IN-VIVO
MONITORING OF WORKERS IN THE NUCLEAR FUEL CYCLE. ICRP-78 APPLICATION.
Maria Antonia Lopez, Teresa Navarro
Internal Dosimetry-CIEMAT
Avda. Complutense 22, 28040 Madrid, Spain
The hazards associated with Uranium processing
include both, radiological and chemical toxicity. The periodic assessment of
individual internal exposures in occupational workers must be associated to
realistic monitoring programs adapted to the annual dose limit established in
the last ICRP recommendations. Taking into account the imminent change of the
Regulations in Spain, the sensitivity of the detection system to use as In-Vivo
technique becomes the clue of the matter.
Spanish workers involved in nuclear fuel
fabrication processes using 3,6% enriched Uranium are included in a routine
monitoring program of In-Vivo measurements to detect insoluble 238U
(234Th) and 235U in lungs. The Lung measurements with a
LE Ge detector system (four germanium detectors array) are performed once in a
year at the WBC facility of CIEMAT, for a counting time of 2700 s.
The “LEGe“ Lung Counting system consists of
four Canberra Low Energy Ge detectors mounted, two each, as arrays in two ACTII
cryostats. The active area of each detector is of 3800 mm2, with a
diameter of 70 mm and a thickness of 25 mm; each device is equipped with a
Carbon Epoxy window, 0.5 mm thick. The ABACOS-GPC software of GENIE-PC provides
a complete set of operating procedures to analyse subjects, to perform calibration
functions and quality assurance operations.
Figure 1 shows the four “LE Ge” detectors mounted on an adjustable support which allows the proper placement of the devices above the chest of the counting subject in a reclined dentist chair geometry.
The “LE Ge” detector system is defined by the
software as four inputs or individual detectors, with an analysis Energy range
of 10-1000 KeV. Three Detector Groups are defined by the system for efficiency
calibration purposes: the All Summed group, as composition of the four
germanium detector spectra, Det1-2, as the sum of the two
individual detectors over the right lung, and the Det3-4 group corresponding
with the left lung. The final result of a lung measurement is analysed from
individual detectors and for the three composed spectra, obtaining information
about the distribution of the contaminant in the lung area
1.-
Multi-Efficiency Calibration
Lung calibration is performed with a LLNL torso
phantom and five overlay plates, covering a Chest Wall Thickness range from
1.67 to 4.27 cm. The Multiple Efficiency Curve Calibration was carried out
using a source containing 241Am and 152Eu, providing
photon emissions with energies between 13 keV and 1400 keV.
Figure
2 shows the results of the Multi-Efficiency Calibration
performed for the All Summed Detector Group:
Efficiency
calibrations were obtained for the 3 detector groups from the composition of
the corresponding spectra. The final combined efficiency calibration with data
from all the five CWT configurations were approved to use in subsequent
analyses, performing the proper efficiency correction for CWT values ranging
from 1.67 cm to 4.27 cm or extrapolating to CWT values lower than 1.7 cm or
higher than 4.27 cm. The mathematical functions of counting efficiency vs.
energy were empirically determined as polynomials Ln Eff = An(LnE)n,
whose coefficients can be seen in Table 1:
Table1:- Efficiency
Calibration- All Summed (sum of spectra from the 4 LE Ge detectors)
Ln Eff= A0 + A 1(LnE) +
A 2(LnE)2 + A 3(LnE)3 + A 4(LnE)4
+ A 5(LnE)5 + A 6(LnE)6
CWT |
A0 |
A1 |
A2 |
A3 |
A4 |
A5 |
A6 |
CORE(1.67cm) |
60.90 |
-128.04 |
81.96 |
-25.03 |
4.04 |
-0.33 |
0.01 |
1B (2.57 cm) |
-475.41 |
544.50 |
-263.48 |
68.03 |
-9.85 |
0.76 |
-0.02 |
2B (3.27 cm) |
-425.06 |
479.97 |
-229.78 |
58.79 |
-8.44 |
0.64 |
-0.02 |
3B (3.57 cm) |
-333.34 |
360.00 |
-166.55 |
41.49 |
-5.84 |
0.44 |
-0.01 |
4B (4.27 cm) |
-430.38 |
477.29 |
-224.82 |
56.67 |
-8.03 |
0.60 |
-0.02 |
2.- Uranium Efficiency Calibration
To guarantee the best
accuracy of the results in the Routine Monitoring of workers involved in the
Nuclear fuel fabrication Industry, Efficiency calibrations using Uranium
sources have been obtained.
As calibration sources, WBC facility of CIEMAT
counts with two pairs of lungs, one with 12 kBq of Natural Uranium and the
other with 500 Bq (very low Activity) of 3% enriched Uranium. Counting times
were selected in order to obtain uncertainties of 5% associated to the peak
areas.
A set of phantom measurements, using a counting
time of 7200 s, were performed using Natural Uranium Lungs and Type B overlay
plates (simulating 50% muscle, 50% adipose tissue); a chest wall thickness
range between 1.67 cm and 4.27 cm were considered. The spectra were analysed
from ABACOS-GPC software through Genie-PC, in order to identify the peaks and
to determinate the areas (counts) associated to each of the emissions of
interest to be used in the specific Uranium calibration.
Empirical calibration functions
of the Counting Efficiency (Eff) depending on the Chest Wall Thickness (CWT) of
the subject were assessed for the emissions of 63 keV and 93 keV from 234Th
and 144 keV, 186 keV and 205 keV from 235U. The calibration
functions Eff(c/g)= f(CWT), with CWT in cm, is obtained from the fit of 5
experimental pairs of data of Efficiency and the corresponding Chest Wall
Thickness.
3.-
Assessment of Activities. Routine monitoring of workers involved in the
fabrication of nuclear fuel elements:
The shielded room of the Whole Body Counting facility of CIEMAT is the unique Laboratory in Spain with capability for In-Vivo measurement of Actinides deposited in Lungs.
In the routine monitoring of workers involved in the fabrication of nuclear fuel elements for nuclear power plants, the procedure for the assessment of Activity deposited in Lung is as follows. An In-Vivo Lung measurement with the “LE Ge” detector system is performed for a counting time of 2700 s. The detection and identification of radionuclides inhaled is carried out from the Peak Analysis of the collected spectra, using the General Calibration described above, through the ABACOS-GPC software. Any isotope present in the body emitting in an Energy range between 10 keV and 1000 keV will be perfectly detected and identified; the Multi-Efficiency function will be used for the first assessment of the Activities.
In case of detection of internal contamination
by Uranium, if the 63 keV emission of 234Th is identified, the
measurement is firstly reported with great accuracy from the Activity of 238U
in Lungs using the General Calibration and the Multi-Efficiency function. A
more complete study will be carried out trying to improve the assessment of the
Activity associated to both, 234Th and 235U. The Specific
Efficiency Calibration obtained from Uranium sources will be used through
GENIE-2000 in connection with a “home-made” software developed in Visual Basic.
The corresponding Efficiency correction for the peaks detected (63 keV emission
of 234Th and 144 keV, 186 keV emissions from 235U) using
the specific Uranium calibration are performed and new values of Activities are
obtained for both Uranium isotopes, 238U and 235U.
MDA calculations are performed by ABACOS-GPC software, using Multi-Efficiency Efficiency calibration and applying the methods developed by Currie and ANSI N13.30 criteria, with a 95% confidence factor.
The values of Chest Wall Thickness (CWT)
belonging to the occupational workers included in different monitoring programs
for lung counting are recorded in a Dosimetric Database. Most of the people
(70%) have CWT’s included in the interval between 2.57 and 3.27 cm. The LE Ge
detector system of CIEMAT supplies MDA’s of 5 Bq for 235U and 38 Bq
for 234Th, associated to a Chest Wall Thickness of 2.6 cm, for a
counting time of 2700 s.
If the frequency continues to be once in a year
and the Annual Dose Limit is reduced to 20 mSv, the dosimetric implications of
the new situation must be studied.
The recycling model for Uranium recommended in
ICRP-69 is based on the generic alkaline earth model given in ICRP-67. As it is
shown in ICRP-78 Publication, the model describes in detail the kinetics of
Uranium in bone (main site of deposition and retention) and considers as well
the retention in liver, kidneys and other soft tissues. It takes account of
initial uptake onto bone surfaces, transfer from surface to bone volume, and
recycling from bone and other tissues to plasma.
ICRP-78 dosimetric data reported for the
inhalation of Type S Uranium (high insoluble compounds such as UO2
or U3O8) have been applied in this study. Considering
both, Dose coefficients e(50) Sv/Bq calculated taking into account the new
respiratory tract Model (ICRP-66), and the new Annual Dose Limits of 20 mSv
(ICRP-60), the assessment of the ALI associated to 3,6% enriched Uranium gives
the value of 1000 Bq of U(3,6%) as Annual Limit of Intake.
Dose assessments associated to the inhalation
of 3.6% enriched Uranium compounds are firstly performed from the retention of 235U
in Lungs. Taking into account a Minimum Detectable Activity of 5-6 Bq of 235U
and considering a routine monitoring frequency of one in a year with the
assumption of time of Intake in the half of the interval, the assessment of
Intake associated to the MDA results in 150 Bq of 235U. Thus, for
this Type S isotope we are considering the detection of minimum occupational
Intakes associated to a 15% of the new Maximum permitted Dose, around 3
mSv.
One of the consequences of this new situation
is to study the possibility of establishing a new monitoring program increasing
the frequency of the In-Vivo lung measurements; the objective is to improve the
accuracy in the dose assessment in case of Uranium inhalation applying new ICRP
recommendations. These dosimetric implications cause important disturbing
factors in the National Uranium Company that must also be taken in
consideration. More than a hundred workers are involved in the nuclear fuel
fabrication process. These workers are located more than 400 km far away from
the WBC facility of CIEMAT, where the lung measurements are performed. Each
employee has to leave his workplace and to be replaced in the fabrication
process for one day when is required for the annual lung measurement. Thus, not
only the economic consequences but also the difficulty of procedure represent
additional problems for the company.
To be able to detect Intakes corresponding to
15% of Annual Dose Limit seems acceptable from a point of view of occupational
monitoring for radiation protection purposes, in the framework of last ICRP
recommendations. Before making any decision, human and economic factors must
also be considered.
The main interest of lung retention
measurements is the realism in the assessment of workers exposure. The
technique is not adequate for the routine monitoring for exposures of most
soluble compounds. In case of internal exposure to insoluble uranium compounds
through inhalation, taking into account the uncertainty associated to the
exposure time, a higher measurement frequency will be studied for a more
realistic dose assessment.
1. ANSI N 13.30- Standard performance
criteria for radiobioassay . August, 1995 version.
2. ICRP-60- 1990 Recommendations of
the International Commission on Radiological Protection.
3. ICRP-66- Human Respiratory Tract
Model for Radiological Protection
4. ICRP-67- Age-dependent Dose to Members of the Public from Intakes
of radionuclides: Part 2.
5. ICRP-69- Age-dependent Dose to Members of the Public from Intakes
of radionuclides: Part 3.
6.
ICRP-78- Individual
Monitoring for Internal Exposure of workers. Replacement of ICRP Publication
54.
7. Groff, D and Booth, L.F. Factory calibration of the Canberra scanning
bed lung counter for General Electric-Wilmington, NC. September, 1995
8. Lopez Ponte M.A., T. N. Bravo; A Low Energy Germanium detector system for
lung counting at the WBC facility of CIEMAT. Radiation Protection and
Dosimetry Vol 89, Nos 3-4, pp. 221-227
(2000)
9. Model S434 AbacosGPC (Canberra
Industries). User’s Manual. 1997.