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Inhalation Dose Assessment: Risk Assessment of Airborne Particulates to Workers in the Florida Phosphate Industry

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Title:
Inhalation Dose Assessment: Risk Assessment of Airborne Particulates to Workers in the Florida Phosphate Industry
Creator:
KIM, KWANG PYO ( Author, Primary )
Copyright Date:
2008

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Subjects / Keywords:
Aerosols ( jstor )
Dosage ( jstor )
Inhalation ( jstor )
Particle density ( jstor )
Particle size classes ( jstor )
Particle size distribution ( jstor )
Phosphates ( jstor )
Radioactive decay ( jstor )
Radionuclides ( jstor )
Solubility ( jstor )

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University of Florida
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University of Florida
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Copyright Kwang Pyo Kim. Permission granted to the University of Florida to digitize, archive and distribute this item for non-profit research and educational purposes. Any reuse of this item in excess of fair use or other copyright exemptions requires permission of the copyright holder.
Embargo Date:
12/31/2006
Resource Identifier:
793286987 ( OCLC )

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INHALATION DOSE ASSESSMENT: RISK ASSESSMENT OF AIRBORNE PARTICULATES TO WORKERS IN TH E FLORIDA PHOSPH ATE INDUSTRY By KWANG PYO KIM A DISSERTATION PRESENTED TO THE GRADUATE SCHOOL OF THE UNIVERSITY OF FLOR IDA IN PARTIAL FULFILLMENT OF THE REQUIREMENTS FOR THE DEGREE OF DOCTOR OF PHILOSOPHY UNIVERSITY OF FLORIDA 2005

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Copyright 2005 by Kwang Pyo Kim

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This document is dedicated to Dr. Willia m Emmett Bolch who passed away on December 26, 2003.

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iv ACKNOWLEDGMENTS The study could not have been accomplished without the valuable assistance of many individuals. Here, I want to acknowledge those people once more. I would like to express my deep appreciati on to my advisor, Dr. Wesley E. Bolch, who not only guided and supported me as my ad visor but also encouraged and challenged me throughout my academic program. I appreci ate the opportunity that he has provided me to study at University of Florida and pursue my interest in health physics. I wish to express my gratitude to Dr. Chang-Yu Wu. He provided me expertise and direction to complete this study. Great appreciation should go to Dr. William Emmett Bolch, who passed away on December 26, 2003. His long experience and wide and deep knowledge in the field of environmental physics led me to successful completion of this study. My thanks also go to my committee members, Dr. David E. Hintenlang and Dr. Guenther Hochhaus, for reading this dissertation and providing many valuable comments that improved this study. This work was supported by Grant #0005-062R and #03-05-064 from The Florida Institute of Phosphate Research (FIPR) to the University of Florida. Dr. Brian K. Birky, Research Director for Public Health, was very supportive of this research, and his guidance and advice were important for its successful completion. Many samplings were made in different plan ts and areas to obtain site-specific air sampling data. This activity was greatly f acilitated by many indivi duals: Allan Pratt, Michael Messina, and Robert Se llers in CF industries, Debra Waters, Ryan Richards,

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v Melody R. Foley, Tara Crews, Taylor D. Abel, Nelson Singletary, Santino Provenzano, and Flint Barnes in Mosaic Company, Mart in St. John and Dennis Killebrew in PCS Phosphate, and Ron Brunk and Foster Thorpe in US Agri-Chemicals. They provided me safety training, escort, and valuable info rmation. Wesley Nall, Tom McNally, and Robert Ammons in Health Unit in Winter Ha ven, Florida, conducted pre-sampling. Their efforts are greatly appreciated. Many thanks go out to all of my collea gues at the Department of Nuclear and Radiological Engineering, especially Dr . Chulhaeng Huh, Dr. Choonsik Lee, Dr. Eunyoung Han, Choonik Lee, Hosang Jin, and Heeteak Chung for their help and friendship. Yu-Mei Hsu at Departments of Environmental Engineering Sciences instructed me in the use of ion chromatogra phy and helped its operation. I also wish to acknowledge the help of depa rtment faculty and staff in Nuclear and Radiological Engineering. I thank my and my wifeÂ’s parents and family for their support. My mother deserves to have my respect and love. She gave love, unconditional support, and encouragement for me to live right. Last , I would like to tha nk my wife for her understanding, patience, and love.

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vi TABLE OF CONTENTS Upage TACKNOWLEDGMENTST.................................................................................................iv TLIST OF TABLEST.............................................................................................................ix TLIST OF FIGUREST..........................................................................................................xii TABSTRACTT.....................................................................................................................xivT CHAPTER 1 INTRODUCTION...................................................................................................1 1.1 Objective......................................................................................................1 1.1.1 Source and Magnitude of Problem..................................................1 1.1.2 Review of Pertinent Literature and Related Work...........................2 1.1.3 Specific Goals..................................................................................3 1.1.4 Impact of Goals................................................................................3 1.2 Methodology................................................................................................4 1.2.1 Task 1: Particle Characterization.....................................................4 1.2.2 Task 2: Effective Dose Scaling Factors...........................................5 1.2.3 Task 3: Dose Assessment via Characterization of Particle Size Distribution......................................................................................6 1.2.4 Task 4: Particle Solubility in Lung Fluid.........................................7 1.2.5 Task 5: Risk Assessment to Workers...............................................7 2 CHARACTERIZATION OF RADIOACTIVE AEROSOLS IN FLORIDA PHOSPHATE PROCESSING FACILITIES.........................................................11 2.1 Introduction................................................................................................11 2.2 Materials and Methods...............................................................................15 2.2.1 Dose Sensitivity to Aerosol Parameters.........................................15 2.2.2 Particle Size Distribution...............................................................16 2.2.3 Particle Density, Shape and Elemental Composition....................18 2.2.4 Particle Radioactivity.....................................................................19 2.3 Results and Discussion..............................................................................20 2.3.1 Dose Sensitivity to Particle Properties...........................................20 2.3.2 Particle Size Distribution...............................................................22 2.3.3 Particle Density, Shape, and Elemental Composition...................25

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vii 2.3.4 Particle Radioactivity.....................................................................28 2.4 Conclusions................................................................................................30 3 EFFECTIVE DOSE SCALING FACTORS FOR USE WITH CASCADE IMPACTOR SAMPLING DATA IN TENORM INHALATION EXPOSURES45 3.1 Introduction................................................................................................45 3.2 Materials and Methods...............................................................................48 3.2.1 Cutoff Size of the Cascade Impactor.............................................48 3.2.2 Other Particle Properties................................................................49 3.2.3 Radioactivity Distribution within a Sub-stage...............................50 3.2.4 TENORM Radionuclides...............................................................51 3.2.5 Dose Calculation............................................................................52 3.3 Results and Discussion..............................................................................54 3.3.1 Inhalation Dose Coefficients – Variation as a Function of Particle Size...................................................................................54 3.3.2 Effective Dose Scaling Fact ors for Uniform Distributions............57 3.3.3 Effective Dose Scaling Fact ors for Linear Distributions...............59 3.3.4 Applications of Effective Do se Scaling Factor to Measured Cascade Impactor Data..................................................................61 3.4 Conclusions................................................................................................63 4 EFFECTIVE DOSE SCALING FACTORS FOR USE WITH CASCADE IMPACTOR SAMPLING DATA IN EXPOSURES TO URANIUM SERIES (APPLICATION TO IMBA PROGRAM)............................................................87 4.1 Introduction................................................................................................87 4.2 Materials and Methods...............................................................................88 4.3 Results and Discussion..............................................................................89 4.3.1 Inhalation Dose Coefficients..........................................................89 4.3.2 Comparison of Inhalation Dose Coefficients from IMBA and LUDEP...........................................................................................91 4.3.3 Effective Dose Scaling Factors......................................................93 4.4 Conclusions................................................................................................94 5 INFLUENCE OF PARTICLE SIZE DISTRIBUTION ON INHALATION DOSES TO WORKERS IN THE FL ORIDA PHOSPHATE INDUSTRY........108 5.1 Introduction..............................................................................................108 5.2 Materials and Methods.............................................................................109 5.2.1 Inhalation Dose Coefficients........................................................110 5.2.2 Inhalation Effective Doses to Workers........................................112 5.3 Results and Discussion............................................................................114 5.3.1 Inhalation Dose Coefficients........................................................114 5.3.2 Inhalation Effective Dose to Workers..........................................118 5.4 Conclusions..............................................................................................121

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viii 6 DETERMINATION OF LUNG SOLUBI LITY OF RADIONUCLIDES IN PARTICLES IN FLORIDA PHOSP HATE PROCESSING FACILITIES.........137 6.1 Introduction..............................................................................................137 6.2 Materials and Methods.............................................................................141 6.2.1 Dose Sensitivity to Radionuclide Solubility................................141 6.2.2 Tested Samples............................................................................141 6.2.3 In vitro Solubility Test................................................................142 6.2.4 Solubility of Surrounding Material..............................................143 6.2.5 Solubility of Uranium, Thorium, and Lead.................................144 6.2.6 Data Analysis...............................................................................145 6.3 Results and Discussion............................................................................147 6.3.1 Dose Sensitivity to Radionuclide Solubility................................147 6.3.2 Lung Solubility of Particles.........................................................148 6.4 Conclusions..............................................................................................151 7 RISK ASSESSMENT OF AIRBORNE PARTICULATES TO WORKERS IN THE FLORIDA PHOSPHATE INDUSTRY......................................................160 7.1 Introduction..............................................................................................160 7.2 Materials and Methods.............................................................................161 7.3 Results and Discussion............................................................................162 7.4 Conclusions..............................................................................................164 8 CONCLUSIONS AND RECOMMENDATIONS..............................................170 8.1 Conclusions..............................................................................................170 8.2 Recommendations....................................................................................174 APPENDIX A PARTICLE SAMPLING AND RADIOAC TIVITY DATA USED FOR DOSE ASSESSMENT....................................................................................................175 B EFFECTIVE DOSE SCALING FACTORS APPLICABLE TO PARTICLES IN THE FLORIDA PHOSPH ATE CHEMICAL PLANTS................................195 LIST OF REFERENCES.................................................................................................201 BIOGRAPHICAL SKETCH...........................................................................................210

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ix LIST OF TABLES UTable U Upage U 2-1 ICRP 66 HRTM default aerosol parame ters and input aerosol parameters for dose sensitivity study...............................................................................................32 2-2 Active operational systems and ventilati on conditions during airborne particle sampling at the various storage areas.......................................................................33 2-3 Mass density of bulk products, settled particles, and airborne particles..................34 2-4 Composition information for dry product ingredients.............................................35 2-5 Phosphorus, silicon, and sulfur comp osition in different sized particles.................36 2-6 P238PU, P226PRa, and P210PPb radioactivity concentrations in bulk dry products, settled particles, and airborne particles................................................................................37 3-1 Particle size information for each impactor stage of the University of Washington Mark III cascade impactor...................................................................65 3-2 Reference physiological parameter values for reference worker.............................66 3-3 GI tract absorption factors (fB1B) for uranium series elements....................................67 3-4 LUDEP decay-chain option for each radionuclide..................................................68 3-5 Inhalation effective dose scaling factors ( SFBEB) for a uniform activity distribution..69 3-6 Inhalation effective dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 2:1)............................................................................................70 3-7 Inhalation effective dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 5:1)............................................................................................71 3-8 Inhalation effective dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:2)............................................................................................72 3-9 Inhalation effective dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:5)............................................................................................73

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x 3-10 Example inhalation dose assessmen t under the assumption of mono-size, uniform, or linearly changing radioactiv ity distribution per impactor stage............74 4-1 Inhalation effective dose scaling factors ( SFBEB) for a uniform activity distribution..96 4-2 Inhalation effective dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 2:1)............................................................................................97 4-3 Inhalation effective dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 5:1)............................................................................................98 4-4 Inhalation effective dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:2)............................................................................................99 4-5 Inhalation effective dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:5)..........................................................................................100 6-1 Samples employed for solubility test.....................................................................153 6-2 Composition of serum ultrafiltrate simulant..........................................................154 6-3 Retention fitting data of P238PU, P232PTh, and P208PPb in serum ultrafiltrate as a function of time....................................................................................................................155 7-1 Inhalation effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants.........................................................................166 7-2 Dose rate and annual inhalation dos e to workers in the Florida phosphate chemical plants due to particle inhalation..............................................................168 7-3 Occupancy times for workers at dry pr oduct, shipping, and storage areas in the Florida phosphate chemical plants.........................................................................169 A-1 Particle size distributi on and radionuclide concentr ation (Granulator area)..........177 A-2 Particle size distribut ion and radionuclide concen tration (Storage area)...............184 A-3 Particle size distribut ion and radionuclide concen tration (Shipping area).............190 B-1 Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for unifo rm radioactivity distribution.........................196 B-2 Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly d ecreasing radioactivit y distribution (AR = 2:1)......................................................................................................................197 B-3 Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly d ecreasing radioactivit y distribution (AR = 5:1)......................................................................................................................198

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xi B-4 Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly in creasing radioactivit y distribution (AR = 1:2)......................................................................................................................199 B-5 Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly in creasing radioactivit y distribution (AR = 1:5)......................................................................................................................200

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xii LIST OF FIGURES UFigure U Upage U 1-1 Anatomic regions of the ICRP Publication 66 HRTM..............................................8 1-2 Particle deposition compartments a nd clearance pathways in ICRP 66 HRTM........9 1-3 Series of tasks for risk assessment of airborne particulates to workers in the phosphate industry....................................................................................................10 2-1 Series of processes for phosphate product manufacture..........................................38 2-2 P238PU decay series.......................................................................................................39 2-3 Inhalation dose coefficients for various particle parameters....................................40 2-4 Airborne particle size distribution............................................................................41 2-5 SEM analysis of different sized airborne particles...................................................43 2-6 EDXS count peak for different sized particles.........................................................44 3-1 Schematic diagram of the University of Washington Mark III cascade impactor...75 3-2 Radioactivity distribution as a function of aerodynamic particle size.....................76 3-3 Inhalation dose coefficients for P238PU decay series....................................................77 3-4 Particle deposition fraction on each sub-region of the respiratory tract...................81 3-5 Inhalation dose coefficient curve of P238PU as a function of particle size...................82 3-6 Effective dose scaling factor of P238PU as a function of activity ratio.........................84 3-7 Application of radioactiv ity distribution using radioactivity measurement data of cascade impactor samples........................................................................................86 4-1 Inhalation dose coefficien ts for light exertion per unit intake of a uranium series radionuclide............................................................................................................101 4-2 Effective dose and tissue-weighted equivalent doses to bone marrow, bone surface, lung, and liver of TypeF P238PU as a function of particle size......................104

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xiii 4-3 Biokinetic model of uraniu m in ICRP Publication 30...........................................105 4-4 Biokinetic model of uraniu m in ICRP Publication 69...........................................106 4-5 Biokinetic model prediction of uraniu m contents in bone, kidneys, liver, and other soft tissues as a function of time after injection into blood..........................107 5-1 Inhalation dose coefficients for P238PU decay series for particle properties of the Florida phosphate industry.....................................................................................124 5-2 Effective dose and weighted equivale nt dose to each tissue per unit intake..........128 5-3 Effective dose rate to workers in Fl orida phosphate chemical plants due to particle inhalation...................................................................................................131 5-4 Annual total effective dose to workers at Florida phosphate chemical plants for radionuclide-specific absorption types F, M, and S...............................................133 5-5 Annual total effective dose to workers at Florida phosphate chemical plants for conservatively assumed radionucli de-specific absorption types............................135 6-1 Schematic of system for in vitro solubility testing.................................................156 6-2 Inhalation dose coefficients fo r absorption types F, M, and S...............................157 6-3 Fraction of remaining phosphate (POB4P B-3P) in particles as a function of time in serum ultrafiltrate...................................................................................................158 6-4 Fraction of remaining P238PU, P232PTh, and P208PPb in particles as a function of time in serum ultrafiltrate...................................................................................................159

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xiv Abstract of Dissertation Pres ented to the Graduate School of the University of Florida in Partial Fulfillment of the Requirements for the Degree of Doctor of Philosophy INHALATION DOSE ASSESSMENT: RISK ASSESSMENT OF AIRBORNE PARTICULATES TO WORKERS IN TH E FLORIDA PHOSPH ATE INDUSTRY By Kwang Pyo Kim December 2005 Chair: Wesley E. Bolch Major Department: Nuclear and Radiological Engineering Health risks to workers in the Florida phosphate industry due to inhalation of particulates containing radioactiv e materials were investigated. Site-specific particle data were established, and were integrated into a full dosimetry assessment. Particle mass concentration varies widely by plants and locations over 1 to 3 orders of magnitude. A majority of samplings indicate that particle mass is concentrated at large particle sizes. A unity shape factor was a ssigned for dose assessment after particle shape analysis. Mass density of the particles ranges from 1.6 to 1.7 g cmP-3P. Radioactivity concentrations of P238PU, P226PRa, and P210PPb in products or particles range 1088 – 4151, 30 – 141, and 248 – 3204 Bq kgP-1P, respectively. An application method of cascade impactor sampling data to dose assessment was presented. Effective dose scaling factors ge nerated in this study reduce computational efforts for dose assessment.

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xv Inhalation dose was calculated using site-s pecific particle size distribution data. Under the least conservative assumption of radi onuclide-specific lung solubility, effective doses at granulator, storage, and sh ipping areas are 0.31 ± 0.12, 0.27 ± 0.07, and 0.22 ± 0.02 mSv yearP-1P, respectively. In contrast, under th e most conservative assumption the effective doses are 2.24 ± 2.53, 1.26 ± 1.19, and 0.56 ± 0.36 mSv yearP-1P at the same areas. Solubility of uranium, thorium, and lead in particles were determined by an in vitro solubility test. More than 94% of phos phate is dissolved in 1 day. However, P238PU, P232PTh, and P208PPb are not dissolved rapidl y with the surrounding matri x. Retention of uranium and lead is similar to type M materials, and retention of thorium to type S materials as defined in the default clearance parameters of the ICRP 66 HRTM. All measured data and dose calculation methods were integrated into a full internal dosimetry assessment. Site-specific effective doses range 0.24 – 0.68 mSv yearP-1P at granulator area, 0.21 – 0.49 mSv yearP-1P at storage area, and 0.21 – 0.27 mSv yearP-1P at shipping area. All results are smaller than dose limit of the general public. Compared with occupational dose limit, effective dose to workers in the Florida phosphate industry is extremely unlikely to approach or exceed the limit.

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1 CHAPTER 1 INTRODUCTION 1.1 Objective The overall objective of this investigation is to evaluate the health risks to workers in the Florida phosphate industry due to inhala tion of particulates containing radioactive materials. This investigati on is limited to areas where tec hnologically enhanced naturally occurring radioactive materials (TENORM) ar e present. In order to accomplish this research objective, the study employs the mo st current particle lung deposition and clearance models published by the Internati onal Commission on Radi ological Protection (ICRP) and collects data to addres s that modelÂ’s input parameters. 1.1.1 Source and Magnitude of Problem The phosphate industry utilizes various natural resources including phosphate ore to manufacture fertilizer and/or animal feed. The radionuclid es present in these natural resources can be concentrated during th e manufacturing process. Elevated and uncontrolled concentrations of radionuclides may pose risks to workers through either external gamma-ray exposure or internal exposure through inhalation of radioactive airborne particles. The processing used in the phosphate industry inevitably generates these airborne particles. Inhalation of ra dioactive airborne particles can potentially elevate the risk to exposed workers. The health risks to workers in the phosphate industry resulting from chronic i nhalation of particulates containing radioactive materials, and/or potentially toxic chemicals, have not been adequately addressed. For the risk assessment, particle properties should be ch aracterized. This study broadens the airborne

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2 particulate database in Florida phosphate chem ical plants. Database collection categories include particle size distribution, particle sh ape, particle elementa l composition, particle density, radionuclide concentrati on in particles, and lung solu bility of radionuclides in particles at different chemical plants a nd locations. The established databases are integrated to individualized site-s pecific dose calculation to workers. 1.1.2 Review of Pertinent Literature and Related Work There has been no related work with this study. Birky et al . conducted intensive study about radiation exposure to workers in the Florida phosphate industry (Birky et al . 1998). The study included internal exposure due to inhaled particulates as well as external exposure. The study adopted conservative assumptions due to the lack of particle information needed for dose assessm ent. Particle size, shape, density, and solubility are important parameters in dose modeling. However, there was no consideration about these prope rties in this 1998 study. Defa ult particle properties were thus assumed for dose assessment. These assumptions tended to increase the dose uncertainty and to skew dose distributions towards possibly unrealistic higher values. Authors recommended: “a more targeted st udy be conducted to reduce uncertainties in that (inhalation) dose component.” The curre nt study provides parame ter data necessary to replace the conservative assumptions used in the previous study and compress the range of uncertainty expressed in the inhalation dose distribu tions. The results of this study are of extreme importance and value; that is, to provide the evidence necessary to confidently determine the need for respir atory protection programs and required or voluntary respirator use in the phosphate industry. In 1994, the ICRP issued Publication 66, Human Respiratory Tr act Model (HRTM) for Radiological Protection (ICRP 1994b). This model represents a substantial

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3 improvement over the existing ICRP Publica tion 30 model that was originally published in 1966 (ICRP 1979). These improvements in clude: (1) facility to estimate regional doses to anatomic substructures of the lung (as opposed to a single averaged lung dose), (2) improved consistency with morphologica l, physiological, a nd radiobiological characteristics of the respiratory tract, (3) ability to estimate doses to individuals of the population other than Reference Man (e.g., wome n and children), (4) ability to consider variations in exertion and cha nges in mouth versus nose breath ing, (5) ability to explicitly consider measured distributions of partic le size, and (6) abi lity to utilize unique information on clearance characteristics (lung flui d solubility, etc.) for specific materials. These features make the ICRP 66 HRTM uniquely applicable to spec ific inhalation dose assessments. Fig. 1-1 shows the anatomic feat ures of the ICRP 66 lung model, while Fig. 1-2 shows the corresponding deposition co mpartments and clearance pathways. 1.1.3 Specific Goals Specific goals of this study ar e: (1) to establish a databa se of particle properties, including particle size distribution, shape, elemental composition, density, and radionuclide concentration, (2) to generate i nhalation dose coefficients and effective dose scaling factors for use with sampling data in inhalation exposures, (3) to establish a database of lung solubility of radionuclides in particles, and (4 ) to assess doses to workers in the Florida phosphate industry us ing integrated database s established in the previous stages. 1.1.4 Impact of Goals Completed databases and dose evaluations generated in this study may be used by industry staff and regulators alike to determ ine the need and appr opriate levels of

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4 administrative controls, engineering contro ls, and respiratory pr otection required to assure that health risks to workers are kept as low as reasonably achievable. 1.2 Methodology In order to accomplish the overall obj ective and specific goals, the study was divided into 5 specific tasks: (1) particle characterization (2 ) generation of effective dose scaling factors, (3) dose assessment via char acterization of particle size distribution, (4) determination of lung solubility of radionuclides in particles, and (5) risk assessment to workers due to radiation exposure. The sequen ce of task procedures is illustrated in Fig 1-3. 1.2.1 Task 1: Particle Characterization Particle properties, including particle size distribution, shape, elemental composition, density, and radionuclide concentr ation in particles were characterized. These are input particle parameters for HR TM models to determine particle deposition and clearance in the respirator y tract. In the absence of specific information, ICRP 66 HRTM suggests default reference values. Ho wever, the reference values were obtained by pooling data from several studies. In reality , particle properties ar e widely distributed. Consequently, using recommended default valu es can potentially skew dose estimates to unrealistic values in individual exposure cases . For this reason, site-specific particle properties should be determined whenever possi ble to reduce this sour ce of bias in the dose assessment. Air particulate samplings were conducted at dry product manufacturing areas, storage warehouses, and shipping areas within 6 different phosphate processing plants in either the central or northern regions of Florida. Partic le size distribution was characterized using a multi-stage cascade im pactor. In addition, a high-volume sampler was used to collect airborne particles for both particle density and ra dioactivity analyses.

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5 Limited numbers of the air samples were furthe r analyzed with regard to particle shape, chemical composition, and density using S canning Electron Microscopy (SEM), Energy Dispersive X-ray Spectroscopy (EDXS), and pycnometer, respectively. The types of radionuclides and their strengths in part icles were measured using gamma-ray spectroscopy. 1.2.2 Task 2: Effective Dose Scaling Factors Application method of sampling data to dose assessment and concept of inhalation effective dose scaling factors ( SFBEB) will be introduced. Once particle properties are characterized, radiation doses due to particle inhalation can be calculated using the ICRP 66 HRTM. Due to the complexity of the ICRP 66 model, one must generally resort to a computer program to apply the model for a specific exposure scenario. Currently, two computer programs are available, includi ng Lung Dose Evaluation Program (LUDEP) and Integrated Modules for Bioassay Analysis (IMBA). In the computer codes, both organ absorbed doses and the effective dose are computed readily only in cases where radioactivity is distributed log-normally. If the airborne particle size distribution is characterized via cascade impactor measurem ents and found not to follow a log-normal distribution, the dose calculati on must be performed via a summation of doses calculated separately for each impactor size range. Cas cade impactor yields one radioactivity value for each stage (i.e., particle size range). One of three options is available to directly use the measured data for dose assessment: (1 ) radioactivity concen tration on a mono-size particle size, (2) uniform ra dioactivity distribution across each size interval, (3) linear increasing or decreasing radioac tivity distribution. A series of inhalation effective dose scaling factors, defined as the ratio of th e effective dose given unde r options 2 and 3 to the effective dose given by option 1, are presen ted. Consequently, these scaling factors

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6 allow one to run the computer codes with less computationally de manding option 1 while then reporting doses under the more realis tic description of options 2 and 3. LUDEP employs biokinetic models in ICRP Publica tion 30 while IMBA uses recent biokinetic models in ICRP Publications 56, 67, 69, a nd 71. Scaling factors generated by LUDEP and IMBA are given in Chapte rs 3 and 4, respectively. 1.2.3 Task 3: Dose Assessment via Characteri zation of Particle Size Distribution Airborne particle measurement data fr om Task 1 and dose assessment methodology from Tasks 2 and 3 were used to calculate dos e to workers. In a given exposure scenario, a worker may be exposed to individual radionuclides in the P238PU series, or to segments of the decay chain that remain under secular equi librium. The dose due to inhalation of particles encompassing a decay-chain series can be easily calculated through summation of the doses received by each decay-chain member. Furthermore, progeny generated by the decay of inhaled radionuclides will furthe r contribute to internal dose; hence, in-vivo in-growth of progeny must also be taken into ac count. It is thus usef ul to generate data regarding inhalation effective doses from each radionuclide independently as they can be modified according to their unique particle properties and radioactivity concentrations. In this task, inhalation dose coefficients for the P238PU series were calculated as a function of particle size and absorption type. No specific information on lung solubility of radionuclides was utilized in this task, and thus dose calculations were conducted for default absorption types given in ICRP Pub lication 66. This study further reviewed the inhalation dose parameter sensitivity to inve stigate the relative importance of particle solubility.

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7 1.2.4 Task 4: Particle Solubility in Lung Fluid In this task, the solubilites of TENOR M radionuclides in the lung fluid were determined. The sensitivity analysis in Task 3 indicated that proper knowledge of absorption type was one of the more critical parameters for inhalation dose calculation. Conservative assumptions on the radionuclide absorption types could skew internal dose by factors of 7 to 22 in the Florida phosphate industry. In the absence of specific information, the ICRP 66 HRTM suggests defau lt values of material absorption to blood for three general classifications: F fast, M moderate, and S – slow. Absorption types of radionuclides given in references depe nd on their physicochemical forms. In the phosphate industry, the physicochemical forms of the radionuclides inhaled are not well defined. In addition, the radi oactive material is a minor constituent of the inhaled particles. In this case, ab sorption of the radionuclide into blood may be determined by the properties of the radionuc lide-containing matrix rather than by the radionuclide compound type (ICRP 1995b). The features of TENORM make it difficult to estimate the absorption type of radionuc lides in the particles in the phosphate industry. An in vitro solubility test was employed to directly m easure radionuclide solubility and thus avoid excessively conservative assumptions in worker dose assessments. 1.2.5 Task 5: Risk Asse ssment to Workers In the final task of this study, all data obtained directly by sampling and analysis were integrated into a full internal dosime try assessment to worker s by both facility and operational location. With external dose estimates documented in the previous TENORM studies, total effective dose was es timated as an indi cator of radiation exposure risk to workers in the Florida phosphate industry.

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8 Figure 1-1. Anatomic regions of the ICRP P ublication 66 HRTM (Adapted from Figure 1 in ICRP Publication 66)

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9 Figure 1-2. Particle deposition compartments and clearance pathways in ICRP 66 HRTM

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10 Figure 1-3. Series of tasks for risk assessment of airborne particulat es to workers in the phosphate industry

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11 CHAPTER 2 CHARACTERIZATION OF RADIOACTIVE AEROSOLS IN FLORIDA PHOSPHATE PROCESSING FACILITIES 2.1 Introduction The phosphate industry produces fertilizer , animal feed, and phosphoric acid using phosphate ore, which contains Naturally O ccurring Radioactive Materials or NORM. The industrial practices utilizing natural resour ces generate aerosols in the workplace that concentrate NORM to a degree. Risks to workers from aerosol emissions from the phosphate industry include radiation exposur es resulting from aerosol inhalation and external gamma-ray exposure. Concentrated radionuclides as a result of human industrial practice are referred to as Technologically Enhanced Naturally Occurring Radioactive Material or TENORM. TENORM issues ar e mainly focused on waste streams from industrial processes and phosphate fertilizers owing to their widespread use (EPA 1991). The majority of published studies on TENORM in the phosphate indust ry to date have been conducted for the purpose of environm ental protection (Boot he 1977; Burnett and Elzerman 2001; Haridasan et al . 2002; Hull and Burnett 1996; Paridaens and Vanmarcke 2001). A comprehensive integrated study on wo rker exposures from TENORM in the Florida phosphate industry was carried out by Birky et al . (Birky et al . 1998). That study collected extensive data on wo rker exposures at different processing areas including mines and chemical plants. External exposur e from airborne and settled particles as well as internal exposure due to airborne partic le inhalation and dige stion was assessed for

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12 workers in different job classi fications. The study demonstrat ed that particle inhalation was a main source of radiati on exposure to workers. The study employed high-volume samplers, settled down dust, and inhalation dose co efficients as given in Publication 61 of the International Commission on Radiological Protection, or ICRP (ICRP 1991; ICRP 1994a). The authors pointed out that in thei r study “no adjustment for particle size” was made and that “the inhalation dos es were conservative” as they tended to be “greater than actual doses.” Other recent st udies were conducted by Gafvert et al . and Lipsztein et al . (Gafvert et al . 2001; Lipsztein et al . 2001). In these studies, detailed information on the particle size distribution and ot her particle characteristics were not available for more accurate and potentially less conservative asse ssments of worker inhalation dose. In 1994, the ICRP issued Publication 66, Human Respiratory Tr act Model (HRTM) for Radiological Protection (ICRP 1994b). Th e features of ICRP 66 HRTM enable the user to individualize the inha lation dose assessment for workers exposed to occupational or environmental aerosols. For dose assessmen t, the properties of the aerosols in the exposure scenario should be characterize d according to both their radiological and physicochemical properties. These aerosol characteristics determine the deposition fractions within each sub-region of respir atory tract, and clearance rates from these respiratory regions, and local va lues of absorbed dose to radi osensitive cell layers. Input aerosol parameters include the aerosol size distribution, mass density, and shape factor. The other input parameter is the absorption type, which denotes the rate at which the aerosol particles dissolve within the lung flui ds and are thus cleared to pulmonary blood vessels. The ICRP 66 HRTM recommends th at whenever possible, site-specific information on aerosol physicochemical properti es should be measured and then used in

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13 the worker dose assessment (ICRP 1994b). In the absence of aerosol information, it is assumed that radioactive materials are distri buted log-normally in particle size with a geometric mean given by the Activity Medi an Aerodynamic Diameter (AMAD) and an associated Geometric Standard Deviation (G SD). Recommended aerosol default values are as follows: AMADs of 5 m and 1 m, respectively, for workplace exposures and general public exposures, a mass density of 3 g cmP-3P, and a particle shape factor of 1.5 (see Table 2-1). Inhalation doses are strongly influenced by the particle size distribution. Consequently, many studies have been c onducted to measure the AMAD in both the workplace and environment. Dorrian and Ba iley reviewed about 400 measurements of AMAD in the workplace and environment (Dorrian 1997; Dorrian and Bailey 1995; Dorrian and Bailey 1996). Median values of AMAD for all workplaces and different industries ranged from 4.0 to 7.3 m, while those for all artificial environmental aerosols, Chernobyl fallout, and natural P7PBe aerosols ranged from 0.6 to 1.5 m. Most measurements followed log-normal distributions . The authors concl uded, therefore, that the reference particle sizes in ICRP Pub lication 66 for occupational and environmental exposures are appropriate. Although medi an values of the AMAD throughout various workplace locations are similar to default va lues recommended in ICRP Publication 66, it should be noted that measured values of AM AD were widely distributed and ranged from as small as 0.12 m to as large as 25 m in specific workplace environments. Similarly, sampled AMADs for environmental aerosols ranged from 0.3 m to 18 m. The particle-size distribution depends on the mech anisms for aerosol generation, the time for coagulation or gravitational settling, and various other ex ternal conditions. The AMAD

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14 in a specific single area can greatly differ from the median values obtained by pooling data from several workplace studies. Consequent ly, use of default valu es rather than sitespecific data can potentially skew dose esti mates to unrealistic values in individual exposure cases. For this reason, site-specific aerosol particle size distributions should be determined whenever possible to reduce this source of bias in the dose assessment. In addition to particle size, particle shap e and density are important parameters to estimate particle behavior inside of the human respiratory tract. Particle deposition and clearance mechanisms depend on two characte rizations of the sampled particle: its aerodynamic diameter and its thermodynamic diam eter. Particle shape and density are parameters relating these two types of particle sizes: 0() ()ae thae thCd dd Cd , Eq. 2-1 where dBthB and dBaeB are the thermodynamic and aerodynamic particle diameters, is particle shape factor, B0B is unit density (1 g cmP-3P), is the actual particle density, and C(dBthB) and C(dBaeB) are slip correction factors at these two diameters (Hinds 1999). The aerosol particles might be composed of product, ambient impurities, and substances from other areas. Therefore, elemental composition analyses of particles with different sizes make it possible to trace the source of aerosols and rela te other particle pr operties with size. The radioactivity concentrati ons of phosphate materials, including matrix, products, by-products, and waste have been the subj ect of many previous studies (Burnett et al . 1995; EPA 1977; EPA 1978; Guimond 1978; Guimond and Windham 1975; Hull and Burnett 1996; Laiche and Scott 1991; Owen and Hyder 1980; Roessler et al . 1979; Warren et al . 2001). The principal radionuclides cont ained in phosphate material are of

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15 the P238PU decay series. It has been noted that these radionuclides are initially in secular equilibrium within the matrix, yet are con centrated non-uniformly within product or byproduct materials after various process steps. The objectives of the present study were to make direct measurements of the particle size distribution, mass density, shap e, chemical composition, and radioactivity concentrations at worker environments within several facilities in the central and north Florida phosphate industry. The database of particle size and phys iochemical properties thus established is needed for more worke r-specific assessments of radiation inhalation exposures devoid of potentially conservative default assumptions. Th e data collected will also provide important information to the phos phate industry and to state regulators as they assess the need for any changes in esta blished respiratory a nd other radiological protection policies. 2.2 Materials and Methods 2.2.1 Dose Sensitivity to Aerosol Parameters A sensitivity study was first conducted to determine the degree to which various aerosol physicochemical propert ies contribute to uncertainties in the effective dose to phosphate workers during inhalation exposures of P238PU and P230PTh aerosols. Values of the 50-year committed effective dose per unit inta ke (e.g., effective dose coefficient) to workers under light exertion were computed using the Integrated Modules for Bioassay Analysis (IMBA) code of James et al. (James et al. 2003). The IMBA code is based upon the ICRP 66 HRTM as well as the more r ecent elemental biokinetic models published by the ICRP. The program explicitly accounts for the in-growth of decay progeny following inhalation of the P238PU and P230PTh parent aerosol particle.

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16 Table 2-1 lists the aerosol input parameters considered. Initia lly, the effective dose coefficient was calculated using ICRP default aerosol parameters. In the ICRP 66 model, the particle size distribution is characterized by a single lognormal distribution with an activity median aerodynamic di ameter (AMAD) of either 5 m for workplace exposures or 1 m for environmental exposures to members of the general public. In both cases, the distribution is further characterized by a geometric standard deviation (GSD) given as a function of the correspondi ng activity mean thermodynamic diameter (AMTD). Other ICRP 66 default parameters include a particle density of 3 g cmP-3P, a shape factor of 1.5, and a moderate level of solubility (type M) for P238PU and a slow level of solubility (type S) for P230PTh within the lung fluids follo wing particle deposition. Next, the IMBA code was use to generate P238PU and P230PTh effective dose coefficients in which each aerosol parameter was varied over a reasonably broad but realistic span of values. The particle size distribution wa s allowed to vary from an AMAD of 0.01 m to an AMAD of 100 m while the GSD was set to unity (mono-size distribution) or to 6.2, roughly the square of the ICRP de fault values. The particle de nsity was then set to values of either 0.7 or 11 g cmP-3P, while the particle shape factor was given a value of either 1.0 (spherical) or 2.0 (elong ated) (Hinds 1999). 2.2.2 Particle Size Distribution Fig. 2-1 briefly depicts the various industrial processes in the Florida phosphate industry. A previous study has shown that dry product manufacturing area, dry product storage warehouse, and shipping area have th e highest potential for worker radiation exposure (Birky et al . 1998). Final phosphate products are generated in dry product manufacturing areas, which accommodate sy stems for product granulation, screening,

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17 and drying. The products are loaded directly for shipping, or are moved to a storage warehouse for later shipment. All processe s are associated with aerosol-generating activities either from automated mechanical sy stems or mobile heavy equipment. In this study, air samples were collected at granulator, st orage, and shipping areas within 6 different phosphate processing plants in either the central or norther n regions of Florida. A PMB2.5 Bdichotomous sampler was employed fo r pre-sampling studies to estimate particle concentrations. These measurements were used in turn to estimate maximum cascade impactor sampling times at each wo rker area. By so doing, impactor stage overload and subsequent partic le-bounce effects could be avoided. A University of Washington (UW) Mark III Cascade Impactor was used to characterize the particle size distribution (Marple 2004; Pe gnam and Pilat 1992; Pilat et al . 1970). The UW Mark III impactor consists of 7 impactor stages followed by a final collection filter thus partitioning particles into 8 different size ranges (Pilat 1998). The cutoff size of each stage was calculated by the methods of Marp le and Willeke and employing the value of 0.49 for the square root of the Stokes number (John 1999; Marple and Willeke 1976; Rader and Marple 1985; Sethi and John 1993) . The aerodynamic cutoff sizes were calculated as 21.50, 9.38, 3.57, 1.76, 0.97, 0.51, and 0.26 m for the 1PstP to 7PthP impactor stages at a flow rate of 15 L minP-1P. The upper particle size li mit of the cascade sampler was taken to be 100 m following that for the total susp ended particulate high-volume air samplers (EPA 1999). The lower limit of partic les collected on the fi nal collection filter was taken to be 0.03 m, a value typical of those employed in similar studies (Divita et al 1996; Howell et al 1998; Marley et al 2000; Wagner and Leith 2001).

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18 A polycarbonate screen membrane film was placed over the cascade impactor substrate to facilitate particle removal as need ed for subsequent analyses of particle shape and elemental composition. The samplings were conducted as close as possible to current worker locations, and the height of the inlet nozzle of the cascade impactor was set at a nominal breathing height of 1.5 m. The cascade impactor was operated at a flow rate of 15 L minP-1P for 3 to 27 hours depending on the pr edetermined aerosol concentration at each area. Duplicate samplings were conducted at all locations. After sampling, impactor substrates containing aerosols were removed from the imp actor, desiccated for 24 hours, and then weighed using a microbala nce (Sartorius, MC210S). Additionally, a PMB10B high-volume sampler (Sierra-Andersen, Model 1200) was used to collect aerosols with an aerodynamic diameter 10 m as needed for analyses of both particle density and radioactivity content. Samplings without the PMB10B orifice were also conducted to collect aerosols at larger sizes ( 100 m). 2.2.3 Particle Density, Shape, and Elemental Composition The mass density measurements using a pycnometer (Quanatachrome, Ultrapycnometer 1000) were made of (1) bulk dry products (including monoammonium phosphate or MAP, and diammonium phosphate or DAP), (2) settled particles from granulator areas, and (3) ai rborne particles collected by high-volume sampler at granulator areas within the 6 Florida phospha te processing plants. A limited number of air samples collected by cascade impactor were further analy zed for both particle shape and elemental composition using Scanni ng Electron Microsco py (SEM) and Energy Dispersive X-ray Spectroscopy (EDXS), respect ively. Particles of 3 size ranges from 3 different areas at 2 different plants we re analyzed for this portion of the study.

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19 2.2.4 Particle Radioactivity The mass fraction of uranium in dry pr oducts ranges from 0.014% to 0.025% (Guimond 1978; NCRP 1987; Owe n and Hyder 1980; Roessler et al . 1979). The small mass of air samples collected at each cascade impactor stage resulted in radioactivity loadings that were generally below detect able limits using high-efficiency gamma-ray spectroscopy. Consequently, ra dioactivity measurements were carried out us ing bulk dry products, settled particles, a nd air samples collected by PMB10B high-volume sampler with aerodynamic diameters smaller than 10 m and 100 m. A well-type High Purity Germanium (HPGe) detector was employed for radioactive measurement because of its hi gh photon detection efficiency. The gammaray spectroscopy system was calibrated with respect to both photon energy and photon detection efficiency using a uranium ore stan dard from the U.S. Department of Energy New Brunswick Laboratory. Phosphate ores contain P238PU series as well as naturally occurring P40PK (Birky et al . 1998). The entire P238PU series is depicted in Fig 2-1. The contribution of P40PK to inhalation dose is negligible compared to the P238PU series. Radioactivity measurements were primarily focused on P238PU, P226PRa, and P210PPb as any deviation from natural decay-chain equilibrium would typically occur at the longer-lived daughters such as P226PRa and P210PPb. Collected samples were sealed for 30 days in small plastic vials suitable for counting in the HP Ge well detector to re-establish secular equilibrium between P226PRa and its P222PRn progeny prior to measurement. In this manner, higher yield gamma-ray photons from P214PPb and P214PBi can be used to estimate the radioactivity of the P226PRa parent within the sample. Following the 30-day hold period, the samples were counted for 24 to 48 hours.

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20 2.3 Results and Discussion 2.3.1 Dose Sensitivity to Particle Properties Figure 2-3 displays the inhala tion dose coefficients of P238PU and P230PTh as a function of each aerosol characteristic given in Table 2-1. The si ngle solid circle and solid triangle indicate dose coefficients for ICRP default aerosol values for general public exposures and workplace exposures, respectivel y. The default effective doses per unit intake of P238PU and P230PTh for the general public are noted to be 1.53 and 1.78 times that for radiation workers due to the smaller size di stribution and thus d eeper lung deposition assumed for environmental aerosols w ithin the ICRP 66 inhalation model. Dose coefficients were further calcula ted as a function of AMAD with other aerosol parameters held cons tant at their ICRP 66 defau lt values. In general, the inhalation dose coefficient decreases with increasing AMAD over the size range 0.01 to 100 m. For P238PU, aerosols with 0.01 and 0.1 m AMAD log-normal distributions result in effective dose coefficients that are 9 and 3 times that of a 1 m AMAD distribution, respectively. Conversely, effective dose s for aerosols with AMADs of 10 and 100 m are 1/2 and 1/12 of those for 1 m AMAD distributions. Inhalation of P230PTh with 0.01, 0.1, 10, and 100 m AMADs, respectively, results in effective doses 5 and 3 times higher, or 1/4 and 1/10 times lower than that of 1 m AMAD P230PTh. If the aerosol is present as single m onodisperse particles (GSD=1), Figure 2-3 indicates that the inhalation dose coefficient decreases with increasing particle size down to a particle size of ~0.6 m, and subsequently increases to a maximum value at a particle size of ~2.5 m. After reaching this maximum, th e effective dose per unit intake again decreases with increasing particle size. The shape of the effective dose coefficient is

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21 related to the regional lung deposition of aerosol particles as a functi on of particle size. Aerosols deposited deeper with in the respiratory tract (e.g., alveolar region) result in longer retention times and thus higher radiat ion absorbed doses in comparison to those deposited primarily within th e upper regions of the respirat ory tract (e.g., extra-thoracic region), as the latter is removed to the gastrointestinal (GI) tract faster through mucociliary action. The regional depositi on fraction of mono-size aerosols (GSD=1) shows an initial decrease with particle size, an increase to a maximum value, and then a decrease once again for all sub-regions of th e respiratory tract, including the extrathoracic airways, the bronchial, br onchiolar, and alveolar-int erstitial regions (Kim et al. 2005). This local maximum in the effective dose coefficient within the 1 to 10 m size region is dampened considerably as the ae rosol size distribution widens. Particle deposition in the respiratory trac t depends on two characterizations of the aerosol particles: the aerodynamic diameter and the thermodynamic diameter given in Eq. 2-1. A higher mass density and smaller shap e factor thus decreases the thermodynamic diameter for the same aerodynamic size and allo ws for deeper penetration of the aerosol within the lung airways, which results in a hi gher lung (and effective) dose per unit intake. As demonstrated in Figure 2-3, once the aerosol size is we ll characterized (AMAD value), inhalation effective dose coefficients for P238PU and P230PTh are shown to vary no more than a factor 2 from ICRP 66 default values with changes in particle mass density, particle shape factor, or distributional GSD. This effective dose sensitivity study indicat es that the aerosol size distribution is one of important input parameters for indivi dual and site-specific dose assessments. The focus of the present study is thus to provide site-specific data on all aerosol

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22 characteristics within the Florida phospha te industry as needed for worker dose assessment, with particular emphasis on the particle size distribution. 2.3.2 Particle Size Distribution Fig. 2-4 displays aerosol size distributions at granulator, storage, and shipping areas across all 6 Florida phosphate facilities particip ating in the study. For clarity, the distributions are depicted as c onnected scatter plots rather than as histograms. Each data point thus represents the aerosol mass con centration at the geometric mean of the impactor stage. The data points shown ar e mean values of duplicate samples. For each worker area, aerosol size measurements are shown not to follow a normal distributional shape on a logarithmic size sc ale (i.e., lognormal distribution). Aerosol mass is increasingly concentrated at larger aero sol sizes for the majority of the plants and operational areas of the study. Aerosol mass c oncentrations in the larger size intervals are generally higher at granul ator and storage areas, w ith values dependent upon the operational systems, mechanical activity levels , ventilation, and buildin g structures at the various sampling areas. In general, operat ional systems in granulator areas include granulators, screeners, drye rs, and conveyers that all run during product manufacturing and thus generate aerosols in the work envi ronment. Those in storage areas include conveyors, reclaimers, and heavy equipment such as bobcat trackers or payloaders. Conveyors move products from granulator areas to storage areas or from storage areas to shipping areas for cargo loading. Various type s of heavy equipment are used to move products inside storage areas for shipping and th ey typically generate aerosols to a greater extent than conveyors or reclaimers. Furtherm ore, ventilation is an important factor in the observed aerosol mass concentration. Dilutio n of highly concentrated aerosols to the

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23 ambient air via opened doors within the stor age facility can greatly reduce aerosol mass concentrations. The gradient of aerosol mass concentration w ith aerosol size (sl opes in Fig. 2-4) is greater in both the granulator areas (except at plant F), and in those storage areas with high levels of operational activity (plants D and F). Within larger size intervals, aerosol mass concentrations at granulator, storage, and shipping areas vary approximately 2 – 3 orders of magnitude. In cont rast, variations in concentration within the smallest size intervals generally decrease at all three operational areas. For granulator areas, aerosol mass concentr ations show a similar trend except at plant F where a roughly size-independent dist ribution is noted. In addition, the concentration variance between plants (excludi ng plant F) is less than those seen at storage or shipping areas. Gene rally, the granulator area is operated mainly under steadystate conditions. Several systems including gr anulator, dryer, scr een, and conveyor are operated such that all introduce aerosols to the worker breathing environment. Granulator areas are isolated from ambient ai r, and thus external conditions such as weather do not influence indoor aerosol mass concentrations. The maximum mass concentration differences among the various pl ants are 1.5 orders of magnitude at large aerosol sizes (excluding plant F) , and about 1 order of magnitude at smaller aerosol sizes. The aerosol mass concentrations at plant F are much lower than those at other plants at the larger aerosol sizes. Unlike the other phos phate plants in this study, Plant F employs a scrubber system, which includes a fan to move aerosols and fumes through water flowing into plastic packing. The aerosols and fumes contacting the water are removed from the air-stream. Nonetheless, aerosol mass concentrations at sizes of a few tenths of

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24 a micrometer are comparable to those at other pl ants. In fact, in the smallest aerosol size range (0.03 – 0.26 m), the aerosol mass concentration at Plant F is noted to be higher than that found at other facilities. Aerosol mass concentrations at storage ar eas show several different trends in comparison to those at granulat or areas. The variation is wide at a given aerosol size range. The maximum concentration difference at large aerosol sizes is more than a factor of 1000. For plants D and F, aerosols are mainly concentrated at larger sizes, where they achieve magnitudes comparable to those seen at granulator areas. For the other plants, large-sized aerosol mass concentrations ar e much lower than those seen at their granulator areas. Air sampling results at storage areas are hi ghly dependent on spec ific types of work activities and other conditions within the buildings. There are times when only conveyer systems are in operation for the product st ack. Alternatively, heavy equipment is sometimes in operation for product loading or moving. The wide differences of activity levels in the warehouses account for the wide variations in aerosol mass concentrations observed within the storage areas of Fig. 2-4B . During low-level activity, fewer aerosols are generated. Large-sized aerosol concentr ations decrease significantly due to rapid settling. If the warehouse is ope n to ambient air, the mass con centration decrease is more pronounced due to dilution. Table 2-2 show s the types of operational systems and ventilation conditions at storage areas during air sa mpling. The aerosol mass concentration is the highest at plant D where the warehous e doors were always closed during operation of heavy equipment. For ot her plants, only conveyor systems were in operation during aerosol sampling, or the door s were open to outside air during heavy

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25 equipment operation. For Plant A-b (2), ther e was no mechanical act ivity at the time of sampling, and thus the aerosol conc entration was extremely low. Aerosol mass concentrations are generally the lowest at shipping areas about one order of magnitude less than found at granulat or areas for a given ae rosol size interval. The concentrations show a wide variability depending on the specific plant in question. The maximum mass concentration difference is about a factor of 25 for large aerosols and about a factor of 10 for small aerosols. A car-loading system is the only operational device that can generate aerosols in the shi pping area. In addition, the shipping areas are open to the ambient environment where conditions such as wind or rain can strongly influence the aerosol mass concentration. Ae rosols generated from the car-loader are then quickly diluted in the atmosphere, and ar e thus generally lower than at either the storage or granulator areas. 2.3.3 Particle Density, Shape, and Elemental Composition Table 2-3 lists the measured densities of bulk products, settled particles, and air samples. As there was no discernable de nsity difference between the bulk products MAP and DAP, the data are given only for the differe nt sample types. The mass densities of bulk products, settled particles, and airborne particles smaller than 100 m are identical at ~1.7 g cmP-3P, whereas the mass densities of ai rborne particles smaller than 10 m were found to be slightly less than this value. Particle shape analysis was conducted on ai r samples from 3 different areas within 2 different chemical processing plants. Fi g. 2-5 shows particles of different sizes collected at the granulator area of plant A. Since the particles were compressed during impaction on the filter (e.g., Fig. 2-5B), it wa s difficult to differentiate each individual

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26 particle from the aggregate particle mass. Single particles can be found at the edges of the impactor stage, and thus images of these edge particles were used for particle shape analysis (e.g., Figs. 2-5A and 2-5C). In genera l, all particles of various sizes appeared as spheroids or as rough spherical fragments. No fiber-like particles were found. Particle shape influences the particle drag force, a nd is thus characterized by the dynamic shape factor defined as the ratio of the resistance forc e of the true non-spheri cal particle to that of a spherical particle of unit density (Hinds 1999). Dynamic shape factors for occupational aerosols typically range in valu e from 1.1 to 1.9 (Mercer 1973). In ICRP Publication 66, a reference dynamic shape fact or of 1.5 is suggested for occupational radioactive aeros ols. Johnson et al . calculated dynamic shap e factors of several nonspherical particles, and noted th at they ranged from unity (spheric al particles) to values of 1.09 to 1.23 for cylindrical particles having ax ial ratios of 2 to 5, respectively (Johnson et al . 1987). Axial ratios of most particles observed in this study did not exceed a value of 2. Consequently, a shape factor of unity can be assumed for the airborne particles in the Florida phosphate industry as needed for inhalation dose assessment. Another benefit of the shape analysis is to verify the cutoff size at each stage of the Mark III cascade impactor. If a particle ha s a dynamic shape factor of unity, the volume equivalent diameter ranges at the 1PstP, 4PthP, and 7PthP stages are 17.0 – 79.1 m, 1.4 – 2.8 m, and 0.2 – 0.4 m, respectively. The geometric size ranges of particles observed under SEM were found within these calculated values. Table 2-4 shows the chemical form of dry products which include primarily nitrogen and phosphorus. The dry products also contain other impurities including calcium, iron, aluminum, magnesium, sulfur, and silicon, although the mass fraction of

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27 these elements are around 10% (CF 2003a ; CF 2003b; Mosaic 2003a; Mosaic 2003b; PCS 2003a; PCS 2003b). Fig 2-6 shows EDXS count peaks for sampled airborne particles. Large characteristic X-ray peak s are found at energies corresponding to carbon, oxygen, and phosphorus for large-sized (1PstP impactor stage) and medium-sized (4PthP impactor stage) particles. For small-sized (7PthP impactor stage) particles, the phosphorus peak is significantly depressed, while domi nant peaks are noted at X-ray energies corresponding to both sulfur and silicon. The carbon peaks come from the coating material required as part of sample prepar ation prior to SEM scanning, and can thus be disregarded in the composition analyses. Qualitative trends are noted in the data of Table 2-5 showing the elemental composition of sampled airborne particles via EDXS for different particle size ranges. The product types from plant A and D are MAP and DAP, respectively. The airborne particles contain various impuritie s, with silicon and sulfur be ing the more dominant ones. The particles collected on the 1st and 4th impactor stages contain ~12 to 28% phosphorus. The fraction of silicon and sulfur in these sa mples is very low in comparison. For smallsized particles, the fraction of phosphorus decreases to between 0.7 and 2.9%, while the fraction of silicon and sulfur increases to ~20% for the majority of samples (except that from the storage area in plant D). The resu lts imply two aerosol sources in this work environment. The first contri butes to large-sized aerosols whose main component is the phosphate product. The other contributes to small-sized aerosols whose main components are sulfur and silicon. Neverthe less, the phosphorus fraction of small-sized particles (7th impactor stage) from the storage area at plant D is still more than 20%. As mentioned earlier, the storage area in plant D is operated at a high level of mechanical

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28 activity without ventila tion, thus showing very high aeros ol mass concentrations. The aerosol contribution from phospha te materials due to high leve ls of mechanical activity exceeds that from sulfur or silicon, and thus the phosphate concentra tion is still high for this particular facility. Phosphate products are derived from reaction between phosphoric acid and ammonium gas, where the phosphor ic acid portion cont ributes the TENORM radioactivity in the product. In contrast, ammonium is not typically associated with natural radioactivity (N CRP 1987). As a result, the radioactivity of the sampled aerosols is thought to be primarily co rrelated with the amount of phos phorus in the aerosol sample. 2.3.4 Particle Radioactivity Table 2-6 shows radionuclide c oncentrations in bulk produc ts, settled particles, and airborne particles. For bulk dry products and settled particles, the radioactivity concentration of P238PU is significantly higher th an that found for either P226PRa or P210PPb. The radioactivity concentrations of P210PPb are ~5-11 times that of P226PRa. During the chemical process for phosphoric acid genera tion, a redistribution of the P238PU decay series occurs from its original state of secular equili brium within the phosphate rock (Burnett et al . 1995; Guimond and Windham 1975; Laic he and Scott 1991; Lardinoye et al . 1982; Roessler et al . 1979). In this redistribution, uranium selectively localizes within product materials, while radium tends to local ize in the by-product material known as phosphogypsum. The results given in Table 2-6 show a similar trend. For P210PPb, different theories for its lo calization have been proposed. Roessler suggested that P210PPb localizes in the bulk dry product, while Horton et al . indicated it localizes in phosphogypsum (Horton et al . 1988; Roessler 1990). In this study, it is seen that P210PPb does not localize in the bulk dry product, but in by-produc t materials during chemical processing. The radionuclide c oncentrations in bulk products are in the range of those

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29 reported previously (P226PRa in DAP and P226PRa and P238PU in MAP) or slightly higher than previously reported (P238PU in MAP) (Guimond 1978; N CRP 1987; Owen and Hyder 1980; Roessler et al . 1979). For settled particles and air borne particles, the radioactivity concentrations of P238PU and P226PRa in most samples are within the range of reported data except for P238PU in settled DAP dust, which is slightly hi gher than that previously reported. The radioactivity concentration of P238PU in airborne particles from northern Florida are lower than those from central Florida, which is likely due to its lower radioactivity concentration in source material or phosphate matrix in that region of the state (Roessler et al . 1979). Between MAP and DAP, no significant radi oactivity differences are found for bulk dry products, settled dust, and sa mpled aerosols. Although mean P238PU radioactivity concentrations of airborne particles are slig htly less than those in bulk products and settled particles, the difference is insignifi cant considering deviat ion among samples. The radioactivity c oncentrations of P226PRa in airborne particles are higher than those in bulk products and settled particles, but st ill within one standard deviation. The radioactivity c oncentrations of P210PPb in bulk products and settled dust are also similar. However, significantly high er radioactivity concentrations of P210PPb are found in the sampled aerosols. The P210PPb radioactivity of MAP aerosols (3204 Bq kgP-1P) is even higher than the P238PU radioactivity con centration (2039 Bq kgP-1P). The increased P210PPb radioactivity is suggested to come from the deposition of ambient airborne radon-decay products in these worker environments. The half-lives of radon progeny, including P218PPo, P214PPb, P214PBi, and P214PPo, are very short, and thus they decay to P210PPb during aerosol suspension. The attachment of radon progeny to aerosols has been studied extensively

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30 (Butterweck et al 1992; Kesten et al 1993; Rein eking et al 1992; Reineking et al 1994). These authors report that more than 90% of radon decay products are attached to aerosols in the indoor condition, and that the aerosol at tached fraction is cl ose to unity in the working condition. 2.4 Conclusions An accurate assessment of worker risk due to inhalation of TENORM aerosols requires detailed and site-specifi c knowledge of a range of ae rosol properties. In this study, TENORM-containing aerosols in the Florida phosphate industry have been characterized to include the particle size distribution, mass densit y, shape, elemental composition, and radioactivity concentrati on. In addition, a sensitivity study was conducted regarding the variations in the effective dose per unit intake of P238PU and P230PTh aerosol particles with corresponding variations in each aerosol parameter value. Of particular importance is the aerosol size distribution. Aerosol size distributions do not follow a lognormal pattern, where aerosol mass concentration is noted to increase with incr easing particle size. The airborne particle concentration at each plant and location varies widely over 1 to 1.5 orders of magnitude at granulator and shipping areas, and over a factor of a 1000 at storage areas. Mass densities of particles ra nge about 1.6 to 1.7 g cmP-3P. The shapes of particles appear as spheroids or rough spherical fragments. A hi gh fraction of phosphorus is associated with both large (21 100 m) and medium (1.76 3.57 m) -sized airborne particles. For small-sized airborne particles (0.2 0.4 m), sulfur and silicon are found to be dominant over elemental phosphorus. Radi oactivity concentrations of P238PU in products and particles are much higher than those of P226PRa, since P238PU selectively localizes to the

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31 product from phosphate source material while P226PRa tends to localize in by-product materials. Radioactiv ity concentrations of P210PPb in both bulk dry products and settled particles are between those of P238PU and P226PRa. However, its con centration significantly increases in airborne particle s due to attachment of radon de cay products on the particles. The database established in this study can thus be used for radiation exposure assessments to workers and to members of the general public. In addition, the information is useful to phosphate companie s and regulators in th eir consideration of health protection programs and policies re garding both respirat ory and radiological protection.

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32 Table 2-1. ICRP 66 HRTM defau lt aerosol parameters and i nput aerosol parameters for dose sensitivity study Aerosol Parameters ICRP Default Values Input Parameters for Dose Calculation Aerosol size DistributionP aP Density Shape factor Absorption typeP bP AMAD = 5 m (workplace) AMAD = 1 m (general public) log-normal distribution GSDBICRP66B=1+1.5[1-(100AMTDP1.5 P P+1)P-1P]P P = 2.50 for 5 m AMAD = 2.47 for 1 m AMAD 3 g cmP-3P 1.5 type M (P238PU)P P type S (P230PTh) AMAD = 0.01 – 100 m GSD = 1 and GSDBICRP66P B2P 0.7 and 11 g cmP-3P 1 and 2 type M (P238PU) type S (P230PTh)P P Pa PAMTD is Activity Median Thermodynamic Diameter. Pb PIn the lack of information, type M is recommended for uranium and type S is recommended for thorium in ICRP 71 Publication (ICRP 1995b).

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33 Table 2-2. Active operational systems and ventilation conditions during airborne particle sampling at the various storage areas Plant PaP Running System Ventilation A-a Conveyor, Reclaimer Closed A-b (1) None Open A-b (2) Payloader Open B Payloader Open C Conveyor Open D-a Reclaimer Closed D-b Payloader Closed E Payloader Open F Payloader Open PaP Capital alphabets are plant identifiers. Plant A-a and Plant A-b refer to the same company plant (Plant A), but two different lo cations within the plant. A-b (1) and A-b (2) indicate two different samplings at the same location.

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34 Table 2-3. Mass density of bulk products, set tled particles, and airborne particles Type of Sample Number of Samples Mean Density (g cmP-3P) Bulk product 11 1.72 ± 0.08 Settled dust 11 1.71 ± 0.07 Aerosol ( 100 m) 4 1.70 ± 0.06 Aerosol ( 10 m) 6 1.59 ± 0.07

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35 Table 2-4. Composition information for dry pr oduct ingredients (Mosaic 2003a; Mosaic 2003b; PCS 2003a; PCS 2003b) Component (w%) Product Type Chemical Formula N P Other Elements or Compounds MAP NHB4BHB2BPOB4B 10~11 22 DAP (NHB4B)B2BHPOB4B 18 20 Calcium, Magnesium, Iron and Aluminum Sulfates, Phosphates, Silicates, and Fluorides

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36 Table 2-5. Phosphorus, silicon, and sulfur co mposition in different sized particles Component (w%) Plant A (MAP) Plant D (DAP) Area Impactor Stage P Si S P Si S 1PstP 28.3 0.3 1.7 4PthP 21.7 3.0 3.0 Granulator 7PthP 2.9 13.3 8.9 1PstP 17.3 6.0 1.9 19.4 5.5 2.8 4PthP 13.2 2.9 0.4 24.1 0.3 2.2 Storage 7PthP 0.7 10.7 13.0 21.5 0.3 7.4 1PstP 25.9 1.3 1.5 27.3 2.4 4.3 4PthP 26.6 1.5 2.0 12.3 2.5 7.1 Shipping 7PthP 2.5 5.7 19.3 1.7 13.4 21.9

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37 Table 2-6. P238PU, P226PRa, and P210PPb radioactivity concentrations in bulk dry products, settled particles, and airborne particles. Me an and one standard derivation are given for n > 1. Mean Radioactivity Concentration (Bq kgP-1P) Product Type: Sample nP238PU P226PRa P210PPb Central Florida DAP: PaPPublished values 1702 – 3064 22 – 448 Bulk product 4 3337 ± 892 41 ± 22 248 ± 141 Settled particle 3 4151 ± 104730 ± 4 344 ± 118 Airborne particle ( 100 m) 3 2419 ± 629 68 ± 61 596 ± 352 Airborne particle ( 10 m) 4 2575 ± 821 93 ± 56 955 ± 274 MAP: PaPPublished values 1702 – 3064 52 – 799 Bulk product 4 2335 ± 381 56 ± 19 311 ± 163 Settled particle 4 2623 ± 725 44 ± 19 333 ± 263 Airborne particle ( 100 m) 1 1843 95 2313 Airborne particle ( 10 m) 1 2039 141 3204 Northern Florida DAP: PaPPublished values 936 19 Airborne particle ( 100 m) 1 1322 44 639 Airborne particle ( 10 m) 1 1088 30 944 Pa PData from studies by Guimond (1978), Roessler et al . (1979), Owen and Hyder (1980), NCRP (1987), and Birky et al . (1998).

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38 Figure 2-1. Series of processes for phosphate product manufacture. Rectangular and oval elements of the chart represent, resp ectively, the type of process and the phosphate material at each process.

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39 Figure 2-2. P238PU decay series

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40 Figure 2-3 Inhalation dose coefficients for various particle parameters: (A) P238PU and (B) P230PTh. The other parameters follow the ICRP 66 HRTM default particle values except as indicated in the legend.

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41 Figure 2-4. Airborne particle si ze distribution: (A) granulator area, (B) storage area, and (C) shipping area. Capital alphabets in legend are plant identifiers. Plant A-a and Plant A-b refer to the same compa ny plant (Plant A), but two different locations within the plant.

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42 Figure 2-4. Continued

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43 Figure 2-5. SEM analysis of diffe rent sized airborne particle s: (A) Particles collected on the 1PstP impactor stage, (B) 4PthP impactor stage, and (C) 7PthP impactor stage.

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44 Figure 2-6. EDXS count peak for different si zed particles: (A) la rge-sized particles (1PstP impactor stage), (B) medium-sized particles (4PthP impactor stage), and (C) small-sized particles (7PthP impactor stage).

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45 CHAPTER 3 EFFECTIVE DOSE SCALING FACTORS FO R USE WITH CASCADE IMPACTOR SAMPLING DATA IN TENORM INHALATION EXPOSURES 3.1 Introduction One of the most important dose modifying parameters of inhaled particles is particle size distribution. It determines particle inhalabili ty and deposition fractions in sub-regions of the respiratory tract (ICRP 1994). Particle si ze distributions within the workplace and environment can be determined through air sampling via cascade impactors. The cascade impactor is a multistage impaction device used to separate airborne particles according to their aerodynamic size. As shown schematically in Fig. 31, the airborne particle to be sampled is dr awn through a series of progressively narrower jets (hence higher speeds), each followed by an impaction surface or collection plate. The air stream passes through the first je t, flows around the impaction surface which obstructs the flow in the fo rward direction, then through th e next jet and so on through each succeeding stage. When the air str eam curves to flow around the obstructing impaction surface, those particles with exce ss inertia will impact while the others will remain airborne and continue to flow to the ne xt stage. After traveling to the last stage, the stream is drawn through a final collection filter. The aerodynamic cut diameter, defined as the mean particle size below whic h particles may pass to the following stage, can be controlled by changing the air flow rate (John 1999; Marple and Willeke 1976; Pilat 1998).

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46 Once particle properties are characterized, radiation dose due to particle inhalation can be calculated using the ICRP 66 Human Respiratory Tract Model (HRTM). Due to the complexity of the ICRP 66 HRTM, one must generally resort to a computer program to apply the model for a specific exposu re scenario. In 1993 and 1995, the National Radiological Protection Board (NRPB) of Great Britain released its Lung Dose Evaluation Program (LUDEP) so ftware, allowing the use of the ICRP 66 HRTM for dose assessment (Jarvis et al . 1993; Jarvis et al . 1996). Using LUDEP, the dose and dose rate to respiratory tract tissues, as well as othe r systemic organs of the body, are computed following a given intake. In the LUDEP code , values of lung and effective dose are computed readily only in cases where radioa ctivity is (1) distributed log-normally as described by the particle AMAD, or (2) concentrated at a sing le particle size (Geometric standard deviation-GSD = 1). If the aerosol size distributi on is characterized via cascade impactor measurement and found not to follow a log-normal distribution, the dose calculation must be performed via a summation of doses calculated separately using the particle-size ranges for the impactor. For ma terials leaving the respiratory tract, LUDEP employs the ICRP 30 biokinetic models (ICRP 1979). Air sampling with the cascade impactor a llows one to determine the radioactivity concentration for each impactor stage (i.e., particle size range). Generally, the measured data are first fit to a continuous function (e.g., log-normal distribution) and then the function is used directly in th e inhalation dose assessment. Wh en the statistical fit is poor, one must revert to using the measured data directly, and one of three options is thus available: (1) assume that the radioactivity measured on each stage is concentrated at a single particle size (e.g., the medi an particle size of each size interval), (2) assume that

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47 the radioactivity is distributed uniformly acro ss each size interval, or (3) assume that the radioactivity is linearly distributed across each particle-size interval in either an increasing or decreasing pattern. For option 1, the entire particle size distribution is rather unrealistically represen ted by a series of delta functi ons – one for each impactor stage. Option 2 is equally valid from a m easurement viewpoint, yet has the advantage of sampling particle-size variations in the e ffective dose across each size interval. This advantage of option 2 is further enhanced under option 3 through the assignment of a linear activity slope across each si ze interval. As noted earlier, linear distributions are not permitted as input to the LUDEP code, and thus they must be further approximated as a series of N discrete particle sizes acro ss the impactor size range each containing a fraction of the total radioactivity measured for the impactor stage. The type of distribution to be assumed can be guided by the relative change in activity between adjacent impactor stages. While options 2 and 3 thus appear to be more realistic, these approaches require more computational effort. For example, if N = 10 and the cascade impactor had 8 stages, options 2 and 3 woul d require 80 executions of the code, while option 1 would require only 8. In the present study, we present a series of inhalation effective dose scaling factors, SFBEB, defined as the ratio of the effective dos e given under options 2 and 3 (uniform and linear size distribution per stage) to the effective do se given by option 1 (mono-size distribution per stage). Cons equently, these scaling factor s allow one to run the LUDEP code with cascade impactor data using the less computationall y demanding option 1, while then reporting doses under the more realistic descriptio n of options 2 and 3. These scaling factors are given for several radionuclides of the P238PU series, for different stages

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48 (i.e., particle size ranges), and for differen t default solubility types F (fast), M (moderate), and S (slow) – as defined in the ICRP 66 HRTM. 3.2 Materials and Methods 3.2.1 Cutoff Size of the Cascade Impactor In this study, the University of Washingt on Mark III cascade impactor was selected as the representative air sampler, alth ough the proposed methodology can easily be applied to other impactor de vices (Pilat 1998). The Mark III impactor consists of 7 particle impactor stages fo llowed by a final collection filt er as shown in Fig. 3-1. Consequently, the size distribution of the radi oactive aerosol can be partitioned across 8 different particle-s ize ranges. The aerodynamic cut diameter, defining the beginning of each size interval, is a function of various jet featur es and particle properties. For a round nozzle design, the aerodynamic cut diameter at each impactor stag e is given by the following expression (John 1999; Marple and Willeke 1976): 3 50 509 4pnWStk DC Q , Eq. 3-1 where DB50B is the particle diameter at 50% collection efficiency, C is the Cunningham slip correction factor, n is number of the jet nozzles in the sampler, is the fluid viscosity, W is the nozzle diameter, StkB50B is the Stokes number at 50% collection efficiency, BpB is the particle density, and Q is the jet flow rate. The cutoff sizes, and thus the particle size intervals, can be adjusted by c ontrolling the jet flow rate ( Q ). In this study, the cutoff sizes were calculated at a flow rate of 15 L minP-1P (a rough approximation to the ventilation rate of ICRP reference workers) and are listed in Tabl e 3-1. In Eq. 3-1, a value of 0.49 was employed as the square root of StkB50B (John 1999; Rader and Marple

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49 1985; Sethi and John 1993). The upper partic le size limit of the cascade sampler was taken to be 100 m following that for the total susp ended particulate (TSP) high-volume air sampler (EPA 1999). The lower limit of particles collected on the Teflon membrane final collection filter was taken to be 0.03 m, a value typical of those employed in similar studies (Divita et al . 1996; Howell et al . 1998; Marley et al . 2000; Wagner and Leith 2001). 3.2.2 Other Particle Properties Particle deposition in the respiratory tract occurs through several types of mechanisms. They include gravitational settling, initial impa ction, and Brownian diffusion. These mechanisms are governed by two different types of particle sizes, aerodynamic diameter and thermodynamic diamet er. Particle shape and density in addition to particle size are the parameters that relate the two sizes. Reference values of particle properties are given in ICRP Public ation 66. These properties, however, can vary greatly depending on the aerosol type, its ge neration mechanism, and other external conditions. In particular, the aerosol pr operties in the environment depend on the surrounding matrix. Both a partic le density and a shape factor of unity rather than ICRP 66 HRTM reference values were used for dos e calculation in this study. Initially, this study was conducted to calculate dose to work ers in the phosphate i ndustry. The values employed in this study are closer to real prop erties of phosphate particles than the ICRP reference values. In additi on, it holds these para meters constant, thus helping to understand dose change as a f unction of particle size.

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50 3.2.3 Radioactivity Distribution within a Sub-stage Gamma-spectroscopy measurements of aerosol particles collected at each impactor stage yield a single radioactiv ity value (e.g., total activity across the size range assigned to that impactor stage) (see Fig. 3-2A). Consequently, radioactiv ity as a function of particle size across the impactor-stage size rang e is not measured and must be assumed in the inhalation dose assessment. Generally, a mono-size distribution is typically applied as shown in Fig. 3-2B, where the radioactivity of collected particles is assigned to a single representative particle size (e.g., the geometric mean of the upper and lower size limits – see Table 3-1). This discrete appr oach is clearly an approximation, and thus other assumptions may be considered. In Fig. 3-2C, the measured impactor-stage radioactivity is assumed to be uniformly dist ributed across the size interval, while in Fig. 3-2D and 3-2E, the radioactiv ity is assumed to linearly increase or linearly decrease across that same size interval. The LUDEP program does not permit a linear size distribution assignment. Consequently, each linear distribution must be further approximated by multiple sub-stages as demons trated in Fig. 3-2F. In this method, each impactor stage is divided into multiple sub-stages, and the measured radioactivity then follows a mono-size distribution within each substage. Mono-size distributions for each sub-stage particle can be approximated with in the LUDEP code through assignment of a log-normal distribution with a geometric standard deviation of unity. If the number of sub-stages within a given impactor stag e is large enough, a li near radioactivity distribution can be accurately depicted. Ho wever, an increase in the number of substages also increases the computational effo rt per exposure scenario as LUDEP must be run separately for each particle sub-stage.

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51 In the present study, each stage of the Mark III cascade impactor is divided geometrically into 10 sub-stages. For a uni form distribution, one-t enth the activity measured within each impactor stage is as signed to each of the ten sub-stages. The activity ratio (AR), defined as the activity a ssigned to sub-stage 1 divided by the activity assigned to sub-stage 10, is t hus unity for the uniform distri bution. Four other sub-stage distributions were additionally considered: lin early decreasing distri butions with ARs of 1:2 and 1:5 (shallow and steep slopes, respectiv ely), and linearly in creasing distributions with ARs of 2:1 and 5:1 (shallow and steep sl opes, respectively). These particular ratios were selected as representing minimum and maximum slopes seen across the cascade impactor size intervals within a true 1 m or 5 m AMAD log-normal distribution (ICRP 66 defaults). Effective dose scaling factors, SFBEB, can thus be developed as defined by the ratio of the effective dose dete rmined under a uniform or linear distribution to that under a single mono-size distribution for each stage of the Mark III cascade impactor. 3.2.4 TENORM Radionuclides Industrial practices ut ilizing natural resources may c oncentrate naturally occurring radionuclides to a degree that can potentially pose risks to workers or the general public. These past or present human practices t hus result in the exposures to TENORM radionuclides. Radionuclides of the uranium and thorium series are major constituents of TENORM, while the actinium series generally does not contribute to TENORM due to its low abundance in natural source materials. The issues of TENORM are mainly focused on mechanical and chemical concentration mechanisms, deposition locations, and waste streams from industrial processes such as the manufacture of phospha te fertilizers (Birky et al . 1998). In this study, e ffective dose scaling factor s were calculated for the P238PU

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52 decay series due to their rela tive importance in the exposur e to workers and the general public within the Florida phosphate industry (Birky et al . 1998). 3.2.5 Dose Calculation Fifty-year committed equivalent doses and committed effective doses were calculated for occupational wo rkers under light exertion follo wing inhalation of airborne particles containing P238PU decay series using the LUDEP 2.0 code. The worker’s physiological and activity-related exposure para meters were taken as reference values given in the ICRP 66 HRTM (default values in the LUDEP code – see Table 3-2). In addition, ICRP-66 default values for particle transport rates between compartments and absorption rates to blood were also used w ithin the LUDEP 2.0 code. To take advantage of more recent information on radionuclide bi okinetic behavior, revised values of the GI tract absorption factor or fB1B given in the ICRP Publicati on 71 were employed in lieu of those given in ICRP Publication 30 (see Table 3-3). The P238PU decay series is shown in Fig. 2-2. In a given exposure scenario, a worker may be exposed to individual radionuclides in the series or to portions of the decay chain in secular equilibrium. The dose due to inhalation of particulates encompassing a decaychain series can be easily cal culated through summation of the doses delivered by each decay-chain member. Furthermore, pr ogeny generated by the decay of inhaled radionuclides will further contribute to intern al dose; hence, the in-growth of progeny must also be taken into account. For ex ample, inhalation of particles containing equilibrium concentrations of P210PPb and P210PBi will result in organ doses from the inhaled parents, as well as from daughters produced within the body following inhalation (e.g., P210PBi and P210PPo from inhaled P210PPb, as well as P210PPo from inhaled P210PBi). However, it

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53 should be noted that the behavior of a given radionuclide produced during in-vivo ingrowth differs from its biokinetic behavior when present only within th e inhaled aerosol. In ICRP Publication 30, daughter ra dionuclides are assumed to behave metabolically like the parent radionuclide excep t in noted cases for radioiodines or noble gases (ICRP 1979). The main parameter for internal dose calculation the number of decays in a source organ or tissue is eas ily calculated by the relation between the inhaled radionuclide and its prog eny. In the more recent I CRP models, the biokinetic behavior of the decay product is descri bed more fully (ICRP 1989; ICRP 1993; ICRP 1995a; ICRP 1995b). For example, if radioac tive progeny are produced in the tissues of trabecular or cortical bone, they are assu med to follow the behavior of the parent radionuclide during their removal from the bone volume. On the other hand, the decay progeny generated in other orga ns or tissues are assumed to follow their own biokinetic behavior distinct from that of the parent radionuclide. The biokinetic behavior of ga s-phase radionuclides differs from that of solid-phase radionuclides. Inhaled or in-growth P226PRa decays to P222PRn, which is removed from the human body rapidly as it is a non-reactive nob le gas. More recent ICRP biokinetic models assume that radon produced in soft ti ssues and on the bone surfaces is removed to blood at a transfer rate of 100 dayP-1P and that P222PRn entering blood from other compartments is removed from the human body by exhalation at a transfer rate of 1 minP-1P (ICRP 1993). In addition, a removal rate of 100 dayP-1P is assigned for P222PRn produced in the respiratory tract (ICRP 1995b). The P222PRn removal rate is much greater than the P222PRn decay constant of 0.18 dayP-1P. Therefore, it can be assumed that P222PRn generated outside of the skeleton and its progeny contribute very little to internal dose.

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54 Internal doses due to the inha lation of a radionuclide in the P238PU decay series depend in part on the physiological beha vior of its progeny. When a progeny radionuclide has a long physical half-life or is removed from the body rapidly, its contribution to internal dose is small or ne gligible. When the phys ical half-life of a progeny radionuclide is short compared to its parent, most of the generated radionuclide will decay within the same organ or tissue as its parent. Consequently, the number of disintegrations of the proge ny radionuclide in a compartmen t can be assumed to be the same as that of the parent radionuclide. Fo r intermediate cases, the number of decays of a progeny radionuclide is some fraction of those of the parent that depends on the magnitude of the physical half-life and the removal rate. These three decay-chain dose scenarios are represented in the LUDEP 2.0 code as the (1) no-merge, (2) merge, and (3) super-merge decay-chain options, respectively. The biological half-life of the uranium series in an organ or tissue ranges from minut es to years. The decay-chain option should be determined considering the progeny physical half-life and the biological removal halflife within each compartment. The dose cal culation options utili zed in this study are shown in Table 3-4. The no-merge option was selected for (1) P234PU and P230PTh due to the long physical half-lives of th eir initial daughters, (2) for P226PRa due to the rapid removal rate of P222PRn, and (3) for P210PPo due to its decay to stable P206PPb. The merge option was selected for P234PTh and its short-lived daughters P234mPPa and P234PPa. Finally, the super-merge option was applied to P238PU, P210PPb, and P210PBi. 3.3 Results and Discussion 3.3.1 Inhalation Dose Coefficients – Variat ion as a Function of Particle Size Fig. 3-3 displays values of the inhalation dose coefficient to reference workers for the P238PU series as a function of particle size a nd absorption type. A ll values in Fig. 3-3

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55 were calculated using LUDEP 2.0. For absorption type F (dotted lines), P230PTh gives the highest effective dose per unit inta ke (Fig. 3-3D). Radionuclides of P210PPb, P210PPo, P226PRa, P234PU, and P238PU contribute the next highest values of effective dose in that order. The dose coefficients for P234PTh and P210PBi are about 4 orders of magnitude smaller than those for P230PTh for type F materials. For absorption type M materials, P230PTh still contributes the greatest to the effective dose in relation to other radionuclides in the P238PU series. The inhalation dose coefficients for type M P238PU, P234PU, P226PRa, and P210PPo are very close to one another. The dose coefficients for P210PBi and P234PTh are smaller than those of P230PTh for type M materials by 2 and 4 orders of magnitude, respectively. For absorption type S materials, the inhalation dose coefficient of P230PTh approaches that for the other ra dionuclides with the exception of P210PBi and P234PTh. The inhalation dose coefficient depends on the type of radionuclide and its absorption type. The radionuclide physical ha lf-life, its distribu tion to and retention within organs and tissues of the body, a nd its absorption type determine how many nuclear transformations occur w ithin these source tissues. In addition, the type of emitted particle and its energy determines values of the radiation weighting factors (wBRB) and thus its contribution to the effective dose. Hi gh inhalation dose coefficients are seen for P230PTh due its long physical half-life, the fact that it is an alpha-particle emitter, and its high fractional distribution to and reten tion in the skeletal tissues. The P234PTh, however, decays through beta-particle emission (with a lower wBRB) and has a relatively short physical halflife; consequently, its inhalation dose coefficient is small in spite of its high skeletal distribution and retention. The P210PBi is an alpha-particle emitter with a short physical

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56 half-life. Its retention within the kidneys and all other source orga ns is also relatively short. Particle deposition and clearance mechanisms in the respiratory tract are directly related to particle size. Initially, the inha lation dose decreases w ith increasing particle size down to a partic le size of about 0.6 m, and then increases to a maximum value at a particle size of about 2.5 m (see Figs. 3-3A to 3-3H). Af ter reaching this maximum, the effective dose per unit intake decreases on ce again with increasing particle size. Furthermore, differences in inhalation doses between large and small-sized particles also depend upon the absorption type of the inha led radionuclide. For radionuclides of absorption types M or S, the effec tive dose due to the inhalation of 0.1 m particles is about an order of magnitude larger than that due to inhalation of 50 m particles. These differences are less pronounced for absorption type F radionuclides. Fig. 3-4 shows the fractional deposition of inhaled particles in various regions of the human respiratory tract as a function of particle size for reference worker (spherical particles of unit density). The effective dose due to par ticle inhalation depends on the regional particle deposition fraction, although apportionment factors of thoracic target tissues (bronchial, bronchiolar, alveolar-interstitial regions) are equally assigned in ICRP Publication 66 (ICRP 1994). Particle density of 1 g cmP-3P, unity aerodynamic shape factor, and unity geometric standard deviation were applied to standard worker for calculation. ETB1B, ETB2B, BB, bb, and Al represent first extr athoracic region (anterior nose), second extrathoracic region (posteri or nasal passages, larynx, pha rynx, and oral cavity), bronchial region, bronchiolar re gion, and alveolar-i nterstitial region, respectively. The particles in each respiratory tract region contribute to the effective dose to a certain

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57 degree as determined by their regional clea rance rate and tissue radiosensitivity. Consequently, all regional deposition fractions s hould be considered in the analysis of the inhalation dose coefficient as a function of pa rticle size. Examination of the particle deposition fraction as a functi on of particle size in Fig. 34 indicates that the curve roughly reflects the particle size dependence of the inhalation dose coefficient (Figs. 33A to 3-3H). However, particles deposited in the deeper regions of respiratory tract yield higher doses due to their slower clearance rates as compared to particles deposited in the upper regions of the respiratory tract. Fig. 3-4 shows that the highest deposition fraction for small particles occurs within the alveolar-interstitial region followed by the bronchiolar regions of the l ungs. Consequently, the inha lation of smaller particles contributes more to the worker effective dose than inhalation of larger particles in the environment. Larger particles within the inhaled aerosol mainly deposit within the extrathoracic regions. 3.3.2 Effective Dose Scaling Factors for Uniform Distributions For the University of Wash ington Mark III cascade impactor, the radioactivity of airborne particulates may be assessed within eight different partic le size ranges as given in Table 3-1. One may use these data with the LUDEP code for inhalation dose assessment by assigning the radioactivity measur ed per stage at its geometric mean, thus obligating the user to run the dose assessmen t for eight different particle sizes. If, however, a uniform distribution were assume d per particle stage (approximated by 10 sub-stages per stage), a total of 80 LUDEP calculations would be required. Through the application of dose scaling factors, howeve r, one may perform the dose calculation for the mono-size distribution assumption (8 L UDEP calculations), and then scale these

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58 doses to those given by more realistic assump tions of activity versus particle size across each impactor stage. Table 3-5 displays values of th e effective dose scaling factor SFBEB needed to convert the inhalation effective dose calculated fo r a mono-size radioactivity distribution per impactor stage to that calculated for a uni form radioactivity distribution per impactor stage. Uniform activity distributions are s hown to yield slightly higher values of effective dose than values given for mono-size distributions, except for the 4PthP and 5PthP stage particle sizes (1.25 4.54 m) where the effective dose is smaller by ~1-2%. For the vast majority of the stages, dose diffe rences are insignifican t between these two assumed activity distributions. Larger dose differences (scaling f actors between 1.15 and 1.17) are found, however, for end-fi lter particles (0.03 to 0.35 m) for all absorption types (F, M, and S) and fo r all radionuclides in the P238PU series. Furthermore, higher scaling factors are additiona lly seen for particles in the size range of 4.54 to 11.88 m (3PrdP impactor stage) for type M radionuclides (scaling factors of 1.11 to 1.28) and for type S radionuclides (scaling factors of 1.09 to 1.24). To better understand under wh at conditions the effective dose scaling factor given in Table 3-5 may or may not differ from unit y, we plot the inhalation dose coefficient for P238PU as a function of particle size in Figures 3-5A to 3-5C for type F, M, and S materials, respectively. Vertical lines correspond to th e cutoff sizes for each impactor stage of the Mark III sampler. Values of SFBEB approach unity under one of two conditions. First, values of SFBEB approach unity when the inhalation dos e coefficient varies only modestly with particle size across the size interva l; consequently, assigning all measured radioactivity to a single representative particle size, or 1/10th that activity to multiple

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59 particle sizes across the size interval, will not alter the reported effective dose. This condition is seen in the 1PstP impactor stage for P238PU. Second, values of SFBEB approach unity when a generally linear varia tion in the dose coefficient ex ists across the logarithmic impactor size interval (either increasing or decreasing) where highe r-than-average dose contributions from particles on the one end of the interval are compensated for by lowerthan-average dose contributions from par ticles on the opposite end. This condition is seen, for example, in the 6PthP impactor stages for all three absorption classes of P238PU. Neither condition is applicable to the endfilter stage, where the inhalation dose coefficient varies non-linearly across this si ze interval, and sub-stage particles smaller than the geometric mean contribute a higher fr action of the total effective dose than do sub-stage particles larger than the geometric mean (0.1 m in Table 3-1). We further note that a significant non-linear dose coefficient pattern is also evident across the 3PrdP impactor stage (dashed circles in Fig. 3-5), but only for type M and t ype S materials. For type F uranium, the dose coefficient in Fig. 3-5A is shown to vary relatively linearly across this particular si ze interval (4.54 to 11.88 m), and thus a dose compensating effect occurs for type F particles. Values of SFBEB are unity for all st age 3 radionuclides for type F materials in Table 3-5, yet they range from 1.21 – 1.28 for type M and from 1.09 – 1.24 for type S radionuclides. 3.3.3 Effective Dose Scaling Fact ors for Linear Distributions In cases where radioactivity measurements for three consecutive impactor stages (e.g., stages 2, 3, and 4) suggest that radioactiv ity in the central impactor stage (e.g., stage 3) decreases linearly across its particle-size ra nge, values of SFBEB may be applied as given in Tables 3-6 or 3-7. Values in Table 36 are for shallow sub-stage activity gradients

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60 (AR of 2:1), while those in Table 3-7 are fo r steep sub-stage activ ity gradients (AR of 5:1). Significant ( 10%) corrections to the effective dose are noted to occur in both Tables 3-6 and 3-7 for genera lly the same combinations of impactor stage and absorption type as given in Table 3-5. For the smallest particle-size range in the cascade impactor (end-filter particles of 0.03 to 0.35 m), values of SFBEB are 1.28 – 1.30 and 1.41 – 1.43 for sub-stage activity ratios of 2:1 and 5:1, respectively. Adjustments to 3PrdP stage contributions of effective dose are also requi red for both type M and type S radionuclides, with the highest value of SFBEB noted to be 1.56 for type M P210PBi (see Table 3-7). Finally, it is noted that for a sub-stage activity ratio of 5:1, 10% adjustments to the effective dose are also required for 4PthP stage particles for select ra dionuclides of types M and S. When cascade impactor measurements indi cate that a linearly increasing activitysize distribution should be assumed for a given impactor stage, values of SFBEB may be applied as given in either Ta ble 3-8 (AR of 1:2) or in Ta ble 3-9 (AR of 1:5). For the former case, only six combinations of radi onuclide, absorption type, and impactor stage indicate an adjustment to the effective dose exceeding ±10%. The maximum value of SFBEB in Table 3-8 is only 1.13 (type M P210PBi for 3PrdP stage particles), a nd corrections to the effective dose for end-stage particles are now shown to be only ~3-4%. When the assumed sub-stage activity ratio is increas ed to 1:5 (see Table 3-9), values of SFBEB are 0.9 for end-stage particles (all absorptions types and radionuclides), 0.85 – 0.91 for the 4PthP stage particles of type M ra dionuclides, and 0.86 – 0.91 for the 4PthP stage particles of type S radionuclides. In contrast, upward adjust ments of 11% are required to the effective dose for stage 6 particles for type F radionuclides (see Table 3-9).

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61 In Fig. 3-6, we plot the relative change in SFBEB from that determined for a uniform distribution (Table 3-5), to th at for a linearly increasing or linearly decreasing sub-stage activity distribution. Activity ratios as hi gh as 20:1 (linear decrease) and 1:20 (linear increase) are considered for type F, type M, and type S particles of P238PU. For type F uranium (Fig. 3-6A), significant changes in SFBEB (> ±25%) are noted for end-filter particles at ARs exceeding 5:1 and 1:5. For t ype M and S uranium (Figs. 3-7B and 3-7C), significant changes in SFBEB (> ±25%) are noted for both end-filter particles and for stage 3 particles. Data from Fig. 3-6 can thus be used to adjust values of SFBEB assembled for uniform sub-stage activity distribu tions to those applicable to linear distributions of either increasing or decreasing activity across the impactor stage size interval. 3.3.4 Applications of Effective Dose Scalin g Factor to Measured Cascade Impactor Data The application of effective dose scaling factors given in Tables 3-5 to 3-9 is best demonstrated through an example problem. Ta ble 3-10 displays a series of simulated cascade impactor measurements of P238PU activity (type S) as collect ed at a flow rate of 15 L minP-1P. The measured activity values at each impactor stage are given in column 2 of Table 3-10, and are plotted in Fig. 3-7. The inhalation dose assessment can be performed by considering the measured activity per impactor stage to be distributed across each size interval as either: (1) a m ono-size distribution, (2) a uniform distribution, or (3) a linear distribution of variable slope . When a mono-size distribution is assumed, the effective dose is obtained by summation of the effectiv e doses contributed by each particle size. These contributions are in turn given as the product of the impactor -stage activity and the LUDEP inhalation dose coefficient assessed at the geometric-mean particle size (see Table 3-1). In the second option, one may assume a uniform distribution of activity per

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62 impactor stage, thus treating the measuremen ts in a histogram fashion (dashed lines in Fig. 3-7). Here, effec tive dose scaling factors SFBEB from Table 3-5 are additionally applied, and the dose estimate is revised upward from 1.30 mSv (mono-size distribution) to 1.33 mSv (uniform distribution). The incr ease is driven primarily by a 12% and 16% enhancement of the effective dose contributions by stage 3 and stage F particles. Application of the third option requires assignment of activity ratios for each impactor stage. The activity ratio for impact or stages N = 2 to 7 may be approximated as follows: 112NN NN NAA A A AR , Eq. 3-3 where ABNB, ABN-1B, and ABN+1B, are the measured activities of th e N, (N-1), and (N+1) stages of the cascade impactor series. For stages 1 a nd F, a uniform distribution is assumed for particles larger and smaller than their geometric mean, respectively (as no other information is available). Using Eq. 3-3, revised scaling factors are assigned either directly from the data of Tables 3-5 to 3-9, or via linear interpolat ion of tabular data. This approach thus yields a revised dose estimate of 1.40 mSv, an increase driven primarily by values of SFBEB of 1.23 and 1.29 for particles in stages 3 and F. Application of Eq. 3-3 to the measured activity within stages 3, 4, and 5 results in an average AR of 2.9:1 for stage 4 particles. It can be argued that perhaps a more justified approach is to assume the peak activity within stage 4 is uniformly distributed (AR of 1:1 with a SFBEB of 0.95). If this change is implemented, the dose estimate is revised to 1.38 mSv, and is slightly closer to that estimated under opti on 1 (1.30 mSv) or op tion 2 (1.33 mSv).

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63 Values of SFBEB given in this study can be direc tly applied in the inhalation dose assessment provided that the cutoff sizes for each stage are reasonably close to those given here for the Mark III impactor operated at a flow rate of 15 L minP-1P. If the Mark III cascade impactor is operated at a different flow rate or other types of impactors are employed for particle sampling, the data in Fig. 3-3 can be used to readjust scaling factors using revised particle-siz e ranges for each impactor stage. 3.4 Conclusions Air sampling with multi-stage cascade impactors enables one to assess airborne radioactivity as a function of particle size, significantly enhancing the accuracy of the dose assessment. For each stage of the im pactor, a single measurement of activity by radionuclide is determined. While the standard assumption is made that this radioactivity can be collapsed at a single representative particle size within the stage’s particle size range, a more realistic assumption would be that the measured activity is linearly distributed across that size range with either a zero, positive, or nega tive slope – the exact choice determined by changes in measured ac tivity across neighboring impactor stages. The concept of an effective dose scaling factor, SFBEB, is introduced whereby (1) the former approach can be used (which requires less computational effort using the LUDEP code), and (2) the resulting values of effective dose per impactor stage can then be rescaled to values appropriate to a linear radi oactivity distribution per stage. For the Mark III cascade impactor operated at a sampling rate of 15 L minP-1P, significant corrections (greater than 10%) are necessary within only 1 or 2 of the particle size ranges. For material ra pidly dissolved within the l ung fluids (type F absorption), contributions to the effective dose from the smallest particles sampled at the endcollection filter must be scaled upward or downward by factors of 1.17, 1.30, 1.43, 1.04,

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64 or 0.91 to approximate effective doses under li near distributions w ith sub-stage activity ratios of 1:1, 2:1, 5:1, 1:2, and 1:5, respect ively. In this size interval (~0.03 to 0.35 m), the dose coefficient varies non-linearly with particle size, and thus the assumption of a mono-size radioactivity di stribution is generally not appropriate. For P238PU series materials of absorption type M or S, greater than 10% corrections must be made to effective dose contributions from end filter particles (~0.03 to 0.35 m), as well as from 3rd stage particles (~4.5 to 12 m) or 4th stage particles (~2.2 to 4.5 m). The largest values of SFBEB were noted for linearly decreasing radi oactivity distributi ons of high slope (activity ratio 5:1): 1.41 to 1.44 for end-filter st age particles (all solubility types) and 1.24 to 1.56 for 3rd stage particles (types M and S) . The smallest values of SFBEB were noted for linearly increasing distributions of high slope (activity rati o 1:5): 0.90 to 0.91 for endfilter stage particles (all solubil ity types) and 0.9 to 0.85 for 4PthP stage particles (types M and S). When a uniform activity distributi on is to be applied, corrections to dose estimates for other than 3PrdP or end-filter stage particles ar e not necessary as values of the effective dose between the two approaches differ only by ± 2-4%.

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65 Table 3-1. Particle size information for each impactor stage of the University of Washington Mark III cascade impactor. The values were calculated for air flow rate of 15 L minP-1P. Impactor Stage DB50P B P( m) Particle Size Range ( m) Geometric Mean ( m) 1 27.21 27.21 100.00 52.16 2 11.88 11.88 27.21 17.98 3 4.54 4.54 11.88 7.34 4 2.24 2.24 4.54 3.19 5 1.25 1.25 2.24 1.67 6 0.66 0.66 1.25 0.91 7 0.35 0.35 0.66 0.48 End filter 0.03 0.35 0.10

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66 Table 3-2. Reference physiological paramete r values for reference worker (ICRP 1994). Parameters Values Physiological parameters Total lung capacity Functional residual capacity Vital capacity Dead space Height Weight Activity related parameters Light exercise Ventilation rate Respiration frequency Tidal volume Rest or Sitting Ventilation rate Respiration frequency Tidal volume 6.98 L 3.30 L 5.02 L 0.146 L 176 cm 73 kg 31.3 % 1.5 mP3P hP-1 20 minP-1 1.25 L 68.8 % 0.54 mP3P hP-1 12 minP-1 0.75 L

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67 Table 3-3. GI tract absorption factors (fB1B) for uranium series elements. Employed fB1B ValuePaP fB1 BValue in ICRP Publication 30 Element Type F Type M Type S Class D Class W Class Y Uranium 0.02 0.02 0.002 0.05 0.05 0.002 Thorium 0.0005 0.0005 0.0005 0.0002 0.0002 Radium 0.2 0.1 0.01 0.2 Lead 0.2 0.1 0.001 0.2 Polonium 0.1 0.1 0.01 0.1 0.1 Bismuth 0.05 0.05 0.01 0.05 0.05 Pa PFor dose calculation, revised fB1B values given in ICRP P ublication 71 were employed (ICRP 1995b). The given values are for the adult model.

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68 Table 3-4. LUDEP decay-chain option for each radionuclide Radionuclide Decay-chain Option Radionuclide Decay-chain Option P238PU Super-merge P226PRa No-merge P234PTh Merge P210PPb Super-merge P234PU No-merge P210PBi Super-merge P230PTh No-merge P210PPo No-merge

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69 Table 3-5. Inhalation effec tive dose scaling factors ( SFBEB) for a uniform activity distribution Effective Dose Scaling Factor SFBE P Ba (Uniform radioactivity dist ribution per impactor stage)P P Absorption Type Impactor Stage P238PU P234PTh P234PU P230PTh P226PRa P210PPb P210PBi P210PPo F 1 1.01 1.01 1.01 1.01 1.01 1.01 1.01 1.01 2 1.01 1.02 1.02 1.02 1.02 1.02 1.02 1.02 3 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 4 0.98 0.99 0.99 0.98 0.99 1.00 0.99 0.99 5 0.98 0.99 0.98 0.98 0.99 0.99 0.99 0.98 6 1.02 1.02 1.02 1.02 1.02 1.02 1.02 1.02 7 1.05 1.04 1.04 1.04 1.04 1.04 1.04 1.04 F 1.17 1.17 1.17 1.17 1.17 1.17 1.17 1.17 M 1 1.01 1.01 1.01 1.01 1.01 1.01 1.01 1.01 2 1.02 1.03 1.03 1.02 1.02 1.02 1.03 1.03 3 1.21 1.12 1.22 1.11 1.21 1.11 1.28 1.22 4 1.03 0.99 1.01 0.99 1.00 0.99 0.99 0.99 5 0.99 0.99 0.99 0.98 0.99 0.98 0.99 0.99 6 1.03 1.02 1.03 1.01 1.03 1.01 1.04 1.04 7 1.02 1.04 1.03 1.03 1.03 1.03 1.03 1.03 F 1.15 1.16 1.15 1.17 1.15 1.16 1.15 1.15 S 1 1.01 1.01 1.01 1.01 1.01 1.01 1.01 1.01 2 1.02 1.02 1.02 1.02 1.02 1.02 1.03 1.03 3 1.12 1.10 1.12 1.09 1.12 1.09 1.22 1.24 4 0.95 0.99 0.93 0.99 0.93 0.99 1.01 1.00 5 0.98 0.99 0.98 0.98 0.98 1.09 0.99 0.99 6 1.01 1.02 1.01 1.01 1.01 1.00 1.03 1.03 7 1.03 1.04 1.03 1.03 1.03 1.03 1.03 1.03 F 1.16 1.16 1.16 1.17 1.16 1.17 1.15 1.15 PaP Defined as the ratio of the committed effective dose for a uniform radioactivity distribution per impactor stage to that fo r a mono-size radioactiv ity distribution per impactor stage. Values of SFBEB given here are specific to particle size ranges of the University of Washington Mark III cascade im pactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).

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70 Table 3-6. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 2:1) Effective Dose Scaling Factor SFBEP B a (Linearly decreasing sub-st age distribution with ARP Pof 2:1)PbP Absorption Type Impactor Stage P238PU P234PTh P234PU P230PTh P226PRa P210PPb P210PBi P210PPo F 1 1.02 1.02 1.02 1.02 1.02 1.02 1.02 1.02 2 1.04 1.04 1.04 1.04 1.04 1.04 1.04 1.04 3 1.03 1.02 1.03 1.03 1.03 1.03 1.02 1.03 4 0.99 0.99 0.99 0.99 0.99 1.00 0.99 0.99 5 0.97 0.96 0.97 0.96 0.96 0.96 0.96 0.97 6 0.97 0.97 0.97 0.97 0.97 0.97 0.97 0.97 7 1.02 1.01 1.01 1.01 1.01 1.01 1.01 1.01 F 1.30 1.30 1.30 1.30 1.30 1.30 1.30 1.30 M 1 1.02 1.02 1.02 1.02 1.02 1.02 1.02 1.02 2 1.04 1.05 1.05 1.05 1.05 1.05 1.06 1.06 3 1.32 1.19 1.33 1.19 1.32 1.18 1.42 1.34 4 1.09 1.03 1.08 1.03 1.07 1.03 1.06 1.06 5 0.97 0.97 0.96 0.97 0.96 0.97 0.96 0.96 6 1.00 0.98 1.00 0.98 1.00 0.98 1.00 1.00 7 1.03 1.02 1.03 1.02 1.03 1.02 1.04 1.04 F 1.28 1.29 1.28 1.30 1.28 1.29 1.28 1.28 S 1 1.02 1.02 1.02 1.02 1.02 1.02 1.02 1.02 2 1.04 1.05 1.04 1.05 1.04 1.04 1.06 1.06 3 1.20 1.18 1.20 1.16 1.20 1.16 1.34 1.37 4 0.98 1.04 0.96 1.04 0.96 1.02 1.07 1.07 5 0.97 0.97 0.97 0.97 0.97 1.07 0.96 0.96 6 0.98 0.98 0.98 0.98 0.98 0.98 1.00 1.00 7 1.02 1.02 1.02 1.02 1.02 1.02 1.03 1.04 F 1.29 1.29 1.29 1.30 1.29 1.30 1.28 1.28 Pa PDefined as the ratio of the committed effective dose for a linearly decreasing (AR = 2:1) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). PbP B BRatio of the activity of the 1PstP sub-stage to the activity of the 10PthP (last) sub-stage.

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71 Table 3-7. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 5:1) Effective Dose Scaling Factor SFBEP B a (Linearly decreasing sub-st age distribution with ARP Pof 5:1)PbP Absorption Type Impactor Stage P238PU P234PTh P234PU P230PTh P226PRa P210PPb P210PBi P210PPo F 1 1.02 1.02 1.02 1.02 1.02 1.02 1.02 1.02 2 1.06 1.06 1.07 1.07 1.07 1.07 1.06 1.07 3 1.06 1.04 1.06 1.06 1.05 1.05 1.04 1.05 4 1.00 1.00 1.00 1.00 1.00 1.01 1.00 1.00 5 0.95 0.94 0.95 0.95 0.94 0.94 0.94 0.95 6 0.93 0.93 0.93 0.93 0.93 0.93 0.93 0.93 7 1.00 0.98 0.99 0.99 0.99 0.99 0.98 0.99 F 1.43 1.43 1.43 1.43 1.43 1.43 1.43 1.43 M 1 1.03 1.02 1.03 1.02 1.03 1.03 1.03 1.03 2 1.07 1.08 1.08 1.07 1.08 1.08 1.09 1.09 3 1.44 1.27 1.45 1.26 1.43 1.26 1.56 1.47 4 1.16 1.08 1.14 1.06 1.13 1.07 1.13 1.13 5 0.94 0.95 0.94 0.96 0.94 0.95 0.94 0.93 6 0.97 0.94 0.96 0.94 0.96 0.95 0.96 0.96 7 1.04 1.01 1.04 1.01 1.04 1.01 1.04 1.04 F 1.41 1.42 1.41 1.43 1.41 1.43 1.40 1.40 S 1 1.02 1.03 1.02 1.02 1.02 1.02 1.03 1.03 2 1.07 1.07 1.07 1.07 1.07 1.06 1.09 1.09 3 1.28 1.26 1.29 1.24 1.29 1.23 1.45 1.50 4 1.02 1.09 0.99 1.10 0.99 1.06 1.13 1.13 5 0.96 0.95 0.96 0.96 0.96 1.04 0.94 0.94 6 0.95 0.94 0.95 0.95 0.95 0.95 0.96 0.96 7 1.02 1.01 1.02 1.02 1.02 1.01 1.04 1.04 F 1.43 1.42 1.42 1.43 1.42 1.44 1.41 1.41 PaP B BDefined as the ratio of the committed effective dose for a linearly decreasing (AR = 5:1) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). PbP B BRatio of the activity of the 1PstP sub-stage to the activity of the 10PthP (last) sub-stage.

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72 Table 3-8. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:2) Effective Dose Scaling Factor SFBEP B a (Linearly increasing sub-st age distribution with ARP Pof 1:2)PbP Absorption Type Impactor Stage P238PU P234PTh P234PU P230PTh P226PRa P210PPb P210PBi P210PPo F 1 1.01 1.01 1.01 1.01 1.01 1.01 1.01 1.01 2 0.99 0.99 0.99 0.99 0.99 0.99 0.99 0.99 3 0.98 0.97 0.98 0.98 0.98 0.98 0.97 0.98 4 0.98 0.98 0.98 0.97 0.98 0.99 0.98 0.98 5 1.00 1.01 1.00 1.00 1.01 1.01 1.01 1.00 6 1.06 1.07 1.06 1.06 1.06 1.06 1.07 1.06 7 1.07 1.07 1.06 1.06 1.06 1.07 1.07 1.06 F 1.04 1.04 1.04 1.04 1.04 1.04 1.04 1.04 M 1 1.01 1.01 1.01 1.01 1.01 1.01 1.01 1.01 2 0.99 1.00 1.00 1.00 1.00 1.00 1.00 1.00 3 1.09 1.04 1.10 1.03 1.09 1.03 1.13 1.09 4 0.97 0.94 0.95 0.95 0.94 0.94 0.92 0.92 5 1.01 1.01 1.01 0.99 1.01 1.00 1.02 1.02 6 1.06 1.06 1.07 1.04 1.07 1.05 1.07 1.07 7 1.01 1.05 1.02 1.04 1.02 1.04 1.02 1.02 F 1.03 1.03 1.02 1.04 1.02 1.03 1.02 1.02 S 1 1.01 1.01 1.01 1.01 1.01 1.01 1.01 1.01 2 0.99 0.99 0.99 1.00 0.99 0.99 1.00 1.00 3 1.04 1.02 1.04 1.01 1.04 1.02 1.10 1.11 4 0.91 0.94 0.90 0.93 0.90 0.95 0.95 0.93 5 0.99 1.00 0.99 0.99 0.99 1.12 1.01 1.02 6 1.04 1.06 1.04 1.04 1.04 1.03 1.07 1.07 7 1.03 1.05 1.03 1.04 1.03 1.04 1.02 1.02 F 1.03 1.03 1.03 1.03 1.03 1.04 1.02 1.02 PaP B BDefined as the ratio of the committed effective dose for a linearly increasing (AR = 1:2) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). PbP B BRatio of the activity of the 1PstP sub-stage to the activity of the 10PthP (last) sub-stage.

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73 Table 3-9. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:5) Effective Dose Scaling Factor SFBEP B a (Linearly increasing sub-st age distribution with ARP Pof 1:5)PbP Absorption Type Impactor Stage P238PU P234PTh P234PU P230PTh P226PRa P210PPb P210PBi P210PPo F 1 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 2 0.97 0.97 0.97 0.97 0.97 0.97 0.97 0.97 3 0.95 0.95 0.95 0.95 0.95 0.95 0.95 0.95 4 0.97 0.97 0.97 0.96 0.97 0.99 0.97 0.97 5 1.02 1.03 1.02 1.02 1.03 1.03 1.03 1.02 6 1.11 1.11 1.11 1.11 1.11 1.11 1.11 1.11 7 1.10 1.10 1.08 1.09 1.09 1.09 1.10 1.09 F 0.91 0.91 0.91 0.91 0.91 0.91 0.91 0.91 M 1 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 2 0.97 0.97 0.97 0.97 0.97 0.97 0.97 0.97 3 0.98 0.96 0.98 0.95 0.98 0.95 0.99 0.97 4 0.90 0.89 0.89 0.91 0.87 0.90 0.85 0.85 5 1.04 1.02 1.04 1.00 1.04 1.01 1.05 1.05 6 1.10 1.10 1.10 1.07 1.10 1.08 1.11 1.11 7 1.00 1.06 1.01 1.06 1.01 1.05 1.01 1.01 F 0.90 0.90 0.90 0.90 0.90 0.90 0.90 0.90 S 1 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 2 0.97 0.97 0.97 0.97 0.97 0.97 0.97 0.97 3 0.96 0.94 0.96 0.94 0.96 0.96 0.98 0.98 4 0.88 0.89 0.86 0.87 0.86 0.91 0.89 0.87 5 1.00 1.02 1.00 1.00 1.01 1.14 1.04 1.04 6 1.07 1.10 1.07 1.07 1.07 1.06 1.10 1.11 7 1.03 1.06 1.04 1.04 1.04 1.05 1.02 1.01 F 0.90 0.90 0.90 0.90 0.90 0.90 0.90 0.90 PaP B BDefined as the ratio of the committed effective dose for a linearly increasing (AR = 1:5) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). Pb PRatio of the activity of the 1st sub-stage to the activity of the 10th (last) sub-stage.

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74 Table 3-10. Example inhalation dose asse ssment under the assumption of mono-size, uniform, or linearly changing radioactivit y distribution per impactor stage. Dose was calculated for type S P238PU. Mono-Size Distribution Uniform Sub-Stage Distribution Impac -tor Stage Measured Particle Activity (Bq) Inhalation Dose Coefficient (mSv BqP-1P) Dose (mSv) SFBEB Source Dose (mSv) 1 7.5 1.04 x 10P-3P 7.79 x 10P-3P 1.01 Table 3-4 7.87 x 10P-3P 2 7.5 1.33 x 10P-3P 9.97 x 10P-3P 1.02 Table 3-4 1.02 x 10P-2P 3 10 2.50 x 10P-3P 2.50 x 10P-2P 1.12 Table 3-4 2.80 x 10P-2P 4 50 8.54 x 10P-3P 4.27 x 10P-1P 0.95 Table 3-4 4.05 x 10P-1P 5 40 9.04 x 10P-3P 3.62 x 10P-1P 0.98 Table 3-4 3.54 x 10P-1P 6 10 6.21 x 10P-3P 6.21 x 10P-2P 1.01 Table 3-4 6.27 x 10P-2P 7 10 4.50 x 10P-3P 4.50 x 10P-2P 1.03 Table 3-4 4.63 x 10P-2P F 30 1.20 x 10P-2P 3.59 x 10P-1P 1.16 Table 3-4 4.16 x 10P-1P Total 1.30 1.33 Variable Linear Sub-Stage Distribution (Linear at Peak) AR (Average) Revised SFBEB Source Dose (mSv) 1 7.5 1.04 x 10P-3P 1 : 1 1.01 Table 3-5 7.87 x 10P-3P 2 7.5 1.33 x 10P-3P 1.2 : 1 1.024 Tables 3-5, 3-6 1.02 x 10P-2P 3 10 2.50 x 10P-3P 3.2 : 1 1.232 Tables 3-6, 3-7 3.08 x 10P-2P 4 50 8.54 x 10P-3P 2.9 : 1 0.992 Tables 3-6, 3-7 4.23 x 10P-1P 5 40 9.04 x 10P-3P 1 : 1.9 0.989 Tables 3-5, 3-8 3.58 x 10P-1P 6 10 6.21 x 10P-3P 1 : 1.6 1.028 Table 3-5, 3-8 6.38 x 10P-2P 7 10 4.50 x 10P-3P 2 : 1 1.02 Table 3-6 4.59 x 10P-2P F 30 1.20 x 10P-2P 2 : 1 1.29 Table 3-6 4.62 x 10P-1P Total 1.40 Variable Linear Sub-Stage Distribution (Uniform at Peak) AR (Average) Revised SFBEB Source Dose (mSv) 1 7.5 1.04 x 10P-3P 1 : 1 1.01 Table 3-5 7.87 x 10P-3P 2 7.5 1.33 x 10P-3P 1.2 : 1 1.024 Tables 3-5, 3-6 1.02 x 10P-2P 3 10 2.50 x 10P-3P 3.2 : 1 1.232 Tables 3-6, 3-7 3.08 x 10P-2P 4 50 8.54 x 10P-3P 1 : 1 0.95 Table 3-5 4.05 x 10P-1P 5 40 9.04 x 10P-3P 1 : 1.9 0.989 Tables 3-5, 3-8 3.58 x 10P-1P 6 10 6.21 x 10P-3P 1 : 1.6 1.028 Table 3-5, 3-8 6.38 x 10P-2P 7 10 4.50 x 10P-3P 2 : 1 1.02 Table 3-6 4.59 x 10P-2P F 30 1.20 x 10P-2P 2 : 1 1.29 Table 3-6 4.62 x 10P-1P Total 1.38

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75 Figure 3-1. Schematic diagram of the Un iversity of Washin gton Mark III cascade impactor. Solid lines and dashed lines indicate air stream cu rve and trajectory of impacted particles, respectively.

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76 Figure 3-2. Radioactivity dist ribution as a function of aer odynamic particle size. (A) Radioactivity measurements of partic les collected by a cascade impactor (dashed circle focuses on one of the impactor stages used to sample this distribution). Four possible assumpti ons can then be made regarding the distribution of measured activity across the impactor stage size-interval: (B) mono-size distribution, (C) uniform di stribution, (D) linearly decreasing distribution, and (E) linear ly increasing distribution for each impactor stage. (F) The latter distributions may be approximated as a series of 10 sub-stages each containing a different fraction of the total activity measured at that impactor stage.

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77 Figure 3-3. Inhalation do se coefficients for P238PU decay series: (A) P238PU, (B) P234PTh, (C) P234PU, (D) P230PTh, (E) P226PRa, (F) P210PPb, (G) P210PBi, and (H) P210PPo. Dotted, dashed, and solid lines represent ab sorption types F, M and S, respectively. 50-year committed effective doses for the standard worker were calculated using LUDEP 2.0 software. A particle density of 1 g cmP-3P, an aerodynamic shape factor of 1.0, and a geometric standa rd deviation of 1.0 were assumed for calculation.

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78 Figure 3-3. Continued

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79 Figure 3-3. Continued

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80 Figure 3-3. Continued

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81 Figure 3-4. Particle deposition fraction on each sub-region of the respiratory tract

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82 Figure 3-5. Inhalation dose coefficient curve of P238PU as a function of particle size: (A) type F, (B) type M, (C) type S. Vert ical lines represent upper and lower size limits of each impactor stage.

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83 Figure 3-5. Continued

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84 Figure 3-6. Effective dos e scaling factor of P238PU as a function of activity ratio: (A) type F, (B) type M, and (C) t ype S. The value of SFBEB is given as a difference from the value of uniform radioactivity distribution.

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85 TFigure 3-6.ContinuedT

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86 Figure 3-7. Application of radi oactivity distribution using ra dioactivity measurement data of cascade impactor samples. Dots, da shed lines, and solid lines indicate mono-size radioactivity dist ribution, uniform radioac tivity distribution, and linear radioactivity distribution of variable slopes, respectively.

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87 CHAPTER 4 EFFECTIVE DOSE SCALING FACTORS FOR USE WITH CASCADE IMPACTOR SAMPLING DATA IN EXPOSURES TO URANIUM SERIES (APPLICATION TO IMBA PROGRAM) 4.1 Introduction Effective dose scaling factors, SFBEB, for use with cascade impactor sampling data in TENORM inhalation exposures are presented in Chapter 3. If the particle size distribution is characterized by air sampling with a cascade im pactor, the measured data are directly used in the i nhalation dose assessment. Sampling with cascade impactor yields one value of mass or radioactivity for each impactor stage. Therefore, the size distribution within each imp actor stage is unknown. Three options were suggested for the use of measured data in dose assessment: (1) mono-size radioactiv ity distribution, (2) uniform radioactivity distribution, and (3) li nearly increasing or d ecreasing radioactivity distribution. The effective dose scaling factor is defined as th e ratio of the effective dose given under uniform and linear radioactivity size distributi on per cascade impactor stage to the effective dose given by mono-size radi oactivity distribution per stage. These scaling factors demand less computational e ffort and yield results corresponding to the more realistic description of options 2 and 3. The previous study used the Lung Dose Ev aluation Program (LUDEP) software to generate the scaling factors. LUDEP software (Jarvis et al . 1993; Jarvis et al . 1996) implements the most recent ICRP 66 Hu man Respiratory Tract Model (HRTM). However, it employs the older biokinetic mode l as given in ICRP Publication 30 (ICRP 1979). New radionuclide biokinetic models have been develo ped in ICRP Publications

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88 56, 67, 69, and 71 since the release of ICRP Publication 30 (ICRP 1979; ICRP 1980; ICRP 1981; ICRP 1989; ICRP 1993; ICRP 1995a; ICRP 1995b). In 2003, ACJ & Associates Inc. and the National Radiologi cal Protection Board (NRPB) developed and released Integrated Modules for Bioassay Analysis (IMBA) Pr ofessional by making a compilation of IMBA Expert USDOE Edition and CANDU Edition (James et al . 2003; James et al . 2004a; James et al . 2004b). The IMBA program enables the calculation of particle inhalation dose using the most re cent ICRP 66 HRTM and the latest ICRP biokinetic models (Strom 2003). The availability of the IMBA program de mands a supplement or revision of the scaling factors introduced by Kim et al (Kim et al . 2005). The present study presents inhalation dose coefficients as a function of particle size and a series of inhalation effective dose scaling factors. The scaling factors are give n for several radionuclides of the P238PU series, for different stages, and for differe nt absorption types. In addition, dose results from IMBA and LUDEP are compared from the aspect of biokinetic models employed in each program. 4.2 Materials and Methods The University of Washington Mark III cascade impactor was selected as a representative air sampler. It consists of 7 impactor stages and one final collection filter thus partitioning particles in to 8 different size ranges (Pilat 1998). The aerodynamic cutoff sizes of each stage are 27.21, 11.88, 4.54, 2.24, 1.25, 0.66, and 0.35 m from the 1PstP to the 7PthP impactor stage at an operating flow rate of 15 L minP-1P. The upper and lower particle size limits were taken to be 100 and 0.03 m, which are typical values employed in similar studies (Divita et al . 1996; EPA 1999; Howell et al . 1998; Marley et al . 2000;

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89 Wagner and Leith 2001). A shape factor and a mass density of unity were employed to match the aerodynamic diameter and the ther modynamic diameter for simplicity. The mono-size radioactivity distribution was described by assigning each stage radioactivity to a sing le particle size, which is the geometric mean of upper and lower size limits of the impactor stage. Uniform and linear radioactivity distributions were described by dividing each impactor stage ge ometrically into 10 multiple sub-stages, and then assigning fractional radioac tivity to the sub-stages. IMBA Professional (Versi on 3.0) includes 75 radionuclides while the LUDEP program includes all radionuclides in ICRP Publication 38 or Oak Ridge National Lab (ORNL) database (ICRP 1983; James et al . 2003). Among the P238PU series radionuclides, P238PU, P234PU, P230PTh, P226PRa, and P210PPo are available in IMBA softwa re. Therefore, effective dose scaling factors were calculated for t hose radionuclides while the previous study (Kim et al ., 2005) included P234PTh, P210PPb, and P210PBi as well. The biokinetic model of bismuth has not been revised since ICRP Pub lication 30. Therefore, its effective dose scaling factors generated by LUDEP software can be used for dose assessment. 4.3 Results and Discussion 4.3.1 Inhalation Dose Coefficients Fig. 4-1 shows fifty-year committed effective doses for occupational workers under light exertion following exposure to radioactive aerosols of P238PU series as a function of particle size and absorption type. For uraniu m, radium, and polonium, types S, M, and F yield the highest doses in that or der for particles smaller than ~7 m. The trend is opposite to that for P230PTh, where type F yields higher valu es than type M and S materials. The radionuclides P238PU, P234PU, P230PTh, P226PRa, and P210PPo are alpha-emitters. The alphaparticle energy of these radionuclides represen ts more than 99% of the released energy

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90 during radioactive decay. For most organs, dosimetry models assume that all alpha particle energy is absorbed in the organ wh ere the particle is released (ICRP 1979). Therefore, the dose due to inhalation of P238PU series is strongly dependent on their source organs and number of decays, which are dete rmined by particle deposition, clearance, and radionuclide behavior afte r absorption into blood as give n by their biokinetic model. Type S radionuclides are dissolved into bl ood slowly in comparison to type F and M radionuclides. They stay in the respiratory tract longer thus resulting in a higher dose to the lungs. If the radioactive material after ab sorption into blood is excreted rapidly from body, the doses to other organs are much less than the dose to the l ungs. Consequently, a high fraction of the total effectiv e dose results from the lung dose . This case is applied to the radionuclides of P238PU, P234PU, P226PRa, and P210PPo. If a high fraction of absorbed radioactive material is retain ed in a certain organs or tissues for a long time, longer residence in the respiratory tract does not n ecessarily result in a higher effective dose. Particles deposited in the respiratory tract are cleared by several mechanisms: (1) transport to the gastrointes tinal (GI) tract, (2 ) transport to regional lymph nodes (LN), and (3) absorption into blood (ICRP 1994). Thes e mechanisms are competitive. Type S material results in higher factional removal to the GI tract thus reducing the absorption fraction to blood, retention fr action in other organs, and e ffective dose. The biokinetic model introduced in ICRP Publication 67 shows that thorium c ontent in bone after injection is ~50% in 1 day, increases to ~70% in 2000 days, and then decreases with time (ICRP 1993). Most of thorium goes to bone and is retained for a long time after absorption. Therefore, faster dissolution of P230PTh reduces the removal through the GI tract and results in a higher effective dose.

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91 4.3.2 Comparison of Inhalation Dose Co efficients from IMBA and LUDEP Inhalation dose coefficients calculated us ing the LUDEP software are also plotted to compare the results from IMBA software in Fig 4-1. The IMBA software yields a higher dose in comparison to the LUDEP software for the same dose calculation scenarios. The difference in effective dose between the two programs is generally small, less than 2% except P230PTh (3-11%) for type S materials a nd increases up to by a factor of 5 for P238PU and P234PU, 1.3 for P230PTh, 1.9 for P226PRa, and 1.2 for P210PPo for type F. These dose differences come from the different biokine tic models employed in each program. The difference is highlighted below for P238PU. The longer the alpha-emitting radionuclide resides in the respiratory tract, the more th e lung dose contributes to the effective dose. Tissue-weighted equivalent doses to lung and extra-thoracic (ET) regions are more than 98% of the effective dose for type S P238PU. Therefore, the different biokinetic models in the two programs do not cause large differences in effective dose. If the radionuclide is absorbed quickly into blood, the dose to lungs is small and thus the other organs contribute more to the effective dose. Do se contributions of lung and ET regions are below 8% of the effective dose for type F P238PU. For this case, the biokinetic behavior after absorption into blood play s an important role in the e ffective dose. Fig 4-2 shows tissue-weighted equivalent doses to several organs and effect ive doses calculated by both the IMBA and LUDEP codes. The IMBA software, in comparison to the LUDEP software, yields a higher dose to red bone ma rrow, bone surface and liver by a factor of 2.7, 1.7, and 106, respectively. It yields a dose that is ~27 times higher to most soft organs including breast, stomach, gonads, skin , thyroid, urinary bladder and esophagus. For lungs, the dose is 27 times highe r for large-sized particles (~ 50 m) and 10 times

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92 higher for small size particles (~ 0.1 m). Bone is the main location of uranium retention and thus red bone marrow and the bone surface receive a higher dose than seen in other organs. Fig 4-2 indicates that a high fraction of effective dos es comes from the doses to these two tissues. The main difference of these two codes is the fractional dose contribution of other organs or tissues. For type F P238PU, about 90% of the effective dose comes from those tissues in the LUDEP soft ware. However, the fraction decreases to 39% in the IMBA program. It means that P238PU retention in other or gans and tissues is much higher in the ICRP 69 biokinetic m odel than given by the ICRP 30 biokinetic model. Figs. 4-3 and 4-4 illustrate the uranium biokinetic models in ICRP 30 and 69 (ICRP 1979; ICRP 1995a). The old biokinetic mode l in ICRP Publication 30 is based on retention instead of systemic physiology, a nd it cannot explain uranium behavior at the early time after absorption to blood. This model assumes that uranium entering the transfer compartment goes to bone with fracti ons of 0.2 and 0.023 and remains there with biological half-lives of 20 and 5000 days. Fractions of 0.12 and 0.00052 go to the kidney with biological half-lives of 6 and 1500 days. Fractions of 0.12 and 0.00052 go to all other tissues with biol ogical half-lives of 5 and 1500 da ys. The remaining fraction is excreted with a biological half-life of 6 hour s in blood. In contrast, the new biokinetic model of uranium is a recycling model. Uran ium absorbed to blood is transferred to bone, kidney, liver, other soft tissues, and excret ion routes. The deposited uranium in each tissue can also re-enter the blood stream. Afte r that, it is transfe rred to tissues or is excreted according to the transf er rates. Because bone is th e main site of deposition and retention, it is divided in detail into b one surface, exchangeable bone volume, and non-

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93 exchangeable bone volume for each cortical a nd trabecular bone. Fig. 4-5 shows the model prediction of uranium contents in bone, kidneys, liver, and the other soft tissues as a function of time after inject ion into blood. Th e retention fractions based on the new biokinetic model are higher than those based on the old biokinetic model for all tissues. The difference is great for soft tissues and liver . Liver is included in other tissues in the old biokinetic model, while it is described separately from othe r soft tissues in the latest model. The equivalent dose to each tissu e is proportional to the number of radionuclide decays in the tissue. Therefore, the number of uranium decays for 50 years in soft tissues was calculated. The decay number difference between the two biokine tic models is the same as the dose difference in soft tissues. 4.3.3 Effective Dose Scaling Factors Tables 4-1 4-5 show effective dose s caling factors for uniform and linearly changing radioactivity distributi ons. The selection of radio activity distribution type and application of effective dose s caling factors are given in Chapter 3. For a majority of radionuclides, particle size ranges, and absorpti on types, the dose differences in these two approaches are less than 10%, and thus no correc tions in effective dose per particle stage are needed. Significant corrections (greater th an 10%) are necessary within only 1 or 2 of the particle size ranges for most cases. For type F filter-stage particles, the effective dose of the mono-size distribution should be scal ed upward or downward by factors of 1.17, 1.30, 1.43, 1.04, and 0.91 to approximate ef fective doses under the linear radioactivity distribution with sub-stage activity ratios of 1:1, 2:1, 5:1, 1:2, and 1:5, respectively. For type M and S radionuclides, greater than 10% corrections must be made for the 3PrdP and filter stage particles, as well as some 4PthP stage particles for uniform and linearly decreasing radioactivity distri butions. The scaling factors range 1.04 1.24 for the 3PrdP

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94 stage particles and 1.15 1.17 for filter st age particles under the uniform radioactivity distribution and increase 1.16 1.49 (3PrdP stage) and 1.41 1.44 (filter stage) under the linearly decreasing distribution with activity ratio of 5:1. Large dose differences are irregularly found under the linea rly increasing distribution. The correction to the effective dose for end-stage particles is onl y ~3 4% for the activity ratio of 1:2 and scales downward by 9 10% for the ratio of 1:5. The effective dose scaling factors genera ted using IMBA software is nearly identical to those calculated using the LUDEP program. The difference between them is less than 1% for most cases. The effective dos e scaling factor is calculated by comparing the dose at the mean particle size of a stage range and doses at the mean particle size of each sub-stage. The relative dose difference be tween two different particle sizes depends on particle deposition, clearance, and biokinetic models, collectively. However, particle deposition and clearance contribute to the rela tive effective dose difference between two different particle sizes more than biokinet ic model. Since IMBA and LUDEP employ the same particle deposition and clearance models in ICRP 66 HRTM, the values of scaling factors of these two codes ar e close to each other. Howe ver, these two programs employ different biokinetic models thus yieldi ng dose differences up to a factor of 5. 4.4 Conclusions Effective dose scaling factors enable one to improve the accuracy of dose assessment with less computational efforts. The inhalation dose coefficients and dose scaling factors were calculated for P238PU series using the recen t ICRP 66 implementing software, IMBA. The results are compar ed with LUDEP software outputs. Different biokinetic models used in th e IMBA and LUDEP c odes result in dose differences. The IMBA software yields a higher dose in comparison to the LUDEP

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95 software for the same dose calculation scenario . The difference is the largest for type F materials because type F radionuclides yield a relatively small dose to lungs and a high dose to the other organs or tis sues that depend on the biokin etic model. The effective dose difference between these two programs is a factor of 5 for P238PU and P234PU, 1.3 for P230PTh, 1.9 for P226PRa, and 1.2 for P210PPo for type F and less than 2% for type S. The scaling factors generated using the IMBA program are nearly identical to those calculated by the LUDEP program because par ticle deposition and clearance contribute to the relative effective dose difference betw een different particle sizes more than the biokinetic model and both programs use the same particle deposition and clearance models. Significant corrections of the effec tive dose are necessary for only 1 or 2 of the particle size ranges for most cases. They are mainly found in the 3PrdP, 4PthP and filter stages. The effective dose should be scaled upw ard by factors of 1.0 – 1.24 for the 3PrdP stage and 1.15 – 1.17 for the filter stage to approximate the effective doses for uniform distribution. For linearly decreasing distribution, the values range 1.03 – 1.36 (act ivity ratio of 2:1) and 1.05 – 1.49 (activity ra tio of 5:1) for the 3PrdP stage and 1.28 – 1.30 (activity ratio of 2:1) and 1.41 – 1.43 (activity ratio of 5:1) fo r the filter stage. Large dose differences are irregularly found for the linearl y increasing distribution. Th e effective dose for the endstage particles need to be corrected by only ~3 4% for activity ratio of 1:2 and scaled downward by 9-10% for th e ratio of 1:5.

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96 Table 4-1. Inhalation effec tive dose scaling factors ( SFBEB) for a uniform activity distribution Effective Dose Scaling Factor SFBE P Ba (Uniform radioactivity dist ribution per impactor stage)P P Absorption Type Impactor Stage P238PU P234PU P230PTh P226PRa P210PPo F 1 1.01 1.01 1.01 1.01 1.01 2 1.02 1.02 1.02 1.02 1.02 3 1.00 1.00 1.00 1.00 1.00 4 0.98 0.99 0.98 0.98 0.99 5 0.98 0.98 0.99 0.99 0.98 6 1.02 1.02 1.02 1.02 1.02 7 1.04 1.04 1.04 1.04 1.04 F 1.17 1.17 1.17 1.17 1.17 M 1 1.01 1.01 1.01 1.01 1.01 2 1.02 1.03 1.02 1.02 1.03 3 1.19 1.20 1.11 1.19 1.20 4 1.02 1.01 0.99 1.00 0.99 5 0.99 0.99 0.98 0.99 0.99 6 1.03 1.03 1.01 1.03 1.03 7 1.02 1.03 1.03 1.03 1.03 F 1.16 1.16 1.17 1.16 1.15 S 1 1.01 1.01 1.01 1.01 1.01 2 1.02 1.02 1.02 1.02 1.03 3 1.12 1.13 1.04 1.12 1.24 4 0.95 0.92 0.98 0.93 1.00 5 0.98 0.98 0.98 0.98 0.99 6 1.01 1.01 1.01 1.01 1.03 7 1.03 1.03 1.03 1.03 1.03 F 1.17 1.16 1.17 1.17 1.15 PaP Defined as the ratio of the committed effective dose for a uniform radioactivity distribution per impactor stage to that fo r a mono-size radioactiv ity distribution per impactor stage. Values of SFBEB given here are specific to particle size ranges of the University of Washington Mark III cascade im pactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).

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97 Table 4-2. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 2:1) Effective Dose Scaling Factor SFBEP B a (Linearly decreasing sub-st age distribution with ARP Pof 2:1)PbP Absorption Type Impactor Stage P238PU P234PU P230PTh P226PRa P210PPo F 1 1.02 1.02 1.02 1.02 1.02 2 1.04 1.04 1.04 1.04 1.04 3 1.03 1.03 1.03 1.02 1.03 4 0.99 1.00 0.99 0.99 0.99 5 0.97 0.97 0.97 0.97 0.96 6 0.97 0.97 0.97 0.97 0.97 7 1.02 1.01 1.02 1.02 1.01 F 1.30 1.30 1.30 1.30 1.30 M 1 1.02 1.02 1.02 1.02 1.02 2 1.05 1.05 1.05 1.05 1.06 3 1.30 1.31 1.19 1.30 1.31 4 1.07 1.07 1.02 1.06 1.06 5 0.97 0.96 0.97 0.96 0.96 6 1.00 1.00 0.98 1.00 1.00 7 1.03 1.03 1.02 1.03 1.04 F 1.28 1.28 1.30 1.28 1.28 S 1 1.02 1.01 1.02 1.02 1.02 2 1.04 1.04 1.05 1.05 1.06 3 1.20 1.21 1.10 1.20 1.36 4 0.98 0.96 1.04 0.96 1.07 5 0.97 0.97 0.97 0.97 0.96 6 0.98 0.98 0.98 0.98 1.00 7 1.02 1.03 1.02 1.03 1.04 F 1.30 1.30 1.30 1.30 1.28 Pa PDefined as the ratio of the committed effective dose for a linearly decreasing (AR = 2:1) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). PbP B BRatio of the activity of the 1PstP sub-stage to the activity of the 10PthP (last) sub-stage.

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98 Table 4-3. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly decreasing activity distribution (AR of 5:1) Effective Dose Scaling Factor SFBEP B a (Linearly decreasing sub-st age distribution with ARP Pof 2:1)PbP Absorption Type Impactor Stage P238PU P234PU P230PTh P226PRa P210PPo F 1 1.02 1.03 1.02 1.02 1.02 2 1.07 1.07 1.07 1.07 1.07 3 1.06 1.06 1.06 1.05 1.05 4 1.00 1.00 1.00 1.00 1.00 5 0.95 0.95 0.95 0.95 0.94 6 0.93 0.93 0.93 0.93 0.93 7 0.99 0.99 0.99 0.99 0.99 F 1.43 1.43 1.43 1.43 1.43 M 1 1.02 1.03 1.02 1.03 1.03 2 1.08 1.08 1.07 1.08 1.09 3 1.41 1.43 1.26 1.41 1.43 4 1.13 1.13 1.06 1.13 1.12 5 0.94 0.94 0.96 0.94 0.94 6 0.97 0.96 0.94 0.97 0.96 7 1.04 1.04 1.01 1.04 1.05 F 1.41 1.41 1.44 1.41 1.41 S 1 1.02 1.02 1.02 1.03 1.03 2 1.07 1.07 1.07 1.07 1.09 3 1.28 1.29 1.16 1.29 1.49 4 1.02 0.99 1.10 0.99 1.13 5 0.96 0.96 0.97 0.96 0.94 6 0.95 0.95 0.95 0.95 0.96 7 1.02 1.02 1.02 1.02 1.04 F 1.43 1.43 1.43 1.43 1.41 Pa PDefined as the ratio of the committed effective dose for a linearly decreasing (AR = 5:1) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). PbP B BRatio of the activity of the 1PstP sub-stage to the activity of the 10PthP (last) sub-stage.

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99 Table 4-4. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:2) Effective Dose Scaling Factor SFBEP B a (Linearly increasing sub-st age distribution with ARP Pof 1:2)PbP Absorption Type Impactor Stage P238PU P234PU P230PTh P226PRa P210PPo F 1 1.01 1.01 1.01 1.01 1.00 2 0.99 0.99 0.99 0.99 0.99 3 0.98 0.98 0.97 0.97 0.98 4 0.97 0.98 0.97 0.98 0.98 5 1.00 1.00 1.01 1.01 1.00 6 1.06 1.06 1.06 1.06 1.06 7 1.06 1.06 1.06 1.07 1.06 F 1.04 1.04 1.04 1.04 1.04 M 1 1.00 1.01 1.01 1.01 1.01 2 1.00 1.00 0.99 0.99 1.00 3 1.08 1.09 1.03 1.08 1.08 4 0.96 0.95 0.95 0.94 0.92 5 1.01 1.01 0.99 1.01 1.02 6 1.06 1.07 1.04 1.07 1.07 7 1.02 1.02 1.05 1.02 1.02 F 1.03 1.03 1.04 1.03 1.03 S 1 1.01 1.00 1.00 1.01 1.01 2 1.00 0.99 1.00 1.00 1.00 3 1.04 1.05 0.97 1.04 1.11 4 0.91 0.89 0.93 0.90 0.94 5 0.99 0.99 0.99 0.99 1.02 6 1.04 1.04 1.04 1.04 1.07 7 1.03 1.03 1.04 1.03 1.02 F 1.04 1.03 1.04 1.03 1.03 Pa PDefined as the ratio of the committed effective dose for a linearly increasing (AR = 1:2) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). PbP B BRatio of the activity of the 1PstP sub-stage to the activity of the 10PthP (last) sub-stage.

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100 Table 4-5. Inhalation effec tive dose scaling factors ( SFBEB) for a linearly increasing activity distribution (AR of 1:5) Effective Dose Scaling Factor SFBEP B a (Linearly increasing sub-st age distribution with ARP Pof 1:5)PbP Absorption Type Impactor Stage P238PU P234PU P230PTh P226PRa P210PPo F 1 1.00 1.00 1.00 1.00 1.00 2 0.97 0.97 0.97 0.97 0.97 3 0.95 0.95 0.95 0.95 0.95 4 0.97 0.97 0.97 0.97 0.97 5 1.02 1.02 1.03 1.03 1.02 6 1.11 1.11 1.11 1.11 1.11 7 1.09 1.09 1.09 1.10 1.09 F 0.91 0.91 0.91 0.91 0.91 M 1 1.00 1.00 1.00 1.00 1.00 2 0.97 0.97 0.97 0.97 0.97 3 0.98 0.98 0.95 0.97 0.97 4 0.90 0.89 0.91 0.88 0.86 5 1.03 1.04 1.00 1.04 1.05 6 1.10 1.10 1.07 1.10 1.11 7 1.01 1.01 1.06 1.01 1.01 F 0.90 0.90 0.91 0.90 0.90 S 1 1.00 1.00 1.00 1.00 1.00 2 0.97 0.97 0.97 0.97 0.97 3 0.96 0.96 0.91 0.96 0.98 4 0.88 0.86 0.87 0.86 0.87 5 1.00 1.00 1.00 1.00 1.04 6 1.07 1.07 1.07 1.07 1.11 7 1.03 1.04 1.04 1.04 1.01 F 0.90 0.90 0.90 0.90 0.90 Pa PDefined as the ratio of the committed effective dose for a linearly increasing (AR = 1:5) radioactivity distribut ion per impactor stage to that for a mono-size radioactivity distribution per impactor stage. Values of SFBEB given here are speci fic to particle size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P. Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose). PbP B BRatio of the activity of the 1PstP sub-stage to the activity of the 10PthP (last) sub-stage.

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101 Figure 4-1. Inhalation dose co efficients for light exertion per unit intake of a uranium series radionuclide: (A) P238PU, (B) P234PU, (C) P230PTh, (D) P226PRa, and (E) P210PPo. Solid and dashed lines indicate result s from IMBA and LUDEP softwares, respectively.

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102 Figure 4-1. Continued

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103 Figure 4-1. Continued

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104 Figure 4-2. Effective dose a nd tissue-weighted e quivalent doses to bone marrow, bone surface, lung, and liver of type F P238PU as a function of par ticle size. Thick and thin lines indicate results from IMBA and LUDEP programs, respectively.

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105 Figure 4-3. Biokinetic model of uranium in ICRP Publication 30 (ICRP 1979). The number to each tissue compartment is the distribution fraction from the transfer compartment. The number to ex cretion is the biol ogical half-life for each tissue.

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106 Figure 4-4. Biokinetic model of uranium in ICRP Publication 69 (ICRP 1995a). The numbers between two tissues are the transfer rates (dP-1P) from one tissue to the other.

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107 Figure 4-5. Biokinetic model pr ediction of uranium contents in bone, kidneys, liver, and other soft tissues as a function of time after injection into blood. Solid lines are the prediction from new biokinetic model in ICRP Publication 69 and dashed lines from the old biokinetic model in ICRP Publication 30 (ICRP 1979; ICRP 1995a).

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108 CHAPTER 5 INFLUENCE OF PARTICLE SIZE DISTRIBUTION ON INHALATION DOSES TO WORKERS IN THE FLORIDA PHOSPHATE INDUSTRY 5.1 Introduction Risk assessments to the workers in the phosphate industry have been carried out by calculating radiation doses due to external exposure, radon exposure, and particle inhalation (Birky et al . 1998; Gafvert et al . 2001; Johnson and Traub 1996; Khater et al . 2001; Lipsztein et al . 2001). Lipsztein et al. (2001) calculated radiati on exposure to the workers in various phosphate mines in whic h the effective doses due to particle inhalation were calculated via personal air sampler data and dose coefficients given by the ICRP (ICRP 2002). Gafvert et al. (2001) estimated doses to workers at phosphate rock facilities where the effective dose due to dust inhalation wa s estimated through the product of the radioactivity mass concentration (Bq gP-1P), dust concentration (g mP-3P), reference breathing rate (m3 hP-1P), exposure time (hr yP-1P), and inhalation dose coefficient (Sv BqP-1P) given in ICRP Publication 71 (ICRP 1995b). In this st udy, radioactivity concentrations were obtained from raw mate rials, products, and waste, as opposed to those of sampled airborne particles. Recently, a comprehensive integrated study on worker exposures from TENORM in the Florida phosphate indus try was conducted by Birky et al . (1998). External and internal exposures were assessed for workers at mine and chemical plants. The study results showed that particle inhalation wa s a main contributor to worker radiation exposure. Air sampling and inhalation dose conversion factor s given in ICRP

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109 Publication 68 were used for dose calcula tion (ICRP 1994b). The study emphasized “no adjustment for particle size” was made, and that “the inhalation doses are conservative and tend to be greate r than actual doses.” The objective of the present st udy was to reassess values of the effective dose due to TENORM aerosol inhalation across the Flor ida phosphate industry using more refined analysis of sampled aerosols in the worker areas including gr anulator facilities, storage facilities, and shipping areas. Detailed meas urements in Chapter 2 were conducted of the particle size distribution, particle shape a nd density, and radioac tivity concentrations using cascade impactor, dichotomous sampler, and high-volume sampler measurements in duplicate at all thr ee work areas across 6 facilities in the northern and centrals regions of the state. These data were then used to asse ss the inhalation component of the annual effective dose to workers using the ICRP Publication 66 human respiratory tract model (HRTM) (ICRP 1994b) as coded within the LUDEP and IMBA programs. This study further reviewed the inhalation dose sensitiv ity to assumptions on radionuclide-specific lung solubility in TENORM aerosols of the Florida phosphate industry. 5.2 Materials and Methods Particle properties characterized in Chapte r 1 were integrated into a full internal dosimetry assessment of indi vidualized worker doses by both facility and operational location. Radionuclides of concern in this study include those of the P238PU series as shown in Fig. 2-2. In a given exposure scenar io, a worker may be exposed to individual radionuclides in the P238PU series, or to segments of the decay chain that remain under secular equilibrium. The dose due to inhala tion of particles encompassing a decay-chain series can be easily calculated through su mmation of the doses received by each decaychain member. Furthermore, progeny genera ted by the decay of inhaled radionuclides

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110 will further contribute to internal dose; hence, in-vivo in-growth of progeny must also be taken into account. It is thus useful to generate data rega rding inhalation effective doses from each radionuclide independently as they can be modified according to their unique particle properties and radio activity concentrations. Inhala tion dose coefficients for the uranium series were calculated and then dose rates and total effective doses to workers in Florida phosphate processing plants were calculated by plants and locations. 5.2.1 Inhalation Dose Coefficients Fifty-year committed equivalent doses and committed effective doses were calculated for occupational workers under light exerti on following annual chronic inhalation exposures of ra dioactive aerosols of the P238PU series. The biokinetic behavior of gas-phase radionuclides differs from th at of solid-phase radionuclides. P222PRn produced by decay of P226PRa inside the human body is removed to blood and then exhaled quickly (ICRP 1993; ICRP 1995b). Since the P222PRn removal rate is much greater than P222PRn physical decay, it can be assumed that P222PRn and its short-lived progeny originating from inhaled P226PRa contribute very little to the wo rker internal dose and can thus be disregarded in this analysis. Inhalation dose coefficients were thus calculated for the following radionuclides: P238PU, P234PTh, P234PU, P230PTh, P226PRa, P210PPb, P210PBi, and P210PPo. Both the IMBA (James et al . 2003) and LUDEP (Jarvis et al . 1996) programs were used to generate inhalation dose coefficients. While both codes incorporate the full ICRP Publication 66 human respiratory tract mode l, systemic localization of radionuclides entering the blood are handled in LUDEP using the biokinetic models of ICRP Publication 30 (ICRP 1979), while IMBA inco rporates the more recent radionuclide biokinetic models of ICRP Publications 56, 67, 69, and 71 (ICRP 1989; ICRP 1993; ICRP 1995a; ICRP 1995b). A limitation of the IMBA program, however, is the number

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111 of radionuclides in its current library. The IMBA Professi onal (Version 3.0) includes 75 radionuclides, while the LUDEP program include s all radionuclides in either the ICRP Publication 38 (ICRP 1983) or Oak Ridge National Lab (ORNL) database (Eckerman et al . 1994). In this study, inhalation dose coe fficients were generated using the IMBA program when possible. LUDEP was then used for the radionuclides not currently covered in IMBA program: P234PTh, P210PPb, and P210PBi. In the ICRP 66 HRTM, the radioactive aeros ol is assumed to be distributed lognormally with the aerodynamic particle size. The aerosol size distributions in the Florida phosphate industry are generally not found to follow log-normal distributions. Hence, the dose calculation was performed via summation of doses calculated separately using the particle size ranges for the Mark III cascade impactor: s s E E and TTs sHH , Eq. 5-1 where E and HBTB are, respectively, effective dose a nd equivalent dose for tissue or organ T , s is impactor stage (total of 8 for Mark III cascade impactor), EBsB and HBTsB are doses due to inhalation of the particle size range within impactor stage s . On the basis of ICRP 66 HRTM, radionuc lide absorption to blood can be defined explicitly if site-specifi c particle solubility in the lung fluid is known. In cases where it is not, the ICRP 66 model establishe s three general classifications of particle solubility in the lung fluids (i.e., absorp tion types): type F (fast ab sorption), type M (moderate absorption), and type S (slow absorption). Th e absorption types of radionuclides are well summarized in ICRP Publication 71 (ICRP 1995b) , and are classified by the radionuclide chemical form. In the absence of site -specific information, default type M is recommended for most elements to avoid either overor under-estimates of worker dose.

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112 As an exception, type S is recommende d for radioisotopes of thorium. The physicochemical forms of P238PU series radionuclides in Flor ida phosphate aerosols are not known at present, and thus a sensitivity study was employed in which all three default absorption types were assumed (types F, M, and S). 5.2.2 Inhalation Effective Doses to Workers In this study, site-specific inhalation doses were calculated using unique values for the particle size distribution, particle sh ape, particle density , and radioactivity concentration in granulator, storage, and shipping areas for all 6 Florida phosphate facilities. All databases of particle size di stribution and radionuclid e concentration used for dose calculation are attached in Appendix A. Radioactivity concentrations of P238PU, P226PRa, and P210PPb were measured independently for sampled airborne particles and settled par ticles in each facility. Various assumptions were thus applied regardi ng other radionuclides of the P238PU series. The pathway of a given radionuclide in the P238PU decay series during chemical processing of phosphate materials depends on its elemen tal type. For example, both P238PU and P234PU will selectively localize in product materials rath er than in by-products of th e chemical process. NCRP Report No. 65 indicates that both P238PU and P230PTh concentrations in fertilizer materials made from Florida phosphate ore are esse ntially identical (N CRP 1987). Many other studies also indicated that thorium isotopes follow product rather than by-product materials (Guimond and Windham 1975; Hu rst and Arnold 1982; Makweba and Holm 1993). In this study, it was thus assumed that the radi oactivity concentration of P234PU, P234PTh, and P230PTh are equivalent to that measured for P238PU. For both P210PBi and P210PPo, additional considerations were necessary as there are two sources of the parent radionuclide P210PPb within airborne particles. The first source is P210PPb that exists within

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113 the phosphoric acid or bulk pr oduct and is carried into the airborne particulates during airborne particle generation. The second source is P210PPb from the aeroso l attachment of ambient P222PRn progeny in the worker environment. Two options thus exist for dose assessment: (1) assign the P210PBi and P210PPo radioactivity per unit ma ss to that measured for P210PPb in the aerosol samples, or (2) assign the P210PBi and P210PPo radioactivity per unit mass to that measured for P210PPb in settled partic les. In this study, the second option was applied as aerosol suspension times are very sh ort in comparison to the physical half-life of P210PPb, and thus ambient radon-generated P210PPb radioactivity attached to airborne particles does not have sufficient time to yield appreciable in-growth of additional P210PBi and P210PPo progeny. The total effective dose was calculated for all air sampling sites at all 6 Florida phosphate chemical plants. An occupancy f actor of 2000 hours per year was assumed for the dose calculation. The total effective dose is the summation of both the internal committed effective dose and external doses from gamma-ray exposures. In a previous TENORM study, Birky et al . (1998) indicated that workers in some job classifications were noted to receive additional external gamma-ray exposures from TENORM radionuclides. Annual external dose rate of 0.2 mSv yP-1P was used for dose estimation. The total effective dose was further calcula ted under two different absorption type scenarios: (1) that all radionuclides have the same lung solubility ab sorption type (either type F, M, or S) and (2) that the radionuclid es are found in aerosols of the most or least conservative absorption types as given by thei r individual inhalation dose coefficients. For each worker location and facility, the radi oactivity as a function of particle size was constructed based on (1) the radionuclid e concentrations seen via gamma-ray

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114 spectroscopy, (2) the particle size distributi on seen via cascade impactor sampling, and (3) effective dose scaling factors. 5.3 Results and Discussion 5.3.1 Inhalation Dose Coefficients Fig. 5-1 displays inhalation dose coefficien ts for light-activity workers applicable to the P238PU decay series as a functi on of particle size and abso rption type. The values were calculated with particle properties for the Florida phosphate chemical plants. For most P238PU decay-series radionuclides, dose coeffici ents are highest for type S aerosols and lowest for type F aerosols at particle sizes below ~8 m (~1 m for P210PPb). For particles of larger diameters, the dose coe fficients for some type F aerosols are larger than those of type M aerosols (P238PU, P234PU, and P210PPb) or of both type S and type M aerosols (P234PTh and P210PPo). Unlike all other radionuc lides, the dose coefficients for P230PTh are noted to be highest for type F aerosols and lowest for type S aerosols (see below). Inspired particles deposit in the human re spiratory tract durin g both inhalation and exhalation. The deposition component of the ICRP 66 HRTM divides the respiratory tract into four anatomical regions: extrathor acic (ET), bronchial ( BB), bronchiolar (bb), and alveolar-interstitial (Al) regions. The regional deposi tion of inhaled particles is determined primarily by the particle size di stribution. After radioactive particles are deposited on the surfaces of l ung airways, they experience seve ral fates: (1) transport to the gastrointestinal (GI) tract via the pha rynx by mucociliary action, (2) transport to regional lymph nodes (LN) via lymphatic channe ls, and (3) absorption into blood. These mechanisms are competitive, and they occur in various degrees simultaneously depending on the nature of the inhaled particles (particle si ze and shape, numbers or mass of deposited particles, intrinsi c chemical solubility, and cyto toxicity) (Guilm ette 1998).

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115 After absorption of radioactive materials in to blood, they are taken up by systemic tissues, transferred from one ti ssue to another, circulated or retained in the body, and excreted via urine depending on their physicochemical proper ties. The distribution and retention of various radionuclides are described in ICRP Pub lications 30, 56, 67, 69, and 71. Particle deposition, clearance, biokinetic behavior, and radioactive decay all influence the dose due to particle inhalation. If it is assumed that all energy from radioactive materials is absorbed in the tissu e where they reside (applicable to alphaand beta-emitting radionuclides as in the P238PU series), the tissue equivalent dose is proportional to the number of radionuclide tran sformations in each tissue. Consequently, the equivalent dose to the lungs is highest if the radionuclide remains within the respiratory tract for periods longer than seen at other systemic orga ns. Higher lung doses are thus seen for radionuclides with (1) highe r fractional deposition in the deeper regions of the respiratory tract, (2) sl ow absorption to blo od, (3) rapid excretion after absorption to blood, and (4) relatively shor t physical half-lives as comp ared to the respiratory tract residence time. The effective dose is the summation of the tissue-weighted equivalent dose to each tissue, and thus the distribution of a radioactive materi al to each tissue or organ is a major factor determining the overall magnitude of the effective dose. Obviously, if a radioactive mate rial is selectively distributed in tissues or organs with high tissue weighting factors, the final effec tive dose will also be high. By definition, radionuclides of type S aerosols remain in th e respiratory tract for long periods of time, and thus they give higher equi valent doses to the lungs than do radionuclides of type F and type M aerosols. Tissue-weighted equiva lent doses to the lung compose ~99% of the

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116 effective dose for radionuclides of type S aerosols except for thoriu m radioisotopes (88% for P234PTh and 62% for P230PTh) at a particle size of ~1 m. In contrast, the lung contribution to the effective dose decreases to less than 7% for radionuclides of type F aerosols. To better understand tissue -specific contributions to the effective dose by radionuclide and absorption type, Fig. 5-2 disp lays the tissue-weighted equivalent doses for P238PU and P230PTh as a function of particle size and absorption type. For P238PU of type S (Fig. 5-2A) or type M (Fig. 52B) aerosols, the lung dose is by far the highest contributor to effective dose at small and medium-sized particles (< 10 m). Weighted equivalent doses to red bone marrow (RBM) and bone surface cells (BS) are the 2PndP and the 3PrdP highest contributors, respectivel y. However, their contribution to the effective dose is negligible compared to that of the weighted equivalent lung dose. When the particle size increases to over 2 to 5 m, however, the lung dose decr eases rapidly with increasing particle size. Deposition within the extrat horacic airways increases with concomitant decreases in radioactiv ity deposition to the deeper airw ays of the respiratory tract. Equivalent doses to the extrathoracic target tissues (ETB1B, ETB2B and LNBETB) thus contribute to the weighted equivalent dose to remai nder tissues at larger particle sizes for P238PU associated with type M and type S aerosols . The discontinuous ri se in the remainder tissue contribution is triggere d by the “ICRP 60 splitting rule” as described in footnote 3 of Table 5-2 in ICRP Publication 60 (ICRP 1991). For P238PU in type F aerosols, the equivalent doses to systemic organs and tissues contribute substantially more to the total e ffective dose than does the lung equivalent dose. Radionuclides of type F aerosols are quickly absorbed to blood within the respiratory tract. Therefore, th e effective dose is governed by the biokinetic behavior of

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117 these radionuclides following blood absorption, as the lung dose is necessarily small. The current ICRP biokinetic model for uraniu m shows that a high fraction of the element is transferred to the kidneys, bone surf aces, and liver (ICRP 1995a; ICRP 1995b). The model prediction for the uranium uptake in bone for adults is ~15% after 1 day, and decreases to below 6% after 100 days. The uranium uptake in kidneys is ~10% after 1 day, and decreases to 0.1% after 100 days. The equivalent dose is th e highest to the bone surfaces (BS), kidneys (KN), liver (LV), and RBM in that order. When the tissue weighting factor is considered in the calcula tion of the effective dos e, the tissues of the RBM and BS are the 1PstP and 2PndP highest contributors, respectively, to the effective dose. For P230PTh associated with type S aerosols, the lung equivalent dose is the highest contributor to the effective dose for small-si zed particles, while the remainder tissues become the higher contributors at larg e-sized particles (as was the case for P238PU for type M and type S aerosols). For type M aerosols of P230PTh, the lung is no longer the highest contributor to the effective dose. In fact, for type M and type F aerosols of P230PTh, the bone surfaces and red bone marrow receive the highest equivalent doses. After absorption to blood, thorium preferentially lo calizes to bone and is retained for long periods of time. Current ICRP biokinetic m odels of thorium show that the fractional uptake in bone after 1 day is ~50%, increases to 70% in 2000 days, and then decreases thereafter (ICRP 1995a). Consequently, equi valent doses to BS and RBM surpass the lung equivalent dose in spite of the long residence time of P230PTh type M aerosol particles in the respiratory tissues. For similar reasons, the effective dose inhalation coefficient for type S aerosols of P230PTh is lower than those for type M and type F aerosols of P230PTh. Longer residence times in the respiratory tract result in a higher fractional removal to the

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118 GI tract. Conversely, rapid absorption to blood results in higher thorium concentrations in bone with resulting higher va lues of the effective dose. In fact, inhalation effective dose coefficients for type S aerosols are highest for P230PTh of all the radionuclides of the P238PU decay series (note ordinate scales in Figs. 5-1A to 5-1H). The biokinetic behavior of P234PTh is obviously iden tical to that for P230PTh. However, the physical half-life of P234PTh is very short, and thus P234PTh decays rapidly in bone after blood absorption in the lungs. 5.3.2 Inhalation Effective Dose to Workers TDose Rates.T Figs. 5-3A to 5-3C display the e ffective dose rates to workers in the Florida phosphate chemical plants due to par ticle inhalation in gra nulator, storage, and shipping areas, respectively. For each area, effective dose rates are calculated under the assumption that all radionuclides of the P238PU are associated with aerosols of absorption types F, M, or S. These effective dose rate s can thus be used to make individualized dose assessments given explicit knowledge of wo rker occupancy factors and cumulative working times at various processing areas acr oss a given facility. Effective dose rates from internal exposures at gran ulator areas range from 5.9 × 10P-5P to 4.5 × 10P-3P mSv hP-1P for type F aerosols, from 3.6 × 10P-5P to 7.4 × 10P-4P mSv hP-1P for type M aerosols, and from 2.8 × 10P-5P to 4.5 × 10P-4P mSv hP-1P for type S aerosols. The larges t variation in internal effective dose rates is seen in storage areas (Fig. 5-3B) and is a direct result of the wide variability in sampled particle size dist ributions as realized through variations in ventilation and mechanical activity during sampling periods. In the storage areas, internal dose rates vary by factors of 90, 40 and 40 for aerosols of types F, M, and S, respectively. Effective dose rates vary by factors up to 10 depending on the presumed absorption type of the TENORM aerosols. The majority of duplicate samplings at the same worker location show small dose rate differences (less than a factor of 2) for type S aerosols.

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119 The differences in dose rates for locations Aa and C-a in granulator areas, and locations A-b, C, and F at storage areas exceed a factor of 2. Samplings at these latter locations were conducted over a span of several months, and thus aerosol concentrations can vary with in-plant external conditions. Interestingly, inhalation effective dose ra tes given in Fig. 5-3 are shown to be highest for type F aerosols, wh ile the data of Figs. 5-2A to 5-2H indicate that effective doses per unit intake are lowest for type F aerosols for all but one radionuclide of the P238PU series. The unique combination of radiological half-life, sy stemic tissue distribution and retention, and alpha-particle ener gies and yields confirms that P230PTh is a major contributor to the overall inhalation effective dose to P238PU series TENORM aerosol s, and thus this one radionuclide drives the overall trend in inhalation effective dose sensitivity to the lung-fluid solubility of th e phosphate aerosols. TAnnual total effective doses.T Fig. 5-4 indicates the a nnual total effective dose to exposed workers at granulator, storage, and shipping areas for aerosols assumed to be of absorption types F, M, or S. For type S aerosols, no individual dose assessment exceeds the annual non-occupationa l dose limit of 1 mSv yP-1P. In these cases, the internal dose due to particle inhalation is negligible as compared to the external effective dose to the workers at these locations and facilities. Th e total effective doses for type M aerosols are higher in comparison to those for type S: by 7 to 52% at granulator areas, 2 to 33% at storage areas, and 2 to 18% at shipping areas. Only 3 cases (15%) at granulator areas exceed the worker annual dose limit. For type F aerosols, the total effective dose exceeds the annual dose limit of 1 mSv yP-1P for 8 cases (40%) at granul ator areas, 5 cases (31%) at storage areas, and 2 cases (15% ) at shipping areas. Ratios of the annual total effective

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120 dose between type F and type S aerosols ranges from 1.2 to 8.4 at granulator areas, 1.1 to 5.7 at storage areas, and 1.2 to 4.2 at shipping areas. One cannot rule out the po ssibility that the lung-fl uid solubility and blood absorption rate of TENORM aerosols might di ffer among the various radionuclides of the P238PU series. Consequently, it is of further in terest to explore the full range of potential inhalation doses that can be expected under the most conservative to least conservative assumptions of the absorption type by radionuc lide. Based on the data of Fig. 5-1, the most conservative assumption (yielding the hi ghest effective doses) would presume that P230PTh is associated with type F aerosols, wh ile all other radionuc lides are of type S aerosols. The least conservative assumption (yielding the lowest estimates of effective dose) would make opposing assignments of solu bility types. For the least conservative case (open squares in Figs. 5-5A to 5-5C), to tal effective doses for all worker scenarios are below the annual dose limit a nd very close to the assumed uniform external dose rate of 0.2 mSv yP-1P. For the most conservative case (solid circles in Figs 5-5A to 5-5C), the total effective dose exceeds the annual dose limit in 44% of the cases at granulator areas, 31% of the cases at storage areas, and 15% of the cases at shipping areas. In the later case, worker annual doses increase to over 5 mSv at 4 cases at the granulator areas. Internal contributions to the total effective dose can thus potentially vary by factors of 7 to 21 at granulator areas, 8 to 22 at storag e areas, and 8 to 21 at shipping areas. Worker effective doses will in reality be between these two extreme cases. The present study thus indicates the importance of knowing the site-specific lung-flu id solubility of TENORM aerosols at these Flor ida phosphate facilities. Such information can be

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121 important for assessing the adequacy of exis ting radiological prot ection programs and the need for respiratory protection of facility workers. 5.4 Conclusions Workers in the Florida phosphate indust ry receive radiati on exposure both from external gamma-ray photons and from aerosol inhalation of TENORM radionuclides of the P238PU decay series. In this present study, individualized assessments of worker committed effective doses were made using detailed information on the particle size distribution, particle density , particle shape, and radioa ctivity concentrations from sampled aerosols at 6 different phosphate facili ties and at various worker areas within these facilities. Inhalation dose assessments were performed using the ICRP 66 HRTM as implemented in the LUDEP and IMBA computer codes. Effective dose rates to workers in Flor ida phosphate chemical plants due to airborne particle inhalation vary widely by 1 to 2 orders of magnitude depending on workplace particle concentrations. Inhalati on effective dose rates range from 5.9 × 10P-5P to 4.5 × 10P-3P mSv hP-1P (type F aerosols), 3.6 × 10P-5P to 7.4 × 10P-4P mSv hP-1P (type M aerosols), and 2.8 × 10P-5P to 4.5 × 10P-4P mSv hP-1P (type S aerosols) at granul ator areas. At storage areas, inhalation effective dos e rates range from 1.8 × 10P-5P to 1.6 × 10P-3P mSv hP-1P (type F aerosols), 9.0 × 10P-6P to 3.6 × 10P-4P mSv hP-1P (type M aerosols), and 6.1 × 10P-6P to 2.4 × 10P-4P mSv hP-1P (type S aerosols). At shipping areas, inhalation effective dose rates range from 2.8 × 10P-5P to 5.4 × 10P-4P mSv hP-1P (type F aerosols), 8.1 × 10P-6P to 7.9 × 10P-5P mSv hP-1P (type M aerosols), and 5.9 × 10P-6P to 5.3 × 10P-5P mSv hP-1P (type S aerosols). In each case, controlling airborne particle concentrations at these facilities can substantially reduce the committed effective dose rate to workers.

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122 Under the least conservative assumptions for radionuclide absorption types, the annual total effective doses are shown be 0.31 ± 0.12, 0.27 ± 0.07, and 0.22 ± 0.02 mSv yP-1P at granulator, storage, and shipping areas , respectively, and thus all annual worker doses are below the annual limit to the members of the general public of 1 mSv yP-1P. In contrast, the most conservative assumptions of the radionuclide ab sorption type yield annual total effective doses of 2.24 ± 2.53 mSv at granulat or areas, 1.26 ± 1.19 mSv at storage areas, and 0.56 ± 0.36 mSv at ship ping areas, and thus 44%, 31%, and 15% of individual dose assessments at granulator, storage, and shipping areas, respectively, yield worker doses above the annual dose limit. Valu es of inhalation effective dose vary by a factor of between 7 and 22 depending on the absorption types of the radionuclides within sampled aerosols. P230PTh is shown to be a major contributor to the worker effective dose, and thus drives the overall trend in dose sens itivity to assumed valu es of particle lung solubility (e.g., type F aerosols yielding the highest effective dose ra tes in all exposure scenarios). When information on various site-specific particle properties is limited (size distribution, density, shape), c onservative values must be util ized from the standpoint of radiological protection. In a dose sensitivity study, the determination of particle solubility has been shown to be important to minimizing uncertainties in dose estimates. Particle solubility is dependent on many fact ors including particle si ze, shape, number or mass of the deposited particles, intrinsic chemical solubility of the material, and cytotoxicity in the lung tissu es. Consequently, areaand materialspecific particle solubility data obtained from in vitro experiments should be given high importance in

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123 dose assessments for regulatory compliance an d decisions regarding worker respiratory protection.

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124 Figure 5-1 Inhalation do se coefficients for P238PU decay series for particle properties of the Florida phosphate industry: (A) P238PU, (B) P234PTh, (C) P234PU, (D) P230PTh, (E) P226PRa, (F) P210PPb, (G) P210PBi, and (H) P210PPo. Values for P238PU, P234PU, P230PTh, P226PRa, and P210PPo and for P234PTh, P210PPb, and P210PBi were, respectively, calculated using IMBA and LUDEP softwares.

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125 Figure 5-1 Continued

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126 Figure 5-1 Continued

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127 Figure 5-1 Continued

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128 Figure 5-2. Effective dose and weighted equi valent dose to each tissue per unit intake: (A) P238PU (type S), (B) P238PU (type M), (C) P238PU (type F), (D) P230PTh (type S), (E) P230PTh (type M), and (F) P230PTh (type F). Abbreviations represent each tissue: BT (Breast), BS (Bone Surfaces), CL (Colon), ES (Esophagus), GN (Gonads), LN (Lung), LV (Liver), RBM (Red Bone Marrow), REM (Remainder), SK (Skin), ST (Stomach), THY (Thyroi d), and UB (Urinary Bladder).

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129 T T Figure 5-2. Continued

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130 Figure 5-2. Continued

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131 Figure 5-3. Effective dose rate to workers in Florida phosphate chemical plants due to particle inhalation: (A) granulator area, (B) stor age area, and (C) shipping area. Solid circles (), open circles (), and solid triangles () indicate dose rates seen when all P238PU series radionuclides are a ssumed to be of absorption type F, M, or S, respectively. On th e abscissa, upper-case letters, lower-case letters, and numerals indicate plant f acility, sampling locations within the facility, and sample sequence, respectively.

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132 Figure 5-3. Continued

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133 Figure 5-4. Annual total effective dose to wo rkers at Florida phosphate chemical plants for radionuclide-specific absorption types F, M, and S: (A) granulator area, (B) storage area, and (C) shipping area. Values ar e given assuming a uniform external dose of 0.2 mSv and an occu pancy factor of 2000 hours per year. Solid circles (), open circles (), and solid triangles () indicate the annual total effective doses seen when all P238PU series radionuclides are assumed to be of absorption type F, M, or S, respectively.

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134 Figure 5-4. Continued

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135 Figure 5-5. Annual total effective dose to wo rkers at Florida phosphate chemical plants for conservatively assumed radionuclid e-specific absorption types: (A) granulator area, (B) storage area, and (C) shipping area. Values are given assuming a uniform external dose of 0. 2 mSv and an occupancy factor of 2000 hours per year. Solid circles () indicate an annual total effective dose for the most conservative case (type F for P230PTh and type S for all other radionuclides). Open circles () indicate values for the least conservative case (type S for P230PTh and type F other all other radionuclides).

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136 Figure 5-5. Continued

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137 CHAPTER 6 DETERMINATION OF LUNG SOLUBILITY OF RADIONUCLIDES IN PARTICLES IN FLORIDA PHOSPHATE PROCESSING FACILITIES 6.1 Introduction Radioactive materials in phosphate industry processes can give rise to internal exposures to workers and the ge neral public. Phosphate rocks, the source material in the phosphate industry, contain elevated levels of naturally occurring radionuclides. Processing of the materials may increase the concentration of thes e radionuclides often out of secular equilibrium with their pare nts and daughters and/or increase radiation exposure to workers and public through inhala tion, particularly during dusty operations. Concentrated radionuclides as a result of hum an industrial practice are referred to as Technologically Enhanced Na turally Occurring Ra dioactive Material or TENORM. Many studies have been conducted to char acterize the exposure source and to assess the dose to workers in the phosphate industry. Radioactivity concentrations of phosphate materials have been the subject of many previous studies (Burnett et al . 1995; EPA 1977; EPA 1978; Guimond 1978; Guimon d and Windham 1975; Hull and Burnett 1996; Laiche and Scott 1991; Lardinoye et al . 1982; Owen and Hyder 1980; Roessler et al . 1979; Wagner and Leith 2001). They measured radioactivity concentrations of matrix, products, by-product, and waste and charac terized the partitioning of radionuclides during chemical processing. Dose assessmen t studies were conducte d using the data of airborne particle concentrations in the air, aerosol or bulk materi al radioactivity, and inhalation dose coefficients (Birky et al . 1998; Gafvert et al . 2001; Johnson and Traub

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138 1996; Khater et al . 2001; Lipsztein et al . 2001). For most cases, there has been no consideration of partic le properties such as their part icle size and lung solubility. In 1994, the International Commission on Radiological Protection (ICRP) issued Publication 66, Human Respir atory Tract Model (HRTM) fo r Radiological Protection (ICRP 1994). Inhalation dose assessments require knowledge of specific particle properties, including particle si ze distribution, density, shape, and absorption type. These parameters influence both particle deposition an d clearance within the respiratory tract. After radioactive particles ar e deposited on the surfaces of lu ng airways, they experience several fates: (1) transport to the GI tract , (2) transport to regi onal Lymph Nodes (LN), and (3) absorption into blood. In the absence of specific info rmation, default values of material absorption to blood are given for th ree general classifications: F – fast, M – moderate, and S – slow. These default cate gories are in some ways analogous to the inhalation Classes D, W, and Y in the ICRP Publication 30 lung m odel (ICRP 1979). In either case, the ICRP has consistently recommended that, whenever possible, materialspecific absorption parameter values, obtained preferably from in-vivo data, but alternatively from in vitro experiments, should be given preference for exposure and dosimetry calculations. The absorption types of radionuclides are given in ICRP Publications 30 and 71 (ICRP 1979; ICRP 1980; ICRP 1981; ICRP 1995 b). Consider, for example, uranium compounds. Each compound was assigned to a single absorption type depending on its chemical compound type in ICRP Publications 30. Recent review of studies showed considerable variation of the absorption type among various uranium compounds and even in the same compound type (Eids on 1994; ICRP 1995b). Uranium hexafluoride

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139 (UFB6B) and uranyl difluoride (UOB2BFB2B) were assigned to type F. Both type F and M characteristics were found for uranyl nitrate (UOB2B(NOB3B)B2B) and uranium tetrafluoride (UFB4B). The studies about ur anium trioxide (UOB3B), ammonium diuranate (ADU), and uranium octoxide (UB3BOB8B) showed that behavior depends on th e particular process that created the compound. In Chapter 5, doses due to particle in halation were assessed using particle properties characterized in Chapter 1. The doses were calculated for default absorption types given in ICRP 66 HRTM. The sensitivity analysis to ab sorption type indicated that proper knowledge of absorption type was one of the most critical parameters for inhalation dose calculation. Conservative assumption of abso rption types could skew the internal dose by fact ors of 7 to 22. In the phosphate industry, the physicochemical form of the radionuclides inhaled is not well defined. In addition, the radioac tive material is a minor constituent of the inhaled particles. In this case, absorp tion of the radionuclide into blood may be determined by the properties of the radionuc lide-containing matrix rather than by the radionuclide compound type (ICRP 1995b). Lack of information about the radionuclide compound types and low radionuclide concentra tions in particles make it difficult to estimate the absorption type of the radionuclides in the particles in th e phosphate industry. In the lack of information, the selecti on of absorption type M is recommended for all elements except cesium, iodine, and thoriu m in order not to overor under estimate dose in ICRP Publication 71 (ICRP 1995b). De fault type S for thorium and type F for cesium and iodine are recommended because prin cipal forms of the element exhibit type S or F behavior.

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140 The best way to determine the dissolution ra te of specific materials is, however, to use in-vivo human data. In general, the human data are unavailable or insufficient, especially for the pa rticles in non-nuclear industries. As an alternative method, in vitro dissolution tests have been developed and used as research tools to estimate the solubility of particulates and aerosols. Many in vitro radionuclide dissolution studies have been conducted focusing on materials from the nuclear fuel cycle (Allen et al . 1981; Ansoborlo et al . 2002; Cooke and Holt 1974; Damon et al . 1984; Dennis et al . 1982; Eidson and Mewhinney 1980; Eidson and Mewhinney 1983; Heffernan et al . 2001; Metzger and Cole 2004; Metzger et al . 1997; Morrow et al . 1972; Reif 1994). Types of tested materials are wide, including yellow cake, mill tailings, airborne particles, mixed uranium compounds, and nuclear fuel. Other studies were carried out using plutonium (Cheng et al . 2004; Miglio et al . 1977), metal tritides (Cheng et al . 1997; Zhou and Cheng 2003; Zhou and Cheng 2004), contaminated soils (LaMont et al . 2001; Lee et al . 1982), and so on. Recently, the test was broadened to weapons (Guilmette et al . 2002). On the contrary, there are only few studi es about TENORM solubility. Kalkarf et al . estimated lung clearances of radionuclides in coal fly ash and calcined phosphate rock dust (Kalkwarf and Jackson 1984; Kalkwarf et al . 1984). For particles in the phosphate industry, no radionuclide solubil ity study has been reported in spite of its importance to the dose estimation. The objective of the present study is to present lung so lubility of radionuclides contained within the particles from the Fl orida phosphate indust ry. The information

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141 enables the assessment of particle inhalation exposures devoid of potentially conservative default assumptions. 6.2 Materials and Methods 6.2.1 Dose Sensitivity to Radionuclide Solubility A sensitivity study was first conducted to determine the degree to which absorption type contributes to uncerta inties in the effective dose to phosphate workers during inhalation exposures of P238PU and P230PTh aerosols. Study methods are similar to the dose sensitivity study described in Chapter 2. Values of the 50-year committed effective dose per unit intake (e.g., effective dose coefficient) of P238PU and P230PTh to workers under light exertion were computed using the Integrat ed Modules for Bioassay Analysis (IMBA) code of James et al. (James et al. 2003). Initially, the effective dose coefficient was calculated using ICRP default aerosol parameters as given in Table 2-2. The partic le size distribution was allowed to vary from an AMAD of 0.01 m to an AMAD of 100 m with corresponding GS D. Other ICRP 66 default parameters include a particle density of 3 g cmP-3 Pand a shape factor of 1.5. Particle solubility w ithin the lung fluids was changed from the default absorption type (type M for P238PU and type S for P230PTh) to the other types: either type F or S for P238PU and either type F or M for P230PTh. 6.2.2 Tested Samples Settled particles, airborne particles, and bulk products were em ployed for solubility testing. The tested samples are tabulated in Table 6-1. Settled particles are a source of inhalation through re-suspension. Differe nt types of settled particles, including monoammonium phosphate or MAP and diammoni um phosphate or DAP, were obtained from 6 phosphate processing facilities locate d in the central and northern regions of

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142 Florida. Each sample was sieved to remove particles larger than 76 m in diameter. In addition to settled particles, two sample s of airborne particles smaller than 10 m and two types of bulk products were tested for si ze effect on particle solubility. Airborne particles are sampled using a PMB10B high-volume sampler (Sierr a-Andersen, Model 1200). The diameter of bulk produc ts was about 3 mm. 6.2.3 In vitro Solubility Test Various types of in vitro dissolution techniques, incl uding (1) flow systems, (2) static systems, and (3) batch methods, have been used in previous studies (Ansoborlo et al . 1999). The radionuclide mass concentration in phosphate particles is rather low. For example, the mass fraction of uranium in MA P and DAP ranges from 0.014% to 0.025% (Birky et al . 1998). There is a limitation of applic able particle mass for the first two methods. Therefore, the batch method was se lected to increase the amount of particles thus increasing radionuclide concentr ation in solution for analysis. Previous studies showed successful efforts to estimate particle solubility in the lung fluids using serum ultrafiltrate (SUF) (Cheng et al . 2004; Kanapilly et al . 1973). This simulant was used in this study, and its co mposition of SUF is shown in Table 6-2. Reagent grade chemicals are dissolved in 18 M water following the sequence in the table. Each ingredient was completely disso lved before adding th e next ingredient. The batch method system is depicted in Fig. 6-1. Samples weighing 5 g were suspended in a flask containi ng 200 mL SUF, which is imme rsed in a water bath for temperature control at 37 PoPC. The suspension was stirred with a Teflon stir bar. The pH value of the suspension was maintained w ithin the range of 7.3 7.4 except MAP samples during the first 1 day. High fract ional dissolution of MAP within 1 day

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143 decreased pH to 5.6 – 5.8 and thus it was difficult to control the pH values. After extracting the solution after 1 day, the pH values can be controlled within 7.3 – 7.4 by flowing a gas mixture of 5% carbon dioxide (COB2B) and 95% air. The flow rates of COB2B and air were 10 to 50 mL minP-1P and 200 to 1000 mL minP-1P, respectively. Increased COB2B flow rate decreased pH values. Periodically, the suspension was filtered by membrane filter with 0.1 m pore size. The test was conducted for 50 to 85 days. Residues, after dissolution testing, were dissolved with conv entional geochemical methods using a mixture of hydrof luoric (HF) and nitric (HNOB3B) acids (Eggins et al . 1997). About 40 mg residues were placed into Savillex screw-top Te flon vials and then 0.5 mL of optima grade concentrated HF a nd 2.5 mL of optima grade concentrated HNOB3B were added. The vials were heated in an oven at 100 °C for 24 hours and then the samples were evaporated on a hot plate. After complete dryness, 3 mL optima grade concentrated HNOB3B spiked with hydrochloric (HCl) acid was added to the samples and the vials were heated on the hot plate for 24 hours, then ev aporated again. The final residues were re-dissolved with 4 mL 5% HNOB3B and diluted about 2000 times before the analysis. 6.2.4 Solubility of Surrounding Material Solubility of surrounding mate rial was analyzed before de termination of solubility of uranium, thorium, and lead. In ICRP Pub lication 71, it is stated that if radioactive element is present as a minor constituent of inhaled particles, absorption of the radionuclide to body fluids may be controlled by the surrounding matrix rather than the elemental form of the radionuclide (ICRP 1995b). Table 2-4 shows the ingredient composition of MAP and DAP. Dry products and particles contain primary components

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144 (nitrogen and phosphorus), other impurities (calcium, magnesium, iron and aluminum sulfates, phosphates, silicates, and fluo rides), and minor constituents including radioactive elements, which are embedded in p hosphate material. The solubility of the surrounding matrix was determined by measuring the phosphate ion (POB4P B-3P) concentration in leachates. However, it was impossible to measure the phosphate ion concentration in residues because the phosphate ion can be chan ged into another form when dissolved in concentrated acids. The remaining phosphate io ns after dissolution tests for 50 – 85 days would be minute. Therefore, it was assumed that there is no remaining phosphate in the residues. Ion Chromatography (IC) (D ionex, ICS-1500) was employed for ion concentration measurement. 6.2.5 Solubility of Uranium, Thorium, and Lead The solubility of uranium, thorium, and lead in particles was characterized by measuring mass concentrations of P238PU, P232PTh, and P208PPb in leachates and residues. Phosphate product contains P238PU decay series. Redistribution of the P238PU decay series occurs from its original stat e of secular equilibri um within the phosphate rock during the product manufacturing processes. Uranium sele ctively localizes in product materials, while radium and lead tend to localize in th e by-product. The previous dose assessment study in Chapter 5 showed that most inhalati on dose to workers in the Florida phosphate industry was caused by P230PTh in particles. The second a nd the third contributors to dose were P234PU and P238PU. The other radionuclide contributions to dose were trivial due to low inhalation dose coefficients (P234PTh and P210PBi) or low concentrations in particles (P226PRa and P210PPo). The chemical behavior of a given radionuclide, and thus its absorption type within the ICRP 66 clearance model, is give n by the element type rather than its particular isotope. A lthough radioactivity of P234PU, P230PTh, and P210PPb in solution is

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145 comparable to that of P238PU, those mass concentrations are too low to count by massspectroscopy because the physical half-lives of the radionuclides are short in comparison to that of P238PU. The mass concentrations are the highest in P238PU among uranium isotopes, in P232PTh among thorium isotopes, and in P208PPb in lead isotopes. Therefore, the solubilites of uranium, thorium, and lead were determined by P238PU, P232PTh, and P208PPb. The measurement was performed on high resolution inductivel y coupled plasma mass spectrometry (HR-ICP-MS) (Thermo Electron Corp., Finnigan Element 2). Recently, the use of ICP-MS for health physics studies ha s increased due to its high sensitivity and rapid pr ocedure (Bouvier-Capely et al . 2003; Galletti et al . 2003; Karpas et al . 2005; Roth et al . 2005). The intensities of the is otopes were measured at medium resolution. The measurement was calibrated with gravimetrically prepared uranium, thorium, and lead calibration standards. The concentration of P232PTh in most leachate samples was below detection limit of 1.15 g L-1 because SUF already contains P232PTh. For these cases, the concentration of P232PTh was assumed to be the same as the detection limit from the aspect of radiation protection because rapid absorption results in a higher dose for P230PTh. 6.2.6 Data Analysis The remaining fraction at each time inte rval was calculated using the following equation: T TM Remaining fraction= Mi i M , Eq. 6-1

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146 where MBTB is total amount of phosphate ion, P238PU, P232PTh, or P208PPb in particles and MBiB is the dissolved amount at each extraction. MBT Bwas calculated by summation of total dissolved amount during the test and remaining amount in the residues. The kinetics of dissolution was e xpressed by two exponential functions: Remaining fraction = s r s t st rs f efe , Eq. 6-2 where fB and sB are dissolution fractions and rate constants, and t is collapsed time (Ansoborlo et al . 1999). Each exponen tial function represents the rapid or slow dissolution compartment, re presented by subscripts r and s , modeled in the ICRP 66 clearance model (ICRP 1994). The retention fraction was plotted by time using the data of element concentration in solution, and th en the curve was fitted using Sigma plot (SPSS Inc., SigmaPlot V. 8.0) (Cheng et al . 2004). It is necessary to set the rapid dissoluti on rate unless dissolution fractions at early time points within 10 minutes are not characterized. In this study, solution samples were first collected after 1 day. Therefore, it was impossible to predic t solubility behavior before that first collection. Fitting data of the rapid dissolution rate change depending on the first sample collection time. A previ ous dissolution study of plutonium aerosol changed the fitting data of rapid dissolution rate to 100 per day, which is the default value given in ICRP 66 HRTM (Cheng et al . 2004). After the modification, their in vitro measurement data agreed well with their bioassay data. As a result, in this study the rapid dissolution rate constant ( sBrB) was set as 100 per day.

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147 6.3 Results and Discussion 6.3.1 Dose Sensitivity to Radionuclide Solubility Figure 6-2 displays the inhala tion dose coefficients of P238PU and P230PTh for absorption types F, M, and S. The single solid circle a nd solid triangle indicate dose coefficients for ICRP default aerosol values for genera l public exposures and workplace exposures, respectively. Deposited particles of P238PU that are slowly absorbed in the lung fluids (type S) remain in the respiratory tract for long peri ods of time, and thus deliver higher absorbed doses to the lung tissues than do either type M or type F P238PU particles. Once absorbed into pulmonary blood, the P238PU atoms are distributed to the kidney, mineral bone, and other soft tissues of the body. However, th eir organ dose contributi ons to the effective dose are small compared to the absorbed dose to the lung tissues. Therefore, type S, type M, and type F materials resu lt in the highest, next highest , and lowest effective dose coefficients for P238PU, respectively. Type S P238PU results in an effective dose per unit intake that is 2.9 and 3.4 times that of type M for 1 m AMAD distribution and 5 m AMAD distribution, respectively. Conversely, type F P238PU has an effective dose coefficient that is 5.2 (1 m AMAD) and 2.9 (5 m AMAD) times less than Type M P238PU aerosol particles. Dose coefficients for P230PTh are noted to be the highest for type F and the lowest for type S. Like P238PU, type S P230PTh longer stays in the respir atory tract thus giving higher dose to lung. The main differen ce is its biokinetic behavior after absorption into blood. A high fraction of absorbed thorium preferenti ally localizes to bone and is retained for longer periods of time. Longer residence times in the respiratory trac t result in a higher

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148 fractional removal to the GI tr act. Conversely, rapid absorp tion to blood results in higher thorium concentrations in bone with resulting higher effective dose. For this case, the main contributor to effective dose is red bone marrow or bone surface ra ther than lung. Types F and M P230PTh result in higher effective doses by factors of 8.0 and 3.2 (for 1 m AMAD) and 16.7 and 3.9 (for 5 m AMAD). The difference increases with increasing particle size. Type F P230PTh with 100 m AMAD yields effective an dose 42 times that of type S P230PTh of the same size. This effective dose sensitivity study indicates that the abso rption type of an inhaled radionuclide, in addition to the particle size di stribution characterized in Chapter 2, is one of the most critical parameters for dose asse ssment. An incorrect selection of absorption type can overor under-estimate dose to wo rkers by an order of magnitude. 6.3.2 Lung Solubility of Particles Fig. 6-3 shows the remaining fraction of phosphate in the sampled particles as a function of time. More than 94% of phosphate is dissolved within 1 day for all samples. Phosphate solubility differences were not ed among different types of products and particle sizes. Fig. 6-4 shows the average retention of P238PU, P232PTh, and P208PPb in particles for all plants as a function of time. Minor consti tuents of uranium, thorium, and lead in phosphate product particles are not rapidly dissolved with th e surrounding matrix. They are retained in particles fo r longer times, while the surrounding phosphate materials are dissolved rapidly. For most samples, the retention of P238PU is between those under types M and S assumptions given in the ICRP 66 HRTM, although somewhat close to type M rather than type S. Retention of P232PTh is close to type S. As mentioned earlier, the

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149 retention data of P232PTh was obtained under a conservative assumption. In reality, it will be retained longer than the estimation. Ther efore, selection of type S for thorium is reasonable for dose assessment to work ers in the phosphate industry. For P208PPb, the retention trend is similar to that of type M in ICRP 66 HRTM. Sample-specific retention fitting data are given in Table 6-3 where the rP2P values of the fit range from 0.8775 to 0.9966. Ra pid and slow dissolution fractions of P238PU range 3% 14% and 86% – 97 %, respectively. Th e slow dissolution rates range from 0.0022 to 0.0076 dayP-1P. For P232PTh, less than 1.4% is dissolved with a half-life of 10 minutes. The remaining fraction is dissolved with half-lives of more than 866 days. For P208PPb, the slow dissolution fraction ranges as low as 59% to as high as 97 % with dissolution rates of 0.0008 to 0.0183 dayP-1P. The results of radionuclide solubility show no noticeable correlation to product types and particle sizes , rather it is sample-dependant. The variation of solubility among samples is the greatest for P208PPb and the smallest for P232PTh. In general, solubility of radioactive part icles depends on the nature of the inhaled particles, including particle size and shape, intrinsic chemical solubility, generation condition, and other factors. The chemical form of the radionuclide in the sampled particles is one of the most critical parame ters to determine solubility. It is thus instructive to look at radionuclide chemis tries in the phosphate product processes. Phosphate products are produced from phosphate ore. The ore from Florida is a sedimentary apatite commonly known as francolit e, which is described as follows (Van Kauwenbergh 1997): 1046320.4()()xyxyzzzCaNaMgPOCOFF.

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150 The contents of Ca, Na, Mg, P, COB2B, and F describe most franco lites. In addition, the rocks contain minor constituents of radionuclide s such as uranium series. However, their exact position is speculative. Th ey may substitute within the apatite structure, substitute in other mineral associated with the phosphate, or be adsorbed on the surface of these minerals. Divalent calcium may be substituted with divalent uranium, thorium, and lead. The phosphate ore reacts with sulfuric ac id to produce phosphoric acid and hydrated phosphogypsum (Osmond et al . 1984): 104622424234()102010262 CaPOFHSOHOCaSOHOHPOHF . It has been reported that most of the uran ium and thorium goes with the phosphoric acid as soluble salts and most of the radium and lead goes with the phosphogypsum (Birky et al . 1998; Burnett et al . 1995; Guimond and Windham 1975; Laiche and Scott 1991; Lardinoye et al . 1982; Roessler et al . 1979). During the reacti on between phosphate rock and sulfuric acid, uranium remains in liquid as uranyl phosphate, sulfate, and fluoride complexes (Mazzilli et al . 2000). Thorium forms sparingly soluble salts w ith hydroxides, fluoride, and phosphate. Sulfate compounds are re latively soluble. Most of the thorium goes with acid during acidul ation process (Burnett et al . 1995). Radium is localized on phosphogypsum because chemical properties of ra dium are similar to calcium (Lardinoye et al . 1982). Chemically, lead is expected to be lead sulfate, which is quite insoluble. Major granulated products, DAP and MAP, are formed by the reaction of ammonium hydroxide with phosphoric acid. There are a lot of possibilities for uranium, thorium, and lead to form oxides or to combine with pl enty of phosphate ions. However, it is not known how these compounds are fractionated. In addition, some gypsum and other impurities get through the filtration process and into the acid. Therefore, it is difficult to

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151 define the exact chemical form of radionuc lides in phosphate product, which is a source of inhalation particles. If a radionuclide exists in particles as a minor constituent, the solubility of the radionuclide may depend on both solubility of surrounding matrix and the radionuclide itself. Kalkwarf and Jackson tested solubil ity of radionuclides in coal fly ash (Kalkwarf et al . 1984). They proposed two reasons for the slow dissolution. The first one is the chemical form of the radionuclides. Inheren tly insoluble radionuclides on the surface of the fly ash are slowly dissolved in the l ung fluids. The other reason is radionuclide protection by fly ash components. They part ially shield radionucli des from contact with the lung fluids. The latter case cannot be a pplied to the phosphate product particles because most of surrounding matrix is dissolv ed in a short time and thus does not prevent radionuclides from contacting the SUF. If so , there should be a relation between particle size and solubility because the radionuclides in larger sized particles, in comparison with smaller particles, will have less chance to contact with SUF and thus they will dissolve more slowly. Consequently , it is concluded that th e main parameter governing radionuclide solubility in phospha te particles is the chemical forms of the radionuclides rather than that of the surrounding matrix. 6.4 Conclusions Blood absorption of radionuclides in particle s deposited in the re spiratory tract is one of the most critical parameters for i nhalation dose calculation. The absorption to blood can be modeled by data of the radionucli de solubility in the lung fluid. Lung solubility of the radionuclides in the particles from th e Florida phosphate processing facilities has been determined using a batch method.

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152 A fraction of 0.03 – 0.14 of P238PU is dissolved rapidly. The remaining fraction is dissolved with half-lives of 91 – 315 days. The dissolution kinetics is close to that of absorption type M material defined in ICRP 66 HRTM. Less than 0.014 of P232PTh is dissolved rapidly. The remaining fraction is dissolved with half-lives of more than 866 days. The retention data of P232PTh is obtained under a conserva tive assumption. Therefore, selection of type S is reco mmended for dose assessment. The solubility data of P208PPb have a wide variation among samples. The fraction of 0.03 – 0.41 is dissolved rapidly with the remainder being slowly dissolve d with half-lives of 38 – 866 days. Lung solubility of radionuclides in phospha te product particles is mainly governed by their chemical forms. Measurements s how no correlation be tween solubility and product type or particle size. Most of the surrounding matrix of the particles, phosphate, is dissolved in a short time. Therefore, pa rticle size does not influence the solubility. However, it is difficult to define the exact chemical forms of radionuclides and their fractions in particles because there are ma ny complexes for the radionuclides to form during a series of mechanical and chemical processes for product generation. Their characterization will be helpful in understanding the kinetics and mechanisms of the radionuclide solubility in the lung fluid. For the purpose of dose assessment to wo rkers in the phosphate chemical plants and to members of the general public, the spec ific values characterizing the radionuclide solubility can be used at a sp ecific facility. Otherwise, it is recommended to select type M for uranium and lead and type S for thorium.

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153 Table 6-1. Samples employed for solubility test Settled Particles ( 76 m) Airborne Particles ( 10 m) Bulk Products ( 3 mm) Plant MAP DAP MAP DAP MAP DAP Plant A O O O O O Plant B O Plant C O Plant D O O O Plant E O Plant F O

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154 Table 6-2. Composition of serum ultrafiltrate simulant Chemical Formula WeightConcentration (g LP-1P) NaCl 58.44 6.7790 NHB4BCl 53.49 0.5349 NaHCOB3B 84.01 2.2683 NaHB2BPOB4B·HB2BO 137.99 0.1656 NaB3B Citrate·2HB2BO 294.10 0.0588 Glycine 75.07 0.3754 L-cysteine hydrochloride 175.63 0.1756 DTPAPaP 393.34 0.0787 HB2BSOB4B 98.08 0.03 mL LP-1P CaClB2B·2HB2BO 147.02 0.0294 ABDCPbP 471.50 0.1 mL LP-1P Pa PDiethylene-triamine-pentaacetic acid Pb PAlkylbenzyl-dimethyl-ammonium chloride

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155 Table 6-3. Retention fitting data of P238PU, P232PTh, and P20P P8PPb in serum ultrafiltrate as a function of time Retention Parameters Radionuclide Plant Product Type Sample Type fBrB fBsB sBrB rP2P P238U A MAP Settled particles 0.03190.9681 0.0022 0.9162 A DAP Settled particles 0.04500.9550 0.0068 0.9760 A DAP Airborne particles 0.09180.9082 0.0058 0.8777 A MAP Bulk products 0.08070.9193 0.0047 0.9755 A DAP Bulk products 0.07000.9300 0.0045 0.9758 B MAP Settled particles 0.05590.9441 0.0028 0.9588 C DAP Settled particles 0.13790.8621 0.0076 0.9710 D MAP Settled particles 0.08110.9189 0.0035 0.9868 D DAP Settled particles 0.05030.9497 0.0025 0.9575 D MAP Airborne particles 0.06610.9339 0.0030 0.9752 E MAP Settled particles 0.03300.9670 0.0030 0.9799 F DAP Settled particles 0.07940.9206 0.0026 0.9744 P232PTh A MAP Settled particles 0.00530.9947 0.0001 0.9162 A DAP Settled particles 0.01430.9857 0.0008 0.8775 A DAP Airborne particles 0.01110.9889 0.0003 0.9274 A MAP Bulk products 0.01080.9892 0.0008 0.9879 A DAP Bulk products 0.00900.9910 0.0006 0.9858 B MAP Settled particles 0.00800.9920 0.0003 0.9102 C DAP Settled particles 0.00840.9916 0.0007 0.9869 D MAP Settled particles 0.00670.9933 0.0004 0.9816 D DAP Settled particles 0.00590.9941 0.0001 0.9475 D MAP Airborne particles 0.00450.9955 0.0003 0.9846 E MAP Settled particles 0.00080.9992 0.0005 0.9647 F DAP Settled particles 0.00300.9970 0.0002 0.9795 P208PPb A MAP Settled particles 0.01860.9814 0.0012 0.9248 A DAP Settled particles 0.40830.5917 0.0039 0.9966 A DAP Airborne particles 0.06340.9366 0.0022 0.9316 A MAP Bulk products 0.09280.9072 0.0068 0.9924 A DAP Bulk products 0.27460.7254 0.0055 0.9944 B MAP Settled particles 0.09980.9002 0.0045 0.9750 C DAP Settled particles 0.12390.8761 0.0183 0.9904 D MAP Settled particles 0.06460.9354 0.0017 0.9845 D DAP Settled particles 0.08080.9192 0.0008 0.9786 D MAP Airborne particles 0.04630.9537 0.0016 0.9660 E MAP Settled particles 0.03170.9683 0.0121 0.9894 F DAP Settled particles 0.26880.7312 0.0043 0.9925

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156 Figure 6-1. Schematic of system for in vitro solubility testing

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157 Figure 6-2 Inhalation dose co efficients for absorption types F, M, and S: (A) P238PU and (B) P230PTh. The other particle propertie s follow the ICRP 66 HRTM default particle values.

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158 Figure 6-3. Fraction of remaining phosphate (POB4P B-3P) in particles as a function of time in serum ultrafiltrate.

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159 Figure 6-4. Fraction of remaining P238PU, P232PTh, and P208PPb in particles as a function of time in serum ultrafiltrate: (A) P238PU and P232PTh and (B) P208PPb. Retentions of default absorption type materials in ICPR 66 HRTM are depicted for comparison.

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160 CHAPTER 7 RISK ASSESSMENT OF AIRBORNE PART ICULATES TO WORKERS IN THE FLORIDA PHOSPHATE INDUSTRY 7.1 Introduction The overall objective of this investigation is to evaluate the health risks to workers in the Florida phosphate industry due to inhala tion of particulates containing radioactive materials. The health risks can be assessed by estimating radiation exposure to workers. There has been no detailed study of dose a ssessment to workers in the phosphate industry due to particle inhalation. Previ ous studies adopted conservative assumptions due to the lack of particle property information thus having possibility to skew dose to unrealistic values (Gafvert et al . 2001; Lipsztein et al . 2001). A comprehensive integrated study was conducted by Birky et al . (Birky et al . 1998). However, there was no consideration of particle properties, includ ing particle size distri bution, density, shape, and lung solubility of radionuc lides in the particles. In the previous chapters, all particle information needed for inhalation dose assessment was measured. In addition, dos e calculation methods were introduced to apply measured data. They enable to make practical dose assessment to workers in the Florida phosphate industry without skewi ng dose to unrealistic values by adopting conservative assumptions. The objective of the presen t study is to assess inha lation dose due to TENORM aerosol inhalation across the Florida phosphate industry. Specific areas of potential internal dose concern include granulator, storage, and shipping areas in phosphate

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161 chemical plants. In addition, total eff ective doses are estimated to demonstrate compliance with standards or regulations. 7.2 Materials and Methods Inhalation dose assessment was conducte d using all particle database and dose calculation methods established in this study. In Chapter 2, particle property database were established. The properties include partic le size distribution, pa rticle shape, density, and radionuclide concentration in particles at specific working locations in Florida phosphate chemical plants. In Chapters 3 and 4, effective dose scaling factor was introduced to apply airborne pa rticle sampling data to dose assessment. In Chapter 5, dose due to inhalation of particles encompassing P238PU series was assessed via particle properties established in Chapter 1. In Chapte r 6, solubility of ra dionuclides in particles in the lung fluid was determined. All th ese measurement data and dose calculation methods were integrated into a full intern al dosimetry assessment of individualized worker doses by both plants and operational locations. Table 7-1 shows scaling factor s used for dose assessment. Scaling factors given in Chapters 3 and 4 were generated on the basis of particle density of 1.0 g cmP-3 Pand unity particle shape factor. The change of part icle density shifts the cutoff size of each impactor stage thus changing effective dose sc aling factors. Scaling factors specifically applicable to the particles in the Florida phosphate chemical plants were generated using the measured particle properties, particle dens ity of 1.6 g cmP-3P and unity shape factor. Dose contributions of P214PTh and P210PBi were less than 0.1% of total estimated dose to workers due to the inhalati on of particles containing P238PU series because those radionuclides have low inhalation dose coeffi cients. Therefore, scaling factors were generated for the radionuclides of P238PU, P234PU, P230PTh, P226PRa, P210PPb, and P210PPo. Radionuclide

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162 solubility study showed that uranium and lead are dissolved like type M materials defined in ICRP 66 HRTM and thorium is dissolved like type S materials. The values in Table 71 are scaling factors of absorption type M P238PU, P234PU, P226PRa, P210PPb, and P210PPo and absorption type S P230PTh. Scaling factors for all absorpti on types are given in Appendix B. For dose assessment, it was assumed that ra dium and polonium is classified into absorption type M. Solubility of radium and polonium were not determined in solubility study in Chapter 6 due to the low mass concen tration and trivial c ontribution to dose. Initial dose calculations were conducted under the least cons ervative assumption (type F P226PRa and P210PPo), intermediate conservative assumption (type M P226PRa and P210PPo), and the most conservative assumption (type S P226PRa and P210PPo). Dose discrepancy between the least or the most conservative case and the inte rmediate case is less than 2% for the most dose calculation cases. A maximum diffe rence of 6% was found in Plant E. Consideration of P226PRa and P210PPo solubility does not influe nce the dose estimation much and thus inhalation dose calculation was conducted under the assumption of type M P226PRa and P210PPo. 7.3 Results and Discussion Table 7-2 shows particle inha lation dose rates at granul ator, storage, and shipping areas in the Florida phosphate chemical plants. The dose rates range 2.5×10P-5P 2.4×10P-4P (6.8×10P-5P ± 6.0×10P-5P) mSv hP-1P at granulator area, 5.5× 10P-6P – 1.5×10P-4P (4.9×10P-5P ± 5.0×10P5P) mSv hP-1P at storage area, and 4.4×10P-6P – 3.3×10P-5P (1.4×10P-5P ± 9.8×10P-6P) mSv hP-1P at shipping area. The values are widely deviat ed depending on locations. These effective dose rates can be used to make individualiz ed dose assessment with worker’s occupancy factors at various processing ar eas across a given facility.

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163 Estimated annual dose are also given in Table 7-2. Annual particle inhalation dose ranges 0.04 – 0.48 (0.14 ± 0.12) mSv yP-1P at granulator area, 0.01 – 0.29 (0.10 ± 0.10) mSv yP-1 Pat storage area, and 0.01 – 0.07 (0.03 ± 0.02) mSv yP-1P at shipping area. The annual doses are estimated by assuming occupa ncy factor of 2000 hours per year at each area. The results would be conservative since the workers employed in the phosphate industry are involved in severa l types of job assignments an d thus exposure time at the dusty area should be shorter than the assumed time duration. Birky et al . collected information about worker time, motion, and posi tion in order to estimate radiation dose to various workers in the Flor ida phosphate industry (Birky et al . 1998). Only 5 types of job classifications in dry product area among 16 were e xposed at dusty area and 7 at shipping/storage area among 10. The employees in the other job classifications work at office, control room, or open plant thus ha ving low possibility to inhale particles. Exposure time of the employees working at dusty area is summarized in Table 7-3. Most workers are involved in the assignments at dusty area for shorte r than 2000 hours per year. Laborers in both areas are expose d longest for 1875 – 2000 hours per year. An extensive external dosimetry study in the Florida phosphate industry was reviewed because total effective dose is summation of ex ternal dose and internal dose. Birky et al . monitored external exposure using li thium fluoride (LiF) thermoluminescent dosimeter (TLD), and aluminum oxide (AlO) TLD (Birky et al . 1998). They reported four series of statistical results, one from the LiF TLD data set and three from AlO TLD data sets. The 4 series of data sets were av eraged in this study. The averaged external doses are 0.121 (with GSD of 2.27) mSv yP-1P at dry product areas and 0.161 (with GSD of 2.42) mSv yP-1P at shipping areas. The data were obtained from all job classifications.

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164 Consideration of all external dosimetry data showed that certain job classifications consistently incurred doses higher than 0.2 mSv yP-1P. They included granulator operator, chief operator, maintenance mechanics, bobcat operator, laborer, dry products utility operator, DAP supervisor, and DAP car loader at dry product area and payload operator, wet rock operator, maintenance, rock rail s upervisor, rock conveyor operator, car loader, rock tractor operator, and shipping pr ocess operator at shipping area. Total effective dose is a scale to show co mpliance with standards or regulations. Total effective dose limits are 1 mSv yP-1 Pfor members of the p ublic and 50 mSv yP-1 Pfor occupational radiation workers by Florida St ate regulations – Fl orida Administrative Code 64E-5. Although annual exte rnal dose rate of 0.2 mSv yP-1 Pis assumed in the granulator, storage, and shipping areas, total effective dose is much lower than occupational exposure limit. The values are even lower than the dose limit to general public. 7.4 Conclusions Radiation doses to workers due to TE NORM particle inhala tion were assessed across the Florida phosphate chemical plants . Individualized dose assessments were made using site specific particle information. Directly measured particle properties were used for the dose calculation w ithout conservative assumption of the properties. Particle inhalation dose ra tes have wide variations depending on plants and locations, ranging 2.5×10P-5 P – 2.4×10P-4P mSv hP-1P at granulator areas, 5.5× 10P-6P – 1.5×10P-4P mSv hP-1P at storage areas, and 4.4×10P-6P – 3.3×10P-5P mSv hP-1P at shipping areas. Total effective doses range s 0.24 – 0.68 mSv yP-1P at granulator ar eas, 0.21 – 0.49 mSv yP-1P at storage areas, and 0.21 – 0.27 mSv yP-1P at shipping areas. The fi nding of this study is that

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165 effective dose to workers is lower than dose limit to general public regulated by the state of Florida and extremely unlikely to approach or exceed the occupational dose limit.

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166 Table 7-1. Inhalation effec tive dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants Effective Dose Scaling Factor SFBE P BaP Radioactivity Distribution Impactor Stage P238PUPbP P234PUPbP P230PThPbP P226PRaPbP P210PPbPbP P210PPoPbP Uniform 1 1.03 1.02 1.02 1.03 1.02 1.02 distribution 2 1.03 1.04 1.03 1.04 1.02 1.05 3 1.15 1.13 0.95 1.13 1.08 1.15 4 0.96 0.96 0.96 0.96 0.97 0.96 5 1.01 1.01 0.99 1.01 0.99 1.01 6 1.03 1.03 1.02 1.04 1.03 1.04 7 1.02 1.02 1.03 1.02 1.03 1.02 F 1.08 1.08 1.07 1.09 1.07 1.08 Linearly 1 1.04 1.03 1.03 1.04 1.03 1.04 decreasing 2 1.07 1.08 1.06 1.08 1.05 1.10 distribution 3 1.28 1.25 1.03 1.25 1.15 1.29 (AR = 2:1) 4 0.96 0.96 0.98 0.96 0.97 0.95 5 0.98 0.97 0.97 0.97 0.97 0.97 6 1.02 1.02 1.00 1.03 1.00 1.02 7 1.05 1.05 1.05 1.05 1.05 1.05 F 1.19 1.19 1.18 1.19 1.18 1.19 Linearly 1 1.05 1.05 1.04 1.05 1.05 1.05 decreasing 2 1.11 1.12 1.10 1.12 1.08 1.14 distribution 3 1.40 1.38 1.11 1.38 1.23 1.43 (AR = 5:1) 4 0.96 0.96 1.00 0.96 0.98 0.95 5 0.95 0.94 0.95 0.94 0.94 0.94 6 1.01 1.01 0.98 1.02 0.97 1.01 7 1.08 1.08 1.07 1.08 1.07 1.08 F 1.30 1.30 1.28 1.30 1.28 1.30 PaP Values of SFBEB given here are specific to partic le size ranges of the University of Washington Mark III cascade impactor at an air flow rate of 15 L minP-1P, particle density of 1.6 g cmP-3P, and unity particle shape factor. Bo lded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).P b Scaling factors are generated a ssuming absorption type M for 238U, 234U, 226Ra, 210Pb, and 210Po and absorption type S for 230Th. Scaling factors for all absorption types are given in appendix B.

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167 Table 7-1. Continued Effective Dose Scaling Factor SFBE Ba Radioactivity Distribution Impactor Stage 238Ub 234Ub 230Thb 226Rab 210Pbb 210Pob Linearly 1 1.01 1.01 1.01 1.02 1.01 1.01 increasing 2 1.00 1.00 1.00 1.00 0.99 1.01 distribution 3 1.02 1.00 0.87 1.00 1.01 1.01 (AR = 1:2) 4 0.95 0.96 0.94 0.96 0.96 0.96 5 1.04 1.04 1.00 1.04 1.02 1.05 6 1.04 1.05 1.04 1.05 1.06 1.05 7 0.99 0.99 1.01 0.99 1.01 0.99 F 0.98 0.98 0.97 0.98 0.97 0.98 Linearly 1 1.00 1.00 1.00 1.00 1.00 1.00 increasing 2 0.96 0.96 0.96 0.96 0.96 0.96 distribution 3 0.90 0.88 0.79 0.88 0.93 0.86 (AR = 1:5) 4 0.95 0.96 0.92 0.96 0.95 0.96 5 1.07 1.07 1.02 1.07 1.05 1.08 6 1.05 1.06 1.06 1.06 1.08 1.07 7 0.97 0.96 0.99 0.96 0.99 0.96 F 0.87 0.87 0.86 0.87 0.86 0.87

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168Table 7-2. Dose rate and annual inhalation dose to workers in the Florid a phosphate chemical plants due to particle inhalation a Capital alphabets are plant identifiers. Plant A-a and Plant A-b refer to the same company plant (Plant A), but two different locations within the plant. A-b (1 ) and A-b (2) indicate two different samplings at the same location. b Annual dose rate is calculated by assuming an occupancy factor of 2000 hours per year. Inhalation Dose Inhalation Dose Inhalation Dose Dose Rate Annual Dose bDose Rate Annual Dose b Dose Rate Annual Dose bGranulator Area a (mSv h-1) (mSv y-1) Storage Area a ( mSv h-1) (mSv y-1) Shipping Area a ( mSv h-1) (mSv y-1) A-a (1) 2.7E-05 0.05 A-a (1) 1.6E-05 0.03 A-a (1) 2.4E-05 0.047 A-a (2) 5.3E-05 0.11 A-a (2) 2.9E-05 0.06 A-a (2) 3.0E-05 0.060 A-a (3) 5.7E-05 0.11 A-b (1) 5.5E-06 0.01 A-b (1) 5.2E-06 0.010 A-b (1) 4.8E-05 0.10 A-b (2) 2.5E-05 0.05 B (1) 2.1E-05 0.041 A-b (2) 3.2E-05 0.06 B (1) 2.3E-05 0.05 B (2) 3.3E-05 0.067 B (1) 8.6E-05 0.17 B (2) 1.4E-05 0.03 D-a (1) 8.0E-06 0.016 B (2) 8.3E-05 0.17 C (1) 5.6E-05 0.11 D-a (2) 6.6E-06 0.013 C-a (1) 5.3E-05 0.11 C (2) 2.2E-05 0.04 D-b (1) 1.8E-05 0.035 C-a (2) 1.2E-04 0.25 D-a (1) 1.5E-04 0.29 D-b (2) 1.0E-05 0.021 C-b (1) 1.6E-04 0.31 D-a (2) 1.1E-04 0.21 E (1) 6.6E-06 0.013 C-c (1) 2.4E-04 0.48 D-b (1) 1.2E-04 0.23 E (2) 1.3E-05 0.026 C-c (2) 1.7E-04 0.34 D-b (2) 1.5E-04 0.29 F (1) 4.4E-06 0.009 D-a (1) 3.0E-05 0.06 E (1) 5.7E-06 0.01 F (2) 6.4E-06 0.013 D-a (2) 2.7E-05 0.05 E (2) 9.7E-06 0.02 D-b (1) 3.5E-05 0.07 F (1) 5.2E-05 0.10 D-b (2) 3.8E-05 0.08 F (2) 1.2E-05 0.02 E (1) 3.1E-05 0.06 E (2) 2.8E-05 0.06 F (1) 3.0E-05 0.06 F (2) 2.1E-05 0.04 Average 6.8E-05 0.14 4.9E-05 0.10 1.4E-05 0.03 Deviation 6.0E-05 0.12 5.0E-05 0.10 9.8E-06 0.02

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169 Table 7-3. Occupancy times for workers at dr y product, shipping, and storage areas in the Florida phosphate chemical plants Job Classification Working Time at Dusty Area a (h y-1) Job Classification Working Time at Dusty Area a (h y-1) Dry Product Area Laborer Bobcat operator Granulator maintenance mechanic Reclaim process operator Operator trainee 1875 750 200 150 125 Shipping/Storage Area Laborer Reclaim operator Payload operator Chief operator Car loader Locomotive engineer Production & shipping supervisor 2000 1600 600 400 200 125 100 a Source (Birky et al . 1998)

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170 CHAPTER 8 CONCLUSIONS AND RECOMMENDATIONS 8.1 Conclusions Health risks to workers in the Florida phosphate industry due to inhalation of particulates containing radioactive materials have been studied. The heath risks have been investigated by assessing particle inha lation dose and estimating effective dose. Dose assessment due to particle inhalati on requires detailed k nowledge of airborne particle properties. Dose calculation is comp lex and requires use of specialized computer programs. The particle property database is used to populate input parameters to these software programs so that the calculated dos es are as realistic as possible. This study established database of particle size distri bution, particle shape, particle elemental composition, particle density, radionuclide conc entration in particles, and lung solubility of radionuclides in the particle s. In addition, dose calcula tion methods are presented for use with sampling data in i nhalation exposure assessment. All measurement data and dose calculation methods were fully integrated into an internal dosimetry assessment using ICRP 66 HRTM. In Chapter 2, particle properties includi ng size distribution, shape, density, and radionuclide concentration were characterized. The size distribution, density, and shape factor are each quite different from refe rence particle propert ies in ICRP 66 HRTM (ICRP 1994b). Therefore, adoption of dose coe fficients calculated for reference particle properties can skew dose to unr ealistic values. From the aspect of particle size distribution, it does not follow a lognormal patte rn in the phosphate i ndustry, which is not

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171 consistent with other industries. Dose calcula tion without considera tion of particle size distribution results in higher estimates because smaller particle sizes of 1 m AMAD (environmental) and 5 m AMAD (workplace) are used as defaults in ICRP 66 HRTM, in comparison with particles in the Florid a phosphate industry that do not follow this trend. The airborne particle concentrati on at each phosphate chemical processing plant and location varies widely to the magnitude of 3 orders. Chapter 3 explains dose cal culation methods using samp ling data in inhalation exposure and effective dose scaling factors. Traditionally, dose assessments made using internal dosimetry codes such as IMBA and LUDEP utilize air sampling information by assigning the radioactivity measured at each stage as concentrated at a single representative size central to the size interval . In this study, more realistic assumptions that the measured radioactivity distributes uniformly, linearly increases, or linearly decreases across the particle si ze interval for each impactor stage were explored. The concept of an effective dose scaling factor, SFBEB, is thus introduced whereby (1) the former approach can be used (which require s less computational effort using computer codes), and (2) the resulting va lues of effective dose per stage can then be rescaled to values appropriate to a linear radioactivity distribution per stage. For a majority of 238Useries radionuclides, particle size ranges, and absorption classes, differences in these two approaches are less than 10%, and thus no correc tions in effective dose per particle stage are needed. Significant corrections, however, we re noted in select cases. For uniform or linearly decreasing radio activity distributions, they are found in the 3rd and end-filter stage particles with SFBEB ranging from 1.11 to 1.53. When the cascade impactor measurements indicate a linear increase of activity across a given impactor-stage size

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172 range, values of SFBEB range from a high of 1.11 to a low of 0.85. In these cases, the inhalation dose coefficient varies non-linear ly across the particle size range, and the assumption of a mono-size distribution per impactor stage either underestimates ( SFBEB > 1) or overestimates ( SFBEB < 1) that stageÂ’s contribution to the worker effective dose. In Chapter 4, differences between IMBA and LUDEP were explained and effective dose scaling factors were generated using IMBA, the more recent program implementing the ICRP 66 HRTM. Under the same dosimet ry conditions, the IMBA program yields higher effective doses by a factor of 5 for 238U and 234U, 1.3 for 230Th, 1.9 for 226Ra, and 1.2 for 210Po for type F. The effective dose di fference between the two programs is less than 2% for most type S radionuclides. Th e difference is caused by biokinetic models. The IMBA program employs the latest bi okinetic models, while LUDEP employs biokinetic models in ICRP 30. The biokinetic models of uranium indicate that uranium retention and decay in bone and kidney is slightly higher in the new model, but much greater in liver and other soft tissues by an order of 1-2. The scaling factors generated using IMBA are almost identical to thos e calculated by LUDEP because the scaling factor depends more on the particle deposit ion and clearance mode l than the biokinetic model. In Chapter 5, inhalation doses were calculat ed via particle properties characterized in Chapter 1. In addition, dose sensitivity to radionuclide absorpti on type was studied. Effective dose rates to workers in Florida phosphate chemical plants due to airborne particle inhalation vary wide ly by 1 to 2 orders of magn itude depending on workplace airborne particle concentrations. Unde r the least conservative assumptions of radionuclide-specific lung solubility, all a ssessed effective doses are below the annual

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173 limits to the members of the general public (1 mSv y-1). In the most conservative cases, some 44%, 31%, and 15% of individual dose assessments yield worker doses above the annual dose limit. All doses are far below th e annual limit for radiation workers of 50 mSv y-1. Effective dose varies by a factor of 7 22 depending on the absorption types of the radionuclides within sampled aerosols. The study thus demonstrates the importance of site-specific particle solubility data in dose assessments. In Chapter 6, lung solubility of the ra dionuclides in particles from phosphate industry was determined by in vitro testing. Lung fluid was mimicked using chemicals, and lung conditions were maintained by controlling temperature and pH values of the mimicked lung fluid. Uranium, thorium, a nd lead are not dissolved rapidly with the surrounding matrix while most phosphate, the ma in component of surrounding matrix, is dissolved in 1 day. Roughly, uranium and lead are dissolved similarly to type M material defined in ICRP 66 HRTM and thorium disso lves like type S material. There are no noticeable correlations between solubility and product type or particle size. Chemical forms of the radionuclides and their fracti ons in the particles are supposed to be governing parameters of the solubili ty kinetics and mechanisms. In Chapter 7, all measurement data and dose calculation methods were integrated into a full internal dosimetry assessment. Particle inhalation dose ranges 0.04 – 0.48 mSv y-1 at granulator areas, 0.01 – 0.29 mSv y-1 at storage areas, and 0.01 – 0.07 mSv y-1 at shipping areas under the assumption of 2000 hours y-1 occupancy factor. Including an average external dose rate of 0.2 mSv y-1, the effective dose to workers is lower than dose limit to general public and extremely unlikel y to approach or exceed the occupational dose limit.

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174 8.2 Recommendations Most particle properties necessary for inhalation dose assessment were characterized in this study. However, there were limitations of some particle property measurements. Once they are characterized, it will strongly confirm the conclusion, “effective dose to workers is lower than dose limit to general public and extremely unlikely to approach or exceed the occupational dose limit” The element analysis study in Chapter 2 showed different elemental composition by particle size. Sulphur and silicon rather than phosphorus , the main component of the phosphate product, was dominant within small size particles. The result hints that radionuclides may be concentrated differently by particle size. The small mass of air samples collected at each cascade impactor stag e resulted in radioactivity loadings that were generally below detectable limits us ing high-efficiency ga mma-ray spectroscopy. Therefore, it was assumed that ra dionuclide concentration is the same in all size particles. A database of radioactivity li nked to particle size enables one to understand the kinetics of radionuclide concentration on particles and assess dose more accurately. The radioactivities of 238U series members other than 238U, 226Ra, and 210Pb were not measured due to their low gamma emissions . Therefore, concentrations of the other radionuclides were inferred by assuming secular equilibrium with the parent radionuclide. It is a reasonable approach for dose assessm ent. This assumption can be verified for other radionuclides, especially for 230Th and 210Po, using techniques other than gamma spectroscopy.

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175 APPENDIX A PARTICLE SAMPLING AND RADIOACTIVITY DATA USED FOR DOSE ASSESSMENT The radionuclide concentration of particles within an im pactor stage is inferred by the collected particle mass and facility-specifi c radioactivity data of airborne particles and settled particles. There is no difference between 226Ra and 238U concentrations for larger-sized particles ( 100 m) and smaller-sized (10 m) airborne particles collected by the high-volume sampler. Ther efore, average values are used to infer 226Ra and 238U concentrations in particles. It is found that the 210Pb concentration is higher fo r smaller-sized particles. Therefore, the 210Pb radioactivity measurement data of larger-sized airborne particles ( 100 m) is used to estimate radioactivity of larger-sized particles collected on impactor stages from 1 – 2 (particle size ranging 9.4 – 100 m). For smaller-sized particles on impactor stages from 3 – end f ilter (particle size ranging 0.03 – 9.4 m), the 210Pb data of smaller-sized airborne particles ( 10 m) is employed. There are two sources of 210Pb within airborne particles. The first source is 210Pb radioactivity that exists within the p hosphoric acid or bulk produc t and was carried into the airborne particulates during aerosol generation. The second source is 210Pb radioactivity from the aerosol attachment of ambient 222Rn progeny in the worker environment. Therefore, two valu es of the radioactivity data for 210Pb are given in the tables. The first one is inferre d from radioactivity data of airborne particles and denoted as 210PbBairB. The other is inferred from the data of settled particles and denoted as

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176 210PbBsettledB. For dose assessment purpose, 210Pb radioactivity inferred from settled particles can be used to estimate 210Po and 210Bi concentration in airborne particles.

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177Table A-1. Particle size dist ribution and radionuclide con centration (Granulator area) Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling Site Size Range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant A-a 21.50 100.0 8.78E-01 1.53E-032.24E-051.53E-041.71E-04DAP (1st sampling) 9.38 21.50 7.42E-01 8.20E-041.21E-058.23E-059.17E-05 3.57 9.38 3.27E-01 4.18E-046.14E-061.14E-044.67E-05 1.76 3.57 1.94E-01 1.82E-042.67E-064.97E-052.03E-05 0.97 1.76 1.02E-01 7.87E-051.16E-062.15E-058.80E-06 0.51 0.97 1.68E-02 1.40E-052.06E-073.82E-061.56E-06 0.26 0.51 3.42E-02 2.80E-054.11E-077.64E-063.13E-06 0.03 0.26 2.34E-02 7.87E-051.16E-062.15E-058.80E-06 Plant A-a 21.50 100.0 8.40E-01 1.50E-032.20E-051.50E-041.67E-04DAP (2nd sampling) 9.38 21.50 4.22E-01 4.79E-047.03E-064.80E-055.35E-05 3.57 9.38 6.07E-01 7.96E-041.17E-052.17E-048.90E-05 1.76 3.57 3.19E-01 3.07E-044.51E-068.38E-053.43E-05 0.97 1.76 1.68E-01 1.33E-041.96E-063.64E-051.49E-05 0.51 0.97 7.81E-02 6.66E-059.79E-071.82E-057.45E-06 0.26 0.51 8.36E-02 7.01E-051.03E-061.91E-057.84E-06 0.03 0.26 6.67E-02 2.30E-043.37E-066.27E-052.57E-05 Plant A-a 21.50 100.0 5.67E+00 9.84E-031.45E-049.87E-041.10E-03DAP (3rd sampling) 9.38 21.50 1.83E+00 2.03E-032.98E-052.03E-042.27E-04 3.57 9.38 7.08E-01 9.05E-041.33E-052.47E-041.01E-04 1.76 3.57 3.02E-01 2.83E-044.16E-067.73E-053.17E-05 0.97 1.76 1.75E-01 1.35E-041.99E-063.69E-051.51E-05 0.51 0.97 1.07E-01 8.91E-051.31E-062.43E-059.96E-06 0.26 0.51 1.09E-01 8.91E-051.31E-062.43E-059.96E-06 0.03 0.26 1.95E-02 6.54E-059.60E-071.78E-057.31E-06

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178Table A-1. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant A-b 21.50 100.0 7.14E-01 1.24E-031.82E-051.24E-041.39E-04DAP (1st sampling) 9.38 21.50 4.46E-01 4.93E-047.24E-064.94E-055.51E-05 3.57 9.38 4.24E-01 5.42E-047.97E-061.48E-046.06E-05 1.76 3.57 1.95E-01 1.83E-042.69E-064.99E-052.05E-05 0.97 1.76 9.90E-02 7.63E-051.12E-062.08E-058.54E-06 0.51 0.97 1.05E-01 8.75E-051.29E-062.39E-059.78E-06 0.26 0.51 1.09E-01 8.91E-051.31E-062.43E-059.96E-06 0.03 0.26 8.33E-02 2.80E-044.11E-067.64E-053.13E-05 Plant A-b 21.50 100.0 1.30E+00 2.27E-033.33E-052.27E-042.53E-04DAP (2nd sampling) 9.38 21.50 1.94E-01 2.15E-043.16E-062.15E-052.40E-05 3.57 9.38 3.69E-01 4.72E-046.93E-061.29E-045.28E-05 1.76 3.57 1.57E-01 1.47E-042.17E-064.02E-051.65E-05 0.97 1.76 1.30E-01 1.00E-041.47E-062.73E-051.12E-05 0.51 0.97 9.96E-02 8.28E-051.22E-062.26E-059.26E-06 0.26 0.51 9.90E-02 8.09E-051.19E-062.21E-059.05E-06 0.03 0.26 2.88E-02 9.68E-051.42E-062.64E-051.08E-05 Plant B 21.50 100.0 5.13E+00 5.76E-032.68E-041.43E-034.45E-04DAP (1st sampling) 9.38 21.50 4.51E-01 3.22E-041.50E-058.03E-052.49E-05 3.57 9.38 1.56E+00 1.29E-035.98E-057.53E-049.93E-05 1.76 3.57 6.41E-01 3.89E-041.81E-052.28E-043.00E-05 0.97 1.76 7.18E-01 3.58E-041.66E-052.10E-042.76E-05 0.51 0.97 5.54E-01 2.97E-041.38E-051.74E-042.30E-05 0.26 0.51 1.67E-01 8.83E-054.11E-065.17E-056.82E-06 0.03 0.26 1.05E-01 2.28E-041.06E-051.34E-041.76E-05

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179Table A-1. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant B 21.50 100.0 4.43E+00 4.97E-032.31E-041.24E-033.84E-04DAP (2nd sampling) 9.38 21.50 4.06E-01 2.90E-041.35E-057.22E-052.24E-05 3.57 9.38 1.75E+00 1.45E-036.72E-058.47E-041.12E-04 1.76 3.57 7.46E-01 4.52E-042.10E-052.65E-043.49E-05 0.97 1.76 6.15E-01 3.06E-041.42E-051.79E-042.37E-05 0.51 0.97 3.84E-01 2.06E-049.59E-061.21E-041.59E-05 0.26 0.51 1.37E-01 7.21E-053.36E-064.22E-055.57E-06 0.03 0.26 1.05E-01 2.28E-041.06E-051.34E-041.76E-05 Plant C-a 21.50 100.0 1.32E+00 2.62E-036.54E-054.65E-041.56E-04MAP (1st sampling) 9.38 21.50 1.84E-01 2.32E-045.79E-064.12E-051.38E-05 3.57 9.38 2.48E-01 3.62E-049.03E-066.42E-052.15E-05 1.76 3.57 1.34E-01 1.44E-043.59E-062.55E-058.55E-06 0.97 1.76 1.40E-01 1.23E-043.07E-062.18E-057.32E-06 0.51 0.97 1.72E-01 1.63E-044.06E-062.89E-059.69E-06 0.26 0.51 1.20E-01 1.12E-042.79E-061.98E-056.65E-06 0.03 0.26 8.15E-02 3.12E-047.79E-065.54E-051.86E-05 Plant C-a 21.50 100.0 1.09E+01 2.16E-025.38E-043.83E-031.28E-03MAP (2nd sampling) 9.38 21.50 2.72E+00 3.43E-038.56E-056.08E-042.04E-04 3.57 9.38 1.23E+00 1.80E-034.48E-053.18E-041.07E-04 1.76 3.57 2.88E-01 3.08E-047.69E-065.46E-051.83E-05 0.97 1.76 1.92E-01 1.69E-044.21E-062.99E-051.00E-05 0.51 0.97 2.30E-01 2.18E-045.44E-063.87E-051.30E-05 0.26 0.51 2.44E-01 2.27E-045.67E-064.03E-051.35E-05 0.03 0.26 1.15E-01 4.42E-041.10E-057.84E-052.63E-05

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180Table A-1. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant C-b 21.50 100.0 1.07E+01 2.11E-025.27E-043.75E-031.26E-03DAP (1st sampling) 9.38 21.50 1.45E+00 1.83E-034.56E-053.24E-041.09E-04 3.57 9.38 1.95E+00 2.85E-037.10E-055.05E-041.69E-04 1.76 3.57 4.66E-01 4.99E-041.24E-058.85E-052.97E-05 0.97 1.76 2.01E-01 1.77E-044.41E-063.13E-051.05E-05 0.51 0.97 2.00E-01 1.90E-044.73E-063.36E-051.13E-05 0.26 0.51 1.20E-01 1.12E-042.79E-061.98E-056.65E-06 0.03 0.26 1.96E-01 7.52E-041.88E-051.33E-044.47E-05 Plant C-c 21.50 100.0 1.97E+01 3.90E-0 29.73E-046.92E-032.32E-03GTSP (1st sampling) 9.38 21.50 3.36E+00 4.24E-031.06E-047.52E-042.52E-04 3.57 9.38 3.69E+00 5.38E-031.34E-049.54E-043.20E-04 1.76 3.57 1.36E+00 1.46E-033.63E-052.58E-048.66E-05 0.97 1.76 7.73E-01 6.80E-041.70E-051.21E-044.04E-05 0.51 0.97 3.90E-01 3.70E-049.23E-066.56E-052.20E-05 0.26 0.51 2.16E-01 2.01E-045.02E-063.57E-051.20E-05 0.03 0.26 8.27E-02 3.17E-047.90E-065.62E-051.88E-05 Plant C-c 21.50 100.0 1.26E+01 2.50E-0 26.24E-044.44E-031.49E-03GTSP (2nd sampling) 9.38 21.50 1.93E+00 2.44E-036.08E-054.32E-041.45E-04 3.57 9.38 3.84E+00 5.60E-031.40E-049.93E-043.33E-04 1.76 3.57 8.70E-01 9.31E-042.32E-051.65E-045.54E-05 0.97 1.76 4.13E-01 3.63E-049.06E-066.44E-052.16E-05 0.51 0.97 1.54E-01 1.46E-043.65E-062.59E-058.69E-06 0.26 0.51 1.57E-01 1.46E-043.65E-062.59E-058.70E-06 0.03 0.26 6.61E-02 2.53E-046.32E-064.49E-051.51E-05

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181Table A-1. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant D-a 21.50 100.0 2.47E+00 2.72E-031.52E-041.38E-034.11E-04DAP (1st sampling) 9.38 21.50 3.66E-01 2.57E-041.44E-051.30E-043.88E-05 3.57 9.38 5.59E-01 4.54E-042.54E-052.77E-046.85E-05 1.76 3.57 1.65E-01 9.84E-055.50E-066.01E-051.49E-05 0.97 1.76 1.24E-01 6.07E-053.39E-063.71E-059.16E-06 0.51 0.97 8.07E-02 4.26E-052.38E-062.60E-056.44E-06 0.26 0.51 6.84E-02 3.55E-051.98E-062.17E-055.36E-06 0.03 0.26 4.97E-02 1.06E-045.93E-066.48E-051.60E-05 Plant D-a 21.50 100.0 1.25E+00 1.38E-037.73E-056.99E-042.09E-04DAP (2nd sampling) 9.38 21.50 3.42E-01 2.41E-041.34E-051.22E-043.63E-05 3.57 9.38 2.14E-01 1.74E-049.73E-061.06E-042.63E-05 1.76 3.57 1.36E-01 8.13E-054.55E-064.97E-051.23E-05 0.97 1.76 1.76E-01 8.60E-054.81E-065.26E-051.30E-05 0.51 0.97 1.40E-01 7.41E-054.14E-064.53E-051.12E-05 0.26 0.51 8.45E-02 4.39E-052.45E-062.68E-056.63E-06 0.03 0.26 6.20E-02 1.32E-047.40E-068.09E-052.00E-05 Plant D-b 21.50 100.0 2.62E+00 2.89E-031.61E-041.46E-034.36E-04DAP (1st sampling) 9.38 21.50 6.58E-01 4.63E-042.59E-052.34E-046.99E-05 3.57 9.38 3.51E-01 2.85E-041.59E-051.74E-044.30E-05 1.76 3.57 2.18E-01 1.30E-047.27E-067.95E-051.97E-05 0.97 1.76 2.51E-01 1.23E-046.88E-067.52E-051.86E-05 0.51 0.97 2.02E-01 1.07E-045.96E-066.52E-051.61E-05 0.26 0.51 1.06E-01 5.52E-053.08E-063.37E-058.34E-06 0.03 0.26 4.77E-02 1.02E-045.68E-066.21E-051.54E-05

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182Table A-1. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant D-b 21.50 100.0 1.93E+00 2.13E-031.19E-041.08E-033.22E-04DAP (2nd sampling) 9.38 21.50 1.28E+00 9.00E-045.03E-054.55E-041.36E-04 3.57 9.38 4.92E-01 4.00E-042.23E-052.44E-046.04E-05 1.76 3.57 1.92E-01 1.15E-046.41E-067.01E-051.73E-05 0.97 1.76 2.16E-01 1.06E-045.91E-066.47E-051.60E-05 0.51 0.97 1.24E-01 6.57E-053.67E-064.02E-059.93E-06 0.26 0.51 1.20E-01 6.24E-053.49E-063.81E-059.42E-06 0.03 0.26 6.92E-02 1.48E-048.25E-069.02E-052.23E-05 Plant E 21.50 100.0 2.69E+00 2.95E-031.79E-043.52E-031.09E-03MAP (1st sampling) 9.38 21.50 8.23E-01 5.74E-043.49E-056.84E-042.12E-04 3.57 9.38 8.12E-01 6.55E-043.98E-051.08E-032.42E-04 1.76 3.57 2.11E-01 1.25E-047.59E-062.06E-044.61E-05 0.97 1.76 1.18E-01 5.75E-053.49E-069.49E-052.12E-05 0.51 0.97 5.48E-02 2.88E-051.75E-064.75E-051.06E-05 0.26 0.51 7.08E-02 3.65E-052.22E-066.03E-051.35E-05 0.03 0.26 2.14E-02 4.53E-052.75E-067.49E-051.67E-05 Plant E 21.50 100.0 1.79E+00 1.96E-031.19E-042.34E-037.25E-04MAP (2nd sampling) 9.38 21.50 3.73E-01 2.60E-041.58E-053.10E-049.60E-05 3.57 9.38 7.56E-01 6.09E-043.70E-051.01E-032.25E-04 1.76 3.57 1.88E-01 1.11E-046.74E-061.83E-044.10E-05 0.97 1.76 1.01E-01 4.91E-052.98E-068.11E-051.81E-05 0.51 0.97 7.98E-02 4.18E-052.54E-066.91E-051.55E-05 0.26 0.51 8.03E-02 4.14E-052.52E-066.84E-051.53E-05 0.03 0.26 2.98E-02 6.32E-053.84E-061.04E-042.33E-05

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183Table A-1. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant F 21.50 100.0 1.48E-01 1.00E-043.09E-065.32E-055.32E-05DAP (1st sampling) 9.38 21.50 6.76E-02 2.93E-059.01E-071.55E-051.55E-05Scrubber system 3.57 9.38 4.70E-02 2.35E-057.24E-071.84E-051.25E-05operation 1.76 3.57 3.12E-02 1.15E-053.53E-078.99E-066.08E-06 0.97 1.76 3.93E-02 1.19E-053.65E-079.29E-066.28E-06 0.51 0.97 6.11E-02 1.99E-056.13E-071.56E-051.05E-05 0.26 0.51 1.58E-01 5.05E-051.55E-063.96E-052.68E-05 0.03 0.26 2.55E-01 3.35E-041.03E-052.63E-041.78E-04 Plant F 21.50 100.0 1.00E-01 6.83E-052.10E-063.62E-053.62E-05DAP (2nd sampling) 9.38 21.50 4.42E-02 1.91E-055.89E-071.01E-051.01E-05Scrubber system 3.57 9.38 3.63E-02 1.82E-055.60E-071.42E-059.63E-06operation 1.76 3.57 3.54E-02 1.30E-054.01E-071.02E-056.89E-06 0.97 1.76 2.91E-02 8.80E-062.71E-076.89E-064.66E-06 0.51 0.97 5.17E-02 1.68E-055.18E-071.32E-058.92E-06 0.26 0.51 4.96E-02 1.59E-054.89E-071.24E-058.41E-06 0.03 0.26 1.87E-01 2.46E-047.57E-061.93E-041.30E-04

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184Table A-2. Particle size di stribution and radionuclide c oncentration (Storage area) Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant A-a 21.50 100.0 1.53E-01 2.66E-043.91E-062.67E-052.98E-05MAP (1st sampling) 9.38 21.50 6.89E-02 7.62E-051.12E-067.65E-068.52E-06Conveyor and reclaimer 3.57 9.38 1.32E-01 1.69E-042.48E-064.61E-051.89E-05operation 1.76 3.57 4.86E-02 4.56E-056.70E-071.24E-055.10E-06Closed door 0.97 1.76 4.55E-02 3.51E-055.16E-079.59E-063.93E-06 0.51 0.97 2.34E-02 1.94E-052.86E-075.31E-062.17E-06 0.26 0.51 4.02E-02 3.29E-054.83E-078.98E-063.68E-06 0.03 0.26 3.05E-02 1.02E-041.50E-062.80E-051.14E-05 Plant A-a 21.50 100.0 5.34E-02 9.85E-041.45E-059.88E-051.10E-04MAP (2nd sampling) 9.38 21.50 6.10E-02 1.32E-041.94E-061.32E-051.47E-05Conveyor and reclaimer 3.57 9.38 7.48E-02 3.92E-045.77E-061.07E-044.39E-05operation 1.76 3.57 5.81E-02 9.47E-051.39E-062.59E-051.06E-05Closed door 0.97 1.76 1.01E-01 4.48E-056.59E-071.22E-055.01E-06 0.51 0.97 3.07E-01 6.22E-059.14E-071.70E-056.96E-06 0.26 0.51 1.19E-01 4.99E-057.33E-071.36E-055.58E-06 0.03 0.26 5.67E-01 1.80E-042.64E-064.90E-052.01E-05 Plant A-b 21.50 100.0 3.75E-03 6.51E-069.57E-086.53E-077.28E-07DAP (1st sampling) 9.38 21.50 1.10E-02 1.22E-051.79E-071.22E-061.37E-06No activity 3.57 9.38 1.02E-02 1.30E-051.91E-073.56E-061.46E-06Closed door 1.76 3.57 1.47E-02 1.38E-052.03E-073.78E-061.55E-06 0.97 1.76 1.79E-02 1.38E-052.03E-073.78E-061.55E-06 0.51 0.97 1.86E-02 1.55E-052.27E-074.22E-061.73E-06 0.26 0.51 2.79E-02 2.28E-053.35E-076.22E-062.55E-06 0.03 0.26 1.04E-02 3.50E-055.14E-079.56E-063.91E-06

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185Table A-2. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant A-b 21.50 100.0 6.44E-01 1.12E-031.64E-051.12E-041.25E-04DAP (2nd sampling) 9.38 21.50 1.22E-01 1.35E-041.98E-061.35E-051.51E-05Payloader operation 3.57 9.38 2.22E-01 2.84E-044.17E-067.74E-053.17E-05Open door 1.76 3.57 9.06E-02 8.50E-051.25E-062.32E-059.51E-06 0.97 1.76 8.58E-02 6.62E-059.73E-071.81E-057.40E-06 0.51 0.97 1.18E-01 9.82E-051.44E-062.68E-051.10E-05 0.26 0.51 1.22E-01 9.95E-051.46E-062.72E-051.11E-05 0.03 0.26 2.68E-02 8.99E-051.32E-062.45E-051.01E-05 Plant B 21.50 100.0 1.19E+00 1.33E-036.20E-053.32E-041.03E-04MAP (1st sampling) 9.38 21.50 1.50E-01 1.07E-044.98E-062.67E-058.27E-06Payloader operation 3.57 9.38 1.31E-01 1.08E-045.04E-066.35E-058.38E-06Open door 1.76 3.57 1.06E-01 6.42E-052.99E-063.76E-054.96E-06 0.97 1.76 8.60E-02 4.28E-051.99E-062.51E-053.31E-06 0.51 0.97 2.82E-01 1.51E-047.03E-068.86E-051.17E-05 0.26 0.51 1.93E-01 1.02E-044.73E-065.96E-057.86E-06 0.03 0.26 3.70E-02 8.03E-053.74E-064.70E-056.20E-06 Plant B 21.50 100.0 1.81E+00 2.03E-039.43E-055.05E-041.57E-04MAP (2nd sampling) 9.38 21.50 1.00E-01 7.18E-053.34E-061.79E-055.54E-06Payloader operation 3.57 9.38 1.62E-01 1.34E-046.22E-067.83E-051.03E-05Open door 1.76 3.57 7.59E-02 4.60E-052.14E-062.69E-053.55E-06 0.97 1.76 5.42E-02 2.70E-051.26E-061.58E-052.09E-06 0.51 0.97 5.48E-02 2.94E-051.37E-061.72E-052.27E-06 0.26 0.51 4.18E-02 2.21E-051.03E-061.29E-051.71E-06 0.03 0.26 1.47E-02 3.19E-051.48E-061.87E-052.46E-06

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186Table A-2. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant C 21.50 100.0 6.36E-01 1.26E-0 33.14E-052.24E-047.50E-05GTSP (1st sampling) 9.38 21.50 8.47E-02 1.07E-042.66E-061.89E-056.35E-06Conveyor operation 3.57 9.38 1.89E-01 2.75E-046.86E-064.88E-051.64E-05Open door 1.76 3.57 6.95E-02 7.44E-051.85E-061.32E-054.42E-06 0.97 1.76 2.04E-01 1.80E-044.48E-063.19E-051.07E-05 0.51 0.97 5.59E-01 5.31E-041.32E-059.41E-053.16E-05 0.26 0.51 3.31E-01 3.08E-047.69E-065.47E-051.83E-05 0.03 0.26 5.04E-02 1.93E-044.81E-063.42E-051.15E-05 Plant C 21.50 100.0 6.73E-01 1.33E-0 33.32E-052.36E-047.93E-05GTSP (2nd sampling) 9.38 21.50 6.95E-02 8.76E-052.19E-061.55E-055.21E-06Conveyor operation 3.57 9.38 1.08E-01 1.58E-043.93E-062.79E-059.37E-06Open door 1.76 3.57 5.03E-02 5.39E-051.34E-069.55E-063.20E-06 0.97 1.76 4.66E-02 4.10E-051.02E-067.27E-062.44E-06 0.51 0.97 1.05E-01 9.97E-052.49E-061.77E-055.93E-06 0.26 0.51 1.27E-01 1.18E-042.95E-062.10E-057.03E-06 0.03 0.26 2.48E-02 9.49E-052.37E-061.68E-055.64E-06 Plant D-a 21.50 100.0 3.98E+00 4.39E-032.45E-042.22E-036.63E-04DAP (1st sampling) 9.38 21.50 6.37E-01 4.48E-042.50E-052.26E-046.76E-05Reclaimer operation 3.57 9.38 2.20E+00 1.79E-039.97E-051.09E-032.70E-04Closed door 1.76 3.57 1.22E+00 7.27E-044.06E-054.44E-041.10E-04 0.97 1.76 7.09E-01 3.48E-041.94E-052.12E-045.25E-05 0.51 0.97 8.33E-01 4.40E-042.46E-052.69E-046.65E-05 0.26 0.51 7.45E-01 3.87E-042.16E-052.36E-045.84E-05 0.03 0.26 2.92E-01 6.24E-043.49E-053.81E-049.43E-05

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187Table A-2. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant D-a 21.50 100.0 5.21E+00 5.75E-033.21E-042.91E-038.68E-04DAP (2nd sampling) 9.38 21.50 1.02E+00 7.14E-043.99E-053.61E-041.08E-04Reclaimer operation 3.57 9.38 3.45E+00 2.80E-031.56E-041.71E-034.23E-04Closed door 1.76 3.57 1.18E+00 7.02E-043.92E-054.29E-041.06E-04 0.97 1.76 6.95E-01 3.40E-041.90E-052.08E-045.14E-05 0.51 0.97 3.01E-01 1.59E-048.90E-069.73E-052.41E-05 0.26 0.51 2.88E-01 1.50E-048.36E-069.14E-052.26E-05 0.03 0.26 7.86E-02 1.68E-049.37E-061.02E-042.53E-05 Plant D-b 21.50 100.0 1.06E+01 1.17E-026.54E-045.92E-031.77E-03MAP (1st sampling) 9.38 21.50 2.02E+00 1.42E-037.92E-057.17E-042.14E-04Payloader operation 3.57 9.38 2.76E+00 2.24E-031.25E-041.37E-033.38E-04Closed door 1.76 3.57 1.12E+00 6.65E-043.72E-054.07E-041.00E-04 0.97 1.76 5.39E-01 2.64E-041.48E-051.61E-043.99E-05 0.51 0.97 2.68E-01 1.42E-047.92E-068.66E-052.14E-05 0.26 0.51 1.94E-01 1.01E-045.62E-066.15E-051.52E-05 0.03 0.26 8.93E-02 1.91E-041.06E-051.16E-042.88E-05 Plant D-b 21.50 100.0 4.05E+00 4.47E-032.50E-042.26E-036.75E-04MAP (2nd sampling) 9.38 21.50 1.27E+00 8.94E-044.99E-054.52E-041.35E-04Payloader operation 3.57 9.38 6.58E+00 5.35E-032.99E-043.27E-038.07E-04Closed door 1.76 3.57 1.65E+00 9.81E-045.48E-055.99E-041.48E-04 0.97 1.76 7.09E-01 3.48E-041.94E-052.12E-045.25E-05 0.51 0.97 3.55E-01 1.87E-041.05E-051.15E-042.83E-05 0.26 0.51 3.08E-01 1.60E-048.94E-069.78E-052.42E-05 0.03 0.26 1.22E-01 2.61E-041.46E-051.59E-043.94E-05

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188Table A-2. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant E 21.50 100.0 6.21E-02 6.80E-054.13E-068.11E-052.51E-05MAP (1st sampling) 9.38 21.50 3.98E-02 2.78E-051.69E-063.31E-051.03E-05Payloader operation 3.57 9.38 5.55E-03 4.47E-062.72E-077.38E-061.65E-06Open door 1.76 3.57 4.43E-02 2.62E-051.59E-064.33E-059.67E-06 0.97 1.76 3.28E-02 1.60E-059.70E-072.64E-055.90E-06 0.51 0.97 5.24E-02 2.75E-051.67E-064.54E-051.01E-05 0.26 0.51 4.89E-02 2.52E-051.53E-064.17E-059.32E-06 0.03 0.26 9.80E-03 2.08E-051.26E-063.43E-057.67E-06 Plant E 21.50 100.0 4.83E-01 5.29E-043.21E-056.30E-041.95E-04MAP (2nd sampling) 9.38 21.50 7.86E-02 5.48E-053.33E-066.53E-052.02E-05Payloader operation 3.57 9.38 7.92E-02 6.39E-053.88E-061.05E-042.36E-05Open door 1.76 3.57 3.16E-02 1.87E-051.14E-063.09E-056.91E-06 0.97 1.76 3.85E-02 1.87E-051.14E-063.09E-056.91E-06 0.51 0.97 7.58E-02 3.98E-052.42E-066.57E-051.47E-05 0.26 0.51 7.07E-02 3.64E-052.21E-066.02E-051.35E-05 0.03 0.26 1.78E-02 3.78E-052.29E-066.24E-051.40E-05 Plant F 21.50 100.0 2.03E+01 1.38E-024.26E-047.33E-037.33E-03DAP (1st sampling) 9.38 21.50 1.53E+00 6.65E-042.05E-053.52E-043.52E-04Payloader operation 3.57 9.38 3.33E-01 1.67E-045.13E-061.31E-048.83E-05Open door 1.76 3.57 1.03E-01 3.77E-051.16E-062.95E-052.00E-05 0.97 1.76 1.09E-01 3.31E-051.02E-062.59E-051.75E-05 0.51 0.97 1.08E-01 3.52E-051.09E-062.76E-051.87E-05 0.26 0.51 1.67E-01 5.36E-051.65E-064.20E-052.84E-05 0.03 0.26 3.16E-02 4.15E-051.28E-063.25E-052.20E-05

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189Table A-2. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant F 21.50 100.0 3.49E+00 2.37E-037.30E-051.26E-031.26E-03DAP (2nd sampling) 9.38 21.50 2.91E-01 1.26E-043.88E-066.68E-056.68E-05Payloader operation 3.57 9.38 1.44E-01 7.23E-052.23E-065.67E-053.83E-05Open door 1.76 3.57 5.73E-02 2.10E-056.48E-071.65E-051.11E-05 0.97 1.76 5.89E-02 1.78E-055.48E-071.39E-059.42E-06 0.51 0.97 5.46E-02 1.78E-055.48E-071.39E-059.42E-06 0.26 0.51 5.32E-02 1.70E-055.24E-071.33E-059.02E-06

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190Table A-3. Particle size di stribution and radionuclide c oncentration (Shipping area) Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant A-a 21.50 100.0 1.69E+00 2.94E-034.31E-052.95E-043.28E-04MAP (1st sampling) 9.38 21.50 6.70E-01 7.41E-041.09E-057.43E-058.29E-05Rail car loading 3.57 9.38 3.06E-01 3.91E-045.75E-061.07E-044.37E-05 1.76 3.57 7.98E-02 7.49E-051.10E-062.04E-058.37E-06 0.97 1.76 4.85E-02 3.74E-055.50E-071.02E-054.19E-06 0.51 0.97 3.17E-02 2.64E-053.88E-077.21E-062.95E-06 0.26 0.51 3.93E-02 3.22E-054.72E-078.78E-063.60E-06 0.03 0.26 2.37E-02 7.97E-051.17E-062.18E-058.91E-06 Plant A-a 21.50 100.0 1.96E+00 3.40E-035.00E-053.41E-043.81E-04MAP (2nd sampling) 9.38 21.50 3.48E-01 3.84E-045.65E-063.86E-054.30E-05Rail car loading 3.57 9.38 3.97E-01 5.08E-047.46E-061.39E-045.68E-05 1.76 3.57 1.18E-01 1.11E-041.63E-063.03E-051.24E-05 0.97 1.76 6.99E-02 5.39E-057.92E-071.47E-056.02E-06 0.51 0.97 4.98E-03 4.14E-066.09E-081.13E-064.63E-07 0.26 0.51 4.56E-02 3.73E-055.48E-071.02E-054.17E-06 0.03 0.26 3.73E-02 1.25E-041.84E-063.42E-051.40E-05 Plant A-b 21.50 100.0 4.27E-03 2.83E-044.15E-062.84E-053.16E-05DAP (1st sampling) 9.38 21.50 7.10E-03 8.70E-051.28E-068.73E-069.73E-06Rail car loading 3.57 9.38 4.98E-03 9.45E-051.39E-062.58E-051.06E-05 1.76 3.57 1.61E-02 5.22E-057.67E-071.43E-055.84E-06 0.97 1.76 5.57E-02 1.24E-051.83E-073.39E-061.39E-06 0.51 0.97 7.39E-02 4.14E-066.09E-081.13E-064.63E-07 0.26 0.51 7.87E-02 5.80E-068.53E-081.58E-066.49E-07 0.03 0.26 1.63E-01 1.43E-052.11E-073.92E-061.60E-06

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191Table A-3. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant B 21.50 100.0 1.77E+00 1.99E-039.24E-054.94E-041.53E-04MAP (1st sampling) 9.38 21.50 2.93E-01 2.09E-049.72E-065.20E-051.61E-05Rail car loading 3.57 9.38 2.38E-01 1.96E-049.13E-061.15E-041.52E-05 1.76 3.57 1.06E-01 6.43E-052.99E-063.77E-054.97E-06 0.97 1.76 9.45E-02 4.71E-052.19E-062.76E-053.63E-06 0.51 0.97 1.17E-01 6.31E-052.93E-063.69E-054.87E-06 0.26 0.51 7.21E-02 3.81E-051.77E-062.23E-052.94E-06 0.03 0.26 2.83E-02 6.15E-052.86E-063.60E-054.75E-06 Plant B 21.50 100.0 5.04E+00 5.65E-032.63E-041.41E-034.37E-04MAP (2nd sampling) 9.38 21.50 8.13E-01 5.81E-042.70E-051.45E-044.49E-05Rail car loading 3.57 9.38 7.10E-01 5.86E-042.73E-053.43E-044.53E-05 1.76 3.57 1.17E-01 7.11E-053.31E-064.16E-055.49E-06 0.97 1.76 7.17E-02 3.57E-051.66E-062.09E-052.76E-06 0.51 0.97 5.66E-02 3.04E-051.41E-061.78E-052.35E-06 0.26 0.51 5.81E-02 3.07E-051.43E-061.80E-052.37E-06 0.03 0.26 1.63E-02 3.54E-051.65E-062.07E-052.73E-06 Plant D-a 21.50 100.0 6.55E-02 7.23E-054.04E-063.65E-051.09E-05DAP (1st sampling) 9.38 21.50 2.33E-02 1.64E-059.16E-078.29E-062.48E-06Rail car loading 3.57 9.38 3.00E-02 2.44E-051.36E-061.49E-053.68E-06 1.76 3.57 4.09E-02 2.44E-051.36E-061.49E-053.68E-06 0.97 1.76 2.71E-02 1.33E-057.43E-078.13E-062.01E-06 0.51 0.97 3.61E-02 1.91E-051.07E-061.16E-052.88E-06 0.26 0.51 3.33E-02 1.73E-059.66E-071.06E-052.61E-06 0.03 0.26 2.94E-02 6.28E-053.51E-063.84E-059.48E-06

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192Table A-3. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant D-a 21.50 100.0 1.18E-01 1.30E-047.28E-066.59E-051.97E-05DAP (2nd sampling) 9.38 21.50 2.36E-01 1.66E-049.25E-068.37E-052.50E-05Rail car loading 3.57 9.38 1.10E-01 8.96E-055.01E-065.48E-051.35E-05 1.76 3.57 3.72E-02 2.22E-051.24E-061.35E-053.35E-06 0.97 1.76 2.73E-02 1.34E-057.47E-078.17E-062.02E-06 0.51 0.97 3.10E-02 1.64E-059.17E-071.00E-052.48E-06 0.26 0.51 2.34E-02 1.22E-056.79E-077.42E-061.84E-06 0.03 0.26 1.44E-02 3.07E-051.71E-061.87E-054.63E-06 Plant D-b 21.50 100.0 5.85E-01 1.87E-031.04E-049.43E-042.82E-04DAP (1st sampling) 9.38 21.50 1.38E-01 4.71E-042.63E-052.38E-047.11E-05Truck loading 3.57 9.38 2.36E-01 2.48E-041.39E-051.52E-043.75E-05 1.76 3.57 1.52E-01 4.76E-052.66E-062.91E-057.18E-06 0.97 1.76 1.01E-01 2.38E-051.33E-061.45E-053.59E-06 0.51 0.97 1.14E-01 1.68E-059.37E-071.02E-052.53E-06 0.26 0.51 9.13E-02 2.04E-051.14E-061.25E-053.08E-06 0.03 0.26 3.83E-02 5.06E-052.83E-063.09E-057.64E-06 Plant D-b 21.50 100.0 8.21E-01 9.06E-045.06E-054.58E-041.37E-04DAP (2nd sampling) 9.38 21.50 9.78E-02 6.87E-053.84E-063.47E-051.04E-05Truck loading 3.57 9.38 3.10E-01 2.52E-041.41E-051.54E-043.81E-05 1.76 3.57 4.96E-02 2.96E-051.65E-061.81E-054.47E-06 0.97 1.76 4.15E-02 2.03E-051.14E-061.24E-053.07E-06 0.51 0.97 4.96E-02 2.62E-051.46E-061.60E-053.96E-06 0.26 0.51 4.15E-02 2.16E-051.21E-061.32E-053.26E-06 0.03 0.26 1.19E-02 2.54E-051.42E-061.55E-053.83E-06

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193Table A-3. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant E 21.50 100.0 3.26E-02 1.65E-041.00E-051.97E-046.11E-05MAP (1st sampling) 9.38 21.50 8.78E-02 4.40E-052.67E-065.25E-051.63E-05Rail car loading 3.57 9.38 9.56E-02 3.98E-052.42E-066.58E-051.47E-05 1.76 3.57 5.66E-02 1.65E-051.00E-062.72E-056.08E-06 0.97 1.76 5.69E-02 1.35E-058.18E-072.23E-054.98E-06 0.51 0.97 1.01E-01 2.46E-051.49E-064.05E-059.07E-06 0.26 0.51 1.29E-01 2.22E-051.35E-063.66E-058.18E-06 0.03 0.26 3.08E-01 3.38E-052.06E-065.59E-051.25E-05 Plant E 21.50 100.0 2.21E-01 2.43E-041.47E-052.89E-048.96E-05MAP (2nd sampling) 9.38 21.50 1.06E-01 7.37E-054.47E-068.78E-052.72E-05Rail car loading 3.57 9.38 2.90E-02 2.34E-051.42E-063.86E-058.63E-06 1.76 3.57 2.73E-02 1.62E-059.82E-072.67E-055.97E-06 0.97 1.76 5.91E-02 2.88E-051.75E-064.75E-051.06E-05 0.51 0.97 2.06E-02 1.08E-056.55E-071.78E-053.98E-06 0.26 0.51 1.60E-01 8.27E-055.02E-061.36E-043.05E-05 0.03 0.26 3.39E-02 7.19E-054.36E-061.19E-042.65E-05 Plant F 21.50 100.0 2.45E-01 1.66E-045.12E-068.81E-058.81E-05DAP (1st sampling) 9.38 21.50 4.64E-02 2.01E-056.18E-071.06E-051.06E-05Rail car loading 3.57 9.38 6.69E-02 3.35E-051.03E-062.62E-051.77E-05 1.76 3.57 3.46E-02 1.27E-053.91E-079.95E-066.73E-06 0.97 1.76 4.55E-02 1.37E-054.23E-071.08E-057.28E-06 0.51 0.97 5.48E-02 1.79E-055.50E-071.40E-059.46E-06 0.26 0.51 4.77E-02 1.53E-054.70E-071.20E-058.09E-06 0.03 0.26 1.58E-02 2.08E-056.40E-071.63E-051.10E-05

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194Table A-3. Continued Particle Size Distribution Ra dionuclide Concentration (Bq m-3) Sampling site Size range (m) Concentration M/logdBpB (g/L)238U 226Ra 210PbBairB 210PbBsettledB Remarks Plant F 21.50 100.0 1.53E-01 1.04E-043.21E-065.52E-055.52E-05DAP (2nd sampling) 9.38 21.50 3.73E-02 1.62E-054.98E-078.57E-068.57E-06Rail car loading 3.57 9.38 7.50E-02 3.76E-051.16E-062.94E-051.99E-05 1.76 3.57 6.28E-02 2.31E-057.10E-071.81E-051.22E-05 0.97 1.76 7.20E-02 2.18E-056.70E-071.70E-051.15E-05 0.51 0.97 8.45E-02 2.75E-058.47E-072.16E-051.46E-05 0.26 0.51 4.65E-02 1.49E-054.58E-071.17E-057.88E-06 0.03 0.26 2.83E-02 3.72E-051.15E-062.91E-051.97E-05

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195 APPENDIX B EFFECTIVE DOSE SCALING FACTORS APPLICABLE TO PARTICLES IN THE FLORIDA PHOSPHATE CHEMICAL PLANTS The effective dose scaling factor, SFBEB, is the ratio of the e ffective dose determined under a linear radioactivity distribution to that under a single mono-size radioactivity distribution for each stage of the cascade imp actor. When inhalation dose coefficients are given as a function of particle size, eff ective dose scaling factors are calculated by definition: 1 1 2()N i lineari E monoNEfi E SF EE , Eq. B-1 where EBiB is effective dose for particle size at ith sub stage and f(i) represent functions of linear radioactivity di stributions. In the present study, (1) uniform distribution, (2) linearly decreasing distribution with activity ratios (AR) of 2:1 and 5:1 (shallow and steep slopes, respectively), and (3) linearly in creasingly distribution with ARs of 1:2 and 1:5 are considered. Linear ra dioactivity distributions are ap proximated as a series of 10 sub-stages each containing a different fracti on of the total activity measured at that impactor stage. Effective dose scaling factors specifically a pplicable to the particles in the Florida phosphate chemical plants are generated. Values of SFBEB given here are specific to particle size ranges of the University of Wa shington Mark III cascade impactor at an air flow rate of 15 L min-1, particle density of 1.6 g cm-3, and unity particle shape factor.

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196 Table B-1. Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for uni form radioactivity distribution Effective Dose Scaling Factor SFBE Ba Absorption Type Impactor Stage 238U 234U 230Th 226Ra 210Pb 210Po F 1 1.02 1.02 1.02 1.03 1.02 1.02 2 1.01 1.01 1.01 1.01 1.01 1.01 3 1.00 1.00 1.01 1.00 1.00 1.00 4 0.98 0.98 0.97 0.98 0.98 0.98 5 1.00 0.99 0.99 1.00 1.00 1.00 6 1.03 1.03 1.03 1.04 1.04 1.03 7 1.04 1.04 1.04 1.04 1.04 1.04 F 1.07 1.07 1.07 1.07 1.07 1.07 M 1 1.03 1.02 1.02 1.03 1.02 1.02 2 1.03 1.04 1.02 1.04 1.02 1.05 3 1.15 1.13 1.09 1.13 1.08 1.15 4 0.96 0.96 0.97 0.96 0.97 0.96 5 1.01 1.01 0.99 1.01 0.99 1.01 6 1.03 1.03 1.02 1.04 1.03 1.04 7 1.02 1.02 1.03 1.02 1.03 1.02 F 1.08 1.08 1.07 1.09 1.07 1.08 S 1 1.02 1.02 1.02 1.02 1.02 1.05 2 1.02 1.02 1.03 1.03 1.02 1.07 3 1.13 1.13 0.95 1.13 1.08 1.14 4 1.04 1.02 0.96 1.02 0.97 0.96 5 0.99 0.99 0.99 0.99 0.99 1.01 6 1.02 1.03 1.02 1.03 1.02 1.04 7 1.03 1.03 1.03 1.03 1.03 1.02 F 1.08 1.07 1.07 1.07 1.07 1.08 a Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).

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197 Table B-2. Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly decreasing radioac tivity distribution (AR = 2:1) Effective Dose Scaling Factor SFBE Ba Absorption Type Impactor Stage 238U 234U 230Th 226Ra 210Pb 210Po F 1 1.03 1.03 1.03 1.04 1.03 1.03 2 1.04 1.04 1.04 1.04 1.04 1.04 3 1.03 1.03 1.03 1.02 1.03 1.03 4 0.98 0.97 0.97 0.97 0.97 0.97 5 0.96 0.96 0.96 0.96 0.96 0.96 6 0.99 0.99 0.99 0.99 0.99 0.99 7 1.05 1.05 1.05 1.05 1.05 1.05 F 1.18 1.17 1.18 1.18 1.18 1.18 M 1 1.04 1.03 1.03 1.04 1.03 1.04 2 1.07 1.08 1.05 1.08 1.05 1.10 3 1.28 1.25 1.18 1.25 1.15 1.29 4 0.96 0.96 0.99 0.96 0.97 0.95 5 0.98 0.97 0.97 0.97 0.97 0.97 6 1.02 1.02 1.00 1.03 1.00 1.02 7 1.05 1.05 1.05 1.05 1.05 1.05 F 1.19 1.19 1.17 1.19 1.18 1.19 S 1 1.03 1.03 1.03 1.03 1.03 1.06 2 1.05 1.05 1.06 1.06 1.04 1.11 3 1.23 1.24 1.03 1.23 1.15 1.27 4 1.04 1.02 0.98 1.02 0.99 0.96 5 0.97 0.97 0.97 0.97 0.97 0.97 6 1.01 1.01 1.00 1.01 1.00 1.03 7 1.05 1.05 1.05 1.05 1.05 1.05 F 1.18 1.18 1.18 1.18 1.17 1.19 a Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).

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198 Table B-3. Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly decreasing radioac tivity distribution (AR = 5:1) Effective Dose Scaling Factor SFBE Ba Absorption Type Impactor Stage 238U 234U 230Th 226Ra 210Pb 210Po F 1 1.04 1.04 1.04 1.05 1.04 1.05 2 1.06 1.06 1.06 1.06 1.06 1.06 3 1.06 1.06 1.06 1.05 1.05 1.05 4 0.97 0.97 0.97 0.97 0.97 0.97 5 0.93 0.93 0.93 0.93 0.93 0.93 6 0.95 0.95 0.95 0.95 0.95 0.95 7 1.06 1.06 1.06 1.06 1.05 1.06 F 1.28 1.28 1.28 1.28 1.28 1.28 M 1 1.05 1.05 1.04 1.05 1.05 1.05 2 1.11 1.12 1.09 1.12 1.08 1.14 3 1.40 1.38 1.27 1.38 1.23 1.43 4 0.96 0.96 1.00 0.96 0.98 0.95 5 0.95 0.94 0.95 0.94 0.94 0.94 6 1.01 1.01 0.97 1.02 0.97 1.01 7 1.08 1.08 1.07 1.08 1.07 1.08 F 1.30 1.30 1.28 1.30 1.28 1.30 S 1 1.04 1.04 1.04 1.04 1.04 1.07 2 1.08 1.08 1.10 1.09 1.07 1.16 3 1.33 1.34 1.11 1.34 1.22 1.41 4 1.04 1.02 1.00 1.02 1.00 0.95 5 0.95 0.95 0.95 0.95 0.96 0.94 6 0.99 0.99 0.98 0.99 0.97 1.01 7 1.07 1.07 1.07 1.07 1.07 1.08 F 1.29 1.28 1.28 1.28 1.27 1.30 a Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).

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199 Table B-4. Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly increasing radioac tivity distribution (AR = 1:2) Effective Dose Scaling Factor SFBE Ba Absorption Type Impactor Stage 238U 234U 230Th 226Ra 210Pb 210Po F 1 1.01 1.01 1.01 1.02 1.01 1.01 2 0.98 0.98 0.98 0.98 0.98 0.98 3 0.98 0.98 0.98 0.98 0.98 0.98 4 0.98 0.98 0.97 0.98 0.98 0.98 5 1.03 1.03 1.02 1.03 1.03 1.03 6 1.07 1.07 1.07 1.08 1.08 1.08 7 1.03 1.03 1.03 1.03 1.03 1.03 F 0.97 0.97 0.97 0.97 0.97 0.97 M 1 1.01 1.01 1.01 1.02 1.01 1.01 2 1.00 1.00 0.99 1.00 0.99 1.01 3 1.02 1.00 1.00 1.00 1.01 1.01 4 0.95 0.96 0.95 0.96 0.96 0.96 5 1.04 1.04 1.01 1.04 1.02 1.05 6 1.04 1.05 1.05 1.05 1.06 1.05 7 0.99 0.99 1.01 0.99 1.01 0.99 F 0.98 0.98 0.96 0.98 0.97 0.98 S 1 1.01 1.01 1.01 1.01 1.01 1.03 2 0.99 0.99 1.00 0.99 0.99 1.02 3 1.02 1.03 0.87 1.03 1.00 1.00 4 1.04 1.02 0.94 1.02 0.95 0.96 5 1.01 1.01 1.00 1.02 1.00 1.04 6 1.04 1.05 1.04 1.04 1.05 1.05 7 1.00 1.00 1.01 1.00 1.01 0.99 F 0.97 0.97 0.97 0.97 0.96 0.98 a Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).

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200 Table B-5. Effective dose scaling factors ( SFBEB) applicable to particles in the Florida phosphate chemical plants for linearly increasing radioac tivity distribution (AR = 1:5) Effective Dose Scaling Factor SFBE Ba Absorption Type Impactor Stage 238U 234U 230Th 226Ra 210Pb 210Po F 1 1.00 1.00 1.00 1.00 1.00 1.00 2 0.96 0.96 0.96 0.96 0.96 0.96 3 0.95 0.95 0.95 0.96 0.96 0.95 4 0.98 0.98 0.97 0.98 0.98 0.98 5 1.06 1.06 1.05 1.06 1.06 1.06 6 1.11 1.11 1.11 1.12 1.12 1.12 7 1.02 1.02 1.02 1.02 1.02 1.02 F 0.86 0.86 0.86 0.86 0.86 0.86 M 1 1.00 1.00 1.00 1.00 1.00 1.00 2 0.96 0.96 0.96 0.96 0.96 0.96 3 0.90 0.88 0.92 0.88 0.93 0.86 4 0.95 0.96 0.93 0.96 0.95 0.96 5 1.07 1.07 1.03 1.07 1.05 1.08 6 1.05 1.06 1.08 1.06 1.08 1.07 7 0.97 0.96 1.00 0.96 0.99 0.96 F 0.87 0.87 0.86 0.87 0.86 0.87 S 1 1.00 1.00 1.00 1.00 1.00 1.02 2 0.96 0.96 0.96 0.96 0.96 0.98 3 0.92 0.92 0.79 0.92 0.93 0.86 4 1.05 1.02 0.92 1.02 0.94 0.96 5 1.03 1.04 1.02 1.04 1.02 1.08 6 1.06 1.06 1.06 1.06 1.07 1.07 7 0.98 0.98 0.99 0.98 0.99 0.97 F 0.87 0.86 0.86 0.86 0.86 0.87 a Bolded values indicate scaling factors 1.10 or 0.9 (more than a 10% adjustment in the effective dose).

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210 BIOGRAPHICAL SKETCH Kwang Pyo Kim was born in Jeju Island, South Korea, on July 8, 1970, and was raised to his high school years there. He attended the Depart ment of Nuclear Engineering in Kyunghee University and earned a Bachelor of Engineering in 1998. A diploma of honors was received in the commencement. In 2000, he received a Master of Engin eering from Kyunghee University. He served as a graduate research assistant and te aching assistant, involve d in projects titled “Study of the Corrosion Mechanism of Zirc onium Alloys” under a Korea Atomic Energy Research Institute grant and “Pressure Eff ects on High Temperature Zircaloy-4 Oxidation in Steam” under a Korea Nuclear Fuel Co. gran t. He studied the kinetics and mechanism of Zirconium alloys corrosion under diffe rent alloy compositions and different conditions. During these years, experimental and analytical techniques and research methodology were acquired and refined. In 2001, Kwang Pyo Kim enrolled at Univers ity of Florida to pursue a Ph.D. in Nuclear and Radiological Engin eering. He was involved in the Health Physics Program and served as a research assistant. He worked on projects titled “Risk Assessment of Airborne Particulates to Workers in the Phosphate I ndustry” and “Assessment of Airborne Particulate Lung Sol ubility and Internal Dose to Phosphate Workers” under a Florida Institute of Phosphate Research gr ant. He studied ai rborne particulate characterization, lung solubili ty, and dose assessment.