I
u r w r
URANIUM IN VITRO BIOASSAY ACTION LEVEL USED TO SCREEN WORKERS FOR CHRONIC INHALATION INTAKES OF URANIUM MILL TAILINGS* R. H. Reif, J. B. Turner, and D. S. Carlson' mSv (100 mrem) annual effective dose equivalent (U.S. DOE 1988). Complying with the DOE internal dose evaluation requirement (U.S. DOE 1988) would be difficult using separate bioassay of each uranium decay chain radionuclide. Moreover, analyzing all occupational radionuclides would be relatively expensive. Insoluble 230This a prime example of a radionuclide that is difficult to measure with bioassay. 230Th is primarily found in insoluble chemical forms at the DOE Uranium Mill Tailings Remedial Action Project (UMTRAP) sites (Lawrence Livermore National Laboratory 1990) and is assigned inhalation class Y in ICRP Publication 30 (ICRP 1979). Insoluble 230This tenaciously retained in the lung and various systemic compartments (ICRP 1979), and only a small fraction (US. NUREG 1987) of the inhalation intake quantity is excreted via the urinary pathway (ICRP 1988);therefore, it is difficult to detect an inhalation intake of insoluble 230Thcorresponding to 1 mSv (100 mrem) annual effective dose equivalent (AEDE).
Abstract-A uranium in vitro bioassay (urinalysis) action level was derived for use at the Department of Energy's Uranium Mill Tailings Remedial Action Project sites to identify chronic inhalation intakes of uranium mill tailings causing 0.5 mSv (50 mrem) annual effective dose equivalent. All radionuclides in the 238Udecay chain that contribute 1% or more to the annual effective dose equivalentfrom an inhalation intake of uranium mill tailings were included in the derivation of the urinalysis action level. Using a chronic inhalation intake model, the uranium urinalysis action level for a 24-h urine sample, collected on a quarterly schedule, was calculated to be 1.5 pg. Health Phys. 63(4):398-401; 1992 Key words: bioassay; internal dose; tailings, uranium; uranium
INTRODUCTION RESIDUAL uranium isotopes that were not removed during the uranium milling process consist of 238Uand 234Uin the soluble chemical form (Lawrence Livermore National Laboratory 1990). Inhalation intakes of soluble uranium can be easily detected by urine bioassay because the uranium is rapidly cleared from the lung (ICRP 1979), and practically all of the uranium that is taken up by the systemic compartment is excreted via the urinary pathway (ICRP 1988). Because of the sufficiently low minimum detectable activity (MDA) that can be achieved by fluorimetry (ASTM 1986), and the relatively large fraction of an inhalation intake of soluble uranium that deposits in the urine, monitoring for uranium in urine provides a method for monitoring all uranium decay products, given that the activity ratios are known. Uranium urinalysis can be used to meet the U.S. Department of Energy (U.S. DOE) internal dose evaluation requirement of identifying intakes of all radionuclides of occupational sources that correspond to 1
MATERIALS AND METHODS The uranium urinalysis action level was derived for radionuclides from the 238Udecay chain that contribute 1% or more to the AEDE from an inhalation intake of an assumed uranium mill tailings activity mixture in all breathing zones. Annual effective dose conversion factors (ADCF), obtained from Dunning (1 985), were used to determine which 238Udecay chain radionuclides contribute at least 1% of the AEDE per unit intake (refer to Table 1). These radionuclides were used to derive the uranium urinalysis action level and are considered radionuclides of concern for internal dose assessment, herein referred to as the UMTRAP occupational radionuclides. Selected UMTRAP sites that were under remediation by MK-Ferguson Company and Chemical Waste Management Federal Environmental Services were evaluated to identify the site with the highest annual average concentrations of UMTRAP occupational radionuclides in the air. The Grand Junction UMTRAP site (Grand Junction, CO), with an annual average 230Thair concentration of 5.6 x lo-' Bq mL-', was selected for dose assessment.
*Work supported by the U.S. Department of Energy under Contract DE-AC04-83AL18796. CWM Federal Environmental Services, Inc., 2309 Renard Place SE, Suite 300, Albuquerque, NM 87 106. (Manuscript received 22 October 199 1; revised manuscript received 8 May 1992, accepted 15 May 1992) 0017-9078/92/$3.00/0 Copyright 0 1992 Health Physics Society 398
Bioassay screens for uranium mill tailings 0 R. H. REIFET AL.
The uranium decay chain radionuclides associated with airborne radioactive particulates are in approximate equilibrium, except for 222Rnand its short-lived decay products and the first four radionuclides (238U, 234Th,234Pa,and 234U),because most of the first four radionuclides have been removed by the uranium milling process. Due to removal during the milling process, 234Uand 238U(which are in equilibrium with each other) exist at 8.3% of the tailings mixture activity. The percent tailings mixture activity of 238Uand 234U(8.3%) represents an average value that was calculated from UMTRAP site characterization data. The characterization data was obtained from the Durango UMTRAP Environmental Impact Statement (U.S. DOE 1985), Grand Junction UMTRAP Environmental Impact Statement (U.S. DOE 1986), and the Mexican Hat UMTRAP Environmental Assessment (US. DOE 1987). A worker, characteristic of the Reference Man (ICRP 1975), is assumed to be chronically exposed to the average annual 230Thconcentration at the Grand Bq m-3) for 2,000 h while Junction site (5.6 x performing light work activity (ICRP 1975). The inhalation intake of 230Thwas calculated as:
I = CT~XOTBK,
Table 1. Annual dose conversion factorsa Nuclide 238"
234Th 234Pa 234u
Science Applications International Corporation (SAIC), REMedy. A Microcomputer Program for the Evaluation of Radionuclide Intake and Dose from Bioassay or Air Concentration Data, 10210 Campus Point Drive, San Diego, CA 92 12I.
1 Pm AMADb (Sv Bq-I) 2.1 x 10-7 9.5 x 10-9 5.4 x 10-13 2.4 x 10-7 7.6 X 2.0 x 4.3 x 2.4 X 2.1 x 10-9 1.8 x 10-9 3.0 x 10-7
1 Pm AMADb (rem PCi-') 0.77 3.5 x 10-2 2.0 x 10-6 0.87 28.0 7.3 1.6 x 10-3 8.7 x lo-'' 7.6 x 10-3 6.6 x lo-'
Contribution to total dose (%)
1.6
u r w r
URANIUM IN VITRO BIOASSAY ACTION LEVEL USED TO SCREEN WORKERS FOR CHRONIC INHALATION INTAKES OF URANIUM MILL TAILINGS* R. H. Reif, J. B. Turner, and D. S. Carlson' mSv (100 mrem) annual effective dose equivalent (U.S. DOE 1988). Complying with the DOE internal dose evaluation requirement (U.S. DOE 1988) would be difficult using separate bioassay of each uranium decay chain radionuclide. Moreover, analyzing all occupational radionuclides would be relatively expensive. Insoluble 230This a prime example of a radionuclide that is difficult to measure with bioassay. 230Th is primarily found in insoluble chemical forms at the DOE Uranium Mill Tailings Remedial Action Project (UMTRAP) sites (Lawrence Livermore National Laboratory 1990) and is assigned inhalation class Y in ICRP Publication 30 (ICRP 1979). Insoluble 230This tenaciously retained in the lung and various systemic compartments (ICRP 1979), and only a small fraction (US. NUREG 1987) of the inhalation intake quantity is excreted via the urinary pathway (ICRP 1988);therefore, it is difficult to detect an inhalation intake of insoluble 230Thcorresponding to 1 mSv (100 mrem) annual effective dose equivalent (AEDE).
Abstract-A uranium in vitro bioassay (urinalysis) action level was derived for use at the Department of Energy's Uranium Mill Tailings Remedial Action Project sites to identify chronic inhalation intakes of uranium mill tailings causing 0.5 mSv (50 mrem) annual effective dose equivalent. All radionuclides in the 238Udecay chain that contribute 1% or more to the annual effective dose equivalentfrom an inhalation intake of uranium mill tailings were included in the derivation of the urinalysis action level. Using a chronic inhalation intake model, the uranium urinalysis action level for a 24-h urine sample, collected on a quarterly schedule, was calculated to be 1.5 pg. Health Phys. 63(4):398-401; 1992 Key words: bioassay; internal dose; tailings, uranium; uranium
INTRODUCTION RESIDUAL uranium isotopes that were not removed during the uranium milling process consist of 238Uand 234Uin the soluble chemical form (Lawrence Livermore National Laboratory 1990). Inhalation intakes of soluble uranium can be easily detected by urine bioassay because the uranium is rapidly cleared from the lung (ICRP 1979), and practically all of the uranium that is taken up by the systemic compartment is excreted via the urinary pathway (ICRP 1988). Because of the sufficiently low minimum detectable activity (MDA) that can be achieved by fluorimetry (ASTM 1986), and the relatively large fraction of an inhalation intake of soluble uranium that deposits in the urine, monitoring for uranium in urine provides a method for monitoring all uranium decay products, given that the activity ratios are known. Uranium urinalysis can be used to meet the U.S. Department of Energy (U.S. DOE) internal dose evaluation requirement of identifying intakes of all radionuclides of occupational sources that correspond to 1
MATERIALS AND METHODS The uranium urinalysis action level was derived for radionuclides from the 238Udecay chain that contribute 1% or more to the AEDE from an inhalation intake of an assumed uranium mill tailings activity mixture in all breathing zones. Annual effective dose conversion factors (ADCF), obtained from Dunning (1 985), were used to determine which 238Udecay chain radionuclides contribute at least 1% of the AEDE per unit intake (refer to Table 1). These radionuclides were used to derive the uranium urinalysis action level and are considered radionuclides of concern for internal dose assessment, herein referred to as the UMTRAP occupational radionuclides. Selected UMTRAP sites that were under remediation by MK-Ferguson Company and Chemical Waste Management Federal Environmental Services were evaluated to identify the site with the highest annual average concentrations of UMTRAP occupational radionuclides in the air. The Grand Junction UMTRAP site (Grand Junction, CO), with an annual average 230Thair concentration of 5.6 x lo-' Bq mL-', was selected for dose assessment.
*Work supported by the U.S. Department of Energy under Contract DE-AC04-83AL18796. CWM Federal Environmental Services, Inc., 2309 Renard Place SE, Suite 300, Albuquerque, NM 87 106. (Manuscript received 22 October 199 1; revised manuscript received 8 May 1992, accepted 15 May 1992) 0017-9078/92/$3.00/0 Copyright 0 1992 Health Physics Society 398
Bioassay screens for uranium mill tailings 0 R. H. REIFET AL.
The uranium decay chain radionuclides associated with airborne radioactive particulates are in approximate equilibrium, except for 222Rnand its short-lived decay products and the first four radionuclides (238U, 234Th,234Pa,and 234U),because most of the first four radionuclides have been removed by the uranium milling process. Due to removal during the milling process, 234Uand 238U(which are in equilibrium with each other) exist at 8.3% of the tailings mixture activity. The percent tailings mixture activity of 238Uand 234U(8.3%) represents an average value that was calculated from UMTRAP site characterization data. The characterization data was obtained from the Durango UMTRAP Environmental Impact Statement (U.S. DOE 1985), Grand Junction UMTRAP Environmental Impact Statement (U.S. DOE 1986), and the Mexican Hat UMTRAP Environmental Assessment (US. DOE 1987). A worker, characteristic of the Reference Man (ICRP 1975), is assumed to be chronically exposed to the average annual 230Thconcentration at the Grand Bq m-3) for 2,000 h while Junction site (5.6 x performing light work activity (ICRP 1975). The inhalation intake of 230Thwas calculated as:
I = CT~XOTBK,
Table 1. Annual dose conversion factorsa Nuclide 238"
234Th 234Pa 234u
Science Applications International Corporation (SAIC), REMedy. A Microcomputer Program for the Evaluation of Radionuclide Intake and Dose from Bioassay or Air Concentration Data, 10210 Campus Point Drive, San Diego, CA 92 12I.
1 Pm AMADb (Sv Bq-I) 2.1 x 10-7 9.5 x 10-9 5.4 x 10-13 2.4 x 10-7 7.6 X 2.0 x 4.3 x 2.4 X 2.1 x 10-9 1.8 x 10-9 3.0 x 10-7
1 Pm AMADb (rem PCi-') 0.77 3.5 x 10-2 2.0 x 10-6 0.87 28.0 7.3 1.6 x 10-3 8.7 x lo-'' 7.6 x 10-3 6.6 x lo-'
Contribution to total dose (%)
1.6
