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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]