprocal of this value is the fraction of uranium in the 3-in. duct
which is released to the atmosphere. Using the average concentrations of
uranium determined both upstream and downstream of the filtration system,
it is seen that
DF = (2.2 g/hr)/(1.18 x 10~3 g/hr) = 1860.
A DF of 1860, corresponding to retention of 99.95% of the total air-
borne uranium by the filtration system, is 103 times smaller than the
2 x 106 value used to calculate the source term for Case 2, Models 1 and 2,
Table 4.5. This large discrepancy appears to be due partly to the fact
that the URC operation was still in the startup period when samples were
taken for this study. Startup difficulties may have included improper
installation of the HEPA filter.
The annual release of natural uranium from the stack can be calculated
using the average mass emission rates of uranium through the 6.25-in.-diam
duct. At the then-existing production rate of about 50 MTU/yr, it would
require approximately 1.5 days to fill a drum, assuming continuous opera-
tion 24 hr/day and 365 days/yr. At this time, the filled drum is replaced
with an empty one. It requires approximately 25 sec to complete this
change, and this represents the total time that the dust blower is turned
on. Release of activity to the atmosphere from the stack is possible only
while this operation is in progress. Thus, the annual mass emission rate,
A, of natural uranium from the stack may be determined by:
A = (mass emission rate from the 6.25-in.-diam stack) x (total period
of time that the emission occurs in a year).
-------
76
Using the average emission rate of uranium calculated from Table 7.1 and
the above stated criteria, the annual emission rate is estimated to be:
(1.18 x 10~3 g/hr)(25 sec/1.5 days)
(hr/3600 sec)(365.25 days/year) = 2 mg/yr.
This emission rate is significantly less than expected for a prorated source
term for Case 2, Models 1 and 2, in Table 4.5. In particular, 2 mg/yr is
only about 0.5% of the calculated source term (Table 4.5) that would apply
to a production rate of about 50 MTU/yr, the rate prevailing during sampling
operations. Considering the low DF (1860) obtained across the bag and HEPA
filter combination, it is apparent that much less than 1% of the plant
product (11303) enters the filter system. The 1% value was used in construct-
ing Table 4.5, as noted in Sect. 4.5.
7.3.2 Determination of Size Distribution
The particle size range of U308 dust in the drumming atmosphere has
been characterized using the particle size measurement techniques presented
above. Tables 7.3 to 7.5 show the results of duplicate runs at the specific
locations listed. Methods 1 and 2 refer to mass spectrometric and delayed
neutron techniques, respectively, as described in Sect. 7.2.2.
A comparison of the two sampling methods shows good results except in
Table 7.4. Table 7.3 shows that approximately 70% of the total uranium
collected was of a particle size larger than 9.2 y diam for both methods.
Table 7.5 shows that approximately 75% of the total uranium collected was
between 1.0 and 5.5 y in diameter for both methods. The differences between
the two runs may be attributed to selective gravitational settling with
respect to particle size.
In Table 7.4, 95% of the total uranium collected by the second method
was in the particle size range 1.0 to 2.0 y diam. For the first method,
70% of the total uranium collected was greater than 9.2 y. In addition, the
total weight of uranium collected (summation of all stages) by Method 2
was 105 times greater than by Method 1, even though sampling conditions
were the same for both methods. No precise explanation can be given for
the cause of this effect.
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77
Data in Table 4.4 and those in Tables 7.3-7.5 are almost directly
comparable, although different calibration procedures were used. The
average mass-medium-dust particle size, where handled in the concentrate
form (after precipitation) is listed in Table 4.4 as 2.5 y in the room
atmosphere of uranium mill operations. This dust is expected to be
primarily 11363. In Tables 7.3-7.5, approximately 70% of the dust
particles (again primarily t^Og) in the drumming building air varied in
aerodynamic diameter from greater than 9.2 y on the bottom floor beside
the product drum (disregarding Table 7.3, Method 2 results) to a range of
1.0 to 5.5 y on the second floor of the building below the room ventilation
fans.
7.3.3 Determination of uranium concentration in air
Using the spot air sampling technique described above, the concentra-
tions of uranium in the room air and just above the roof of the drumming
building were determined. These results and other pertinent information
are listed in Table 7.6. It was determined that 238U, 235U, and 234U were
in natural isotopic abundance; thus the values presented were calculated
using the specific activity for natural uranium. Except for samples 5 and
7, samples were taken while the plant was operating.
Since a radiological assessment of the occupational health effect from
natural uranium was not addressed in this study, only limited data were
collected to characterize the concentration of natural uranium in the drum-
ming building atmosphere (e.g., ventilation effects and air exchange were
not determined). Of prime concern is the release of uranium to the atmo-
sphere from the roof ventilation fans. Samples 2 and 3 (Table 7.6) represent
the concentrations of uranium in the room atmosphere directly below fan
No. 1. Sample 3 has about twice the concentration of sample 2; this may be
due to several disruptions of plant operation during the assessment study.
The two samples were taken on successive days. The concentration average
of the two samples is 47 x 10~12 yCi/ml.
Samples 4 and 5 (Table 7.6) represent the concentrations of uranium in
the room atmosphere directly below fan No. 2. Note that the concentrations
of the two samples vary by almost a factor of 4. This may be attributed
-------
78
partially to the fact that the calciner unit was shut down for approximately
38 hr prior to sampling. The average concentration of the two samples is
5.32 x 10~12 yCi/ml.
No reasonable explanation can be given for the marked difference
(a factor of 9) between the concentrations below the two fans. Although
samples were taken at different dates, this difference should be minimal
when establishing concentrations in the building atmosphere. However,
disruptions in plant operations (e.g., the calciner unit being shut down,
contamination in the drumming building, leakage) between sampling periods
could have had a dominant effect on the values obtained.
Since the concentrations of uranium in the building atmosphere vary
widely, two cases are presented for estimating source-term releases.
Case 1; Assuming the concentration of natural uranium in
the drumming building atmosphere is 47.43 x 10~12 yCi/ml,
a mass emission rate may be calculated as:
(2 fans) x (13,700 ft3/min, the rated output/fan) x (28.317 ml/ft3) x
(47.43 x 10~12 yCi/ml) x (1.48 g/yCi) = 5.45 x 10~2 g/min.
Assuming the fans run continuously for 24 hr/day for the
entire year, an annual mass emission rate of 2.87 x lO1* g/yr
is calculated.
Case 2; Assuming the concentration of natural uranium in
the drumming building atmosphere is 34.41 x 10~12 yCi/ml,*
an annual mass emission rate of 2.08 x lO1* g/yr is calculated.
Two air samples, Nos. 6 and 7 of Table 7.6, were taken on the roof of
the drumming building by outlet fan No. 1. The burden of 1.49 x 10~12 yCi/
ml in sample 7 should be compared with the 2.27 x 10~12 yCi/ml of sample 5,
taken at essentially the same time after a 38-hr shutdown of the calciner.
It should be noted that there was no filter in the roof ventilation system.
The fact that the roof-air sample was found to contain less uranium than
that in the room air probably is a result of some atmospheric dilution on
^Averaging the three concentrations of uranium determined during normal
plant operation.
-------
79
the roof. Similarly, the 11.75 x 10~12 viCi/ml in roof sample 6 should be
compared with 63.12 and 8.37 x 10~12 yCi/ml, respectively, in room samples 3
and 4.
7.3.4 Radionuclide concentration in soil
Liquid and drainage paths. Forty-eight soil samples were collected
along on-site liquid and drainage paths, as shown in Fig. 7.5, at both
surface and 6-in. depths to characterize the radionuclide concentrations
present. Sample numbers with the prefix FLD refer to drainage path soil,
and those with FLW refer to water and sediment samples. Table 7.7 lists
the average concentrations of 238U and 232Th, tabulated from the analysis
of these samples (excluding anomalies) at both surface and 6-in. depths.
Also listed are their maximum and minimum values and respective sample
locations at the appropriate depths. From the compiled data, it appears
that 238U is approximately in equilibrium with its daughter products through
210Pb. Thus only 238U concentrations are listed. The average 238U con-
centrations established appear to be typical for reclaimed land of this
region.*
The concentrations of 232Th have been determined using a gamma-
spectrometric technique. It was assumed that daughter products through
208T1 were in secular equilibrium with 232Th. This assumption appears
valid even though few data points exist to support it.
Also listed in Table 7.7 are the surface-to-6-in. depth ratios for
both 238U and 232Th concentrations. In determining these values, only
those samples were compared in which concentrations at both depths had
been established for each sample. For both 238U and 232Th, the concentra-
tions at the 6-in. depth are higher than those at the surface by about the
same factor. Other information in the table includes the ratio of uranium
to thorium concentrations at both the surface and 6 in. depths. Only those
samples were compared in which both concentrations had been established for
each sample, and the ratios are about the same for both depths.
*Typical values for 226Ra in reclaimed land of this region vary from
1 to 40 pCi/g.8
-------
Table 7.7.
238U and 232Th concentrations in soil along liquid
and drainage paths on the URC site
239U
232Th
Concentration Concentration
Sampling
depth
(in.)
0
(surface)
6
Ratio
fsurface value"!
L 6- in. value J
Sampling
depth
(in.)
0
(surface)
6
av
and s
av
26.73
2.35
39.40
4.95
0.82
0.12
Activity
av
and s
av
34.03
2.35
31.87
2.98
(pCi/g)
max,
s b , and
max
location
42.64
0.27
FLD19
66.44
0.41
FLD20
1.55
FLD24
ratio, 238U7:
max
and
location
45.65
FLD17
49.38
FLD4
(pCi/g)
min, max, min,
s , and , s , and s . , and „ ..
min No. or av max mm No. of
location0 samples and s location location samples
8.10 17 0.74 1.25 0.34 12
0.08 0.08 0.12 0.07
FLD9 FLD22 FLD9
8.30 14 1.46 4.31 0.35 18
0.09 0.25 0.70 0.05
FLD24 FLD18 FLD24
0.38 9 0.85 1.23 . 0.56 8
FLD8 0.08 FLD24 FLD20
min
and No. of
location samples
23.82 12
FLD9
19.18 16
FLD13
Standard deviation of the mean.
Standard deviation due to counting statistics.
Location of the sample, shown in Fig. 7.5.
oo
o
-------
81
In Table 7.8, anomalous concentrations of 238U with corresponding
226Ra and 232Th concentrations and sample locations are listed. Some dis-
equilibrium appears to exist between 238U and 226Ra. However, activities
of the three nuclides suggest a "hot spot" of soil in which their concentra-
tions are higher than normal for reclaimed land.
Table 7.8. Anomalous radionuclide concentrations in soil along liquid
and drainage paths on the URC site
Sample
location
FLD 8
FLD 8
FLD 18
FLD 18
Sample 238U 226Ra 232Th
depth concentration concentration concentration
(in.) (PCi/g)a (PCi/g)a (PCi/g)a
0 40.33 (0.26) 34.00 (1.76) b
6 106.44 (0.62) 93.00 (1.41) 3.91 (0.68)
Ob b b
6 110.41 (0.62) 104.00 (1.31) 4.31 (0.70)
Standard deviation due to counting statistics is given in parentheses.
No data.
100-Foot grid plot. Forty-nine soil cores were collected on the site
on an approximately 100-ft grid at the intersection points shown in Fig. 7.4.
The prefix FLS refers to soil core locations. A summary of the sample
analysis is shown in Table 7.9, and the average concentration of 238U present
at the 1, 3, 5, and 12 in. depths along with their maximum and minimum values
and respective sample locations are listed. When comparing these results
to on-site drainage soil, good correlation seems to exist between the
average 238U surface concentration in the drains and the 5-in.-deep core
samples (26.73 and 27.76 pCi/g respectively). Also, the average 238U con-
centration at the 6 in. depth is in close agreement with the 12-in.-deep
core sample (39.40 and 36.83 respectively). Since the drainage paths have
been excavated to several inches, these correlations appear reasonable.
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82
The results show that the concentrations increased with depth at least
to 12 in. In most cases, these values appear to be within typical 238U
concentration ranges found in reclaimed lands of this region. In several
cases, as is indicated from Table 7.9 by the maximum values obtained at
each depth, 238U concentrations appear to be greater than expected for
reclaimed lands. Of the 196 samples analyzed, 18 contained greater than
60 pCi of 238U per gram. Of these 18, 6 contained between 100 and 120 pCi
of 238U per gram.
Table.7.9. Concentrations of 238U in core samples at 1-, 3-, 5-, and
12-in. depths on the URC site
238U concentrations (pCi/g)
Statistic
Av , and
s a
av
Max, b
s , and
max
Q
location
Min
s . , abd
mm
location
No. of samples
1-in.
depth
22.82
2.75
106.44
9,55
FLS41
5.75
0.07
FLS28
49
3-in.
depth
22.89
2.12
90.25
0.52
FLS6
1.51
0.04
FLS26
49
5-in.
depth
27.76
3.31
119.00
0.67
FLS33
1.02
0.05
FLS7
49
12-in.
depth
36.83
4.74
118.01
0.65
FLS26
0.42
0.02
FLS7 .
49
Standard deviation of the mean.
Standard deviation due to counting statistics.
•>
'Location of the sample, shown in Fig. 7.4.
In situ gamma-spectra measurements. Eight in situ gamma spectra were
accumulated on the URC site at specific locations shown in Fig. 7.6. The
measuring system is capable of detecting photons from naturally occurring
radioactive nuclides in soil within an area of approximately 100 m2. Soft-
ware was developed for the analysis of these photon spectra taking into
-------
83
account the detector's response as a function of energy. Those radionuclides
identified from these spectra and their average concentrations in the soil
are listed in Table 7.10. Concentrations are listed in both units of pCi/cc,
which is determined by the measurement technique used, and in units of pCi/g,
which assumes a soil density of 1.60 g/cc. This density factor has been
established for general use in calculations involving in situ gamma-ray
4
measurements.
Table 7.10 shows that the average 231*Th concentration is greater than
the average 226Ra concentration by a factor of 1.9. Since 231+Th (24.1 day
half-life) is very likely to be in secular equilibrium with its parent
238U, these data may be compared with the 238U concentrations in core samples.
When comparing specific in situ measurements with core samples within the
surrounding area, fairly close relationships are observed between the in
situ 226Ra data and the 238U concentrations in soil in almost all cases.
The in situ 231+Th concentrations are somewhat higher than the surrounding
238U concentrations in almost all cases. Biasing of the in situ gamma
measurements by extraneous photons from decay of on-site chemicals (e.g.,
uranium-enriched phosphoric acid stored in tanks and I^Og powder in the drum-
ming building) is a highly probable cause of this effect. Expected °Ra
concentrations in many of these chemicals are less than the equilibrium
3 Th values. Also, a more accurate density factor for this type soil
would help reduce this suspected bias. Nonetheless, both 23ttTh and 22eRa
values are within concentration ranges expected for reclaimed lands of
this region.
Concentrations of 232Th are not listed, because the required assumption
that its daughter products be in secular equilibrium with this parent is
not fulfilled. As Table 7.10 reveals, 212Pb concentrations are greater
than 228Ra and 208T1 concentrations, which essentially are equilibrium
values; however, the 228Ra and 2^8T1 concentrations are in close agreement
with the average 232Th concentration found in on-site surface drainage soil
(Table 7.7).
-------
Table 7.10. Terrestrial radionuclide concentrations identified by in situ gamma
measurement techniques
Statistic
Av , and
a c
sav
Max, d
smax , and
location6
Min d
location
No. of
samples
23I|
pCi/cc
55.01
12.27
111.59
8.41
FLGS5
12.46
3.58
FLGS3
8
Th
pCi/ga
34.38
7.67
69.74
5.26
7.79
2.24
8
226
pCi/cc
29.11
4.81
54.10
1.67
FLGS5
13.60
0.51
FLGS3
8
Ra
pCi/g
18.19
3.01
33.81
1.04
8.50
0.32
8
228Ra
pCi/cc
0.72
0.11
1.19
0.21
FLGS4
0.42
0.11
FLGS8
7
pCi/g
0.45
0.07
0.74
0.13
0.26
0.07
7
212Pb
pCi/cc
3.46
1.17
4.63
0.89
FLGS5
2.29
0.36
FLGS4
2
pCi/g
2.16
0.73
2.89
0.56
1.43
0.23
2
PCi/cc
0.26
0.03
0.42
0.03
FLGS4
0.15
0.01
FLGS3
7
21
pCi/g
0.163
0.02
0.263
0.02
0.09,,
0.01
7
) 8 rp-l
calculated as
0.36 x 228Ra valueb
PCi/g
0.162
0.03
0.266
0.05
O-09" oo
0.03 -p-
7
Values obtained assuming a soil density of 1.60 g/cc.
b0.36 equals the fraction of 212Bi which decays to 208T1.
CStandard deviation of the mean.
Standard deviation due to counting statistics.
Location of the sample, shown in Fig. 7.6.
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85
7.3.5 Radionuclide concentrations in water and sediment
Eight water and corresponding sediment samples were collected at on-
site standing water locations (in drainage paths) and at a pond at locations
shown in Fig. 7.5. Radionuclide concentrations in these samples are given
in Table 7.11. Since variance in the individual sample-counting statistics
is as great as the variance between the samples, only maximum and minimum
values for 230Th and 210Pb are listed. These water concentrations represent
the soluble portions of each radionuclide because the samples were filtered
prior to analyses. Radionuclide concentrations in undissolved and suspended
matter were not determined because a sediment sample was collected at each
sampling site. The extreme disequilibrium between 238U and its daughters
in water is typical for natural waters where uranium is much more soluble
than radium. For comparison, maximum permissible concentrations in water
(MPCw) for unrestricted areas are also listed in the table for each
radionuclide. These values are based on exposure limits to an individual
of the general population. An additional factor of 1/3 for the total popula-
9
tion has not been applied.
The average radionuclide concentrations in sediment, also listed in
Table 7.11, reveal uranium and its daughters to be very nearly in equilibrium.
The large variance between the radionuclide concentrations at each end of
the pond (maximum and minimum value) was not as evident in the water con-
centrations where mixing is more complete. The 232Th concentration in
sediment was 0.26 ±.01 pCi/g, based on the analysis of only one sample.
Approximate equilibrium between 232Th and its daughters was evident in this
sample.
In Table 7.12, the average radionuclide concentrations in water and
sediment are listed for on-site standing water in drainage paths, excluding
anomalies. Once again, extreme disequilibrium between 238U and its
daughter concentrations in water is evident. Comparison of these values
with the specific values of MPCw, also listed in Table 7.12, reveals their
relatively low concentrations. Except for 230Th, the average radionuclide
concentrations in drainage-path waters are slightly greater than in the
pond water.
-------
Table 7.11. Hadionuelide concentrations of Che uranium decay series
in pond water and sediment on the URC site
Sample
type
and
concentration
Water,
fCi/ml
Sediment,
PCi/g
d
MFC ,
w
fCi/ml
Sample
type
concentration
Water
fCi/ml
Sediment,
pCi/g
HPCd,
fCi/ml
2 38,,
Concentration
max,
a , and
av max
and s a location0
18.52 18.66
0.14 3.79
FLW1
7.05 12.20
5.58 0.11
FLW2
40,000
?. ?. 6 ,
Concentration
max,
s , and
av max
and s location
av
0.44 0.45
0.02 0.05
FLW1
6.26 10.50
4.25 0.33
FLW2
30
min,
b
s . , and „
min No. of
location0 samples
18.38 2
7.57
FLW2
1 . 39 2
0.04
FLW1
pin,
s . , and , c
mm No. of
location samples
0.42 2
0.05
FLW2
2.01 2
0.03
FLW1
2 3 DTI,
Concentration
max,
S , and
av max
and s location
e 1.35
0.90
FLW1
7.45 12.30
4.85 0.45
FLW2
2000
Concentration
max,
s , and
av max
and .9 location
av
e 2.60
2.50
FLW1
7.34 12.30
4.95 1.08
FLW2
100
min,
s . , and ,.
mm rJo. of
location samples
0.41 2
0.45
FLW2
2.60 2
0.13
FLW1
min,
s . , and . r
mm ho. of
location samples
<2.25 2
FLW2
2.39 2
0.50
FLW1
00
Standard deviation of the mean.
Standard deviation due to counting statistics.
°Location of the sample, shown in Fig. 7.5.
MFC is the maximum permissible radionuclide concentration allowed in water as defined in Title 10, Code of Federal Regulations,
Part 20, Appendix B, column 2 of Table 2 (Jan. 1, 1977).
6Not determined.
-------
Table 7.12. Radionuclide concentrations of the uranium decay series In standing water
and sediment along liquid drainage paths on the URC site
Sample
type
and
concentration
Water,
fCi/ml
Sed iment ,
pCi/g
MPCd,
w
fCi/ml
Samp ! ^
type
and
concent rai: ion
Water
fCi/ml
Sediment
pCi/g
IIPCJJ,
fCi/ml
238U
Concentration
max, tnin,
av s , and s . , and
max mm
and s location location0
av
29.18 49.58 20.08
5.41 4.14 3.81
FLW6 FLW55.8
51.56 113.39 1.07
J9.00 0.60 0.01
FLW6 FLW4
40,000
226Ra
Concentration
max, mm,
.s1 , and rf . , and
av max min
and s location0 location0
av
0.90 1.19 J.05
0.22 0.03 0.02
FLW6 FLW5
10.39 19.8 2.05
2.64 0.37 0.06
FLU5 FLW8
30
2 30
Concentration
max,
s . and
No. or av max
samples and s location
av
5 e 0 . 90
0.45
FLW6
6 9.8 19.7
2.51 1.26
FLW5
2000
2iopb
Concentration
max,
« , and
max
No. of av
samples and s location
av
5 e 6.3
3.6
FLW7
6 12.70 20.5
3.33 1.17
FLW5
100
Th
min,
s . , and
mm
location
0.32
0.36
FLW8
1.3
0.27
FLW8
mm,
smin'and
location
<1.35
FLW3
1.53
0.45
FLW8
No. of
samples
5
6
No. of
samples
5
6
'Standard deviation of the mean.
Standnrd deviation due to counting statistics-
Location of the sample shown in Fig. 7.5.
MFC is the maximum permissible radionuclide concentration allowed in water as defined in Title 10, Code of Federal Regulations,
Par¥ 20, Appendix B, column 2 of Table 2 (Jan. 1, 1977).
Not determined -
00
-------
88
The radionuclide concentrations in sediment reveal disequilibrium
between 238U and its daughters. Comparison between the average 238U con-
centrations in drainage and pond sediment reveal a marked difference (51.56
and 7.05 pCi/g respectively). Even though not as great, this difference
is evident when comparing the concentrations of 238U in drainage sediment
samples with drainage soil samples (Table 7.7) and core samples (Table 7.9).
Slight contamination from phosphoric acid spillage is suspected. Concentra-
tions of 232Th could not be established.
The radionuclide concentrations in water of sample FLW3 appear to be
anomalous. The concentrations of 238U, 230Th, 226Ra, and 210Pb are
915.67 ± 10.12, 0.45 ± 0.45, 8.62 ± 0.22, and 13.50 ± 9.00 fCi/ml respec-
tively. Values after the ± sign are standard deviations. The corresponding
radionuclide concentrations in sediment were well within the range listed
in Table 7.12. The geographical location of this sample, shown in Fig. 7.5,
reveals that the drainage path terminates at this point. Thus, accumulation
of both dissolved and undissolved radionuclides discharged along its path
is probable. These values are still well within the MPCw's stated in
Table 7.12.
7.3.6 Background measurement
A radiological assessment was made of the area surrounding the URC
property so that the local environment could be characterized. Four water
and corresponding sediment samples were collected at locations shown in
Fig. 7.3 and were analyzed for the specific radionuclides mentioned for
the on-site survey. In addition to this, gamma spectrometric analysis was
performed on all sediment samples. These data are summarized in Table 7.13.
The soluble off-site 238U concentrations are lower than those in both on-
site pond and standing water (Tables 7.11 and 7.12 respectively). This
difference, excluding anomalies, is not significant when compared to the
MPCw value for 238U. The average soluble 226Ra concentration in on-site
drainage paths also appears to be elevated when compared with the off-site
average concentration (0.90 and 0.32 respectively).
Off-site sediment samples reveal that 238U is in near equilibrium with
its daughter products, which is the case for the sediment concentrations
-------
Table 7.13.
Radionuclide concentrations of the uranium decay series in standing water and sediment
off the URC site
Sample
type and
concentration
Water
fCi/ml
Sediment
pCi/g
d
MFC ,
w
fCi/ml
23eu
Concentration
max,
s ,b and
av max
and s location
8.39 11.03
1.07 3.70
FLW12
20.15 29.45
4.03 0.21
FLW11
40,000
226Ra
min,
rain' No. of
location samples
6.16 4
3.63
FLW11
10 . 78 /:
0.10
FLW12
Concentration
Sample
type and
concentration
Water
fCi/ml
Sediment
PCi/g
MPCd,
w
fCi/ml
max,
s , and
av max
and s- location
av
0.32 0.52
0.08 0.06
FLW12
17.9 27.1
3.6 0.26
FLW11
30
min,
s . , and .. .
mm No . of
location samples
0.19 4
0.04
FLW10
11.2 4
0.13
FLW12
23oTh
Concentration
max,
s , and
av max
and s location
e 0.45
0.45
FLW10
17.05 24.3
2.96 3.19
FLW11
2000
2iopb
min,
s . , and
mm
location
0.09
0.32
FLW9
10.00
0.81
FLW12
No. of
samples
4
4
Concentration
av max and min and
and s location
av
e 7.2
3.10
FLW12
18.0 27.4
3.54 3.3
FLW11
100
location
<2.25
FLW10
10.7
1.13
FLW12
No of.
samples
4
4
Standard deviation of the mean.
Standard deviation due to counting statistics.
Location of the sample, shown in Fig. 7.3.
MFC is the maximum permissible radionuclide concentration allowed in water as defined in Title 10, Code of
Federal Regulations, Part 20, Appendix B, column 2 of Table 2 (Jan. 1, 1977).
Not determined.
00
-------
90
in the on-site pond as well (see Table 7.11). However, this does not apply
to the sediment samples in on-site drainage paths (Table 7.12). Elevated
concentrations of 238U along the drainage paths suggest possible contamina-
tion by spillage of uranium-bearing phosphoric acid. The average 232Th
concentration in two off-site sediment samples is 0.76 ± 0.23 pCi/g.
Fourteen soil samples (shown in Fig. 7.3) were collected at both sur-
•
face and 6-in. depths on undisturbed (reclaimed) land surrounding the URC
plant. A representative portion of these samples was analyzed for 230Th,
210Pb, 235U, and 238U to verify natural isotopic abundance. A summary of
the data for 238U and 232Th is presented in Table 7.14. From this compila-
tion, it was determined that 238U is in near-equilibrium with daughter
products through 210Pb. Concentrations of 232Th are listed and assumed to
be in equilibrium with daughter products. Only a few data were available
to confirm this assumption.
Concentrations of 238U in sample FLOS3 (Fig. 7.3) appeared to be
anomalous when compared to the concentrations found in other off-site
samples. The surface- and 6-in.-deep concentrations were 49.59 ± 0.34 and
64.46 ± 0.40 pCi/g respectively. The 232Th concentration at both these
depths was 1.41 ± 0.30 pCi/g.
Apparently, the 238U concentrations at the surface and 6-in. depth
are similar in Table 7.14. However, when comparing only those sampling
sites for which uranium concentrations have been determined for each sample,
a surface-to-6-in.-depth ratio of 2.21 ± 0.49 is established. Comparing the
232Th concentrations at both depths in the same manner described above, the
observed ratio is 0.94 ± 0.08. Uranium to thorium ratios have also been
established this way and are presented in Table 7.14 for both depths.
The average external gamma-exposure rate in this region is 14 yR/hr
with a standard error of 2.9. The maximum exposure was 40 ± 4.7 yR/hr
detected at FLOS3, and the minimum was 3.8 ± 3.0 yR/hr at sample location
FLOS10.
A significant contrast can be drawn when comparing the radionuclide
concentrations in soil off- and on-site. The average off-site 238U con-
centration at the surface and at a depth of 6 in. was about 6 pCi/g, with
-------
Table 7.14.
238U and 232Th concentrations in soil off the URC site
Sampling
depth
(in.)
0
(surface)
6
Ratio
["surface value"!
|6-in. value J
Sampling
depth
(in.)
0
(surface)
6
av
and s
4.99
1.85
5.94
3.37
2.21
0.49
av
and s
12.30
2.69
5.22
1.67
238U
Concentration
(pCi/g)
max,
s , and
max
a location
av
15.83
0.13
FLOS2
27.42
0.18
FLOS1
2.82
FLOS 11
Activity ratio,
max and
location
av
23.98
FLOS 2
13.44
FLOS13
232Th
Concentration
(pCi/g)
min, max, min,
s . , and s ,and s . , and ,, _
mm No. of av max mm No. of
location samples and s location location samples
av ^
0.90 8 0.35 0.66 0.19 10
0.03 0.04 0.06 0.02
FLOS12 FLOS2 FLOS9
0.42 8 0.29 0.47 0.09 7
0.03 0.05 0.02 0.02
FLOS10 FLOSS FLOS13
0.77 4 0.94 1.18 0.68 4
FLOS3 0.08 FLOSS FLOSS
238u/232Th
min and No. of
location samples
2.81 8
FLOS12
1.88 7
FLOS4
Standard deviation of the mean.
Standard deviation due to counting statistics.
"Location of the sample, shown in Fig. 7.3.
-------
92
238U in approximate equilibrium with its decay products (tabulated from
16 samples). For core samples taken on-site, the average 238U concentra-
tion varied from approximately 23 to 37 pCi/g, increasing with depth to
12 in. The average 238U concentration in soil along on-site drainage paths
was about 27 pCi/g at the surface and about 39 pCi/g at the 6-in. depth.
Both these values are in good agreement with the 238U concentrations in
on-site cores. When comparing the uranium- to- thorium ratio between soil
collected off-site and on-site (soil in drainage paths) , it is seen that
the ratio for on-site samples is approximately three times higher than off-
site samples.
A final point concerns the maximum concentration of 238U found both
on and off the property. Of 18 values off-site, the highest concentration
in soil (and sediment) was 64 ± 0.40 pCi/g (FLOS3 at the 6-in. depth).
On the site, several cases (10 of 252 values) existed in which 238U con-
centrations were greater than 100 pCi/g in soil and sediment. Although
more complete data collection and statistical analysis would be required
to permit strong conclusions, some contamination of the property is
suspected.
7.4 Conclusions
Background radionuclide concentrations (especially 238U) on the
grounds and surrounding area have been characterized so that further trends
can be established. In most cases, these concentrations are typical for
reclaimed lands of this region. Slight contamination of the on-site
grounds is suspected.
Atmospheric release of particulate U308 from both the room ventilation
fans and stack is evident. The observed concentrations of natural uranium
at the point of release are close to the maximum permissible concentration
in air for an individual of the general public. Dispersion of these
releases in the atmosphere (particularly the effluent measured in the stack)
will reduce these concentrations significantly. Permanent records of all
data analyses are kept on file.
-------
93
7.5 References for Section 7
1. M. Durham and W. D. Balfour, University of Florida, Department of
Environmental Engineering, personal communication, July 1977.
2. F. F. Dyer, J. F. Emery, and G. W. Leddicotte, Comprehensive Study
of the Neutron Activation Analysis of Uranium by Delayed Neutron
Counting, ORNL-3342 (October 1962).
3. M. Lippmann, "Filter Media and Filter Holders for Air Sampling,"
pp. N2-4 in Air Sampling Instruments for Evaluation of Atmospheric
Contaminants, 4th ed., American Conference of Governmental Indus-
trial Hygienists, Cincinnati, Ohio, 1972.
4. H. W. Dickson, G. D. Kerr, P. T. Perdue, and S. A. Abdullah,
"Environmental Gamma-Ray Measurements Using In Situ and Core
Sampling Techniques," Health Phys. 30, 221 (1976).
5. E. B. Wagner and G. S. Hurst, "A Geiger-Mueller Gamma-Ray Dosimeter
with Low Neutron Sensitivity," Health Phys. _5, 20 (1961).
6. Code of Federal Regulations, Title 10, Part 20, Appendix B,
Table 2, Col. 2 (Jan. 1, 1977).
7. W. B. Harris, A. J. Breslin, H. Glauberman, and M. S. Weinstein,
"Environmental Hazards Associated with the Milling of Uranium
Ore," Arch. Ind. Health 20_, 366 (1959). (See Table 8 on p. 374.)
8. W. D. Rowe, Preliminary Findings, Radon Daughter Levels in
Structures Constructed on Reclaimed Florida Phosphate Land,
pp. 3-6, ORP/CSD-75-4 (September 1975).
9. National Council on Radiation Protection and Measurements, Basic
Radiation Criteria, Report No. 39, NCRP Publications, Washington,
D.C. (1971).
-------
94
8. ENVIRONMENTAL IMPACT
Radiological impacts of model processes for recovery of uranium from
WP phosphoric acid were assessed by calculating radiation doses to indi-
viduals and populations surrounding two representative sites in central
Florida. A previous study on milling of uranium ores is closely related
to the present study. Thus the assessment methodology used in this study
is similar to the one used for milling. There are, however, some differ-
ences in the processes. The major difference is that in the processes for
recovery of uranium from phosphoric acid, the uranium is already dissolved,
2
with no additional mining, leaching, or tailings disposal required. The
calcination and packaging operations are nearly the same for milling and
for the uranium recovery plants, but the operations are on a smaller scale
for uranium recovery plants. Dusting in the calcination and packaging
operations is the major problem in uranium recovery from phosphoric acid
with natural uranium and thorium (Sect. 4) constituting the major radio-
nuclides of the source term for calculation of impacts.
The removal of uranium from phosphoric acid also has a positive impact
because the removal of uranium and thorium reduces the radiological impact
of fertilizer usage. A detailed radiological assessment of the impact of
reduction in uranium content of fertilizers would require a much larger
effort than is possible in this study; thus, we have limited our investiga-
tion to a few general cases. These cases, described in Sect. 8.3 below,
were chosen primarily to provide an order of magnitude estimate of the
radiological impact of fertilizer usage for comparison with the local
impact of uranium recovery plants.
8.1 Radiological Impact of Airborne Effluents
During Operations
Uranium and thorium isotopes in particulate form (Sect. 4) will be
released from normal operations of model plants (Sect. 3) for recovery of
uranium from phosphoric acid. The only exception is a small contribution
from uranium and thorium decay products if the product (11303 or UF^)* is
*The isotopes are of more importance than the chemical form since conserva-
tive assumptions regarding solubilities are made for purposes of radio-
logical assessments. »^
-------
95
stored for a significant length of time before shipment (see Figs. 4.3
and 4.4) or if the solvent extraction product is not calcined or dried
immediately. The doses to individuals and populations surrounding the
model plants will thus result primarily from these isotopes and their
daughters which build up in the environment during and after plant
operation.
No dose calculations are presented for liquid streams because there
will be no intentional release of any liquid waste during normal operating
conditions. For the model plants, all areas where uranium-bearing solutions
could be lost due to tank rupture are curbed so that the product value can
be recovered and put back into the system. Other liquid releases (e.g.
organics in the black or green acid) are disposed of through the waste
streams of the WP phosphoric acid plant. Another potential source of
liquid release is an accident during transport of solutions between WP
modules and the central plant for Model 1 (Sect. 4.1.1). However, there
is less than one chance in 20 that one accident resulting in the release
of about 70 mCi (Table 4.5) of uranium will occur during the 30 to 40 yr
operating life of a plant. The potential consequences of such a release
were not analyzed in detail since analyses of accidents are outside the
scope of this assessment (Sect. 3). However, the concentration of uranium
in the released phosphoric acid (6.8 yCi/liter) would be only about 200
times the maximum permissible concentration in drinking water (0.03 pCi/
liter) for the general population (10 CFR 20). Thus, it would not appear
that such an accident would have significant radiological impact on the
general population. A more conclusive analysis would, of course, require
a much larger effort than was possible for this study.
8.1.1 Models and assumptions
The source terms given in Tables 4.5 and 4.6 provide input for models
3 4
and computer codes ' that have been developed at ORNL for assessing the
dose to man due to environmental releases of radioactive materials. The
specific code used to assess the impacts of atmospheric release of uranium
and thorium from uranium recovery plants (Sect. 3) is the AIRDOS-II computer
3
code. This code estimates atmospheric dispersion and surface deposition
-------
96
of released radionuclides as a function of direction and distance from a
facility; it also estimates doses to man through inhalation, air immersion,
exposure to contaminated ground, food ingestion, and water immersion.
When sufficient data on environmental transport or in vivo metabolism of
radionuclides are not available, dose estimates are made higher than expected
by adopting conservative assumptions regarding transport parameters. Using
AIRDOS-II, doses to total body, Gl tract, bone, thyroid, lungs, muscle,
kidneys, liver, spleen, testes, and ovaries may be estimated. The particular
organ-dose estimates will depend on the radioisotope under consideration and
its relevant radiological properties. The output of AIRDOS-II is either
annual population doses (person-rem/yr) or the highest annual individual
doses in the assessment area (rem/year)—or both—based on a continuous
release of the radionuclides of concern. These doses are summarized in out-
put tables by nuclides, modes of exposure, and organs (Sect. 8.1.3). The
location of the highest individual organ doses can be specified for each
assessment area.
Input information necessary to apply AIRDOS-II is described in detail
in ref. 3. Only the information specific to the assessment of uranium
recovery from phosphoric acid will be discussed here. Specific meteoro-
logical, population, and agricultural data are needed for each site.
8.1.2 Site-specific meteorological, population, and agricultural data
Given plant characteristics, release rates, and meteorologic conditions,
3
the AIRDOS-II computer code calculates approximate annual average concentra-
tions of nuclides of interest in the air at various distances and directions
from the plant. For particulate releases, atmospheric dilution factors
(x/Q* values) at ground level are used in conjunction with deposition
velocities to estimate air concentrations at ground level and rates of dry
134
deposition on ground surfaces. ' ' Wet deposition rates are calculated
using average x/Q' values in the vertical column above ground surfaces
together with appropriate scavenging coefficients. General plant and meteoro-
logical information for the model uranium recovery plants in Florida are
given in Tables 8.1 and 8.2. References to discussions of the information
are given in Table 8.1. The more detailed wind data and population and
-------
97
Table 8.1. General meteorological and plant information supplied to
AIRDOS-II program3 for model uranium recovery plants in Florida
References
Average air temperature (°K)
Average vertical temperature gradient (°K/m)
In stability class E
In stability class F
In stability class G
Rainfall rate (inches/yr)
Height of lid (m)
Gravitational fall velocity (m/sec)
Deposition velocity (m/sec)
Scavenging coefficient (sec"1)
Effective decay constant in plume (day"1)
U-234
U-235
U-238
Th-227
Th-228
Th-230
Th-231
Th-232
Th-234
Stack height (m)
Stack diameter (m)
Effluent velocity (m/sec)
Rate of heat emission (cal/sec)
295.4
0.0728
0.1090
0.1455
54
1070
0
0.01
3.1E-5
7.6E-9
2.7E-12
4.3E-13
3.8E-2
9.9E-4
2.4E-8
6.5E-1
1.3E-13
2.9E-2
10
0
0
0
3,6
5,6
6
3
3,4
1,4
1,3,4
Calculate!
Assigned vali
Assigned vali
No plume risi
No plume risi
-------
98
Table 8.2. Minimum and maximum x/Q' values for ground-level release
at the Florida sites3
Distance (km)
0.8
1.2
2.4
4.0
5.6
7.2
12.1
20.1
32.2
48.3
64.4
80.5
Maximum x/Q1
5.1E-6
3.1E-6
9.9E-7
3.4E-7
1.7E-7
l.OE-7
3.4E-8
1.1 E-8
3.9E-9
1.4E-9
6.8E-10
3.7E-10
Minimum x/Q '
7.7E-7
4.2E-7
1.3E-7
4.7E-8
2.4E-8
1.4E-8
5.5E-9
2.1E-9
8.3E-10
3.7E-10
2.0E-10
1.2E-10
Q1 is a reduced release rate that takes into account removal of
radioactive nuclides from the plume (by decay and deposition processes)
as it moves downwind from the release point. A complete definition is
given by Moore^.
-------
99
agricultural data needed to estimate external and internal doses to man
through the various pathways are discussed below.
Meteorological data. Meteorological data on wind direction and speed
for seven stability classes were obtained for Tampa, Florida, from the U.S.
Department of Commerce. These data, which were converted for use in the
AIRDOS-II code, are given in Tables 8.3-8.5. Atmospheric dilution values
(x/Q* values) were calculated for sectors in 16 compass directions bounded
by radial distances of 0.5, 1, 2, 3, 4, 5, 10, 15, 25, 35, 45, and 55 miles
surrounding points of release about 30 miles east of Tampa near Mulberry,
Florida (Site 1), and about 10 miles south of Tampa (Site 2). Maximum and
minimum x/Q' values are given in Table 8.2 as a function of distance from
the sites. All x/Q' values for any direction fall within the limits shown
in Table 8.2. Most uranium recovery plants will be located in Polk County,
Florida (nearer to Site 1), but complete meteorological data are not
available for the reference sites. A comparison of average annual data
for areas in south-central Florida indicated that the Tampa data are represen-
tative of the area. Partial data for Bartow, Florida (near Site 1) are
available but do not include data for the hours of 4:00 PM to 8:00 AM, the
period when more stable conditions and less dispersion and higher concentra-
tions may occur. South-central Florida also exhibits a flat terrain, and
wind patterns are not expected to shift dramatically over distances of
50 miles. Thus the complete Tampa data were judged to be more appropriate
for assessment of the Florida sites than the incomplete Bartow data.
Population data. Population data for the sectors used to calculate
X/Q' values were obtained from the PANS computer code for the two sites in
Florida. The population data for the two sites are given in Tables 8.6 and
8.7. The PANS population may be inaccurate within a 5-mile radius, because
it is inherently assumed that the entire population of an enumeration
district is located at its centroid. Within the 5-mile radius, it is neces-
sary to check the PANS output against the actual population distribution.
Land-use maps, discussed under agricultural data below, were used to check
the PANS output, especially for areas near the plant sites. Potential
population growth is not considered for this assessment, but the appropriate
population growth model could be used to modify the dose estimates given in
-------
100
Sector
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
Table 8.3.
A
0.0041
0.0014
0.0013
0.0010
0.0010
0.0
0.0024
0.0032
0.0031
0.0019
0.0
0.0
0.0027
0.0040
0.0075
0.0
Frequency of atmospheric stability classes for each direction
(fraction of time in each stability class)
B
0.0641
0.0600
0.0536
0.0574
0.0429
0.0253
0.0368
0.0602
0.0496
0.0442
0.0269
0.0508
0.1110
0.1655
0.1437
0.0883
C
0.1524
0.1783
0.1477
0.1224
0.1173
0.1169
0.0968
0.1174
0.0982
0.0949
0.0701
0.1449
0.2917
0.3024
0.2487
0.2463
D
0.4967
0.4129
0.3828
0.3346
0.3245
0.3130
0.3689
0.3544
0.4142
0.4332
0.3820
0.4531
0.4225
0.3403
0.2970
0.4661
E
0.1380
0.1786
0.1826
0.1890
0.2054
0.1979
0.1884
0.1986
0.1891
0.1958
0.2449
0.1882
0.0927
0.0864
0.1551
0.1231
F
0.1017
0.1424
0.1508
0.2142
0.2467
0.2731
0.2367
0.2057
0.1944
0.1694
0.2286
0.1282
0.0530
0.0778
0.1198
0.0618
G
0.0431
0.0264
0.0811
0.0813
0.0623
0.0738
0.0702
0.0606
0.0515
0.0605
0.0476
0.0349
0.0264
0.0235
0.0281
0.0144
Table 8.4. Frequencies of wind directions and reciprocal-averaged wind speeds
Wind speeds for each stability class
(meters/sec)
n i nu tuwai u~ r i tr^ucn^y
A
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
059
055
060
076
153
100
068
050
051
042
047
038
090
040
033
037
1.
1.
1.
1.
1.
0.
1.
1.
1.
1.
0.
0.
1.
1.
1.
0.
31
99
99
99
99
0
99
99
99
99
0
0
31
99
87
0
B
2.10
2.45
2.27
1.85
2.40
2.09
2.07
2.44
2.43
1.77
1.91
2.15
2.82
3.23
2.63
2.39
C
3.73
3.49
3.41
3.53
3.32
3.40
3.37
3.37
3.30
3.69
4.01
3.61
4.43
3.65
3.87
3.86
4
4
4
4
3
3
4
4
4
4
5
4
5
4
4
4
D
.06
.27
.04
.11
.95
.97
.25
.23
.20
.86
.03
.76
.06
.66
.00
.66
E
2.97
3.00
3.26
3.08
3.09
3.20
3.29
3.40
3.44
3.28
3.39
3.26
3.37
3.21
2.89
3.10
F
1.80
1.76
1.83
1.84
1.95
1.90
1.97
1.85
1.86
1.77
2.07
1.81
1.62
1.82
1.84
1.56
G
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
0.77
Directions are numbered counterclockwise starting at 1 for due north.
-------
101
Table 8.5. Frequencies of wind directions and true-average wind speeds
Wind toward0 Frequency
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
0.059
0.055
0.060
0.076
0.153
0.100
0.068
0.050
0.051
0.042
0.047
0.038
0.090
0.040
0.033
0.037
aWind directions are numbered
Sector
Table 8.6.
surrounding
a
0-1 1-2 2-3
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11 -SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
0 0
0 0
0 0
0 0
0 0
0 1407
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
Wind speeds for each stability class
(meters/sec)
A B C D E F G
1.82 2.94 4.41 5.09 3.15 2.24 0.77
2.35 3.12 4.12 5.06 3.20 2.22 0.77
2.35 3.10 4.03 4.85 3.49 2.26 0.77
2.35 2.71 4.14 4.89 3.29 2.26 0.77
2.35 3.12 4.06 4.79 3.30 2.33 0.77
0.0 2.96 4.10 4.93 3.43 2.30 0.77
2.35 2.75 4.10 5.23 3.53 2.34 0.77
2.35 3.13 4.09 5.08 3.64 2.27 0.77
2.35 3.05 4.06 5.19 3.68 2.28 0.77
2.35 2.72 4.67 5.84 3.51 2.22 0.77
0.0 2.70 4.84 5.99 3.63 2.38 0.77
0.0 2.90 4.62 5.99 3.49 2.25 0.77
1.82 3.53 4.91 5.65 3.61 2.12 0.77
2.35 3.69 4.19 5.32 3.44 2.25 0.77
2.29 3.44 4.38 5.06 3.05 2.27 0.77
0.0 3.17 4.67 5.55 3.31 2.07 0.77
counterclockwise starting at 1 for due north.
Population at successive distances and directions
model uranium recovery plant near Mulberry, Florida
Site 1
Radial distance (miles)
3-4 4-5 5-10 10-15 15-25 25-35 35-45 45-55
0 0 5213 50420 17647 1272 0 8144
2687 14 1217 3797 6523 13036 16237 4697
0 0 1608 4756 23497 1637 5433 6408
00 0 958 30642 236626 64290 68629
000 0 4596 72065 172761 310325
0 0 0 1703 3363 11464 25333 37764
0 0 1374 0 0 1842 11708 124533
000 0 0 0 790 632
0 0 176 0 631 814 6319 5457
00 0 0 10569 2875 931 353
000 4100 0 4200 11153 3891
0 0 929 1975 0 11685 602 717
00 0 511 10385 4910 0 309
0 0 9793 4365 8870 3126 0 3316
0 0 3098 2777 45502 13038 9670 28466
0 533 1641 4285 9728 1173 12 19894
-------
102
Table 8.7. Population at successive distances and directions
surrounding a model uranium recovery plant located near Tampa, Florida
Site 2
Sector
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11-SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
Radial distance (miles)
0-1
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
1-2
0
0
0
0
0
0
0
0
0
382
554
0
0
0
0
0
2-3
1959
0
0
0
0
0
0
0
0
0
0
3246
0
0
1004
0
3-4
95
0
0
0
0
0
0
0
0
0
0
0
601
0
1437
132
4-5
1933
0
1098
0
0
0
0
0
0
0
0
0
1624
0
0
4611
5-10
26638
68270
61066
42529
91639
0
0
0
1966
0
1975
0
1717
8282
9715
4212
10-15
29691
74030
34967
0
5261
11598
0
3099
4247
2611
0
1291
0
4049
4589
7626
15-25
4419
11660
6092
61949
130979
211873
5393
6238
2999
0
97
3077
2566
18433
15988
1096
25-35
4231
4099
39603
42518
36048
0
5191
85612
40915
0
0
455
18001
81263
5199
14654
35-45
7594
9253
809
0
0
0
0
46174
18251
599
3886
13066
12367
53939
1708
13350
45-55
9641
78
0
0
0
0
0
0
0
432
2603
2129
21239
18215
0
5506
Sect. 8.1.8. Population growth should also be considered in a broader
assessment, as discussed in Sects. 3 and 9.
Agricultural data. AIRDOS-II estimates ingestion doses resulting
from deposition of radionuclides on cropland and pasture separately for
vegetable, beef, and milk consumption. Input data are the number of beef
cattle, dairy cattle, and square meters of area on which vegetable crops
are produced for each sector around a release point. The sectors are
defined in the same manner as for x/Q' calculations and population
distribution.
The necessary data for the two sites in Florida were obtained on a
county-by-county basis for every county or part of a county within 55 miles
of the two sites. Many of the counties supplied detailed information to
our direct requests. Fortunately, the counties that would be expected to
experience the greatest impact responded with the most detailed data.
These data were in the form of land-use maps that were used to construct
-------
103
a master map of the areas surrounding each site. An overlay of the master
map consisted of the sectors bounded by the 16 compass directions and
radial distances, for which x/Q' values were calculated, and allowed the
assignment of the needed information for each sector. When detailed data
were unavailable for a county, an average for the whole county or for that
part of the county within the assessment area was used. None of the counties
where averaging was used contained a population center, and some judgment
was applied on the basis of maps of the area to ensure that averaging was
adequate. The data are given in Tables 8.8-8.13.
An important Florida crop is citrus fruits and this category is not
incorporated into the AIRDOS-II computer code. However, a study of Florida
citrus fruits by EPA may provide a basis for an analysis of the contribution
of this food crop to human internal dose.
Table 8.8. Beef cattle at successive distances and directions surrounding
a model uranium recovery plant near Mulberry, Florida
Site 1
Sector
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11-SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
Radial distance (miles)
0-1
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
1-2
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
2-3
0
81
81
40
20
0
0
0
0
0
0
0
0
0
0
0
3-4
40
81
81
162
81
0
0
0
0
0
0
0
0
0
0
0
4-5
162
162
162
162
162
0
0
0
0
0
0
0
0
0
0
0
5-10
485
1130
484
0
0
593
1179
0
0
0
0
0
0
323
646
485
10--15
485
2478
0
0
0
890
3558
1482
0
0
0
0
0
485
808
969
15-25
1615
7762
5930
0
0
593
8411
5042
0
1648
2826
0
0
0
0
0
25-35
90
3969
3097
0
0
1018
5180
5457
4004
7183
11068
4071
3230
7752
2907
0
35-45
2145
5807
5585
0
108
2683
6542
7430
8038
2440
6287
9217
7106
6175
7440
574
45-55
1240
2985
7049
3377
900
0
3819
5059
5452
6376
6853
9607
7200
11280
3624
1548
-------
104
Table 8.9. Beef cattle at successive distances and directions surrounding
a model uranium recovery plant near Tampa, Florida
Site 2
Sector
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11-SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
Radial distance (miles)
0-1
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
Table
1-2
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
8.10.
a
2-3
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3-4
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
4-5
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
5-10
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
10-15
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
15-25
735
514
180
288
468
0
0
1826
2822
2846
9132
7709
831
1494
1661
2887
25-35
3969
4116
2124
162
108
0
0
332
5146
5146
3581
0
604
2567
1576
4190
35-45
4107
4052
147
0
0
0
0
151
5453
6110
332
3768
755
302
0
5132
45-55
2460
960
0
0
0
0
0
0
5187
6395
10107
17662
0
5545
1953
1180
Milk cattle at successive distances and directions surrounding
model uranium recovery plant near Mulberry, Florida
Site 1
Sector
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11-SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
Radial distance (miles)
0-1
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
1-2
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
2-3
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3-4
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
4-5
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
5-10
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
10-15
0
0
0
0
0
0
0
0
0
0
0
0
77
0
0
39
15-25
671
749
958
2395
6706
958
206
376
0
119
3110
864
516
284
310
387
25-35
1084
429
958
4311
1437
1258
1034
1074
0
408
938
320
0
0
103
1187
35-45
135
630
1463
4790
140
300
1165
1166
0
17
547
1280
0
114
353
2079
45-55
225
292
1290
1200
1176
0
259
24
0
90
668
495
180
282
537
540
-------
105
Table 8.11. Milk cattle at successive distances and directions surrounding
a model uranium recovery plant near Tampa, Florida
Site 2
Sector
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11-SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
Table !
Sector
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11-SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
Radial distance (miles)
0-1
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
1-2
479
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
2-3
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3-4 4-5
239
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3.12. Food crops (square meters)
a model uranium recovery
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
5-10
479
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
10-15
718
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
15-25
659
514
180
288
468
0
0
1826
2822
2846
9132
7709
831
1494
1661
2887
at successive distances and
plant near Mulberry, Florida
Site 1
25-35
810
4116
2124
162
108
0
0
332
5146
5146
3581
0
604
2567
1576
4190
35-45
766
4052
147
0
0
0
0
151
5453
6110
332
3768
755
302
0
5132
45-55
328
960
0
0
0
0
0
0
5187
6395
10107
17662
0
5545
1953
1180
directions surrounding
Radial distance (miles)
0-1
0,
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
1-2
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
2-3
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3-4
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
4-5
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
5-10
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
10-15
0
6.4E8
1.3E6
0
0
0
0
0
0
0
0
0
3.0E4
0
0
1.5E4
15-25
2.9E5
8.1E4
7.1E6
7.7E6
0
1.3E6
1.1 E6
2.2E6
1.7E6
9.5E5
8.1E4
3.4E5
3.8E5
1.1 E5
1.2E5
1.5E5
25-35
4.3E8
6.6E4
0
0
0
3.9E6
4.3E6
4.3E6
2.0E6
0
0
2.5E5
1.6E5
4.9E5
2.2E5
4.6E5
35-45
0
2.4E5
0
1.9E6
4.1E3
1.9E6
4.4E6
4.7E6
2.6E6
2.5E6
7.0E5
7.1E5
4.4E5
6.8E5
1.0E6
4.2E6
45-55
1.5E6
1.1 E6
4.9E5
3.9E6
3.6E3
0
7.7E5
8.1E4
1.9E6
2.0E6
7.2E5
7.5E5
9.1E5
1.4E6
9.3E6
7.0E6
-------
106
Table 8.13. Food crops (square meters) at successive distances and directions
surrounding a model uranium recovery plant near Tampa, Florida
Site 2
Sector
Radial distance (miles)
0-1 1-2 2-3
1-N
2-NNW
3-NW
4-WNW
5-W
6-WSW
7-SW
8-SSW
9-S
10-SSE
11-SE
12-ESE
13-E
14-ENE
15-NE
16-NNE
8.1.3
0
0
0
0
0
0
0
0
o -
0
0
0
0
0
0
0
Radiation
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
doses
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3-4
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
4-5
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
from airborne
5-10
0
0
0
0
0
0
0
0
6.4E5
0
0
0
0
0
1.3E6
6.4E5
10-15
0
0
0
0
0
0
0
6.4E5
2.6E6
0
0
1.3E6
0
0
6.4E5
6.4E6
15-25
0
1.3E6
6.4E5
1.2E3
2.0E3
0
0
3.4E6
2.3E6
0
1.4E6
6.7E5
0
0
3.9E6
4.5E6
25-35
0
0
0
8.1E2
4.0E2
0
0
2.7E5
4.1E6
0
4.1E6
3.5E6
8.9E4
2.3E5
1.6E5
1.0E5
35-45
8.3E5
7.6E5
0
0
0
0
0
4.0E4
1.5E6
0
4.7E6
3.4E6
2.3E6
6.1E5
7.9E5
6.5E5
45-55
8.3E5
3.2E5
0
0
0
0
0
0
2.1E6
0
3.1E6
2.5E6
3.0E5
9.4E5
9.3E5
3.6E6
effluents
Annual total body and organ doses were calculated by AIRDOS-II for the
maximally exposed individual and to the population within 50 miles, assuming
a continuous release from the model uranium recovery facilities. Exposure
pathways included inhalation, ingestion, surface exposure, air immersion,
and swimming. Organ doses were calculated for bone, lungs, kidneys, liver,
testes, ovaries, thyroid, GI tract, muscle, and spleen. This report has a
limited discussion of AIRDOS-II application because detailed descriptions
3 4
of the models and their application are available. ' Pathways that contrib-
uted less than 0.1% and organs that received doses less than, and in some
cases equal to, total body doses are arbitrarily eliminated from further
discussion. Isotopes of uranium and thorium (source term for Model 3,
Case 1, crude UF^) that generally contributed less than 1% to organ doses
are similarly eliminated from further discussion. Resuspension of deposited
activity during plant operation makes an insignificant contribution to
-------
107
internal doses compared to direct inhalation of activity. Long-term post-
operational doses are discussed in Sect. 8.1.4. Total body and organ dose
conversion factors for the major nuclides and major pathways selected
according to the above criteria are given in Table 8.14. These dose conver-
sion factors were taken from ref. 4 where their derivation and application
are discussed in detail. For unit uptakes of the uranium and thorium
isotopes, 230Th inhalation results in bone doses about a factor of 260
times higher than for the uranium isotopes. In general, the inhalation
pathway results in the highest total body or organ doses for unit uptake
of the isotopes (bone when 230Th is present, and lungs when 230Th is not
present). Determination of actual uptake involves many variables, the
values for which are already incorporated into the AIRDOS-II and other
3 4
codes developed at ORNL, and include sample calculations. '
Maximum individual doses. Annual total body and organ doses to the
maximally exposed individual for the major nuclides and major pathways are
given in Table 8.15 for the "worst case" release from a model uranium
recovery plant (Model 3, Case 1, crude UF^ product). The maximally exposed
Table 8.14. Total body and organ dose conversion factors for major nuclides
emitted from model uranium recovery plants
Nuclides
23tu
2351J
238U
230Th
Pathway
Inhalation
Ingestion
Surface
Inhalation
Ingestion
Surface
Inhalation
Ingestion
Surface
Inhalation
Ingestion
Surface
Total body
(rem/yCi)
1.3
5.1E-2
5.4E-4
1.2
4.8E-2
5.7E-2
1.2
4.5E-2
4.4E-3
1.4E+2
5.7E-2
7.0E-4
Organ (rem/pCi)
Bone
2.1E+1
8.3E-1
6.4E-4
2.0E+1
7.9E-1
8.9E-2
1.9E+1
7.6E-1
7.8E-3
5.2E+3
2.0
l.OE-3
Lungs
5.4E+1
5.1E-2
2.3E-4
5.0E+1
4.8E-2
5.0E-2
4.7E+1
4.5E-2
3.6E-3
6.4E+2
5.7E-2
4.1E-4
Kidneys Liver Testes
5.0
2.0E-1
1.8E-4
4.7
1.8E-1
4.1E-2
4.4
1.7E-1
3.1E-3
1.1E+3 2.3E+2
5.6E-1 1.2E-1
3.3E-4 4.0E-4
-------
108
Table 8.15. Annual total body and organ doses to the maximally exposed
individual0 for major nuclides^ and major pathways'3
Model 3, Case 1, crude UF4 product, 1000 MTU/yrd
Nuclide
23>*u
235U
23SU
230Th
Pathway
Inhalation
Surface
Ingestion
Inhalation
Surface
Ingestion
Inhalation
Surface
Ingestion
Inhalation
Surface
Ingestion
Total body
dose (mrem)
6.0E-2
4.5E-2
1.4E-1
2.5E-3
2.2E-1
5.8E-3
5.3E-2
3.6E-1
1.2E-1
6.5
5.8E-2
1.3E-1
Organ dose (mrem)
Bone
9.6E-1
5.3E-2
2.2
4.2E-2
3.3E-1
9.6E-2
8.8E-1
6.5E-1
2.0
2.4E+2
8.6E-2
4.8
Lungs
2.4
1.9E-2
1.4E-1
l.OE-1
1.9E-1
5.8E-3
2.1
3.0E-1
1.2E-1
2.9E+1
3.4E-2
1.3E-1
Kidneys
2.3E-1
1.5E-2
5.3E-1
9.7E-3
1.5E-1
2.2E-2
2.0E-1
2.6E-1
4.6E-1
5.1E+1
2.8E-2
1.3
Liver
6.0E-2
1.4E-2
1.4E-1
2.5E-3
1.6E-1
5.8E-3
5.3E-2
2.6E-1
1.2E-1
l.OE+1
2.7E-2
2.7E-1
Testes
6.0E-2
3.8E-2
1.4E-1
2.5E-3
2.6E-1
5.8E-3
5.3E-2
3.4E-1
1.2E-1
6.4
5.1E-2
1.3E-1
alndividual is located 0.5 mile downwind of site and all food is purchased
at that site. However, inhalation is the major pathway when thorium
isotopes are present.
Bother thorium isotopes contributed less than ~\% of total body or organ
doses except for lungs where 232Th contributed 1.4% and 228Th contributed
21.
cAir submersion and swimming contributed less than 0.1% to doses.
^Model 3, Case 1, semirefined U308 may be obtained by deleting 230Th from
this Table. Models 1 and 2, Case 1 are 0.1 of the semirefined U308 values.
Case 2 values are 5 x 10"1* times the values obtained in Case 1.
individual is a hypothetical individual who spends 100% of his time in the
open (unprotected by buildings, etc.) at a location 0.5 mile downwind from
the plant. It is also assumed that 100% of his food is produced at that
site. As indicated in Table 8.15, data for the other model uranium
recovery processes and radwaste treatment cases may be obtained as follows:
Model 3, Case 1, semirefined U308 may be obtained by deleting
the 230Th data. Models 1 and 2, Case 1 are 10% of the semi-
refined 11309 values, and Case 2 values are 0.05% of Case 1
values. Doses to the maximally exposed individuals are the
same for both sites, because Tampa meteorological data were
used for both sites, and it is assumed that any combination
of the model processes and radwaste treatment cases could be
located at either site. The maximum annual dose to bone from
-------
109
inhalation of 230Th was 240 mrem, as shown in Table 8.15,
for a 1000-MTU/yr product rate.
Maximum annual total body and organ doses to individuals summed over
all pathways and all isotopes are given in Table 8.16 for all model processes
and radwaste treatment cases. It is evident from Tables 8.6 and 8.7 that
bone dose due to inhalation is the major concern when 230Th is present.
When 230Th is not present, ingestion bone doses and inhalation lung doses
are about the same. These results support the assumption that 100% local
food consumption for the maximally exposed individual does not have a large
influence on maximum doses. If there were no local consumption of locally
grown food, doses would be changed only slightly with the presence of 230Th,
and about 50% without 230Th. Residence time at the location would have a
much greater effect on the doses, as would radwaste treatment cases. The
maximum annual dose to bone through all pathways is 250 mrem, as shown in
Table 8.16.
Table 8.16. Maximum annual doses'3 to individuals^ from airborne effluents
of model uranium recovery plants in Florida
Site 1 and Site 2, 1000 MTU/yr°
Model and
radwaste treatment
case
Models 1 and 2
Case 1
Case 2
Model 3, semirefined U308
Case 1
Case 2
Model 3, crude UF4
Case 1
Case 2
Maximum
total body
(mrem)
l.OE-1
5.0E-5
1.0
5.0E-4
7.7
3.9E-3
dose Bone
7.2E-1
3.6E-4
7.2
3.6E-3
252.0
1.3E-1
Maximum adult organ doses (mrem)
Lungs
5.4E-1
2.7E-4
5.4
2.7E-3
36.0
1.8E-2
Kidneys
1.9E-1
9.5E-5
1.9
9.5E-4
54.5
2.7E-2
Liver
8.2E-2
4.1E-5
8.2E-1
4.1E-4
11.8
5.9E-3
Testes
l.OE-1
5.0E-5
1.0
5.0E-4
7.7
3.9E-3
50-yr dose commitment from exposure to effluents during 1 yr.
Maximum dose at 0.5-mile downwind from plant site.
^Maximum individual dose is the same for both sites. Site 1 is near
Mulberry, Florida and Site 2 is near Tampa, Florida.
-------
110
For Models 1, 2, and 3, a properly Installed and maintained HEPA filter
(Sect. 4.5.4) would reduce all maximum doses to less than 0.1 mrem/yr. The
incremental cost of Case 2 over Case 1 is discussed in Sect. 6, and cost
vs impact is discussed in Sect. 9.
Population doses. Annual total body and organ doses for the populations
within 55 miles of Site 1 and Site 2 for all model processes and radwaste
treatment cases are given in Tables 8.17 and 8.18 for a 1000-MTU/yr product
rate. For inhalation and ingestion, these doses are 50-yr dose commitments
caused by continuous intake of the isotopes for 1 yr. External doses are
annual doses from 1 yr of exposure. The doses apply for the period during
operation of the plant. Models used and illustrative applications are
discussed in refs. 3 and 4. The annual doses and costs of control technolo-
gies (Sect. 6) give estimates of cost per person-rem for reducing doses
(Sect. 9). Postoperational doses are discussed in Sect. 8.1.4. The rela-
tive contributions of the various pathways and isotopes to population doses
are similar to those for maximum individual doses. However, magnitudes of
population doses differ for the two sites because of different population
and food production distributions. In estimating population doses, any
additional quantity (supplemental to locally produced supply) of food
required to feed the population is assumed to come from outside the assess-
ment area in an uncontaminated state. Doses due to food exported from the
area are not considered. The maximum population bone dose for all pathways
and all isotopes is 125 person-rem for Site 1 (near Mulberry, Florida) and
426 person-rem for Site 2 (near Tampa, Florida), as shown in Tables 8.17 and
8.18. These annual population doses can be multiplied by years of plant
operation (perhaps 30 to 40 yr) to obtain total population doses during
plant operation. These doses should also be multiplied by the appropriate
factor (0.2 for a 200-MTU/yr product rate) for product rates less than 1000
MTU/yr.
Population growth projections for south-central Florida through the
year 2000 are as follows:
1970 1,798,000
1980 2,787,000
1990 3,769,000
2000 4,556,000
-------
Ill
Table 8.17. Annual dose to the population from airborne effluents
of a model uranium recovery plant in Florida
Site 1, 1000 MTU/yr
Model and Population
Population adult organ doses (man-rem)
radwaste treatment total body dose „
case (man-rem)
Models 1 and 2
Case 1
Case 2
Model 3, semi refined U^OS
Case 1
Case 2
Model 3, crude UFU
Case 1
Case 2
7.1E-2
3.5E-5
7.1E-1
3.5E-4
4.0
2.0E-3
5.0E-1
2.5E-4
5.0
2.5E-3
125.0
6.2E-2
Lungs
2.8E-1
1.4E-4
2.8
1.4E-3
17.7
8.8E-3
Kidneys
1.3E-1
6.5E-5
1.3
6.5E-4
27.3
1.4E-2
Liver
5.7E-2
2.8E-5
5.7E-1
2.8E-4
6.0
3.0E-3
Testes
7.2E-2
3.6E-5
7.2E-1
3.6E-4
4.0
2.0E-3
Table 8.18. Annual dose to the population from airborne
effluents of model uranium recovery plants in Florida
Site 2, 1000 MTU/yr
Model and
radwaste treatment
case
Models 1 and 2
Case 1
Case 2
Model 3, semi refined U^C
Case 1
Case 2
Model 3, crude UFU
Case 1
Case 2
Population
total body dose
(man-rem)
1.7E-1
8.5E-5
k
1.7
8.5E-4
13.3
6.6E-3
Bone
5.9E-1
2.9E-4
5.9
2.9E-3
425.9
2.1E-1
Population
Lungs
9.4E-1
4.7E-4
9.4
4.7E-3
62.8
3.1E-2
adult organ doses (man-rem)
Kidneys
1.8E-1
9.0E-5
1.8
9.0E-4
91.7
4.6E-2
Liver
1.3E-1
6.5E-5
1.3
6.5E-4
20.1
l.OE-2
Testes
1.8E-1
9.0E-5
'1.8
9.0E-4
13.3
6.6E-3
-------
112
The projections apply to Bureau of Economic Analysis (BEA) region number 37,
which includes the counties where phosphate mining and uranium recovery
Q
facilities are likely to be located. The model, MULTIREGION, is used to
obtain the projections. Finer structure in population projections (i.e.,
county-level projections) are highly speculative and are essentially
impossible for smaller areas near the plants. For purposes of estimating
the increase in population dose due to population growth, we assume that
the present population distribution will remain constant, but that the total
population will grow according to the above projections.
Based on these assumptions and on an annual population dose of about
426 person-rem in 1980 for Site 2, Case 1, control, the annual dose would
be about 770 person-rem in 1990 and about 1100 person-rem in 2000. The
population dose during the period 1980 to 2000 would be about 15,400 person-
rem as compared to 8520 person-rem if zero population growth is assumed.
Population growth thus does not substantially alter the analysis given
herein in light of other large uncertainties in projecting dose over long
periods of time.
8.1.4 Postoperational source term
Model uranium recovery plants release small amounts of uranium and its
decay products during each year of operation. These radionuclides continue
to expose populations long after the plant has ceased operation, but a lack
of sufficient information makes accurate predictions difficult; however,
the estimates of dose are likely to be well above actual doses. It is of
interest to compare the estimated 20-yr release from the worst case (highest
release) uranium mill with the highest release from uranium recovery plants,
as shown in Table 8.19. A 30- or 40-yr release time (perhaps plant operating
life) would increase the estimated releases, and thus doses, by a factor of
1.5 and 2 respectively. The highest release rates for uranium recovery
plants were for Model 3, Case 1, crude UF^ product. The relevant comparison
is between uranium mill process dusts and uranium recovery dusts, since no
tailings pile is associated with uranium recovery per se. However, the
total radioactivity from tailings piles and mill process dusts are given
so that doses can be properly apportioned.
-------
113
Table 8.19. Curies of long-lived radionuclides released during 20-yr
operations of mills and uranium-recovery facilities3
Radionuclide
23«.u
238u
226Ra
230Th
222Rn
Mill dusts
1.79EOO
1.79EOO
1.24E-1
9.01E-2
-
Recovery
0.66
0.66
6.6E-3
0.66
-
Tailings
pile»
2.38E-2
2.38E-2
3.13E-1
3.18E-1
8.43E+3^
Total for
mill
1.81 EDO
1.81EOO
4.37E-1
4.08E-1
8.43E+3
Surface activity0 (Ci/m2)
Mill Recovery
8.90E-11 3.2E-11
8.90E-11 3.2E-11
2.15E-11 3.2E-14
2.00E-11 3.2E-11
-
Worst cases, a Wyoming mill using the alkaline leach process and a uranium recovery plant
with UF4 as product. The basis is approximately 1000-MTU/yr product rate.
Tailings dust resuspended from an average of 20 acres of dry beach over the 20-year life of
the mill and from 25 acres of untreated tailings for 2 years following mill closures and
before the final earth cover is placed.
C2.033 x 1010 m2 in an area of 50-mile radius.
Continuous annual release of 222Rn from a 128-acre tailings pile with a 6-inch earth cover
after mill has closed.
It was assumed that all of the activity released during mill opera-
tions was uniformly spread over a 50-mile radius from the mill. No specific
meteorological, population, or agricultural data were adopted for dose
calculations. Under these conditions, it is possible to compare potential
doses for uranium recovery operations with those calculated for uranium
mills. Such a generic assessment omits many details, but lack of detailed
information leaves little choice in the matter.
8.1.5 Postoperational pathways of exposure
Resuspended air activity, ingestion, and exposure via contaminated
ground were considered as the pathways of exposure. For each pathway,
conservative assumptions such as no loss of deposited radionuclides from
the soil, no downward movement of the radionuclides beyond the root zone,
and average values for resuspension factors and uptake fractions to plants
were used to estimate average individual adult doses in the assessment
area. Movement of some of the radionuclides added to soils probably does
occur, especially in areas such as Florida where average to heavy rainfall
9
occurs. According to Guimond, as much as 60% of natural uranium deposited
-------
114
on farmlands by fertilization may eventually be carried by surface runoff
to rivers and, eventually, to oceans. In arid regions where uranium mines
and mills are located, transport of uranium may be much less and the
conservative assumptions may be more appropriate. Nevertheless, in this
crude generic postoperational assessment, it will be assumed that no move-
ment occurs even though such an assumption is even more conservative for
uranium recovery in Florida than for uranium milling in Wyoming and New
Mexico.
8.1.6 Dose estimates
Radiation doses to individuals residing within 50 miles of a model
mill have been estimated for total body, bone, and lungs. All radiation
doses from ingestion and inhalation are 50-yr dose commitments from 1 yr
of exposure (i.e., the dose an individual will accrue over a 50-yr period
from 1-yr of intake). External doses are annual doses from 1 yr of
exposure. Exposure is assumed to be continuous until radioactive decay
of the isotope of concern; no other removal processes are considered.
Radiation doses to individuals residing within a 50-mile radius of a model
uranium recovery plant may be estimated under the same conditions and
assumptions used by Sears et al. The doses determined for the case of
uranium milling are apportioned to uranium recovery according to relative
release over a 20-yr period for each radioisotope. Details of the sites
will differ, but the individual dose estimates are more conservative for
Florida than for Wyoming or New Mexico because of climatic differences.
Population dose for the Florida site, given comparable individual
doses, would be higher because the population within a 50-mile radius of
the Florida site is about 30 times the population surrounding the Wyoming
site used in the uranium milling study. Population dose is estimated here
by multiplying the previously determined doses for the Wyoming site by the
ratios of the source terms and the ratios of the total populations within
7.85 x 103 miles surrounding the Florida and Wyoming sites. Individual
doses will, of course, be much higher near the release point than the
averages calculated here and much lower for individuals near 50 miles from
the release point. However, the average individual doses are extremely
low, and multiplying the average individual doses by the total population
-------
115
in the area may be a representative estimate of population dose. The
population dose, under the assumption of a linear, nonthreshold dose
response relationship, is the appropriate quantifier for estimating health
effects.
Individual and organ doses. Annual doses to an individual living in
a uniformly contaminated area of 7.85 x 10 sq miles surrounding a model
uranium mill and a model uranium recovery plant are given in Tables 8.20
and 8.21. The uranium mill data suggest the following comparisons with
respect to relative organ doses: (1) approximately 2% of total body dose
is due to inhalation of resuspended activity; (2) except for radon released
from the tailings pile, the inhalation lung dose to the average individual
is insignificant; (3) bone dose due to inhalation is about an order of
magnitude lower than bone dose due to ingestion, approximately 90% of
inhalation bone dose is from 230Th, and approximately 95% of ingestion
bone dose is from 226Ra; (4) in the absence of significant 222Rn, bone dose
(7.5 yrem/yr) is about an order of magnitude higher than lung dose (0.7
yrem/yr).
The relative organ doses from uranium recovery are different than
from uranium milling because the 226Ra release is much lower (1.5%) than
the release from the mill, and the 23^Th release is about 1.6 times higher
than the mill release. For the uranium recovery plant (1) approximately
10% of the total body dose is due to inhalation of resuspended activity;
(2) the inhalation lung dose is still very low; (3) in contrast to the
case for uranium mills, bone dose due to inhalation is about four times
higher than for ingestion, essentially 100% of inhalation bone dose is
from 230Th, about 30% of ingestion bone dose is due to 226Ra, and approxi-
mately 40% of ingestion bone dose is due to 230Th; (4) bone dose (1.5 yrem/
yr) is about 30 times higher than lung dose (0.05 yrem/yr).
The above comparisons show that thorium and 225Ra through ingestion
and thorium through inhalation have a major effect on organ doses. The
uranium isotopes play a more important role in total body dose, as can be
seen in Tables 8.20 and 8.21.
-------
Table 8.20. Major radionuclides and exposure modes contributing to the annual total body dosea to the average individual
after a uranium mill or uranium recovery facility is closed until significant decay of radionuclides occurs
Exposure mode
Radionuclide
23"»u
2381J
226Ra
230Th
Total
Submersion
Mill
1.2E-12
7.1E-12
1.3E-12
2.6E-13
9.9E-12
in air (mrem)
U-recovery
4.4E-13
2.6E-12
1.9E-14
4.2E-13
3.5E-12
Contaminated
Mill
2.3E-4
3.8E-4
2.5E-5
4.4E-5
6.8E-4
ground (mrem)
U-recovery
8.5E-5
1.4E-4
3.7E-7
7.0E-5
2.9E-4
Inhalation
Mill
8.6E-7
7.6E-7
6.5E-6
2.1E-5
2.9E-5
(mrem)
U-recovery
3.2E-7
2.8E-7
9.7E-8
3.4E-5
3.5E-5
Ingestion
Mill
6.9E-6
6.1E-6
6.6E-4
2.3E-6
6.7E-4
(mrem)
U-recovery
2.5E-6
2.3E-6
9.9E-6
3.7E-6
1.8E-5
Total
Mill
2.4E-4
3.9E-4
6.9E-4
6.7E-5
1.4E-3
(mrem)
U-recovery
8.9E-5
1.4E-4
1E-5
1.1 E-4
3.5E-4
Dose after 20 years of mill or uranium recovery plant operation from radon and the radioactive materials that were dispersed during 20 years of
operation. The uranium mill was located in Wyoming; the uranium recovery plant in Florida. Uranium mill data from Sears et al.1 The basis
is 1000 MTU/yr.
-------
Table 8.21. Annual dosea to the average individual after mill or uranium recovery plant
is closed until significant decay of radionuclides occurs
Total body dose (mrem)
Radionucl ide
23-»|J
238u
226Ra
230Th
222Rn
Total
Mill
2.4E-4
3.9E-4
6.9E-4
6.7E-5
-
1.4E-3
U- recovery
0.89E-4
1.4E-4
1E-5
1.1 E-4
-
3.5E-4
Inhalation
Mill
1.4E-5
1.3E-5
6.3E-5
7.6E-4
-
8.5E-4
U-recovery
5.2E-6
4.8E-6
9E-7
1.2E-3
-
1.2E-3
Organ
Bone
dose (mrem) per exposure mode
Lung
Ingestion
Mill
1.1 E-4
l.OE-4
6.4E-3
8.3E-5
-
6.7E-3
U-recovery
4.1E-5
3.7E-5
9.6E-5
1.3E-4
-
3E-4
Inhalation
Mill
3.5E-5
3.1E-5
5.5E-5
7.7E-6
3.4E-1
3.4E-1
U-recovery
1 .3E-5
1.1E-5
8E-7
1.2E-5
-
3.7E-5
Ingestion
Mill
6.9E-6
6.1E-6
6.6E-4
2.3E-6
-
6.7E-4
U-recovery
2.5E-6
2.2E-6
9.9E-6
3.7E-6
-
1.8E-5
QDose after 20 years of mill-or uranium-recovery plant operation from radon and the radioactive materials that were dispersed during 20 years of
operation. A comparison of releases for uranium mills versus uranium recovery plants is given in Table 8.19. The uranium mill was located in
Wyoming; the uranium recovery plant in Florida. Uranium mill data from Sears et al.1 The basis is 1000 MTU/yr.
-------
118
Population doses. Doses to the population surrounding a model mill
and a model uranium recovery plant are given in Table 8.22. In each case,
the population dose is the average individual total body or organ dose,
given in Table 8.21, multiplied by the total populations within a 55-mile
radius of the plant sites. The population within 55 miles of the mill
is about 5.3 x 10^ persons, but the population within 55 miles of the
uranium recovery plant is about 1.6 x 105. Projection of population
distribution or growth over a time frame where significant decay of the
long-lived isotopes will occur is impossible. However, a population density
of ten times that used for the above estimates would lead to population
doses no greater than 0.1% of population doses due to background concentra-
tions of the same radionuclides. The annual bone dose to the population
surrounding the model uranium recovery plant (Model 3, crude UF^ product)
is estimated to be 2.3 person-rem/yr (Table 8.22). Such an annual popula-
tion dose rate may be assumed to occur for millions of years if no other
radionuclide removal processes are assumed.
8.2 Positive Radiological Impacts
Radioactive elements in fertilizers contribute to population doses
by various pathways including inhalation, ingestion of foods and water,
9-13
and direct radiation. The relative radionuclide content of phosphate
9
fertilizers is given in Table 8.23. Removal of any of these radionuclides
from fertilizers would result in a reduction in long-term population doses.
Uranium and its long-lived daughters will persist in the environment for
millions of years; thus the potential dose reduction could be substantial.
However, there is no methodology for estimating potential doses over such
long periods. A simple, conservative approach is usually taken in such
cases and is also used here.
All the radionuclides of concern in fertilizers are natural radio-
nuclides, and the contribution of these to natural background exposures
has been studied extensively. Therefore, in this analysis, natural
background concentrations and related exposures will be used as a baseline
for scoping the impact of removal of radionuclides from fertilizers.
-------
119
The dose or dose rate (depending on. the units of background doses),
D., due to addition of radionuclide i to soil through fertilizer appli-
cation is
D. = (C./C.JD., ,
i v i ib ib'
where C. is the concentration of radioisotope i in soil due to fertilizer
application, C is the background concentration of radioisotope i, and
D is the dose from radioisotope i at its background concentration.
Table 8.22. Annual dose to the population0 after 20 years of operation
of model mills and uranium recovery plants
Nucl ide
234J
238u
226Ra
230Th
222Rn
Total
Total body
Mill
1.3E-2
2.1E-2
3.6E-2
3.6E-3
-
7.4E-2
dose (man-rem)
U-recovery
1.4E-1
2.2E-1
1.6E-2
1.7E-1
-
5.4E-1
Bone dose (man-rem)
Mill
6.7E-3
6.1E-3
3.4E-1
4.5E-2
-
4.0E-1
U-recovery
7.1E-2
6.5E-2
1.5E-1
2
-
2.3
Lung dose (man-rem)
Mill
2.2E-3
1.9E-3
3.8E-2
5.3E-4
1.8E1
1.8E1
U-recovery
2.7E-2
2.0E-2
1.7E-2
2.4E-2
-
8.5E-2
aDose to population is average total body and organ dose out to a distance of 50 miles. Actual
population within a 55-mile radius of the Wyoming model mill was 5.3 x lO4 (Sears et al. ) and
1.55 x 106 for the Florida uranium recovery plant. The basis is 1000 MTU/yr.
Table 8.23. Natural radioactivity concentrations in fertilizer materials
made from Florida phosphates (pCi/g)a
Material
Normal superphosphate
Diammonium phosphates
Concentrated superphosphate
Monoammonium phosphates
Phosphoric acid
Gypsum
Ra-226
21.3
5.6
21.0
5.0
<1.0
33.0
U-238
20.1
63.0
58.0
55.0
25.3
6.0
Th-230
18.0
65.0
48.0
50.0
28.3
13.0
Th-232
0.6
0.4
1.3
1.7
3.1
0.3
aData from Guimond.9
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120
Determination of C. requires knowledge of uptake and removal of each
radioisotope from a given soil area or volume. Detailed information is not
available for the natural radionuclides of concern over the time periods
of interest. In such a case, generalizing and simplifying assumptions are
made such that C. and, therefore, D. are maximized. The major simplifying
assumption made here is that the only removal process for the natural
radioisotopes deposited in fertilizers is mixing to plow-layer depth.
The assumption made in comparing the impacts of radionuclides in
fertilizers which have been extracted, processed, and returned to the
environment by man is that they will eventually behave like their counter-
parts in nature. Uncertainties of the chemical form, soil characteristics,
climate, and application rates render any dose calculations only qualitative,
especially for the long time periods of concern. Such uncertainties necessi-
tate assumptions that tend to maximize potential negative impacts (increases
in dose). However, the same assumptions also tend to maximize the positive
impact (reductions in dose), and more credit is taken for reduction in
impact than may be realistically justifiable.
In the case of the recovery of uranium from WP phosphoric acid, the
negative impacts have been discussed in Sect. 8.1 as the local impact of
uranium and thorium released from the recovery plants. The positive impact
to be considered is that in removing these natural radionuclides from WP
phosphoric acid, they will be less widely distributed onto farmlands, and
urban gardens and grasses where they may be more readily available for
assimilation into food chains.
8.2.1 Source terms
The contributions of phosphate fertilizer-derived uranium, thorium,
and their decay products to radiation doses will vary greatly because of
many factors, including rainfall, the extent of plowing, the characteristics
of the soil, and, finally, the amount of phosphate used. Recommended
phosphate fertilization of farmlands in Tennessee varies from as little as
20 to 40 Ib P205/acre (22 to 45 kg P205/ha) for lespedeza and soybeans to
to 150 Ib/acre (100 to 170 kg/ha) for burley tobacco. In contrast with
these recommendations, only 67% of 1,840,000 acres planted in soybeans in
-------
121
1976 received any phosphate fertilizer; an average 43 Ib
(48 kg/ha) was used on land actually fertilized. In the United States,
43,240,000 acres was planted in soybeans; only 29% of this area received
phosphate at an average 36 Ib/acre (40 kg/ha) . Various other data from
17-24
the Tennessee Valley Authority and from the U.S. Department of Agriculture
indicate that crop and pasture lands receive from 0 to 200 Ib
(225 kg/ha), with an average (including nonfertilized land of 31.8 Ib
acre (35.6 kg/ha) for the 202 million acres planted in the four major crops -
cotton, soybeans, corn, and wheat. For comparison, an average 68.3 kg
ha was used in 1973-1974 in the Federal Republic of Germany, and the maximum
13
was 1525 kg/ha.
Data listed above and the value 1 MTU/2480 MT P205 (Table 3.1) may be
used to calculate the following quantities of uranium annually spread onto
farmlands in the United States: (1) an average of 12.8 Ib of uranium per
1000 acres (14.4 kg of uranium per 1000 ha) on all land used to grow the
four major crops - cotton, soybeans, corn, and wheat - corresponding to
31.8 Ib P205/acre; a maximum of about 80 Ib of uranium per 1000 acres
(90 kg of uranium per 1000 ha), corresponding to 200 Ib P205/acre.
Calculations of uranium (and thorium in the case of Model 3) source-
term reductions due to removal of uranium from WP acid are based on the most
recent data concerning ?2®5 production and consumption, rather than capaci-
ties; this is a conservative assumption in the positive impact sense.
Such data are presented for the total United States rather than just for
21-25
central Florida; however, relative values will apply to production of
WP P2C>5 from central Florida, because about 70% of the nation's rock capacity
is located there (Table 8.24). In 1975, the actual production of WP acid
was 6,889,000 short tons (ST) as P2°5» containing approximately 2800 short
tons of uranium (STU) on the basis of 2480 tons of P205/ton of uranium;
for comparison, the production capacity was 8,638,000 ST of P2°s as WP acid
and 3480 STU. Production and capacity data indicate about 80% utilization
of capacity in 1975. Also in 1975, 4,507,000 ST of PzOs was used in the
United States, of which 487,000 ST was applied in solution form. In 1976,
587,000 ST of P205 was applied in solution form out of the total 5,230,000
ST of P2°s applied in the United States. For these two years, about 11%
-------
Table 8.24. Phosphate rock and P2&5 production capacities and
consumption in the United States
Company
Agrico Chem-Will iams
Borden Chemical
Brewster Phosphates
C. F. Industries
Cities Service
Gardinier
W. R. Grace
International Minerals
Mobil Chemical Co.
Poseidon Mines
Swift Chemical
T-A Minerals Corp.
USS Agri-Chem
Central Florida Total
United States Total
Central Florida
United States
Plant location
County
Polk
Hardee
Polk
Polk
Polk
Hardee
Polk
Polk
Polk
Polk
. Polk
Polk
Polk
Polk
Polk
Polk
Polk
Polk
City
Pierce
Fort Green
Tenoroc
Big Four
Brewster
—
Ft. Meade
Ft. Meade
Bonny Lake
Hooker's Prairie
Bonnie
Kingsford
Nichols
Ft. Meade
Lakeland
Bartow
Polk City
Ft. Meade
1973
6000
-
1000
-
3000
2000
-
2500
-
-
9500
1500
3200
-
3000
-
2000
33,700
48,625
10,500
15,100
1974
6000
-
1000
-
3500
_
2000
2500
-
-
9500
1500
3200
-
3000
-
2000
34,200
50,025
10,600
15,600
Annual capacity (thousand short tons
1975
6000
3500
1000
-
3500
.
2000
2500
-
3000
9500
1500
3200
600
3000
-
2000
41,300
57,125
Contained P205
12,800
17,800
1976
6000
3500
1000
-
6300
_
2000
2500
-
3000
9500
1500
3200
600
3000
500
2000
44,600
64,125
at 0.311
13,900
19,900
1977
6000
3500
1000
-
6300
_
2000
2500
2800
3000
9500
1500
3200
600
3000
500
2000
47,400
66,725
tons P205/ton
14,700
20,800
of rock)
1978
6000
3500
1000
-
6300
_
2000
2500
2800
3000
9500
1500
3200
600
3000
500
2000
47,400
66,725
of rockfc
14,700
20,800
1979
6000
3500
1000
1000
6300
2000
_
2000
2500
2800
3000
9500
1500
3200
600
3000
500
2000
50,400
69,725
15,700
21,700
1980
6000
3500
1000
1000
6300
2000
-
2000
2500
2800
3000
9500
1500
3200
600
3000
500
2000
50,400
69,725
15,700
21,700
Annual rock production,
U.S. total, 1000 ST«
42,137
45,686
48,816
Totalc
Used in liquids"
5085
543
Annual consumption, U.S. total (thousand short tons of P205)
5099
533
4507
487
5230
587
aHarre et al.
20
Stowasser,25 summarizing data for 1973, stated that the average grade of phosphate in the U.S. was 13.6% P205, and
marketable rock was 31.1% P205. The average weight recovery of concentrate and rock marketable as mined was 30.2%,
averaged 68.9%. Hignet25 used 34% P205 in many of his calculations.
'Ref. 21 .
dRef. 22 (Tables 1 and 2).
:he average grade of
and the P205 recovery
-------
123
of P205 consumed was applied as solution. The 6,889,000 ST of P205 in
WP acid produced in 1975 corresponds to about 45% of the P205 contained
in the 48,816,000 ST of rock, at 31.1% P205 in marketable rock. 5 For
comparison, expected production capacities listed in Tables 3.1 and 8.24
suggest that about 30% of marketable rock in central Florida and 42% of
rock from the entire United States will be converted to WP phosphoric acid.
The quantity of uranium in WP phosphoric acid that originates in
central Florida is readily calculated for the year ending June 30, 1980 as
the product of the following numbers: (1) 5,500,000 ST of P205 consumed:
(2) the fraction of P205 derived from central Florida is 0.7; (3) the
fraction of the P205 converted to WP acid is 0.3; (4) 1 STU/2480 ST of
P205. The product is about 470 STU. This will be contained in about
1,160,000 ST of P205 as WP acid of which about 600,000 will be applied as
aqueous solutions and 560,000 as solids.
The potential reduction of dose commitments from the recovery of
uranium from phosphates in central Florida is calculated on the basis of
the numbers presented earlier in this section. It should be noted, however,
that thorium is presumed to be retained in the WP acid of Models 1 and 2
but to be contained in the crude UF^ of Model 3. If the second cycle of
purification is performed in Model 3, then thorium will be contained in a
waste stream (perhaps discharged to the gypsum pile) and will not be
returned to the WP phosphoric acid stream.
If the uranium recovery industry grows such that all of the uranium
in WP phosphoric acid is recovered by 1980, then natural uranium in about
30% of the total fertilizer production will be removed; thorium will be
removed in a lesser percentage (Model 3 recovery process), which will
result in a corresponding reduction in radiation dose. If the percentage
of WP phosphoric acid is higher, then the dose reduction will be corre-
spondingly higher. An order of magnitude estimate of the possible dose
reduction can be made by comparison with background doses from the 238U
decay chain.
-------
124
8.2.2 External radiation dose estimates
External gamma dose. The average individual dose rate from exposure
to the 238U decay chain is about 10.3 mrem/yr. This dose rate results
from a surficial uranium concentration of about 1.8 ppm. At the average
fertilizer application rate of 14.4 kg of uranium per 1000 ha given above,
the uranium concentration at a plow-layer depth of 20 cm would be approxi-
mately 5 ppb. At the maximum application rate of 90 kg of uranium per
1000 ha, the plow-layer depth concentration would be approximately 30 ppb.
The fraction 0.3 (see above) of these concentrations that would be avoided
due to removal of uranium from phosphoric acid corresponds to an average of
1.5 ppb and a maximum of 9 ppb. The maximum dose rate due to external
exposure from such concentrations after the decay products had grown in
(especially radon daughters, Table 8.25) would be 9 yrem/yr at average
fertilizer application rates, and 54 yrem/yr at maximum fertilizer applica-
tion rates. Such a calculation assumes no dispersal of uranium other than
mixing to plow-layer depth and that the individual spends 100% of his time
at that location. If the uranium were assumed to accumulate year by year,
then application for about 1100 yr at the average rate and 190 years at
the maximum rate would eventually (when decay products approach equilibrium)
result in doubling the background dose due to external exposure from the
natural uranium series. At present rates of usage, phosphate reserves and,
thus, uranium from this source will be expended in a few hundred years.
Dose rates from radionuclides in fertilizers that have been and are
still being added to soils would be at least three times higher than those
previously calculated and would provide a source term for millions of years.
Thus uranium concentrations could have doubled at the maximum rate of
application in about 60 yr. There is some evidence that a doubling may have
occurred in some potato and tobacco fields.
Of the thorium isotopes received at the first cycle extraction system
(Table 4.3), 230Th makes a major contribution to population and individual
doses surrounding uranium recovery plants (Sect. 8.1.3). However, 230Th
does not contribute significantly to external exposure. Its presence will
lead to the more rapid buildup of 226Ra and its daughters, and the maximum
dose reductions will occur in a few thousand years rather than in a few
-------
125
Table 8.25 External dose conversion factors for exposures to
surface activities of the 238U decay chaina
Isotope
U-238
Th-234
Pam-234
U-234
Th-230
Ra-226
Rn-222
Po-218
Pb-214
Bi-214
Po-214
Pb-210
Bim-210
Po-210
Dose conversion factors
(mrem/yr/MCi/cm2^
Electrons
4.0E-31
4.1E-5
2.1E7
9.6E-26
2.3E-18
3.9E-8
0
0
1.6E6
1.4E7
0
0
7.2E6
0
Gamma
2.1E4
2.1E4
5.5E2
2.6E4
2.2E4
1.2E4
6.1E2
1.4E2
3.9E5
2.2E6
1.7E2
6.6E4
0
0
a
Minor contributors not included.
. 4.
million years. However, most of the 230Th and its daughters initially
present will have decayed by the time equilibrium is attained between
238U and its decay products. Therefore, the maximum dose rate will not
change substantially, but the cumulative dose will be greater because the
dose rate will occur earlier in time. However, the cumulative dose will
not substantially exceed that given above for the 238U series, since the
half-life of 230Th is about 8 x I0k yr and the half-life of 238U is
about 4.5 x 109 yr (Fig. 4.2).
-------
126
External electron dose. External electron doses (essentially skin
dose) result in a different apportionment of doses among isotopes than for
external gamma exposures. Although electron doses (including beta rays)
are not usually considered to contribute significantly to background doses,
it is interesting that exposures from electrons are substantially greater
than from gamma rays. Conservative estimates of the doses in mrem/yr for
a surface activity of 1 yCi/cm2 may be calculated via the EXREM-II computer
4
code A discussion of this code and sample calculations for both external
gamma and electron exposures are given in Ref. 4. The conversion factors
for the natural uranium series are given in Table 8.25, which illustrates
that approximately two-thirds of the electron dose is due to 23ltPa; this
nuclide reaches equilibrium in the 238U series in less than 1 yr. Thus the
electron dose (in contrast to gamma dose) due to a 1-yr application of 238U
in fertilizers, reaches two-thirds of its maximum value rather quickly and
remains relatively constant for several hundred-thousand years. The calcu-
lations given in Table 8.25 assume that all the activity is and remains on
the surface. Using these same assumptions, the surface activity of 23tfPa
from the average application rate (14.4 kg of uranium per 1000 ha) of
fertilizers would be about 4.7 x 10~8 yCi/cm2, and about 2.8 x 10~7 yCi/cm2
at the ma.ximum application rate (90 kg of uranium per 1000 ha). Using these
data and the dose conversion factors of Table 8.25 yields 1 and 6 mrem/yr
at average and maximum application rates, respectively, for a 1 yr applica-
tion.
The above dose estimates are very conservative and are given primarily
to emphasize that external electron exposures should be investigated more
closely. Since 23^Pa has a half-life of only 1.17 min, the annual dose
rate from long-term accumulation through fertilizer application would
approximately equal the dose rate from a 1-yr application. If 231+Pa
electrons could be measured directly, the isotope might serve as a tracer
for 238U turnover in the environment. Its parent, 231fTh, also has a
relatively short half-life of 24.1 days.
8.2.3 Internal dose estimates
The reduction in internal dose due to removal of uranium and its decay
products from WP phosphoric acid may be roughly estimated in a manner similar
to that given above for external exposures.
-------
127
Natural uranium. Uptake of uranium in the daily diet is about 1 yg/
14
day for 1.8-ppm soil concentration. The highest dose rate is to endosteal
cells and is estimated to be about 0.8 mrad/yr. Thus, the dose to endosteal
cells is about 0.5 mrad yr ppm"1 of uranium in the soil. As discussed
above, removal of uranium from the 30% of phosphate rock that is used to
produce WP phosphoric acid corresponds to an annual reduction in the rate
of uranium added to farm soils as fertilizer of 1.5 ppb and 9 ppb max; the
corresponding internal dose reductions are 0.75 yrad/yr and 4.5 yrad/yr
respectively. Accumulation of uranium in fertilized soils (no uranium
removal process) could eventually double doses from background levels in
1100 yr at the average and 190 yr at the maximum application rates, as was
the case for external exposures.
Radium. The average daily uptake of 226Ra in normal background areas
14
is stated to be 1 pCi per gram of calcium. The daily intake of calcium
27
by Reference Man is 1.1 g/day; thus the daily intake of radium is about
1.1 pCi/day or 1 pg/day. The highest dose rate from uptake of 225Ra is
again to endosteal cells and is given as about 1.6 mrad/yr or 6 mrem/yr.
Since 225Ra is expected to be in secular equilibrium in background areas,
the concentration of 226Ra corresponding to 1.8 ppm of uranium in soil
would be approximately 0.6 parts per trillion (ppt) (0.6 pg/g). Thus the
dose to endosteal cells per unit concentration is about 2.7 mrad yr"1
ppt"1 or 10 mrem yr"1 ppt"1. The annual amount of 226Ra which occurs to
only about 1% of its equilibrium value,13 that could be added to the plow
layer of soils from phosphoric acid, would be about 5.1 x 10 18 g of
226Ra/g of soil, and 30.6 x 10~18 g of 22SRa/g of soil at average and
maximum application rates respectively. These concentrations would
correspond to endosteal cell doses of 13.8 nrad/yr and 51 nrem/yr at average
application rates, and 84 nrad/yr and 306 nrem/yr at maximum application
rates.
If normal superphosphate (Table 8.23) were added to soils at the
average and maximum application rates, the endosteal cell doses due to
radium would be approximately 4.6 yrad/yr and 28 yrad/yr at average and
maximum application rates respectively. Since only 30% of phosphate rock
is used for production of WP phosphoric acid, the doses would be 1.4 yrad/yr
-------
128
(0.3 x 4.6) and 8.4 yrad/yr (0.3 x 28) if radium stayed in WP phosphoric
acid. Thus the precipitation of 225Ra during production of WP phosphoric
acid is nearly twice as effective in reducing internal radiation dose as
the subsequent removal of uranium (0.75 and 4.5 yrad/yr).
Thorium. For Models 1 and 2, thorium will remain in the WP phosphoric
acid, but for Model 3, it will appear in the product of the first cycle
extraction (Sect. 8.2.1). For Model 3, crude UF^, thorium will be released
as part of the source term (Table 4.6), but for Model 3, semirefined 11303,
it will be released in a waste stream to the phosphoric acid plant phospo-
gypsum pile. If thorium release is controlled strictly at the plant, that
Model 3 can result in a potentially larger positive impact than Models 1
and 2.
14
There appear to be no data on thorium ingestion in the literature.
However, according to the dose estimates for operation and postoperation of
uranium recovery plants, inhalation bone dose from 230^ ^s a limiting
consideration. Thorium may become airborne during fertilizer application
or may be resuspended after application. The maximally exposed individual
would be the person applying the fertilizer since the fraction of activity
suspended in this case could be much higher than for resuspension. About
30% of WP phosphoric acid is applied in liquid form and about 70% in pellet
or powder form. No information was found on the fraction of fertilizers
that would become airborne during application so that, again, lack of
information precludes quantitative dose estimates. It may be assumed that
exposures to personnel in the phosphate industry would exceed exposures to
28
persons applying fertilizers. The highest potential exposures of workers
in phosphate operations were observed in areas of high concentrations of
9 28
dose and in locations around phosphoric reactor vessels. ' Direct gamma
dose equivalents for workers ranged from 30 to 300 mrem/yr. The maximum
potential dose equivalent rate to lungs was about 5 rem/yr. According to
estimates given in Table 8.15, the bone dose to the maximally exposed
individual of the general population due to inhalation of 230Th is about
ten times higher than the lung dose; this same factor of 10 probably should
be used in estimating the bone dose of workers in the phosphate industry.
Lower occupancy could reduce these doses to industry workers by a factor
9
up to 10. For persons applying fertilizers, exposures would be for only
-------
129
a few days per year, and dust or solution droplet concentrations would be
expected to be much lower than for phosphate industry personnel. Thus,
doses during application (averaged over 1 yr) would probably be less than
those given for maximally exposed individuals (Tables 8.15 and 8.16) near
uranium recovery plants. A more detailed assessment than can be given
here is recommended.
If it is rather arbitrarily assumed that thorium intake is about the
same as uranium and radium (0.68 pCi of uranium/day and 1.1 pCi of radium/
day), then the bone dose due to 23^xh ingestion would be two to three times
greater than for uranium (see dose conversion factors, Table 8.14), or
about 2 and 13 yrad/year at average and maximum fertilizer application
rates respectively.
Resuspension of uranium, radium and thorium. Resuspension may be
treated in.a manner similar to the postoperational cases given in Sect. 8.1.4,
Under postoperational conditions, the highest organ dose, by an order of
magnitude, was bone dose due to inhalation of 230Th. The surface activity
yielding this bone dose (1.2 yrem/yr, see Table 8.21) was about 3.2 x 10"11
Ci/m2 (see Table 8.19). If 230Th in phosphoric acid (Table 8.23) were added
to the soil surface, the surface activity at average and maximum fertilizer
application rates would be about 1.5 x 10~9 Ci/m2 and 9 x 10~^ Ci/m2 respec-
tively. Thus the bone dose due to long-term inhalation from a 1-yr applica-
tion of 23°Th could amount to 57 and 330 yrem/yr at average and maximum
application rates respectively. The comparison for uranium is similar.
Thus, the positive impact of removal of uranium and thorium from phosphoric
acid would appear to offset the local postoperational impact of uranium
recovery many times over. Continuous accumulation of uranium and thorium
with no removal from near-surface soil, or with similar removal in the two
cases, would balance the comparison even further in favor of positive
impacts.
The highest postoperational population dose is about 2 person-rem/yr,
again due to inhalation of 230Th. Lacking details of population distribu-
tions surrounding fertilized soils, now and in the future, we assume that
the population will be at least as large as that which surrounds uranium
recovery facilities. Under this condition, the long-term population bone
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130
dose from a 1-yr application of phosphoric acid derived fertilizers, with
3^Th present, would amount to about 40 to 200 person-rem/yr. Removal of
uranium and thorium from WP phosphoric acid, and avoiding their distribution
in fertilizers for a few years would offset both the postoperational popula-
tion dose and the population dose during operation, regardless of the model
or radwaste treatment case (see Tables 8.17, 8.18, and 8.22). Again, a
more thorough assessment would appear to be justified.
The dose estimates given above are, as stated before, highly conserva-
tive and omit many details. The relative comparison of potential positive
and negative impacts is, to the degree possible within the scope of this
study, based on comparable estimates of dose; however, the absolute
magnitudes may change considerably as more quantitative studies are made.
Based on the qualitative assessment given herein, it is likely that the
positive impact of removal of uranium from phosphoric acid will more than
compensate for the local impact of the uranium recovery facilities them-
selves, especially if the best available radwaste treatment systems are
used at the plants.
8.3 References for Section 8
1. M. B. Sears, R. E. Blanco, R. C. Dahlman, G. S. Hill, A. D. Ryon, and
J. P. Witherspoon, Correlation of Radioactive Waste Treatment Costs
and the Environmental Impact of Waste Effluents in the Nuclear Fuel
Cycle for Use in Establishing "As-Low-as-Practicable" Guides —
Milling of Uranium Ores, ORNL/TM-4903, Vol. 1 (May 1975).
2. F. J. Hurst, W. D. Arnold, and A. D. Ryon, "Progress and Problems of
Recovering Uranium from Wet-Process Phosphoric Acid," presented at
the 26th Annual Meeting of the Fertilizer Industry Roundtable,
Atlanta, Ga., October 1976.
3. R. E. Moore, The AIRDOS-II Computer Code for Estimating Radiation
Dose to Man from Airborne Radionuclides in Areas Surrounding Nuclear
Facilities, ORNL-5245 (April 1977).
4. G. G. Killough and L. R. McKay, A Methodology for Calculating Radia-
tion Doses from Radioactivity Released to the Environment, ORNL-4992
(March 1976).
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131
5. Seasonal and Annual Wind Distribution by Pasquill Stability Classes
STAR Tabulation, National Oceanic and Atmospheric Administration,
U.S. Dept. of Commerce, Tampa, Fla.
6. Climatic Atlas of the United States, National Oceanic and Atmospheric
Administration, U.S. Dept. of Commerce (June 1974).
7. P. R. Coleman and A. A. Broosk, PANS: A Program to Tally Population
by Annuli and Sectors, ORNL/TM-3923 (October 1972).
8. R. J. Olsen et al., MULTIREGION: A Simulation-Forecast Model of BEA
Economic Area Population and Employment, ORNL/RUS-25 (October 1977).
9. R. J. Guimond, "Radiological Aspects of Fertilizer Utilization,"
presented at the Symposium on Public Health Aspects of Radioactivity
in Consumer Products, Atlanta, Ga., February 1977.
10. R. G. Menzel, "Soil-Plant Relationships of Radioactive Elements,"
Health Phys. 11, 1325 (1965).
11. R. G. Menzel, "Uranium, Radium, and Thorium Content in Phosphate
Rocks and their Possible Radiation Hazard," J. Agr. Food Chem. 16, 233
(1968).
12. R. K. Schulz, "Soil Chemistry of Radionuclides," Health Phys. 11,
1317 (1965).
13. H. Pfister, G. Phillipp, and H. Pauly, "Population Dose from Natural
Radionuclides in Phosphate Fertilizers," Radiat. Environ. Biophys. 13,
247 (1976).
14. Radiological Quality of the Environment, U.S. Environmental Protection
Agency, EPA-520/1-76-010 (May 1976).
15. D. T. Oakley, Natural Radiation Exposure in the United States, U.S.
Environmental Protection Agency, EPA-ORP/SID 72-1 (June 1972).
16. J. A. S. Adams and W. M. Lowder (eds.), The Natural Radiation Environ-
ment, University of Chicago Press, Chicago, 1964.
17. Fertilizer Recommendations for Tennessee, Agricultural Extension
Service, University of Tennessee, Publication 381 (rev. Jan. 1977).
18. Crop Production, 1976 Annual Summary, Acreage, Yield, Production,
U.S. Dept. of Agriculture, CrPr 2-1(77) (Jan. 17, 1977).
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132
19. Changes in Farm Production and Efficiency, A Special Issue Featuring
Historical Series, U.S. Dept. of Agriculture, Statistical Bulletin
No. 561 (September 1976).
20. E. A. Harre, M. N. Goodson, and J. D. Bridges, Fertilizer Trends 1976,
Tennessee Valley Authority, Bulletin Y-lll (March 1977).
21. Commercial Fertilizers, Final Consumption for Year Ended June 30, 1976,
U.S. Dept. of Agriculture, SpCr 7 (77) (April 1977).
22. Commercial Fertilizers, Consumption by Class for the Year Ended June 30,
1976, U.S. Dept. of Agriculture, SpCr 7 (77) (May 1977).
23. 1977 Fertilizer Situation, U.S. Dept. of Agriculture FS-7 (January 1977),
24. Agricultural Statistics 1975, U.S. Dept. of Agriculture (1975)
25. W. F. Stowasser, "Phosphate Rock," in Minerals Yearbook 1973, vol. 1.,
U.S. Bureau of Mines.
26. T. P. Hignet, "Characteristics of the World Fertilizer Industry—
Phosphate Fertilizers," Tennessee Valley Authority, Report No. S-442,
prepared for use at United Nations International Symposium on Indus-
trial Development, Athens, Greece, December 1967.
27. International Commission on Radiological Protection, Report of the
Task Group on Reference Man, ICRP Publication No. 23, Pergamon Press,
New York, 1975.
28. S. T. Windham, J. Partridge, and T. Horton, Radiation Dose Estimates
to Phosphate Industry Personnel, EPA 520/5-76-014 (December 1976).
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133
9. OVERVIEW AND RECOMMENDATIONS
9.1 Summary of New Impacts
The previous discussion (especially Sect. 8.2) indicates that the
overall radiological impact of recovery of uranium from WP phosphoric acid
is likely to be positive. That is, reduction in dose (positive impact)
due to removal of uranium (and thorium for Model 3, crude UF^ product) is
likely to exceed the increases in dose (negative impact) on the populations
surrounding the recovery facilities. Such a conclusion is based on
specific processes for uranium recovery from phosphoric acid. The influence
of uranium recovery on. the present impacts of the entire phosphate industry
or on impacts of nuclear fuel cycles that would utilize the recovered uranium
is not assessed. However, it seems reasonable to suppose that the overall
impact would tend to be positive rather than negative. For example, no
mining or milling is required for recovery of uranium from WP phosphoric
acid per se, and occupational exposures of personnel in the phosphate
industry could be reduced. Such possible associations should be studied fur-
ther as discussed in Sect. 9.2.
The largest potential positive and negative impacts are associated with
Model 3, Case 1, crude UF^ product (Sect. 4.8) where thorium isotopes
constitute a major component of the source term (Table 4.6). The maximum
individual dose is about 250 mrem/yr to bone for a 1000-MTU/yr product rate,
primarily due to 230Th inhalation (Table 8.16). The maximum population dose
for a 1000-MTU/yr product rate is about 426 person-rem/yr, which is also to
bone and also primarily due to 230Th (Table 8.9). However, the potential
reduction in dose due to removal of the uranium and thorium isotopes from
WP phosphoric acid, which prevents their application to farm lands and urban
gardens and grasses in fertilizers, can easily exceed the doses to the
populations surrounding the recovery facilities (Sect. 8.2). Furthermore,
addition of Case 2 effluent-control methods (HEPA filters) could reduce
maximum individual doses to about 0.1 mrem/yr and population dose to about
0.2 person-rem/yr. The total annual cost of Case 2 over Case 1 is about
$9000 for any product rate up to 1000 MTU/yr (Sect. 6). Thus a substantial
positive impact could be assured at a cost of about $36/mrem for maximum
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134
individual dose and about $21/person-rem for population dose; individual
doses would be well below established guidelines. Addition of HEPA filters
to plants with smaller product rates would result in higher costs per
person-rem reduction in dose (e.g., about $100/person-rem for a 200-MTU/yr
product rate).
Maximum annual doses to individuals (Table 8.16) and to the population
(Tables 8.17 and 8.18) are much lower for Models 1 and 2, Cases 1 and 2, and
for Model 3, semirefined u^Og product, than they are for Model 3 operating
with crude VF^ as the product. Thus, the benefits to be derived from
improving.Model 3, Case 1, with UF^ as product, greatly exceed those from
improving all of the other model cases. For example, the costs of Case 2
over Case 1 for Model 3, semirefined u^Og, would be about $1250/mrem to
reduce bone dose by 7.2 mrem/yr (Table 8.7), and about $960/person-rem to
reduce population lung dose by 9.4 person-rem/yr (Table 8.18). These cost-
benefit values are much higher than the $36/mrem and $21/person-rem given
above for Model 3 in which the product is crude UF^. All of these costs are
for a 1000-MTU/yr product rate. It should be noted that Models 1 and 2 and
Model 3, semirefined 11303, lead to doses to the surrounding population that
are within established guidelines using Case 1 radwaste treatment methods.
9.2 Information Gaps and Research Needs
Assessment of the radiological impacts of uranium recovery from WP
phosphoric acid involves many uncertainties and assumptions, especially
regarding the long-term environmental fate of the natural radionuclides of
concern. Conservative assumptions are generally made such that estimated
impacts are greater than could realistically be expected in any actual
situation. However, most such uncertainties are common to assessment
science in general and are the subject of ongoing research. Hence, this
discussion is restricted to the particular application of uranium recovery
from phosphoric acid. The primary current need is to expand the site-
specific assessment of the reported uranium recovery facilities to an
integrated assessment of the phosphate industry. Such an integrated assess-
ment should include resource depletion, transportation, siting, and economic,
social, environmental, and health impacts. In addition, recovery of uranium
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135
from phosphoric acid as it relates to the phosphate industry in general
can be compared to conventional mining and milling. When the phosphate
industry as a whole is considered, there are major similarities with
commercial mining and milling that will help to facilitate some comparisons.
The impact of uranium-mill tailings piles may be compared to those of waste
phosphogypsum piles. Homes built on reclaimed phosphate mining lands have
their counterpart in homes where uranium mill tailings were used as fill
material. Epidemiological studies on uranium miners provide basic health-
effect information on some of the radioactive species common to both
processes (primarily radium and decay products). However, making comparisons
between processes and extrapolating information from occupational groups to
the general population is not straightforward, and in some cases, it is
controversial.
Some specific factors related to radiological impacts that will have
to be addressed in conducting an integrated assessment of uranium recovery
from phosphoric acid as contrasted to conventional mining and milling are:
1. dose conversion factors for radium decay products, particularly
radon and decay products;
2. extrapolation of epidemiological information on occupational
groups to the general population;
3. local impacts vs regional, national, and global impacts;
4. positive impacts as well as negative impacts; and
5. differences in exposure pathways, especially for long-lived
isotopes such as 210Pb and 210Po.
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137
10. APPENDIXES
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139
Appendix A. Description of Sampling Train
Since the primary concern was the concentration of
particulate
matter being released into the atmosphere, a sampling train was
developed similar to the standard EPA train. The train used to obtain
isokinetic sampling (shown in Fig. A.I) consists of:
ORNL-DWG 77-18285 R
4 4
fr&T
11
o
p» 1
1
1
i
1 ^ —
1
1 . PROBE
2. PARTICULATE FILTER
3. DRY GAS METER
4. THERMOMETER
5. VACUUM GAGE
6. VACUUM PUMP
7. FLOW-AD JUST VALVE
8. PITOT
9. INCLINE MANOMETER
Fig. A.I. Diagram of particulate sampling train.
1. Two stainless steel probes. (a) For the 6.25-in.-diam duct, the
probe was a l/2-in.-diam nozzle, 28.26 cm long from tip of nozzle
to filter head; and (b) for the 3-in.-diam duct, the probe
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140
consisted of a 0.25-in.-diam nozzle, 22.54-cm long from tip of
nozzle to filter head.
2. Filter and filter-head assembly, which consisted of commercially
*
available 37-mm-diam, 0.8-y membrane filter papers with
A*
commercially available filter head assemblies designed
specifically for this size paper. A mesh-type backing screen
was used to minimize the pressure drop across the filter face.
3. A calibrated dry-gas meter (+2% at the flow rates encountered)
with temperature gages at both the inlet and outlet ports of
the meter.
4. Positive displacement vacuum pump with vacuum gage capable of
flow rates greater than 3 cfm.
The probes were specially designed such that the tip of each probe
fitted snugly into the filter head assembly. A standard Pitot tube
with inclined manometer was connected to the probe in order to ensure
isokinetic sampling during a run. Clogging of the Pitot tube was not
evident since the velocities remained constant during sample collection
at each sampling point in the run. Impingers were not included in
the train because ambient air with noncorrosive gases was sampled
through the train. Moisture content was determined using wet bulb-dry
bulb techniques. Surgical tubing was used at all connections in the
train.
Once a run was completed, the probe was flushed thoroughly with
a 4 M HN03 solution, and the elutriant was combined with the appropriate
filter sample prior to assaying. The samples were then assayed for
abundance of uranium isotopes by mass spectrometric techniques. The
sensitivity of this method for uranium is in the low parts-per-billion
range.
*
Millipore Corporation.
**
Mine Safety Appliances Company.
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141
Appendix B. Pertinent Operating Procedures at URC for
Changing Product Drums (with Respect to Source Sampling)
1. Place a properly marked empty drum on the rollers to the right
of the drum loader.
2. Be certain that the powder valve is turned off.
3. Start the dust collecter blower.
4. With sampling drum cover at hand (see Fig. B.I), raise the drum
loader ("minispout" system) and immediately position the sampling
drum cover on the drum.
ORNL-DWG 77-18286R
DIAM 1.0625"
Fig. B.I. Diagram of sampling cover with sampling ports.
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142
5. Roll the full drum to the left until it clears the drum loader.
Immediately roll the empty drum under the loader and lower the
drum loader into position on the empty drum.
6. Open the powder valve.
7. Stop the dust collector blower. The bag shaker should start
automatically; if it does not, start it manually.
8. Go to the bag filter located on the drumming building second
floor and open the 6-in. powder (butterfly) valve on the bottom
of the bag filter housing. This allows the collected dust to
drop into the drum now being loaded.
9. When the bag shaker stops, close the 6-in. butterfly valve. If
the bag shaker was started manually, close the butterfly valve
before returning downstairs to turn off the shaker.
Replacing a product drum, Steps 3-7, requires approximately 25 sec.
These steps designate that period of time during which releases from
the duct are occurring. Observing these procedures actually being
performed, the following time sequence was developed:
(1) It required 5 sec from starting the dust collector blower
to placing the sampling drum cover on the product drum;
(2) It required 20 sec from the previous step to stopping
the dust collector blower.
Since 25 sec is too short a time to establish steady state in
isokinetic sampling, the following time scheme was allotted (using
the above criteria) for source sampling so that representative
samples of effluent releases could be obtained.
(1) During l/'5 of the total time that a sample was collected
(per sampling point in a run) the product drum remained
uncovered.
(2) During the remaining 4/5 of the time the sampling drum cover
was positioned on the drum.
In addition to these criteria, no sample was taken unless the product
drum was at least half full. A product drum is considered full at the
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143
three-quarter level. Also, samples were taken as nearly as possible
when freshly calcined material was dropped into the drum. This would
help minimize the settling of particulate matter in the product drum.
Appendix C. Description of Ge(Li) Detector System
A holder for twelve 30-cc polyethylene bottles (standard containers
for liquid scintillation samples) and a Ge(Li) detector system in
laboratory counting of radioactivity in environmental samples. During
counting of the samples, the holder is used to position ten of the sample
bottles around the cylindrical surface of the detector, parallel to and
symmetric about its axis, and two additional bottles across the end surface
of the detector, perpendicular to and symmetric with its axis. With a
300-cc sample and a graded shield developed for use with the system, it is
possible to measure 1 pCi/g of 232Th or 226Ra with an error of ±10% or
less.
Pulses are sorted by a 4096-channel analyzer, stored on magnetic
tape, and subsequently entered into a computer program that uses an
iterative least squares method to identify radionuclides corresponding
to those gamma-ray lines found in the sample. The program relies on a
library that contains data on energies of approximately 2500 photons
from 700 nuclides. In identifying and quantifying 226Ra, six principal
gamma-ray lines are analyzed. Most of these are from 214Bi and correspond
to 295, 352, 609, 1120, 1765, and 2204 keV. An estimate of the concen-
tration of 238U is obtained from an analysis of the 93 and 1001 keV
line from its daughter 231*Th.
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ORNL/EPA-2
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187. R. S. Goor, National Science Foundation, 2101 Constitution Ave.,
Washington, DC 20418.
188. R. L. Gotchy, Division of Site Safety and Environmental Analysis,
U.S. Nuclear Regulatory Commission, Washington, DC 20555.
189. C. J. Hardy, Australian Atomic Energy Commission, Research
Establishments, Private Mail Bag, Sutherland, N.S.W. 2232
Australia.
190. W. R. Hardwick, U.S. Bureau of Mines, Eight West Paseo Redondo,
Tucson, AZ 85705.
191. J. H. Harley, Environmental Measurements Laboratory, U.S. Depart-
ment of Energy, 376 Hudson St., New York, NY 10014.
192. E. A. Harre, Tennessee Valley Authority, Muscle Shoals, AL 35660.
193. D. N. Harrington, U.S. Department of Agriculture, GHI Building,
500 12th St., SW, Washington, DC 20250.
194. J. W. Healy, Health Physics Division, Los Alamos Scientific
Laboratory, Box 1663, Los Alamos, NM 87544.
195. F. Heinzman, Department of Environmental Control, 1301 Cattleman
Rd., Sarasota, FL 33577.
196. D. Hendricks, ORP Las Vegas Facility, U.S. Environmental
Protection Agency, P.O. Box 15027, Las Vegas, NV 89114.
197. T. P. Hignett, Tennessee Valley Authority, Muscle Shoals, AL
35660.
198. H. Hooks, Florida Phosphate Council, Suite 24, Executive Plaza,
4406 S. Florida Ave., P.O. Box 5530, Lakeland, FL 33803.
199. R.J.M. Horton, Health Effects Research Laboratory, U.S. Environ-
mental Protection Agency, Research Triangle Park, NC 27709.
200. V. N. Houk, Environmental Health Services Division, Center for
Disease Control, Atlanta, GA 30333.
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201. J. Jared, Institute of Agriculture, University of Tennessee,
Knoxville, TN 37916.
202. K.D.B. Johnson, A.E.R.E., Harwell, Didcot, Oxon., England.
203. W. B. Johnson, Jr., State of Florida, Radiation Health Program,
P.O. Box 6635, Orlando, FL 32803.
204. B. Kahn, Georgia Institute of Technology, Environmental Resources
Center, Atlanta, GA 30332.
205. J. Kastner, Division of Siting, Health, and Safeguards Standards,
U.S. Nuclear Regulatory Commission, Washington, DC 20555.
206. R. H. Kennedy, Environmental Control Technology Division, U.S.
Department of Energy, Washington, DC 20545.
207. R. E. Kidwell, U.S. Department of Transportation, Federal High-
way Administration, Washington, DC 20590.
208. R. L. Kilmer, Department of Food and Resource Economics,
University of Florida, Gainesville, FL 32611.
209. C. L. Klinstiver, U.S. Department of Transportation, Federal
Highway Administration, Washington, DC 20590.
210. P. Kotin, Health, Safety and Environment, Johns-Manville Corp.,
Greenwood Plaza, Denver, CO 80217.
211. A. P. Kouloheris, Gardinier, Inc., P.O. Box 3269, Tampa, FL
33601.
212. T. R. Lash, National Resources Defense Council, Inc., 664
Hamilton Ave., Palo Alto, CA 94301.
213. H. Lawroski, Nuclear Services Corp., 1700 Dell Ave., Campbell,
CA 95008.
214. E. C. Lazar, Standards and Regulation Evaluation Division, U.S.
Environmental Protection Agency, Waterside Mall, 401 M St., SW,
Washington, DC 20460.
215. W. M. Leaders, Uranium Recovery Corporation, P.O. Box 765,
Mulberry, FL 33860.
216. D. A. Lipps, Freeport Chemical Company, Uncle Sam, LA 70792.
217. J. L. Liverman, U.S. Department of Energy, Washington, DC 20545.
218. R. L. Loftness, Electric Power Research Institute, 1750 New York
Ave., NW, Washington, DC 20006.
219. M. F. Lucid, Kerr-McGee Corporation, P.O. Box 25861, Oklahoma
City, OK 73125.
220. P. J. Magno, Fuel Reprocessing and Recycle Branch, U.S. Nuclear
Regulatory Commission, M.S. 29655, 7915 Eastern Ave., Silver
Spring, MD 20910.
221. B. J. Mann, U.S. Environmental Protection Agency, P.O. Box
15027, Las Vegas, NV 89114.
222. M. J. Martinasek, W. R. Grace and Company, Barton, FL 33830.
223. R. 0. McClellan, Lovelace Biomedical and Environmental Research
Institute, 5200 Gibson Blvd., SE, Albuquerque, NM 87108.
224. H.A,C. McKay, A.E.R.E., Harwell, Didcot, Oxon., England.
225. G. McNeil, Head, Environmental Radiation Section, U.S. Environ-
mental Protection Agency, 345 Curtland, Atlanta, GA 30308.
226. G. Meier, Farmland Industries, Pierce, FL 33867.
227. L. A. Meierkord, Wyoming Mineral Corporation, 4406 S. Florida
Ave., Lakeland, FL 33803.
228. D. R. Miller, Physical Research Division, U.S. Department of
Energy, Washington, DC 20545.
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150
229. M. T. Mills, GCA Technical Division, Burlington Rd., Bedford,
MA 01730.
230. W. A. Mills, Criteria and Standards Division (AW-460), Office
of Radiation Programs, U.S. Environmental Protection Agency,
Waterside Mall, 401 M St., SW, Washington, DC 20460.
231. R. B. Minogue, Office of Standards Davelopment, U.S. Nuclear
Regulatory Commission, Washington, DC 20555.
232. K. Z. Morgan, School of Nuclear Engineering, Georgia Institute
of Technology, Atlanta, GA 20332.
233. W. E. Mott, Environmental Control Technology Division, U.S.
Department of Energy, Washington, DC 20545.
234. D. R. Nelson, Office of Radiation Programs, U.S. Environmental
Protection Agency, Waterside Mall, 401 M St., SW, Washington,
DC 20460.
235. R. R. Newton, Biomedical and Environmental Research Division,
U.S. Department of Energy, Washington, DC 20545.
236. W. R. Ney, National Council on Radiation Protection and Measure-
ments, 7910 Woodmont Ave., Bethesda, MD 20014.
237. Y. C. Ng, Biomedical and Environmental Research Division,
Lawrence Livermore Laboratory, P.O. Box 808, Livermore, CA
94550.
238. H. T. Odum, Department of Environmental Engineering, University
of Florida, Gainesville, FL 32601.
239. M. J. Ohanian, Department of Nuclear Engineering Sciences,
University of Florida, Gainesville, FL 32601.
240. G. Palm, Florida Phosphate Council, Suite 24, Executive Plaza,
4406 S. Florida Ave., P.O. Box 5530, Lakeland, FL 33803.
241. D. A. Paul, U.S. Department of Agriculture, GHI Building,
Room 120, 500 12th St., SW, Washington, DC 20250.
242. H. R. Payne, U.S. Environmental Protection Agency, Region IV,
1421 Peachtree St., Atlanta, GA 30309.
243. G. I. Pearman, CSIRO, Division of Atmospheric Physics,
Aspendale, Victoria, Australia.
244. A. J. Pennak, National Lead Company of Ohio, P.O. Box 39158,
Cincinnati, OH 45239.
245. W. H. Pennington, Biomedical and Environmental Research Division,
U.S. Department of Energy, Washington, DC 20545.
246. R. 0. Pohl, Laboratory of Atomic and Solid State Physics, Clark
Hall, Ithaca, NY 14853.
247. B. J. Porter, Louisiana Division of Radiation Control, P.O. Box
14690, Baton Rouge, LA 70808.
248. C. Porter, Eastern Environmental Research Facility, U.S. Environ-
mental Protection Agency, P.O. Box 3009, Montgomery, AL 36109.
249. K. Purushothamon, University of Missouri, Rolla, MO 64501.
250. 0. G. Raabe, Radiobiology Laboratory, University of California,
Davis, CA 95616.
251. D. P. Rail, National Institute of Environmental Health Sciences,
P.O. Box 12233, Research Traingle Park, NC 27709.
252. G. Rausa, Office of Research, U.S. Environmental Protection Agency,
Waterside Mall, 401 M St., SW, Washington, DC 20460.
253. S. L. Reese, Uranium Recovery Corporation, P.O.. Box 765, Mulberry,
FL 33860.
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254. Office of Assistant Manager, Energy Research and Development,
DOE-ORO, Oak Ridge, TN 37830.
255. R. W. Riddle, C F Industries, Plant City, FL 33566.
256. R. J. Ring, Australian Atomic Energy Commission, Research
Establishment, Lucas Heights, Australia.
257. W. J. Robertson, Kerr McGee Corporation, P.O. Box 25861,
Oklahoma City, OK 73125.
258. J. V. Rodricks, Division of Chemistry and Physcis, Food and Drug
Administration, 200 C St., SW (H-FF-151), Washington, DC 20204.
259. C. E. Roessler, 525 NE 4th St., Gainesville, FL 32601.
260. R. C. Ross, United Nuclear Corporation, 101 Executive Blvd.,
Elmsford, NY 10523.
261. L. C. Rouse, Division of Fuel Cycle and Material Safety, U.S.
Nuclear Regulatory Commission, Washington, DC 20555.
262. W. D. Rowe, Office of Radiation Programs, U.S. Environmental
Protection Agency, Waterside Mall, 401 M St., SW, Washington,
DC 20460.
263. G. J. Rubin, IMC Chemicals Corporation, P.O. Box 1035, Mulberry,
FL 33860.
264. E. J. Salmon, National Academy of Sciences, 2101 Constitution
Ave., Washington, DC 20418.
265. S. Samuels, Health, Safety and Environmental Affairs, Industrial
Union Department, AFL-CIO, 815 16th St., NW, Washington, DC
20006.
266. J. Schacter, General Offices, UCND, Oak Ridge, TN 37830.
267. K. J. Schiager, University of Pittsburgh, School of Public
Health, Pittsburgh, PA 15261.
268. A. Schneider, Georgia Institute of Technology, School of Nuclear
Engineering, Atlanta, GA 30332.
269. Th. Friis Soerensen, Research Establishment Risoe, 4000 Roskilde,
Denmark.
270. J. K. Soldat, Pacific Northwest Laboratories, P.O. Box 999,
Richland, WA 99352.
271. H. J. Sonnenberg, U.S. Department of Transportation, Federal
Highway Administration, Washington, DC 20590.
272. B. Spinrad, Oregon State University, Radiation Center, Corvallis,
OR 97331.
273. A. C. Stern, Department of Environmental Sciences and Engineering,
University of North Carolina, 602 Groom Court, Chapel Hill, NC
27514.
274. J. Swinebroad, Biomedical and Environmental Research Division,
U.S. Department of Energy, Washington, DC 20545.
275. L. S. Taylor, National Council on Radiation Protection and
Measurements, 7910 Woodmont Ave., Bethesda, MD 20014.
276. W. S. Twenhofel, U.S. Geological Survey, Federal Center, Denver,
CO 80225.
277. B. F. Warner, U.K.A.E.A., Windscale and Calder Works, Sellafield,
Seascale, Cumberland, England.
278. R. L. Watters, Biomedical and Environmental Research Division,
U.S. Department of Energy, Washington, DC 20545.
279. A. M. Weinberg, Oak Ridge Associated Universities, Oak Ridge, TN
37830.
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280. W. Weinlander, GFK - Institute f. Heise Chemie, Postbox 3640,
D7500, Karlsruhe, W. Germany.
281. F. W. Whicker, Department of Radiology and Radiation Biology,
Colorado State University, Fort Collins, CO 80521.
282. G. H. Whipple, University of Michigan, School of Public Health,
Ann Arbor, MI 48104.
283. G. E. Wilkinson, Gardinier Inc., P.O. Box 3269, Tampa, FL 33601.
284. S. T. Windham, Eastern Environmental Research Facility, U.S.
Environmental Protection Agency, P.O. Box 3009, Montgomery, AL
36109.
285. R. W. Wood, Biomedical and Environmental Research Division, U.S.
Department of Energy, Washington, DC 20545.
286-510. Given distribution as shown in TID-4500 under Environmental
Control Technology and Earth Sciences category (25 copies - NTIS)
6 U.S.GOVERNMENT PRINTING OFFICE: 1979-748-189/359
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