Radiation dose assessment of exposure to depleted uranium

Journal of Exposure Science and Environmental Epidemiology (2009) 19, 502–514 r 2009 Nature Publishing Group All rights reserved 1559-0631/09/$32.00

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Radiation dose assessment of exposure to depleted uranium WEI BO LI, UDO C. GERSTMANN, VERA HO¨LLRIEGL, WILFRIED SZYMCZAK, PAUL ROTH, CHRISTOPH HOESCHEN AND UWE OEH Institute of Radiation Protection, Helmholtz Zentrum Mu¨nchen, German Research Center for Environmental Health (GmbH), Neuherberg, Germany

Depleted uranium (DU) is claimed to contribute to human health problems, known as the Gulf War Syndrome and the Balkan Syndrome. Quantitative radiation dose is required to estimate the health risk of DU materials. The influences of the solubility parameters in the human alimentary tract and the respiratory tract systems and the aerosol particles size on the radiation dose of DU materials were evaluated. The dose conversion factor of daily urinary excretion of DU is provided. The retention and excretion of DU in the human body after a contamination at a wound site were predicted. Dose coefficients of DU after ingestion and inhalation were calculated using the solubility parameters of the DU corrosion products in simulated gastric and simulated lung fluid, which were determined in the Helmholtz Zentrum Mu¨nchen. 238U is the main radiation dose contributor per 1 Bq of DU materials. The dose coefficients of DU materials were estimated to be 3.5 108 and 2.1 106 Sv Bq1 after ingestion and inhalation for members of the public. The ingestion dose coefficient of DU materials is about 75% of the natural uranium value. The inhalation dose coefficient of DU material is in between those for Type M and Type S according to the category for inhaled materials defined by the International Commission on Radiological Protection. Radiation dose possibly received from DU materials can directly be estimated by using the dose conversion factor provided in this study, if daily urinary excretion of DU is measured. Journal of Exposure Science and Environmental Epidemiology (2009) 19, 502–514; doi:10.1038/jes.2008.40; published online 2 July 2008

Keywords: depleted uranium; biokinetic modeling; internal dose; aerosol; wound model; radiaton exposure.

Introduction Depleted uranium (DU) is a byproduct of uranium (U) enrichment processing in nuclear reactors and in production of nuclear weapons. DU is used in both civilian (Betti, 2003) and military (AEPI, 1995) applications. Because of the different percentages of uranium isotopes in DU, its specific activity (14.8 mBq mg1) is approximated 40% lower than that of natural U (25.4 mBq mg1) and considerably lower than that of enriched U (1750 mBq mg1) (Harley et al., 1999; Fulco et al., 2000). DU may have trace amount of 236U (AEPI, 1995), and it was detectable in some of the biological samples collected in Serbia and Montenegro (Jia et al., 2004) and of DU penetrator samples collected in Kosovo (Oeh et al., 2005). DU penetrator is a type of ammunition like a bullet, which is produced by DU metal normally in the 25, 105 and 125 mm kinetic energy cartridges, to penetrate a target (see Figures 1–3 in AEPI, 1995). Typical trace isotopes identified as being present in DU used in munitions and armor manufacture, include 238Pu, 239Pu, 240Pu, 241Am,

1. Address all correspondence to: Dr. WB Li, Institute of Radiation Protection, Helmholtz Zentrum Mu¨nchen, German Research Center for Environmental Health (GmbH), Neuherberg D-85764, Germany. Tel.: þ 49 89 3187 3314. Fax: þ 49 89 3187 2517. E-mail: [email protected] Received 17 March 2008; accepted 24 May 2008; published online 2 July 2008

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Np and 99Tc. These impurities typically add less than 1% to the dose from DU and are therefore not relevant from a radiological or chemical toxicity standpoint (WHO, 2001). So far, DU has been used in military conflict in the Balkans, in Afghanistan and in Iraq. DU is claimed to contribute to health problems, known as the Gulf War Syndrome and the Balkan Syndrome (Bleise et al., 2003; McDiarmid et al., 2004; Squibb and McDiarmid, 2006). United Nations Environment Programme (UNEP) reported the post-conflict environmental assessment of DU in the Balkan region (UNEP/UNCHS, 1999; UNEP, 2001, 2002, 2003) and in Lebanon (UNEP, 2007). The Royal Society and the World Health Organization (WHO) published the possible health hazards of DU munitions (The Royal Society, 2001, 2002; WHO, 2001). These investigations (Priest, 2001; The Royal Society, 2001, 2002; UNEP, 2001; WHO, 2001; Durante and Pugliese, 2003; Roth et al., 2003; Oeh et al., 2007a, b) reported that no significant risk of DU could be found, but recommended that further scientific study should be carried out to reduce the uncertainties related to the assessment of the environmental impacts of DU, especially through exposure of the soldiers and residents. In this study, soldiers and residents are people, who are potentially exposed to DU. In the calculation of DU dose coefficient, the terms of workers and members of the public, which were developed by International Commission on Radiological Protection (ICRP) for referring to two different exposure scenarios, are used. The ICRP developed primarily the

Radiation dose assessment of exposure to DU

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Figure 1. Biokinetic compartmental model for uranium after intake. It combines the systemic biokinetic model of uranium, the HRTM with the GIT model (a) and with the HATM (b). Extrathoracic region: ET1 (anterior nose), ET2 (posterior nasal passages, larynx, pharynx and mouth), ETseq (particle sequestered by macrophages in the lamina propria), LNET (lymph nodes). Thoracic region: BB (bronchial), bb (bronchiolar), AI (alveolar-interstitial), LNTH (lymph nodes). Parameters of sp, st and spt denote the transfer rate from initial state to blood, from transformed state to blood, from initial state to transformed state, respectively. Parameter of fb gives the transfer fraction to bound material. f1 denotes the intestinal absorption fraction in the GIT model, and fA denotes the alimentary tract transfer factor in the HATM. Particles in initial state were deposited in compartments numbered 41–54 in the bold rectangles. They are transferred to other compartments in the lungs and then to the circulation system. ‘‘NOEXCH’’ and ‘‘EXCH’’ denote nonexchangeable and exchangeable bone volume, respectively (ICRP, 1979, 1994a, 1995a, 2006).

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Figure 2. Biokinetic compartmental model for uranium contaminated in a wound site. It coupled the systemic biokinetic model of uranium (ICRP, 1995a) and the biokinetic model for radionuclide-contaminated wounds (NCRP, 2007). ‘‘NOEXCH’’ and ‘‘EXCH’’ denote nonexchangeable and exchangeable bone volume, respectively.

Stomach (ST) Solubility

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biokinetic and dosimetric models for occupational workers (ICRP, 1979) and then applied those models for the general public (ICRP, 1995a, b) by changing the exposure condi504

tions, for example, aerosol size and the physiological parameters, like, lung volume, physical activities, breathing rate and so on. So, the terms of workers and the public, if mentioned in this context, are referred to the dose coefficients, which were calculated using input parameters for different exposure scenarios for workers and members of the public defined in the ICRP Publications (ICRP, 1994a, b, 1995a, b). There is no additional exposure definition for soldiers. The dose coefficients calculated using the above two mentioned exposures are applicable to soldiers. The three main routes of human exposure to DU on the battlefield are inhalation, ingestion and wounding by shrapnel; on impact with an armored vehicle, substantial amounts of DU may be dispersed as particles that can be inhaled and DU fragments may cause shrapnel wounds (The Royal Society, 2001). The contaminated soil, drinking water and foods in the community could be also a potential radiation risk to the residents by ingestion route. DU might become an environmental risk factor, especially if the use of DU produces respirable particles (Durakovic, 1999), it might be cytotoxic and clastogenic to human lung cells (Wise et al., 2007). Miller and McClain (2007) emphasized that although no conclusive epidemiologic data have correlated DU exposure to specific health effects, studies using cultured cells and laboratory rodents continue to suggest the possibility of leukemogenic, genetic, reproductive and neurological effects from chronic exposure. Schimmack et al. (2005, 2007) found that uranium leaching rates increase dramatically from corroding DU penetrators after 3 years. The high 238U concentrations Journal of Exposure Science and Environmental Epidemiology (2009) 19(5)


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Materials and methods

besides the normal routes, DU materials, mostly as fragment, go directly into blood through the wound site of the body. About 0.2%–2% of natural U or DU materials entering the body are absorbed by blood through the gut, the lungs or the lymph nodes and preferentially transferred to organs, for example, kidney, liver, bone (ICRP, 1994b, 1995a, b). The ICRP has developed comprehensive biokinetic models for radionuclides for more than two decades (ICRP, 1979). The systemic biokinetic models of uranium (ICRP, 1995a), coupled with the GIT model of ICRP Publication 30 (ICRP, 1979), the new HATM (ICRP, 2006) and the human respiratory tract model (HRTM) of ICRP Publication 66 (ICRP, 1994a), were used to model the excretion and the retention of 234U, 235U, 236U and 238U in organs or tissues after incorporation (Li et al., 2005). The biokinetic model of U after intake is illustrated in Figure 1. The upper compartmental model (Figure 1a) shows the GIT model (shaded compartments) combined with the systemic and the respiratory tract model, the lower (Figure 1b) for the new HATM. The new HATM, provides a much more realistic physiological description of material in the human alimentary tract, has superseded the simple GIT model. In the GIT model, the absorption of radionuclides in blood is assumed to occur only from small intestine (SI) and indicated as f1 value (ICRP, 1979), whereas this absorption quantity is specified in the HATM as a fraction of the total entering the alimentary tract and denoted as fA (ICRP, 2006). As no information is available on the absorption of U in other regions of the human alimentary tract, it is assumed that all absorption of U takes place in the SI, fSI ¼ fA, where fSI is the fractional transfer from the SI to blood. Therefore, in this study, fA value is used instead of f1. The HATM is more appropriate to the dose calculation using the human voxel-based phantom (Zankl et al., 2005, 2007). However, the specific absorption fractions for the new HATM are not yet fully completed. Therefore, in this study, the dose calculation is based on the GIT model. A new wound model has been jointly developed by NCRP and ICRP (NCRP, 2007); it can ideally be used to assess the retention, excretion and radiation internal dose for those individuals who were injured by DU fragments in the conflict. The wound model, combined with the U systemic model, is shown in Figure 2. There are five compartments that comprise the wound site: soluble (SOL), colloid and intermediate state (CIS), particle aggregates and bound state (PABS), trapped particles and aggregates (TPA) and fragment (FRG). The other two compartments, that is, blood and lymph nodes (LN), receive radionuclide cleared from the wound site.

Biokinetic model of uranium Natural U enters the human body through consumption of foodstuffs, drinking water or through inhalation. Small amount of U can enter the body through the skin. In the war,

Internal dose calculation of DU after ingestion and inhalation The organ equivalent dose, HT, for each U isotope, that is, 234 U, 235U, 236U and 238U, is calculated by summing the

observed in the seepage water highlight the need for further investigations on the transport of 238U through soil, in particular with regard to the potential 238U contamination of groundwater in areas affected by DU weapons. Briner (2006) reported the evidence of potential carcinogenic effects of DU materials. Bertell (2006) concluded that all the questions about DU and Gulf War Syndrome are not yet answered. Long-term studies, monitoring of health effects and further research are recommended (The Royal Society, 2002). To assess the radiation risk of DU materials, the dose coefficients are necessary. Until now, dose coefficients are based on the default parameters published by ICRP (ICRP, 1994b, 1995a, b, 1996). To precisely assess the radiation dose of DU, properties of DU aerosols, behavior of DU oxide produced during penetrator impact and solubility are key parameters (Gerstmann et al., 2008). It is reiterated by ICRP that where material-specific values are available, they should be used in the models for dose assessment (Valentin and Fry, 2003). In this study, ingestion and inhalation dose coefficients for 234 U, 235U, 236U and 238U were calculated using the solubility parameters of the DU materials in simulated gastric juice and simulated lung fluid synthesized at the Radioanalytical Laboratory (RADLAB) in the Helmholtz Zentrum Mu¨nchen, German Research Center for Environmental Health (HMGU). For a practical dose estimation, the dose conversion factor of daily DU urinary excretion is evaluated using the effective dose and the daily urinary excretion after hypothetical ingestion, inhalation and a contamination at a wound site. According to the DU particle size investigated by Capstone Aerosols (Parkhurst et al., 2004), 238U inhalation dose coefficients as a function of aerosol size were calculated. In addition, the new human alimentary tract (ICRP, 2006) model and the biokinetic model for radionuclide-contaminated wounds jointly developed by the National Council on Radiation Protection and Measurements (NCRP) and ICRP (NCRP, 2007) were implemented in the modeling. Retentions of DU material of different categories in the main organs, for example, blood, kidney, bone and excretion were predicted. The nuclear transformation number of U between the gastrointestinal tract (GIT) model and the new human alimentary tract model (HATM) are calculated and compared, and the possible impact of the HATM on the dose coefficient was discussed. Influences of the solubility parameters in the alimentary and respiratory tract systems, and the alimentary tract transfer factor on the dose coefficients were presented.


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product of US (nuclear transformations) and the specific effective energy (SEE(T’S)) over all source organs, S (ICRP, 1995a, b; Li et al., 2006). In the values of the SEEs each type of radiation is weighted by a given radiation weighting factor, wR, which accounts for the different biological effectiveness in inducing late effects. The US of the radionuclide in the source organ or tissue was calculated by integrating the retention function, which was modeled by the SAAM II program (University of Washington, Seattle, WA, USA). The value of SEE(T’S) was calculated by using the SEECAL program (Oak Ridge National Laboratory, Oak Ridge, TN, USA). The effective dose, E, was calculated by the sum of the product of the equivalent dose, HT, and the corresponding tissue weighting factor, wT, over 12 organs and tissues, T, and the product between the equivalent dose of the so called remainder tissue, Hrem, and its weighting factor, wrem (ICRP, 1991). The equivalent dose of the colon, Hcolon, was calculated as 0.57 HULI þ 0.43 HLLI, where HULI and HLLI are the equivalent doses in the walls of the upper large intestine (ULI) and the lower large intestine (LLI), respectively (ICRP, 1995b). The dose to the thymus is taken as a surrogate for the esophagus, in the absence of a dosimetric model for the esophagus (ICRP, 1995a). For the lungs, the apportionment factors for target cells in different lung regions recommended by ICRP were used (ICRP, 1994a). The regional deposition of U in the human respiratory tract for outdoor exposure was taken from ICRP Supporting Guidance 3 (ICRP, 2002). These values take into account the activity median thermodynamic diameter (AMTD) and the activity median aerodynamic diameter (AMAD), and typical breathing rate and daily time schedule for members of the public and the occupational workers. Deposition data of aerosol larger than 20 mm in AMAD were calculated by using the program LUDEP 2.0 (Health Protection Agency, Chilton, UK), which implemented the ICRP HRTM. For inhalation dose calculation, the rapid dissolution fraction, fr, in the lung and its absorption rate, sr, were evaluated based on the measurement performed at our institute in the HMGU (Gerstmann et al., 2008). The simulated lung fluid (‘‘Gamble solution’’) included the following components: L-cysteine (1 mM), ammonium chloride (10 mM), glycine (6 mM), sodium citrate (0.2 mM), calcium chloride (0.2 mM), sodium dihydrogen phosphate (1.2 mM), sodium hydrogen carbonate (27 mM), sodium chloride (116 mM), diethylenetriaminepentaacetic acid (2 mM) and alkylbenzyldimethyl ammonium chloride (50 ppm) (Gamble, 1967). As the duration of the investigation is too short to provide a reliable quantitative value for the slower blood absorption rate, ss, the value published by ICRP for material categories Type M and Type S was used in this study (ICRP, 1994a). The Type M and Type S are materials, which are classified by ICRP, to describe how fast and slow the deposited materials in lung are absorbed into blood, for example, with a moderate rate (Type M) and a slow rate 506

(Type S). To implement these parameters derived from experiments into the modeling, the transfer rates, that is, sp, st, spt were calculated as following (ICRP, 1994a): sp ¼ ss þ fr ðsr ss Þ spt ¼ ð1 fr Þðsr ss Þ st ¼ ss ; where sp is the transfer rate from initial state to blood, st is the transfer rate from transformed state to blood and spt is the transfer rate from initial state to transformed state (Figure 1). For ingestion dose calculation, the solubility of DU material in simulated gastric juice is, in average, about 75% (Gerstmann et al., 2008). The gastric juice was freshly prepared in the laboratory by dissolving 2 g NaCl, 7 ml of 37% HCl and 3.2 g pepsine from the mucosa of pigs in 1 l of Milli-Q water produced by purification water system (Millipore GmbH, Schwalbach, Germany) (Hamel et al., 1998). The compartment SI was split into two compartments, defined as SI soluble (SI-SOL) and insoluble (SI-INSOL) compartments as shown in Figure 3. The dose coefficients were predicted as a function of solubility applying the model presented in Figure 3, this model was also used to model the decay daughters of 238U, for example 234 Th, 231mPa, 231Pa and 231Th, assuming that those decay daughters of the insoluble DU materials are still in the insoluble state. According to our calculations, the contribution of decay products to the ingestion effective dose coefficient of 238U is about 2%, and the maximum contribution of organ equivalent dose is about 4% in red marrow. So, the ingestion effective dose coefficients of DU presented in this study, if the solubility of decay products is different from the parent, give an uncertainty of the ingestion dose coefficient maximum to 2%. The transfer rates of kSI-SOL’ST and kSI-INSOL’ST were calculated as following: kSI-SOL’ST ¼ kSI’ST solubility, and kSI-INSOL’ST ¼ kSI’ST (1-solubility), where kSI’ST ¼ 24 day1, the transfer rate from the stomach to the SI in the GIT model (ICRP, 1979).

Results Dose coefficients of U and DU materials after intake The effective dose coefficients of 234U, 235U, 236U, 238U calculated by using the solubility parameters derived from experiments performed in the HMGU are listed in Table 1 in comparison to the ICRP published values (ICRP, 1994b, 1995a, b, 1996). In addition, the organ equivalent dose coefficients of DU materials using exposure parameters for members of the public after ingestion, inhalation and a contamination at a wound site were presented in Table 2. The ingestion dose coefficients of DU materials are, based on the radioactivity of each U isotope, estimated as Journal of Exposure Science and Environmental Epidemiology (2009) 19(5)


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Table 1. Effective dose coefficients (Sv Bq1) of 234U, 235U, 236U, 238U and DU using the HMGU solubility (75% in stomach) and dissolution (fr ¼ 0.37, sr ¼ 6.7 day1 the in lungs) parameters in comparison to ICRP values (ICRP, 1994b, 1995a, b, 1996). Intake Population

Ingestion Public

Inhalation

Workers

Public

AMAD

F

F

F

Absorption type

F

F

F

fA value

0.02

0.02

1 mm F

0.002

U (ICRP) 4.9x10 in DU (HMGU) 3.8 108

8

4.9 10 3.8 108

235 U (ICRP) 4.7 108 in DU (HMGU) 3.6 108

M

0.02

S

0.02

F

0.002

8.3 10 7.3 109

5.6 10 F

6

3.5 10 2.6 106

4.6 108 3.6 108

8.3 109 7.3 109

5.2 107 F

236

U (ICRP) 4.7 108 in DU (HMGU) 3.6 108

4.6 108 3.6 108

7.9 109 6.9 109

238 U (ICRP) 4.5 108 in DU (HMGU) 3.4 108

4.4 108 3.4 108

4.5 108 3.5 108

4.4 108 3.5 108

DU (ICRP) DU (HMGU)

9

1 mm

7

234

8

Workers

M

0.02 6

5 mm S

0.02

F

0.002

9.4 10 6.1 106

7

5.5 10 F

6

3.1 10 2.3 106

3.1 106 2.3 106

8.5 106 5.5 106

5.1 107 F

5.3 107 F

3.2 106 2.4 106

8.7 106 5.7 106

7.6 109 6.6 109

5.0 107 F

2.9 106 2.1 106

7.7 109 6.7 109

5.0 107 F

2.9 106 2.1 106

M

0.02

0.02

0.002

8.5 10 5.6 106

7

6.4 10 F

6

2.1 10 1.5 106

6.8 106 4.4 106

2.8 106 2.0 106

7.7 106 5.0 106

6.0 107 F

1.8 106 1.3 106

6.1 106 4.0 106

5.2 107 F

2.9 106 2.1 106

7.9 106 5.2 106

6.1 107 F

1.9 106 1.4 106

6.3 106 4.1 106

8.0 106 5.2 106

4.9 107 F

2.6 106 1.9 106

7.3 106 4.8 106

5.8 107 F

1.6 106 1.2 106

5.7 106 2.7 106

8.1 106 5.3 106

4.9 107 F

2.6 106 1.9 106

7.4 106 4.8 106

5.8 107 F

1.6 106 1.2 106

5.8 106 2.9 106

3.5 108 and 6.7 109 Sv Bq1 assuming fA values of 0.02 and 0.002 for the public and workers, respectively. The variation of ingestion organ and effective dose coefficients of DU materials, calculated in the HMGU, as a function of solubility was shown in Figure 4. The ingestion effective dose coefficients of DU materials were reduced by about 75% in comparison to the ICRP values for both fA values of 0.02 and 0.002. The reduction of dose is proportional to the solubility (75%) of DU materials in the gastric fluid. The effective dose coefficient increases from 3.6 109 to 4.8 108 Sv Bq1, by a factor of 13.3, as the solubility increasing from 0% to 100% with a value of 2.4 108 Sv Bq1 at a solubility of 50% (Figure 4). The fraction, that is, fr, of DU materials rapidly dissolved in the simulated lungs environment was demonstrated, by the HMGU experimental data, to be a mean value of 0.37 in comparison to 0.1 and 0.001 for ICRP default clearance classes Type M and Type S. The fractional dissolution rate for this fast phase was estimated, in average, to be 6.7 day1. The use of those HMGU dissolution parameters provided a set of transfer rates used in the modeling (Figure 1): sp ¼ 2.45 day1, spt ¼ 4.24 day1 and st ¼ 0.005 for Type M and st ¼ 0.0001 day1 for Type S. The sp of 2.45 day1 was found to be situated between the classes of Type M (sp ¼ 10 day1) and Type S (sp ¼ 0.1 day1). The spt value of 4.24 day1 is much lower than the value of Type M (spt ¼ 90 day1) and Type S (spt ¼ 100 day1). The inhalation dose coefficients using the exposure condition for the members of the public are 2.1 106 and 5.3 106 Sv Bq1 (AMAD ¼ 1 mm) by using the mean Journal of Exposure Science and Environmental Epidemiology (2009) 19(5)

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S

solubility of 0.37 and the mean transfer rate of 6.7 day1 in the HRTM with fA values of 0.02 and 0.002, respectively. For the exposure condition for occupational workers, the inhalation dose coefficients are 1.9 106 and 4.8 106 Sv Bq1 (AMAD ¼ 1 mm) with fA values of 0.02 and 0.002, respectively; and 1.2 106 and 2.9 106 Sv Bq1 (AMAD ¼ 5 mm) with fA values of 0.02 and 0.002, respectively.

Inhalation dose coefficient of 238U at different aerosol size Inhalation dose coefficients of 238U as a function of aerosol size represented in AMTD (0.01–1 mm) and AMAD (0.1– 100 mm) are shown in Figure 5. In the DU particle range at AMTD domain of 0.01–1 mm, the lungs receive highest organ dose: 2.3 104 Sv Bq1 at 0.01 mm of AMTD and 4.6 105 Sv Bq1 at 1 mm of AMTD, followed by bone surface, kidneys, liver and red marrow. Effective doses, mostly contributed from lungs, range from 2.8 105 Sv Bq1 at 0.01 mm of AMTD, of 9.6 106 Sv Bq1 at 0.1 mm, and of 5.7 106 Sv Bq1 at 1 mm. In general, the effective and organ dose coefficients decrease by a factor of five, as AMTDs increase from 0.01 to 1 mm. The organ dose in ET airways is lower than bone surface as particle size is smaller than 0.1 mm. However, as AMTD increases, more DU particles deposit in ET airways and release higher doses, but still a bit lower than the lungs dose. In the AMAD domain, the distribution of effective and organ doses are similar to the AMTD domain. The dose coefficients decrease by about a factor of 12, as AMADs increase from 0.1 to 100 mm. The equivalent dose to ET airways exceeds the dose to lungs and stays in constant as 507


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Table 2. Dose coefficient (Sv Bq1) of DU materials for ingestion (fr ¼ 0.37, sr ¼ 6.7 day1), inhalation and a contamination in a wound site. Exposure

Ingestion

Inhalation Wounda

fA value

0.02

0.002

0.02

Solubilityb

0.75

0.75

ss ¼ 0.005

ss ¼ 0.0001

Weak

Particle

Fragment

1.9 108 1.9 108 5.3 107 1.9 108 1.9 108 1.9 108 2.0 108 2.1 108 3.4 108 6.4 108 4.7 108 1.9 107 7.3 108 1.9 108 1.9 108 1.9 108 5.6 108 1.9 108 1.9 108 1.9 108 1.9 108 1.9 108 1.9 108 1.9 108 1.9 108 2.1 108 3.5 108

1.9 109 1.9 109 5.3 108 1.9 109 1.9 109 1.9 109 2.9 109 4.4 109 1.7 108 4.8 108 3.1 108 1.9 108 7.3 109 1.9 109 1.9 109 1.9 109 5.6 109 1.9 109 1.9 109 1.9 109 1.9 109 1.9 109 1.9 109 1.9 109 1.9 109 2.1 109 6.7 109

1.4 107 1.4 107 3.9 106 1.4 107 1.4 107 1.4 107 1.4 107 1.4 107 1.4 107 1.5 107 1.4 107 1.4 106 5.4 107 1.4 107 1.4 107 1.4 107 4.2 107 3.8 106 1.6 105 1.4 107 1.4 107 1.4 107 1.4 107 1.4 107 1.4 107 1.5 107 2.1 106

8.0 108 8.0 108 2.3 106 8.0 108 8.0 108 8.0 108 8.0 108 8.0 108 8.5 108 9.5 108 8.9 108 8.4 107 3.1 107 8.0 108 8.0 108 8.0 108 2.4 107 2.0 105 4.3 105 8.0 108 7.9 108 7.9 108 8.0 108 8.0 108 8.0 108 9.7 108 5.3 106

1.2 106 1.2 106 3.5 105 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 105 4.7 105 1.2 106 1.2 106 1.2 106 3.8 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 1.2 106 3.2 105 2.0 106

9.4 107 9.4 107 3.2 105 9.3 107 9.3 107 9.3 107 9.3 107 9.3 107 9.3 107 9.4 107 9.4 107 1.2 105 4.3 106 9.3 107 9.3 107 9.3 107 3.5 106 9.3 107 9.3 107 9.3 107 9.3 107 9.3 107 9.3 107 9.3 107 9.3 107 1.1 106 1.7 106

4.0 108 4.0 108 1.5 106 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 6.4 107 2.0 107 4.0 108 4.0 108 4.0 108 1.7 107 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 4.0 108 4.6 108 7.8 108

Adrenals Bladder wall Bone surface Brain Breast Oesophagus St wall SI wall ULI wall LLI wall Colon Kidneys Liver Muscle Ovaries Pancreas Red marrow ET airways Lungs Skin Spleen Testes Thymus Thyroid Uterus Remainder Effective dose

0.002

a

Data for 238U, cited from NCRP Report No.156 (NCRP, 2007). In gastric fluid (for ingestion) and in simulated lung fluid (for inhalation), respectively.

b

AMAD is larger than 4 mm, and the equivalent dose to the lungs decreases sharply to 1.0 107 Sv Bq1, at 100 mm, lower than the equivalent dose to kidneys and liver.

Effective dose conversion factor of DU per daily urinary excretion The DU conversion factor of daily urinary excretion to effective dose for adult members of the public after ingestion,

DU ingestion dose coefficient (Bq Sv-1)

Figure 4. Ingestion dose coefficient of DU materials as a function of solubility.

Nuclear transformation number with HATM and GIT model To evaluate the impact of the new HATM on the U ingestion dose in comparison to the GIT model, the nuclear transformations (US) of 238U in different type of materials in the alimentary tract were calculated and are presented in Table 3. The daily fecal excretion after ingestion of 238U predicted by the new HATM and the old GIT model was shown in Figure 6. The retention in organ and tissue, for example, blood, kidney, liver, bone and other soft tissue, and the urinary excretion after DU ingestion by using the HATM and the GIT model showed no difference. However, the fecal excretion predicted by GIT model shows a higher value than that using HATM during 4–20 days after DU materials ingestion. The fecal excretion predicted by GIT model shows 69 times higher than that using HATM at day 11.

508

10-6

10-7

Effective dose Bone surface Kidneys Liver Red marrow Colon

10-8

10-9

0

20

40 60 Solubility (%)

80

100



Li et al.

238U

inhalation dose coefficient (Sv Bq-1)

Effective dose Lungs ET airways

Figure 5. Dependence of

Bone surface Kidneys Red marrow

10-3 10-4 10-5 10-6 10-7 10-8 0.01

0.1 AMTD (µm)

1

0.1

1

10

100

AMAD (µm)

238

U inhalation dose coefficient on aerosol particle size in the AMTD and the AMAD.

Table 3. Nuclear transformation number in the human alimentary tract of

238

U after a single ingestion using HATM and GIT model.

Region

Total diet

Solids

Caloric liquid

Noncaloric liquid

GIT model

Mouth Esophagus (fast) Esophagus (slow) Stomach Small intestine Right colon/ULI Left colon/LLI Rectosigmoid

12 6.3 4 4.2 103 1.4 104 4.2 104 4.2 104 4.2 104

15 7.2 4.5 4.5 103 1.4 104 4.2 104 4.2 104 4.2 104

2 4.5 3 2.7 103 1.4 104 4.2 104 4.2 104 4.2 104

2 4.5 3 1.8 103 1.4 104 4.2 104 4.2 104 4.2 104

F F F 3.6 103 1.4 104 4.7 104 8.5 104 F

in different soluble and chemical forms was predicted and shown in Figure 7.

DU daily fecal excretion (Bq d-1)

100 10-1

GIT model HATM

Retention and excretion of 238U after a wound contamination Assuming different categories of DU materials, retentions in organs, for example, blood, bone, kidney and other soft tissues after a contamination at a wound site were predicted (Figure 7a–c), and the urinary excretion was shown in Figure 7d. Soluble DU is retained longer in the body than colloidal, particulate or fragments of DU.

10-2 10-3 10-4 10-5 10-6 10-7 10-8

1

10

100

Time after a single ingestion (day)

Figure 6. Daily fecal excretion of DU materials using GIT model and HATM after a single ingestion.

inhalation and a wound contamination are listed in Table 4. The effective doses of 238U in the categories of weak, particle and fragment for a wound site were adopted from NCRP Report (NCRP, 2007). The daily urinary excretion of 238U Journal of Exposure Science and Environmental Epidemiology (2009) 19(5)

Discussion The parameters derived from the investigation on the DU penetrators, that is, the solubility in the alimentary tract and the dissolution in the respiratory tract were applied into the ICRP biokinetic model to investigate the variation of the dose coefficients. The soluble part of the DU material can change by entering the intestinal region where an increase in pH of the intestinal solution occurs. DU in the soluble state is 509


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Table 4. Conversion factor of daily DU urinary excretion to effective dose, K (Sv per Bq day1), for the exposure condition of adult member of the public at time t after ingestion, inhalation and a contamination in a wound site. Time, t

Ingestion

Inhalation

Wound

(day)

fA ¼ 0.02

fA ¼ 0.002

fA ¼ 0.02

fA ¼ 0.002

Weak

Particle

Fragment

1 3 5 7 10 30 50 70 100 300 500 700 1000 3000 5000 7000 10,000 20,000 30,000 40,000

3.6 106 1.2 104 1.5 104 1.8 104 2.4 104 9.4 104 2.1 103 3.6 103 6.4 103 7.3 102 2.0 101 2.4 101 2.8 101 5.9 101 1.1 100 1.7 100 2.8 100 6.7 100 1.4 101 2.5 101

7.1 106 2.4 104 3.0 104 3.6 104 4.7 104 1.8 103 4.2 103 7.1 103 1.3 102 1.4 101 3.9 101 4.8 101 5.5 101 1.2 100 2.1 100 3.4 100 5.5 100 1.3 101 2.7 101 4.9 101

5.5 105 1.5 103 1.8 103 2.0 103 2.4 103 5.1 103 7.3 103 9.2 103 1.2 102 4.4 102 1.4 101 3.8 101 1.1 100 4.8 100 8.9 100 1.4 101 2.3 101 5.6 101 1.1 102 2.1 102

1.5 104 5.1 103 6.2 103 7.4 103 9.7 103 3.7 102 8.0 102 1.3 101 2.1 101 9.6 101 1.4 100 1.7 100 2.1 100 6.5 100 1.2 101 1.9 101 3.1 101 1.1 102 2.8 102 6.0 102

3.5 106 9.0 105 1.3 104 1.5 104 2.0 104 7.6 104 1.8 103 3.1 103 5.7 103 6.5 102 1.8 101 2.2 101 2.5 101 5.3 101 9.7 101 1.5 100 2.5 100 6.0 100 1.2 101 2.2 101

1.7 102 1.4 102 1.4 102 1.4 102 1.5 102 1.9 102 2.4 102 2.6 102 2.6 102 2.1 102 1.7 102 1.5 102 1.4 102 1.3 102 1.8 102 2.9 102 6.3 102 8.5 101 5.3 100 1.3 101

1.5 100 7.8 102 3.1 102 1.9 102 1.2 102 4.3 103 3.3 103 3.1 103 3.2 103 9.5 103 2.0 102 2.6 102 2.8 102 2.9 102 2.9 102 2.9 102 2.9 102 3.0 102 3.1 102 3.2 102

K ¼ (effective dose of DU)/(daily DU urinary excretion). t ¼ time after intake of DU materials.

then potentially available for uptake across the intestinal lumen to the blood. The actually total fraction absorbed in the alimentary tract, predominately in the small intestinal wall and consequently transferred into blood is described by the fA value. Generally, the ingestion effective dose coefficients of DU materials were reduced by about 75% in comparison to the ICRP values. The ingestion effective dose coefficients presented are lower for workers than the public, which is because the fA value is lower for workers (fA ¼ 0.002) than the public (fA ¼ 0.02). Computational results for inhalation dose coefficients, using the dissolution parameters evaluated in the HMGU, showed a general decrease in comparison to the ICRP values. This observation can be explained by variation of dissolution parameters used, that is, the fraction rapidly dissolved, fr ¼ 0.37, of DU materials in comparison to 0.1 and 0.001 for ICRP default clearance classes Type M and Type S, and the fractional dissolution rate for this fast phase, sr ¼ 6.7 day1, in comparison to the ICRP default values of 100 day1 for both Type M and Type S. Consequently, the effective doses calculated using the HMGU dissolution parameters decreased (Table 1). By inhalation, the total deposition of inhaled aerosols in the whole respiratory tract for workers is 1.7 times higher than that of the public. However, for workers, about 74% of the total inhaled aerosols are assumed to be deposited in the extrathoracic (ET) region, which gives less contribution to the resulting dose coefficient, in comparison to the public, for 510

which only 32% is assumed to be deposited in ET region. The deposition in the alveolar–interstitial region of the lungs for public is two times higher than that for the workers (ICRP, 1994a). Therefore, the dose coefficient of the lungs is higher for the public. This higher equivalent dose in lungs makes a slightly higher effective dose for the public (Table 1). Capstone Aerosols study (Parkhurst et al., 2004) analyzed the aerosol size distribution produced by the DU munitions. The DU particles were collected inside and outside the target vehicles. Taking into account the unimodal and bimodal AMAD analysis models, the interior results range from 0.01 to 70.1 mm, whereas the external results range from 0.16 to 30.1 mm. Accordingly, the effective dose per inhaled unit intake to a person could range from 2.8 105 to 8.5 107 Sv Bq1 inside the vehicles, and from about 7.7 106 to 1.2 106 Sv Bq1 outside the vehicles (Figure 5). Salbu et al. (2003) identified that DU particles are ranging from submicrons to about 30 mm from soil samples collected at Ceja Mountain, Kosovo. This DU particle size is very close to the external results measured by Parkhurst et al. (2004). The possible inhalation dose for the residents in the Ceja Mountain area appears to be in the range of 1.4 105 Sv Bq1 (AMAD ¼ 0.1 mm) to 1.2 106 Sv Bq1 (AMAD ¼ 30 mm). The effective dose above 4 mm of AMAD decreases slowly in contrast to the lung dose because the ET airway, as a remainder organ, receives the highest committed equivalent dose of all organs, and a weighting factor of 0.025 is assigned Journal of Exposure Science and Environmental Epidemiology (2009) 19(5)


Li et al.

10-3 10-4 10-5 10-6 10-7

10-1 retention in bone (Bq)

W M S A C P F

238U

238U

retention in blood (Bq)

10-2

Blood

10-8 10

100

10-4 10-5 10-6 10-7

1000

10-2 10-3 10-4 10-5 10-6 10-7

Kidneys

10-8 1

238U daily

urinary excretion (Bq d-1)

10-1 retention in kidney (Bq)

10-3

10-8 1

238U

10-2

Bone 1

10

100

1000

100 10-1 10-2 10-3 10-4 10-5 10-6 10-7

Daily urinary excretion

10-8

10

100 1000 1 10 100 1000 Time after an accidental contamination (1 Bq) at a wound site (day) Figure 7. Retention of DU materials in organ and tissue (a–c) and daily urinary excretion (d) after an accidental contamination at a wounding site. The line legend symbols, which are shown in (a), are applicable for the others (b–d). The letters of ‘‘W,’’ ‘‘M,’’ ‘‘S,’’ ‘‘A,’’ ‘‘C,’’ ‘‘P’’ and ‘‘F’’ represents the categories of DU materials (defined in the NCRP Report No.156 (NCRP, 2007)): ‘‘Weak soluble,’’ ‘‘Moderate soluble,’’ ‘‘Strong soluble,’’ ‘‘Avid soluble,’’ ‘‘Colloids,’’ ‘‘Particles’’ and ‘‘Fragments,’’ respectively.

to it. This appearance is not shown in the relationship of DU dose coefficient with aerosol monodisperse particle size in the Figure 2 of the publication of Guilmette and Parkhurst (2007). The use of their dose coefficients may give an underestimate of effective dose of DU aerosols at larger AMAD domain, say, 410 mm. Although the new HATM was introduced in the modeling, no difference was found in the systemic organ or tissue retention and the urinary excretion of U in comparison to the results modeled by using the GIT model. However, the U daily fecal excretion is greater using the GIT model than by the HATM during day 4 to day 20 after ingestion (Figure 6). The maximum difference is about 69 times GIT model upon HATM (total diet food) at day 11 after intake. The nuclear transformations, US, in the systemic organs or tissues for 238 U are identical. Comparison between HATM and GIT model, depending on the foods ingested US for stomach becomes larger for total diet and solid intake; it becomes smaller for liquid intake. US in the right colon is slightly smaller in comparison to ULI. However, US in the left colon is half of the LLI. Rectosigmoid receives the same US as the two colons. Moreover, the mouth and esophagus receive several nuclear transformations. The ICRP proposed a new Journal of Exposure Science and Environmental Epidemiology (2009) 19(5)

tissue weighting factor of 0.12 for the remainder tissues in its 2007 Recommendation (ICRP, 2007) in comparison to the factor of 0.05 in old 1990 Recommendation (ICRP, 1991). The oral mucosa is included in the remainder tissues, however, if a high dose is received by the oral mucosa, the effective dose can be affected greatly. Uranium retention after entering into the wound site depends strongly on the form and solubility of the material. The weak soluble 238U follows the same behavior as injected directly into blood. The other soluble materials, for example, moderate, strong and avid have the similar shape of retention and excretion (Figure 7). Materials in colloid, particle and fragment forms demonstrate a much lower concentration in blood, in organs and in urine as well in the earlier times, say, 100 days; however, those materials (except fragments) are close to the curves of soluble materials, especially in bone and kidney in the later days. The excretion in urine of colloid materials is even higher than the soluble materials after 200 days (Figure 7d). If DU materials can be measured in urine of the personnel, who is probably contaminated at a wound site, the effective dose received can be evaluated using the dose conversion factor for wound contaminations listed in Table 4. But the skin injury and its dose should 511


Li et al.

concomitantly be taken into account. Skin dose rate of 83 mSv h1 kBq1 in a DU-contaminated wound site can be used. This value is derived using value of 92 mSv h1 kBq1 for 238U and 84 mSv h1 kBq1 for 235U (NCRP, 2007) and their radioactivity percentage per 1 Bq DU materials. Beta radiation from DU external exposure is very low. As Marshall (2007) reported, the effect of long-term contact of DU with the skin should not result in beta radiation burns. He points out that a dose rate measured by the US Department of Defense is not larger than 2 mSv h1; it can be used to assess the external dose for soldiers in the battlefield. The typical DU area density is much lower than 1 g m2; consequently, the dose from ground shine is insignificant (10 mSva1 per 1 gm2 of DU).

Quality assurance and uncertainty of dose coefficient In the biokinetic modeling and dose calculation, hundreds of parameters are involved and it is important to validate and to compare the results with other codes. The nuclear transformations, US, of 234U, 235U, 236U and 238U accumulated in the later 50 years after intake (e.g. from 25 to 75 years) were compared with results of the DCAL program (Oak Ridge National Laboratory, Oak Ridge, TN, USA), which implemented the independent biokinetic model for decay daughters and widely used in the dosimetric calculation community in USA and in other countries. The US ratios between our and DCAL results are in the range of 0.995–1.001. Bolch et al. (2001, 2003) analyzed the influences of uncertainties of particle deposition and clearance parameters on the nuclear transformations in the ICRP HRTM. It is found that in all the regions of the respiratory tract for an adult male at light exertion, uncertainties in particle deposition fractions are distributed only over a range of about a factor of 2–4 for particle sizes between 0.005 and 0.2 mm. At particle sizes exceeding 1 mm, uncertainties in the deposition fraction within the ET regions are relatively small, but approach a factor of 20 for deposition in the bronchial region. Deposition fractions for particles above 1 mm become very uncertain within the deeper regions of the lungs, that is, bronchiolar and alveolar-interstitial. Of the four ET clearance compartments of the respiratory tract (Figure 1), uncertainties in US were highest within the ETseq tissues. Of the eight clearance compartments within the thoracic regions of the respiratory tract (Figure 1), and for particle sizes below 5 mm, uncertainties in US were highest within the BBseq tissues. Fritsch (2006) reported that uncertainties in aerosol deposition within the main regions of the respiratory tract for the default workers defined by ICRP (1994b), in the range of AMADs between 0.01 mm and 50 mm, are relatively low, for example, for deep lung depositions: the geometric standard deviation (GSD) o3; for other regions: GSD o1.8. Harrison et al. (2007) and Bailey and Puncher (2007) analyzed the uncertainty of the ICRP systemic model and 512

HRTM applied to interpretation of bioassay data for DU, respectively. In their sensitivity analysis of HRTM, representative central values and ranges in parentheses were chosen: fr ¼ 0.15 (0.01–0.5); sr ¼ 3 day1 (0.3–20 day1); ss ¼ 0.002 day1 (0.0001–0.005 day1). These values are close to the HMGU values (fr ¼ 0.37; sr ¼ 6.7 day1) used in our calculation. Chazel et al. (2003) found similar value of DU penetrator on the turret and the glacis in parentheses in Gamble’s solution: fr ¼ 0.57 (0.47); ss ¼ 0.00034 (0.00018) day1, however the value of sr ¼ 0.07 (0.06) day1 is obviously lower. Bailey and Puncher (2007) identified, by the sensitivity analysis, some parameters in all four categories (aerosol size; breathing; absorption to blood and particle transport) for which the range in probable values led to changes from the result using the central value in the dose coefficient more than 20%. The largest effects were found for parameters related to the absorption to blood. The calculation in this study showed the similar consequence that the parameter ss (indicates how fast material is absorbed to blood) affects the dose much stronger than the fA value.

Conclusion In this study, the dose coefficients of 234U, 235U, 236U, 238U after ingestion and inhalation were calculated using the solubility and dissolution parameters, which were evaluated at the RADLAB in the HMGU from the experimental data on DU penetrators. The dose coefficient of DU materials was estimated for the exposure scenarios of the general public and the workers. The 238U inhalation effective doses as a function of AMTD and AMAD are calculated aiming to provide a realistic dose coefficient that can be used for the soldiers, who are inside vehicles and outside vehicles, and for the residents who live in the areas the DU was used. The dose coefficients for inhalation used for inside and outside vehicles are different because the aerosol size measured inside and outside vehicles are different. The inhalation dose coefficients vary largely as the AMTD and AMAD change. The calculations in this study show that there is no significant change in retention and daily urinary excretion of U in the body with the old GIT model and the new HATM. However, the fecal excretion predicted by GIT model shows a higher value than that using HATM during 4–20 days after DU materials ingestion. Retention and excretion of DU materials were predicted using the new wound model developed by NCRP and ICRP. It showed that the category of soluble DU materials is retained much more in the body than other categories. The DU dose coefficients and the dose conversion factor of daily urinary excretion calculated in this study can be used in radiation dose assessment for those soldiers and veterans, who were in the conflict in Kosovo and those local residents. Journal of Exposure Science and Environmental Epidemiology (2009) 19(5)


Acknowledgements This study was in part supported by the German Ministry of Defence under the contract no. InSan I-0701-V-3803. We thank Herwig G. Paretzke for support, and Augusto Giussani and Matthias Greiter for discussions on the model of gastrointestinal absorption of DU materials. We also thank the reviewers for valuable comments.

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