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Non-destructive gamma spectrometric measurements of the isotopic composition of plutonium are routinely carried out in the framework of nuclear safeguards inspections. High-resolution spectra of Pu obtained from planar Ge detectors are de-convoluted by means of special codes such as MGA or FRAM and these results are then applied when determining the total plutonium mass using neutron coincidence measurement techniques. The radiation emission of 242Pu is so weak that its abundance cannot be measured directly; hence isotopic correlations are utilized to estimate it in low and medium burn-up Pu material. The present work investigates the feasibility of determining the total Pu/Am ratio, exploiting X-ray fluorescence lines present in the plutonium spectrum. Combining the total Pu/Am ratio with individual Pu isotopic ratios to 241Am then allows an estimate of the 242Pu to be made. To this end, reference samples of plutonium, with an 241Am content comprised between 5 and 7 %, were measured using a planar germanium detector. The spectra were processed using MGA followed by quantification of the X-ray fluorescence peak of Americium, and then further calculations, including MCNP simulations. The conclusion of the present work is that, when the americium content of the sample is high enough to produce an Am K α 1 line which can be exploited, it is possible to indirectly determine the 242Pu abundance by analysing the X-ray fluorescent lines present in the plutonium gamma spectrum, thus raising the prospect to apply this method to cases for which the algorithms based on empirical correlations cannot be applied. [ABSTRACT FROM AUTHOR]

Abstract: Room-temperature CdZnTe and CdTe detectors have been routinely used in the field of Nuclear Safeguards for many years [Ivanov et al., Development of large volume hemispheric CdZnTe detectors for use in safeguards applications, ESARDA European Safeguards Research and Development Association, Le Corum, Montpellier, France, 1997, p. 447; Czock and Arlt, Nucl. Instr. and Meth. A 458 (2001) 175; Arlt et al., Nucl. Instr. and Meth. A 428 (1999) 127; Lebrun et al., Nucl. Instr. and Meth. A 448 (2000) 598; Aparo et al., Development and implementation of compact gamma spectrometers for spent fuel measurements, in: Proceedings, 21st Annual ESARDA, 1999; Arlt and Rudsquist, Nucl. Instr. and Meth. A 380 (1996) 455; Khusainov et al., High resolution pin type CdTe detectors for the verification of nuclear material, in: Proceedings, 17th Annual ESARDA European Safeguards Research and Development Association, 1995; Mortreau and Berndt, Nucl. Instr. and Meth. A 458 (2001) 183; Ruhter et al., UCRL-JC-130548, 1998; Abbas et al., Nucl. Instr. and Meth. A 405 (1998) 153; Ruhter and Gunnink, Nucl. Instr. and Meth. A 353 (1994) 716]. Due to their performance and small size, they are ideal detectors for hand-held applications such as verification of spent and fresh fuel, U/Pu attribute tests as well as for the determination of 235U enrichment. The hemispherical CdZnTe type produced by RITEC (Riga, Latvia) [Ivanov et al., 1997] is the most widely used detector in the field of inspection. With volumes ranging from 2 to 1500mm3, their spectral performance is such that the use of electronic processing to correct the pulse shape is not required. This paper reports on the work carried out with a large volume (15×15×7.5mm3) and high efficiency hemispherical CdZnTe detector for the determination of 235U enrichment. The measurements were made with certified uranium samples whose enrichment ranging from 0.31% to 92.42%, cover the whole range of in-field measurement conditions. The interposed shield thickness varied from 2mm of aluminium to 16mm of steel. The results showed that, despite the detector instability, it is possible to achieve a good precision within a short counting time and for low enriched uranium, even when thick shielding is interposed between the source and the detector. [Copyright &y& Elsevier]

Abstract: The traditional method used to determine the Uranium enrichment by nondestructive analysis is based on the “enrichment meter principle” [1]. It involves measuring the intensity of the 186keV net peak area of 235U in “quasi-infinite” samples. A prominent factor, which affects the peak intensity, is the presence of gamma absorbing material (e.g., container wall, detector cover) between the sample and the detector. Its effect is taken into consideration in a commonly called “wall thickness” correction factor. Often calculated on the basis of approximations, its performance is adequate for small attenuation factors applicable to the case of narrow beams. However these approximations do not lead to precise results when wide non-parallel beams are attenuated through thick container walls. This paper is dedicated to the calculation by numerical integration of the geometrical correction factor (K wtc) which describes the effective mean path length of the radiation through the absorbing layer. This factor was calculated as a function of various measurement parameters (types and dimensions of the detector, of the collimator and of the shielding) for the most commonly used collimator shapes and detectors. Both coherent scattering (Rayleigh) and incoherent scattering (Compton) are taken into account for the calculation of the radiation interaction within the detector. [Copyright &y& Elsevier]