[en] In our study we compared the following four unfolding codes: CRYSTAL BALL, RFSP-JUeL, SAND-II, and STAY'SL. The characteristics of the four algorithms will not be summarized here. Apart from the algorithm the codes differ in the following aspects: a) the requirements for the input information, b) the requirements for uncertainties for the input values, c) the convergence criterion, and d) the presentation of the output information. Much time and effort was needed to prepare some utility programs to have comparable situations at the input, at the output, and for the convergence. The comparison was performed for various neutron spectra. In the final stage the DOSCROS 77 library was used; this library corresponds to a large extent with the ENDF/B-IV dosimetry file. (orig./RW)

[en] The characterization of appropriate materials for fusion reactors requires a high intensity neutron source which simulates the neutron spectrum and the radiation conditions at the positions of interest in a fusion reactor (first wall). A neutron spectrum of interest is found in the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF). Various ceramic materials and some polycrystalline graphites were irradiated in this facility during two intervals of time in 1986 and 1987. The specimens were accompanied by several sets with activation detectors. This report presents the saturation activities per atom obtained from these sets. (author). 3 refs.; 8 figs.; 10 tabs

[en] Results are given of a study on the consistency between 'integral' and 'differential' cross sections data for four benchmark neutron spectra and 36 neutron reactions of importance for reactor neutron metrology. The energy dependent cross section data and their uncertainty data are obtained from the ENDF/B-V dosimetry file. The reactions have been considered with respect to the following quantities: 1. the precision of the averaged cross sections, for a specified spectrum; 2. the discrepancy between the measured and the calculated average cross section values; 3. the consistency between the measured and calculated average cross section values, described by the chi²-parameter. It was possible to take into account the available cross section covariance information present in the ENDF/B-V dosimetry file. Covariance information on the benchmark flux density spectra was not taken into account in this study

[en] Cross-section libraries in the ENDF/B-V format contain uncertainty information for a number of cross-section sets. The programs UNC 32/33 read this information and convert it to a group structure of interest for further calculations. The output of the programs becomes available in the form of a set of covariance matrices. These programs are useful for estimating the role of covariances in the calculation of reaction rates and adjusted neutron spectra, since the programs provide a convenient tool for deriving covariance matrices of cross-sections in a selected coarse group structure directly from libraries in the ENDF/B-V format. 8 refs.; 9 figs.; 6 tables

[en] This report describes the experience, obtained in the past few years with the determination of the Damage to Activation Ratio (DAR). When this ratio is multiplied by the equivalent fission neutron fluence as determined with activation detectors, one obtains the so-called damage fluence. The latter quantity is of interest when neutron induced property changes resulting from irradiations with different energy spectra have to be compared. (orig.)

[en] A comparison of some integral cross section values for several cross section libraries in the SAND-II format is presented. The integral cross section values are calculated with aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross section libraries used have all the SAND-II format. (author)

[en] A comparison of some integral cross-section values for several cross-section libraries in the SAND-II format is presented. The integral cross-section values are calculated with the aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross-section libraries used have all the SAND-II format. Discrepancies between cross-sections in the different libraries are indicated but not discussed

[en] Some results are presented for the comparison of two versions of the ENDF/B-V dosimetry file. The comparison is made for a fine group structure of the SAND-II type. (author)

[en] A review is given of the experience obtained in the past few years with the following unfolding codes: SAND-II, CRYSTAL BALL, RFSP-JUEL, and STAY'SL. The main emphasis is on the comparison of these codes, based on practical experience with application of these codes under comparable conditions

[en] Neutron selfshielding corrections for activation detectors have been studied with a computer program SELFS which is used in combination with SAND-II, the computer program for determining neutron spectra with the aid of a set of irradiated foil detectors. The method does not apply integral correction factors to the measured activities but applies energy dependent correction factors to the group cross section values. This calculation method has the advantage that correction can be calculated only once for a given foil thickness and can be applied to each neutron spectrum with known shape. The applied formula to calculate the selfshielding is valid under limited conditions but in practice the elementary formula provides data which are accurate enough. The results of calculations have been compared with literature data (e.g., for a 1/E neutron spectrum, and for the neutron spectra of the ΣΣ facility in Mol and CFRMF in Los Alamos). The cross section data required were taken from the SAND-II, ENDF/B and the BNL-325 data sources. The program SELFS is written in FORTRAN and was used on a CDC-6600 computer