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

 

 

INTRODUCTION

 

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.

 

 

“LE GE” SYSTEM AT THE WBC FACILITY OF CIEMAT

 

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

 

EFFICIENCY CALIBRATIONS.  ASSESSMENT OF URANIUM IN LUNGS

 

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. ICRP-78 APPLICATION: DOSIMETRIC IMPLICATIONS

 

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. 

 

 

 

 

CONCLUSION

 

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.

 

 

 

REFERENCES

 

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.