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• Measurement of reaction rates in compact DT generator. • Calculation of neutron flux in targets around compact DT generator. • Determination of DD neutrons in compact DT generator. • Testing of IRDFF-II cross sections in 14.05 MeV neutrons. This paper deals with the measurement of the differential dosimetry cross sections using a small compact D-T neutron generator with 14.05 ± 0.08 MeV neutron emission (1 × 108 n/s into 4π). In spite of low neutron flux, it is possible to get measurable gamma-ray activities by irradiating a larger amount of target material placed in close geometry. The experimentally determined cross sections are in good agreement with the cross sections in the IRDFF-II dosimetry library. The comparison with other nuclear data libraries was performed as well. It is worth noting that the mean standard deviation in IRDFF-II library is about 4 %, while in the case of other data libraries they are from 5.5 % to 7.5 %. This result can be understood as a validation of IRDFF-II using 14.05 MeV neutrons and also a confirmation of the applicability of small compact generators in the measurement of activation cross sections. [ABSTRACT FROM AUTHOR]

• Measurement of reaction rates in VR-1 reactor. • Calculation of reaction rates in IRDFF-II. • Validation of IRDFF-II against experiment. • Testing of various approaches of evaluation of SACS. Cross section data are fundamental quantities which affect the accuracy of all calculations in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations with full uncertainty quantification was developed by the International Atomic Energy Agency and released in January 2020 (https://www.nds.iea.org/IRDFF). A previous version, IRDFF-1.05, was released in 2014 and experimental validation of the newly released cross section by spectrum averaged cross section (SACS) measurements is a high priority task. For such purpose, a neutron dosimeter set containing 5 target foils was activated in 2 independent experiments at the VR-1 reactor of the Czech Technical University in Prague. Care was taken to derive SACS with low uncertainties. New experimental evaluation method is in good agreement with previous approaches based on relative measurements using monitor foils. Good agreement of measured SACS and evaluated IRDFF-II cross sections is observed. Slight overestimation of evaluated ENDF/B-VIII.0 235U(n th ,f) PFNS above 10 MeV is discussed. [ABSTRACT FROM AUTHOR]

The spectral averaged cross section is an important quantity used in a validation of nuclear cross section. When the cross sections are averaged over the neutron standard field (252Cf(s,f) or 235U(n,f) neutron spectrum), they can be used for tuning of evaluations. This kind of quantities is very useful because the data in integral measurements can be determined with a significantly smaller uncertainties than the standard differential data. The experiment was aimed at the spectral average cross sections measurement and was performed in a radial channel of VR-1 reactor (with fuel enrichment 19.75 wt %). The results are in a good agreement within the uncertainties with a previous measurements in LR-0 reactor (with fuel enrichment 3.3 wt %), thus it supports the hypothesis that even significant amount of 238U(n,f) neutrons in the LR-0 reactor spectrum does not have a significant influence. The derived spectral averaged cross sections are as follows: 0.1709 ± 0.0115 mb for 89Y(n,2n), 10.738 ± 0.719 mb for 46Ti(n,p), 17.896 ± 1.181 mb for 47Ti(n,p), 0.294 ± 0.02 mb for 48Ti(n,p), 72.994 ± 4.964 mb for 54Fe(n,p), 0.528 ± 0.036 mb for 63Cu(n,α), 0.444 ± 0.029 mb for 93Nb(n,2n)92Nb* and 0.239 ± 0.016 mb for 58Ni(n,x)57Co. • Measurement of the spectral averaged cross sections. • Irradiation in the radial channel in the VR-1 reactor. • The neutron energy spectrum of the VR-1 reactor calculations. • Results comparison to the IRDFF and to the LR-0 experiments. [ABSTRACT FROM AUTHOR]