THE PROJECT IS CONCERNED WITH, IN ASSOCIATION WITH SEVERAL EUROPEAN LABORATORIES, THE DEVELOPMENT OF A DOSEMETER USING SOLID STATE DETECTORS CAPABLE OF PROVIDING A RESPONSE AS CLOSE AS POSSIBLE TO DOSE EQUIVALENT IN A MIXED FIELD OF IONISING RADIATIONS.
Good sensitivity can be obtained for gamma ray detection in the frame of international requirements for a gamma energy of above 100 keV and an improvement in measurement of lower energy could be achieved by association with at least 2 or 3 tin filters. For the fast neutron channel, the dose response could be flattend with an assocation of 2 different thicknesses of polyethylene. In that case, it would be necessary to associate several of the prototypes envisageable for an area dosimetric system.
The present status of the electronic setup development indicates that a pocket sized dosimeter can be realised both for gamma rays and neutrons (thermal and fast) with the possibility to display and to numerise the contribution of each type of radiation to the total dose.
Calculations have been done on the response of some channels of an electronic pocket dosimeter. A set of computer codes has been used for the theoretical calculations. For X-ray and y-ray detection efficiency curves, the calculations have been done for energies between 20 keV and 1.5 meV, a fixed dose of 1.5 v entering the above silicon surface barrier sensor placed behind a 1.5 mm aluminium converter and 8 different shielding thicknesses of tin foils. These efficiency curves have been calculated for both detection modes: digital counting mode with 3 cutoffs and integral current mode. The calculations have also been performed for several integration times and for 2 lifetimes of the minorites. Fast neutron interactions inside several thicknesses of polyethylene have been simulated and results obtained for the countings of the emitted recoil protons detected by the silicon.
From the results it can be concluded that, for personal dosimetric survey, it is possible to measure X-ray and y-ray doses accurately. It may also be possible to extend the acceptable energy range. For fast neutrons, the sensitivity of the device is lower than for gamma but in all cases such an electronic semiconductor dosimeter h as sensitivities comparable to that of other detectors with the advantage of low bias working possibilities.
1. THEORETICAL STUDIES
AS THE METHODS OF DETECTING RADIATION WITH SILICIUM AND WITH CADMIUM TELLURIDE ARE KNOWN, A MONTE CARLO COMPUTER PROGRAMME WILL BE USED TO DETERMINE THE RESPONSE TO DIFFERENT COMPONENTS OF RADIATION FIELDS (WITH REGARDS TO PARTICLES AND ENERGY); A SYSTEM OF SCATTERERS AND FILTERS WILL BE INCLUDED IN THE CALCULATIONS IN ORDER TO OBTAIN A DEFINED RESPONSE (NORMALISATION TO CAESIUM 137 FOR THE DETECTION OF GAMMA RAYS OF VARIOUS ENERGIES IN SILICIUM, THE RESPONSE OF CADMIUM TELLURIDE TO THERMAL NEUTRONS, EVALUATION OF A CONVERTER TO BE PLACED ON A SILICIUM DETECTOR TO OBTAIN A CONSTANT RESPONSE TO FAST NEUTRONS AS FUNCTION OF ENERGY). IN ORDER TO OBTAIN RESPONSE RANGE FROM LOW DOSES TO HIGH DOSES, INVESTIGATIONS WILL BE CARRIED OUT ON USING THE PULSE MODE AND INTEGRATED CURRENT. THE CALCULATIONS WILL BE MADE ON THE COMPUTER AT THE PHYSIQUE NUCLEAIRE AT THE CENTRE IN LYON; THEY WILL PROVIDE THE THEORETICAL DEFINITION OF THE STRUCTURE OF THE VARIOUS DETECTORS TO BE UTILIZED IN ORDER TO ACHIEVE THE BEST EFFICIENCY FOR THE DETECTION OF THE COMPONENTS OF A FIELD OF MIXED RADIATIONS.
2. EXPERIMENTAL STUDIES
DEPENDING ON THE THEORETICAL RESULTS OBTAINED, THE DETECTORS CHOSEN WILL BE DEVELOPED AT THE PHASE LABORATORY AND FILTERS AND SCATTERERS WILL BE ACCORDINGLY ADAPTED. INITIAL EXAMINATION OF EACH UNIT THUS DEFINED WILL BE MADE WITH THE HELP OF RADIATION SOURCES AVAILABLE AT THE CENTRE DE RECHERCHE NUCLEAIRES: 60-CO FOR GAMMA RAYS, THERMAL NEUTRONS FROM THE REACTOR ULYSSE, FAST NEUTRONS FROM A PUBE-SOURCE OR AMBE-SOURCE.
AT THIS STAGE COLLABORATION WITH THE LABORATORIES ASSOCIATED WITH THE PROJECT WILL BE NECESSARY IN ORDER TO EXAMINE THE VALIDITY OF THE RESPONSE TO DIFFERENT ENERGIES WITH THE HELP OF THEIR CALIBRATED SOURCES.
3. CONSTRUCTION OF THE DOSEMETER
THIS WILL INVOLVE DETERMINING THE STRUCTURE OF THE ASSOCIATED MEASURING ELECTRONICS. THE QUALITY FACTORS OF DIFFERENT RADIATIONS WILL HAVE TO BE TAKEN INTO ACCOUNT AND TO BE INCORPORATED INTO THE READING OF EACH DETECTOR AS DETERMINED BEFORE.
SUBTRACTION OF RESPONSES WILL BE NECESSARY (CONTRIBUTION OF GAMMA RAYS IN THE PRESENCE OF NEUTRONS, FOR EXAMPLE); SUBSEQUENTLY, THE INFORMATION SPECIFIC TO EACH DETECTOR WEIGHTED BY THEIR CONTRIBUTION TO DOSE EQUIVALENT WILL BE ADDED IN SUCH A WAY AS TO OBTAIN A CONVERSION IN GLOBAL DOSE EQUIVALENT (SIEVERT) FOR THE WIDEST POSSIBLE ENERGY SPECTRA AND FOR DIFFERENT PENETRATING RADIATIONS.
USE WILL ALSO BE MADE OF A MICROPROCESSOR WHICH WILL BE PROGRAMMED TO PROVIDE THE RESULTS REQUIRED.