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 ------- Printed in the United States of America. Available from National Technical Information Service U.S. Department of Commerce 5285 Port Royal Road, Springfield, Virginia 22161 Price: Printed Copy $8.00; Microfiche $3.00 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, contractors, subcontractors, or their employees, makes any warranty, express or implied, nor assumes any legal liability or responsibility for any third party's use or the results of such use of any information, apparatus, product or process disclosed in this report, nor represents that its use by such third party would not infringe privately owned rights. ------- 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 ------- 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 ------- 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 ------- 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 ------- 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 ------- 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). ------- 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; ------- 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 ------- 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). ------- 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 ------- 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. ------- 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 ------- 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 ------- 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 ------- 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). ------- 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. ------- 12 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. ------- 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. ------- 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). ------- 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). ------- 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). ------- 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). ------- 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. ------- 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. ------- 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. ------- 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 ------- 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). ------- 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. ------- 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. ------- 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; ------- 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. ------- 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.) ------- 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. ------- 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.) ------- 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. ------- 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.) ------- 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. ------- 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. ------- 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. ------- 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 ------- 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. ------- 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. ------- 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. ------- 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. ------- 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. ------- 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 ------- 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. ------- 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. ------- 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. ------- 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. ------- 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. ------- 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 ------- 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 ------- 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). ------- 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). ------- 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). ------- 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 ------- 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 ------- 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. ------- 137 10. APPENDIXES ------- 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 ------- 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. ------- 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. ------- 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 ------- 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. ------- 145 ORNL/EPA-2 Dist. Category UC-11 1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12-21. 22. 23. 24. 25. 26. 27-36. 37. 38. 39. 40. 41. 42. 43. 44. 45. 46. 47. 48. 49. 50. 95 96 97 98 99 W. S. J. R. K. W. C. G. L. K. D. W. D. E. W. E. R. F. R. J. S. G. F. F. S. 0. G. R. A. R. E. F. • • • INTERNAL DISTRIBUTION D. Arnold, Jr. 51. I. Auerbach 52. A. Auxier 53. E. Brooksbank 54. B. Brown 55. D. Burch 56. A. Burchsted 57. R. Choppin (consultant) 58. J. Colby, Jr. (consultant) 59. E. Cowser 60. J. Grouse 61. Davis, Jr. 62. E. Ferguson 63. J. Frederick 64. Fulkerson 65. L. Gaden, Jr. (consultant) 66. W. Glass 67. F. Haywood 68. F. Hibbs 69. R. Hightower, Jr. 70. G. Hildebrand 71-80. S. Hill 81. 0. Hoffman 82. J. Hurst 83-84. V. Kaye 85-86. L. Keller 87-88. G. Killough 89-90. E. Leuze P. Malinauskas 91-92. E. Moore 93. Newman 94. R. O'Donnell J. S. Olson D. C. Parzyck H. A. Pfuderer H. Postma D. E. Reichle C. R. Richmond J. W. Roddy P. S. Rohwer M. W. Rosenthai T. H. Row E. M. Rupp A. D. Ryon C. D. Scott M. B. Sears E. G. Struxness L. E. Swabb (consultant) K. D. Timmerhaus (consultant) D. G. Trauger P. R. Vanstrum E. B. Wagner P. J. Walsh J. P. Witherspoon R. G. Wymer Central Research Library Document Reference Section RSIC Library ORNL - Y-12 Technical Library Laboratory Records Laboratory Records — RC ORNL Patent Office EXTERNAL DISTRIBUTION A. Agnew, USGS, Conservation Division, 12201 Sunrise Valley Dr., Reston, VA 22092. M. Altobello, Department of Agricultural Economics, University of Arizona, Tuscon, AZ 85721. J. A. Ambler, U.S. Department of Commerce, Bureau of the Census, Industry Division, Room 2208, FOB 4, Suitland, MD 20233. P. A. Andrilenas, U.S. Department of Agriculture, GHI Building, 500 12th St., SW, Washington, DC 20250. A. L. Ayers, Allied General Nuclear Services, P.O. Box 847, Barnwell, SC 29812. ------- 146 100. D. A. Baker, Radiological Health Research, Pacific Northwest Laboratories, P.O. Box 999, Richland, WA 99352. 101. N. F. Barr, Technology Overview Division, U.S. Department of Energy, Washington, DC 20545. 102. D. S. Earth, National Environmental Research Center, U.S. Environmental Protection Agency, P.O. Box 15027, Las Vegas, NV 89114. 103. C. B. Bartlett, Division of Safeguards, Fuel Cycle, and Environ- mental Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 104. M. R. Bateman, Union Carbide Corporation, Corporate Research Laboratory Library, P.O. Box 324, Tuxedo, NY 10987. 105. S. Beard, Exxon Nuclear Co., Field Box 3965, San Francisco, CA 94119. 106. E. Beckjord, Nuclear Power Development Division, U.S. Department of Energy, Washington, DC 20545. 107. R. M. Bernero, Division of Engineering Standards, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 108. W. Berry, IMC Chemicals Corp., P.O. Box 1035, Mulberry, FL 33860. 109. L. H. Bohlinger, Louisiana Division of Radiation Control, P.O. Box 14690, Baton Rouge, LA 70808. 110. D. Bordelon, Farmland Industries, Pierce, FL 33867. 111. A. J. Breslin, Environmental Measurements Laboratory, U.S. Department of Energy, 376 Hudson St., New York, NY 10014. 112. J. A. Broadway, Eastern Environmental Research Facility, P.O. Box 61, Montgomery, AL 36101. 113. L. H. Brooks, General Atomic Co., P.O. Box 81608, San Diego, CA 92138. 114. C. D. Broyles, Test Sciences (7110), Sandia Corporation, P.O. Box 5800, Albuquerque, NM 87115. 115. J. Burchard, Control Technology Branch, U.S. Environmental Protection Agency, Research Triangle Park, NC 27709. 116. W. W. Burr, Jr., Biomedical and Environmental Research Division, U.S. Department of Energy, Washington, DC 20545. 117. J. Butler, Office of Extramural Coordination and Special Projects, National Institute of Occupational Safety and Health, Rockville, MD 20850. 118. M. Calkins, National Environmental Research Center, U.S. Environ- mental Protection Agency, Cincinnati, OH 45268. 119. N. L. Carr, Gulf Science and Technology Company, Pittsburgh, PA 15230. 120. A. B. Carson, General Electric Co., 175 Curtner Ave., San Jose, CA 95100. 121. C. E. Carter, Biomedical and Environmental Research Division, U.S. Department of Energy, Washington, DC 20545. 122. B. V. Chatfield, U.S. Department of Transportation, Federal Highway Administration, Washington, DC 20590. 123. U. Clark, State of Florida, Radiation Health Program, P.O. Box 6635, Orlando, FL 32803. 124. R. J. Cloutier, Special Training Division, Oak Ridge Associated Universities, Oak Ridge, TN 37830. ------- 147 125. R. G. Cochrell, Licensing Division, Westinghouse Electric Corp., P.O. Box 158, Madison, PA 15663. 126. B. L. Cohen, University of Pittsburgh, Department of Physics, Pittsburgh, PA 15261. 127. R. D. Cooper, Office of the Assistant Secretary, Integrated Assessments, U.S. Department of Energy, Washington, DC 20545. 128. M. Corn, Assistant Secretary of Labor for Occupational Safety and Health Administration, Department of Labor, 200 Constitution Ave., NW, Washington, DC 20310. 129. G. Cowper, Health Physics Branch, Chalk River Nuclear Labora- tories, Chalk River, Ontario, Canada. 130. J. Crawford, Reactor Research and Technology Division, U.S. Department of Energy, Washington, DC 20545. 131. B. Crow, W. R. Grace and Company, Bartow, FL 33830. 132. F..L. Culler, Jr., Electric Power Research Institute, 3412 Hillview Ave., Palo Alto, CA 94304. 133. R. E. Cunningham, Division of Fuel Cycle and Material Safety, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 134. J. H. Davis, Tennessee Valley Authority, River Oaks Bldg., Muscle Shoals, AL 35660. 135. J. J. Davis, Division of Safeguards, Fuel Cycle and Environ- mental Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 136. W. Davis, III, Yankee Atomic Electric Company, 20 Turnpike Rd., Westboro, MA 01581. 137. N. J. Diaz, Department of Nuclear Engineering, University of Florida, Gainesville, FL 32601. 138. T. Dillon, Advanced Systems and Material Production Division, U.S. Department of Energy, Washington, DC 20545. 139. Directorate of Health Protection, Commission of tht European Communities, Luxembourg. 140. R. Duffey, Nuclear Engineering Dept., University of Maryland, 5614 Alta Vista Rd., Bethesda, MD 20034. 141. P. B. Dunaway, Office of Effects Evaluation, Nevada Operations Office, U.S. Department of Energy, Washington, DC 20545. 142. D. Durost, U.S. Department of Agriculture, GHI Building, Room 120, 500 12th St., SW, Washington, DC 20250. 143. G. G. Eichholz, Georgia Institute of Technology, School of Nuclear Engineering, Atlanta, GA 30332. 144. M. Eisenbud, New York University Medical Center, 501 First Ave., New York, NY 10016. 145. L. Elikan, Wyoming Mineral Corporation, 4406 S. Florida Ave., Lakeland, FL 33803. 146. H. J. Ettinger, Los Alamos Scientific Laboratory, P.O. Box 1663, Los Alamos, NM 87544. 147. R. D. Evans, 4621 East Crystal Lane, Scottsdale, AZ 85353. 148. H. L. Falk, National Institute of Environmental Health Sciences, P.O. Box 12233, Research Triangle Park, NC 27709. 149. P. Ferrand, Freeport Chemical Company, Uncle Sam, LA 70792. 150. D. Fisher, 213 NW 20th Terrace, Gainesville, FL 32603. ------- 148 151-175. J. E. Fitzgerald, Criteria and Standards Division (AW-460), Office of Radiation Programs, U.S. Environmental Protection Agency, Waterside Mall, 401 M St., SW, Washington, DC 20560. 176. E. H. Fleming, University of California, Lawrence Livermore Laboratory, Mail Stop L-l, Box 808, Livermore, CA 94550. 177. R. H. Flowers, A.E.R.E., Harwell, Didcot, Oxon., England. 178. D. Friedland, Allied Chemical Co., 20 Peabody St., Buffalo, NY 14240. 179. M. Gates, Nevada Operations Office, U.S. Department of Energy, P.O. Box 1676, Las Vegas, NV 99101. 180. J. A. Geer, Dow Chemical, U.S.A., Rocky Flats Division, Golden, CO 80401. 181. F. Gera, C.N.E.N., Viale Regina, Margherita 125, 00198, Rome, Italy. 182. F. A. Gifford, Atmospheric Turbulence and Diffusion Laboratory, P.O. Box E, Oak Ridge, TN 37830. 183. T. P. Gillett, U.S. Department of Commerce, Bureau of Domestic Commerce, Washington, DC 20233. 184. H. Gitterman, Burns and Roe, Inc., Industrial Division, P.O. Box 663, Parmus, NJ 07652. 185. A. S. Goldin, U.S. Environmental Protection Agency, Waterside Mall, 401 M St., SW, Washington, DC 20460. 186. M. Goldman, Radiobiology Laboratory, University of California, Davis, CA 95616. 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. ------- 149 201. 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. ------- 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. ------- 151 254. 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. ------- 152 280. 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 ------- |