The G-1 intercomparison is one of the first undertakings of the International Atomic Energy Agency for the investigation of the performance of the routinely used evaluating programs for gamma ray spectra of semiconductor detectors. The details and the conclusions of this procedure are presented.
A quantitative method for the analysis of -emitters in infinitely thin and thick samples was described in Part I. The calculation of errors in intensity, intensity ratio and activity concentration is discussed here in detail. Different definitions of detection sensitivity are compared and evaluated on the basis of the relative statistical error associated therewith. Dependence between the relative error of the net signal and the required measurement time is deduced and illustrated.
A method for determining bulk or surface activity concentration of -emitters in infinitely thick or thin samples using energy selective detectors without any prior chemical separation or qualitative analysis was introduced in Part I. In addition to that method, a correction procedure is often required in order to compensate for the -sensitivity of the -detector. A system or two — preferably identical —detectors positioned to the front and rear side of a sample has been established. The rear side detector is shielded against -radiation. The -response of both detectors is determined as a function of -energy using monoenergetic calibration standards. Response pulses are selected to form 8 or 16 energy intervals, and separate series of correction coefficients are then computed for each interval and standard energy using the ratio of the respective counts. When a sample having mixed - and -radiation is evaluated, an effective
-energy is first calculated from the pulse height distribution of the rear side detector, the appropriate series of correction coefficients is selected and finally net -counts are generated from the counts of the front side detector. Sensitivities in natural and elevated -background are presented.
The activity concentration of bulk and surface samples contaminated with -emitting radioisotopes is difficult to measure without the a priori knowledge of the nature of the non-gamma emitting components. Beta-emitters cannot be identified from any measured spectral distribution. The counting efficiency of the measuring system changes significantly with -energy so it cannot be assumed to have a single value obtained with a standard source with known energy. Application of an energy selective -detector is introduced for determining bulk and surface activity concentration. Samples of infinitely thick or infinitelythin nature are to be prepared. The distribution of -energy deposited in the detector is registered as counts in 8 or 16 energy intervals. No information is needed on the qualitative composition of the sample. Normalised integral distributions (intensity ratios) are derived from the count rates of the intervals. These distributions are then compared to calibrated intensity ratios obtained with suitable standard sources. An average (effective) counting efficiency is generated from this comparison by a special algorithm. Activity concentration of an unknown sample is then obtained using this average efficiency. Calibration and sensitivity data are presented for different types of bulk and surface samples.
Investigations on the disintegration rate of fission products of238U and239Pu are presented. The intensity of the - and -radiation of fission products were measured continuously in an interval of 1–1300 hours following the fission, offering the possibility for determining the general and specific characteristics of the individual fission products. A universal measuring procedure was elaborated for the rapid in situ determination of the dosimetric features of fission products, which is suitable for the accurate evaluation and prediction of external absorbed dose even in case of fission products of various origin and unknown composition.
Authors:P. Ormai, A. Fritz, J. Solymosi, I. Gresits, E. Hertelendi, Z. Szúcs, N. Vajda, Zs. Molnár and P. Zagyvai
In the execution of disposal of low and intermediate level radioactive wastes, it is important to evaluate accurately the kind and quantity of each radionuclide in the wastes. For such an evaluation, correlation of non-gamma-emitting nuclides based on gamma-emitting nuclides is recommended and regarded as a practical method. This method necessitates a completion of a highly accurate and reliable nondestructive assay system of gamma-emitting nuclides for practical use. In 1992, in support of the new waste disposal program in Hungary, Paks NPP initiated a waste characterization program to determine the radiological properties of its radwastes. A segmented gamma scanning system has been set up to measure the gamma-emitting nuclides in 200 litre low level drums following in-drum compaction. In the framework of the program a radiochemical analysis sub-program was started to determine the long-lived non-gamma emitting radionuclides, mainly those listed in the US regulatory document (10CFR61). The radionuclides of interest have been3H,14C,90Sr,55Fe,59Ni,99Tc,129I and TRUs. Sample preparation techniques and measurement methods have been selected and used. Newly developed or adopted methods have been tested on real liquid radwaste streams such as concentrates, ion-exchange resin and sludge. The measurements taken so far have revealed brand new information and data on radiological composition of waste of WWER-type reactors. In the next stage of the characterisation program attempt will be made for providing correlation factors between the gamma and non-gamma-emitting radionuclides in different waste streams. Short description of the methods and results on waste inventory are given by highlighting the problem areas.