Search

The measured and evaluated excitation functions are fundamental quantities that affect the accuracy of all calculations in nuclear applications. Some cross sections, such as 14N(n,p)14C, have added value for special applications, as these reactions may be responsible for possible contamination in industrial processes such as spent fuel reprocessing. For the validation of the evaluated cross sections, we can rely on the comparison of the calculated spectrum averaged cross sections (SACS) for the given neutron spectrum with the measured SACS value. The benchmark reference neutron field has been identified, characterized, and well validated in the LR-0 special core. A very large set of SACS measurements in the LR0 reference field is measured with low uncertainty, making it an excellent set for deconvolution codes validation. The impact of the gamma-induced reaction on the production yield of neutron-induced reactions was investigated for most of the benchmarked reactions. Gamma competition was found to contribute at most 1% for the 197Au(n,2n) reaction, while being substantially lower for other target isotopes and neutron-induced activation reactions.

Graphite is important reactor material used as a neutron moderator in various reactor designs. It is not only structural material used in previous reactors design but even material to which attention is focused due to the very high-temperature reactor design, which uses graphite as moderator and reflector. Due to the well-characterised cross-section of graphite, it is also a good candidate for neutron shaping experiments. In this kind of experiment, the neutron spectra are modified to the unconventional shape of neutron spectra which can be used to validate cross-section. This paper deals with validation of neutron field in graphite block surrounded by driver core arranged in large special core in LR-0 reactor. The experimentally determined shape of neutron spectra and spatial distribution of neutron flux is compared with the calculation. It is worth noting that the agreement for flux distribution is satisfactory in regions below the moderator level.

The integral experiments covering the neutron leakage from geometrically simple assemblies with a 252Cf source inside are very valuable tools usable in validation of transport cross section data, since geometric uncertainties play a much smaller role in simple geometric assemblies than in complex assemblies as for example reactor pressure vessel geometry. Since 252Cf(s.f.) is standard neutron source, the uncertainties connected with the source neutron spectrum can be even neglected. The paper refers on validation efforts of neutron leakage from stainless steel block ~50 x 50 x50 cm in Research Center Rez. Both the neutron leakage flux at a distance of 1 m from the center of the cubical assembly using stilbene spectrometry and the activation rates at different positions of the assembly were evaluated. In addition to experiments, main sources of uncertainty were identified and evaluated. The results of the stilbene measurements are consistent with the activation measurements results.

Dosimetry cross sections are fundamental quantities essential in determination of neutron fluences in points of interest in technologies under heavy radiation load. The most common is their application to Reactor Pressure Vessel aging management, relating to correct estimation of its residual lifetime. The neutron spectrum in various reactor positions has a similar character as the fission spectrum. Due to this fact, the validation of the neutron dosimetry cross sections in reactor fields, or in a very well-known neutron field of 252Cf(s.f.) is sufficient for ensuring their validity in estimations of neutron fluxes in reactor physics. With an increasing field of applications, as in neutron dosimetry in accelerator-based fields or space applications, requests occurred on detailed validation in spectra different from fission ones. This paper presents the testing of a new methodology for the use of quasi monoenergetic neutron fields, where different sensitivity allows validations of the cross-section in different energy regions than in the fission spectrum. The exact shape of the neutron spectrum in the tested fields is determined by stilbene spectrometry and corrected to scattering by calculation, where applicable. The total flux is determined from Ni and Al flux monitors. The evaluated neutron flux in the target set of activation foils is used for calculation of theoretical reaction rate, which is compared with the experimental value determined from gamma activity. This kind of comparison can be understood as validation. It is worth noting, this methodology applied to the IRDFF-II library shows satisfactory agreement for selected reactions.

The reactor baffle is an important component of a nuclear reactor that fixes the location of the fuel assembly in the reactor core. The main types of baffles are called light or heavy. The light baffle has mostly the form of steel plates with outer space filled by water, and the heavy baffle is mostly a forged steel element. Both concepts have advantages as well as disadvantages. In the case of the light baffle, one does not need to solve the issue of void swelling, but the neutron economy is not ideal, while the heavy baffle has a good neutron economy, but void swelling is an issue. This paper deals with the effect of the heavy VVER-1000 baffle on criticality. Criticality was measured using a well-defined core composed of 6 fuel assemblies moved to a simulator of the VVER-1000 internals, which is located at the LR-0 reactor. The experiments confirm the fact that the water filling the cooling channels in the baffle has a strong neutron absorbing effect. The keff calculated using the ENDF/B-VIII.0 library significantly underpredicts the experiment, whereas calculations using a new evaluation of 56Fe by the IAEA (INDEN collaboration) give a better agreement. Generally, the presented results are suitable for validation of iron cross sections.

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(nth,f) PFNS above 10 MeV is discussed.

The cross section averaged over 235U thermal-neutron induced fission spectrum is a fundamental quantity that can be used in evaluation and validation of nuclear data. Many experiments focused on the determination of Spectrum Averaged Cross Sections (SACS) in 235U(nth,f). Prompt Fission Neutron Spectrum (PFNS) in light water reactors using enriched uranium fuel. In these reactors, there is already some amount of water moderator between the uranium fuel and the irradiated sample. Due to the decrease of hydrogen cross-section with neutron energy, the high energy tail of the reactor spectrum in cores with water moderator may be harder than the pure PFNS. This paper aims to compare the shape of the actual reactor spectrum in various core positions of a research light-water reactor differing each from other by the effective water thickness. The spectrum shape is determined both by calculations and experimentally using various high energy threshold reactions. The impact of the photo-nuclear reactions ({\gamma},n) competing with (n,2n) in production of the same residual nucleus was shown to be less than a percent for most of studied dosimeters. An important exception was found for 197Au(n,2n) 196Au dosimeter irradiated in the outcore channel where a notable photo-neutron contribution to the production of 196Au is caused by the neutron production from the high energy {\gamma}-rays from thermal-neutron capture in 54Fe. The corresponding ENDF/B-VIII.0 data turned out to underestimate such {\gamma}-yield by 40% in comparison with ENDF/B-VI.8. This has improved but however not resolved the disagreement between our measurement and calculations. The remaining deficiency was attributed to the underestimation of the evaluated cross section IAEA/PD-2019 for the 197Au({\gamma},n) cross section near the reaction threshold. The later was confirmed by comparison with existing measured data.

Dosimetry cross sections are fundamental quantities necessary for neutron dosimetry using the neutron activation method. It is worth noting that the uncertainty in cross sections is the major source of uncertainty in calculational predictions using nuclear data in simulations thus, cross section validation is a key issue in any aims for refinement of any predictions. A small compact neutron generator is a promising tool for performing integral experiments and even for differential experiments. This paper deals with the measurement of the differential dosimetry cross sections using a small compact D-T neutron generator with 14.05 MeV neutron emission (10E8 n/s into 4pi). Achieving measurable activation at such a low flux field is allowed by using a larger amount of activation material placed in close measurement geometry during decay gamma measurement. 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. Its worth noting, the mean standard deviation in IRDFF-II library is about 4 %, while in case of other data libraries they are from 5.5 % - 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.

The spectrum averaged cross section (SACS) in standard neutron field is a preferable tool for cross section validation. There are very few measurements involving lutetium neutron cross sections. The presented work uses only neutron standard, i.e., 252Cf(sf) neutron field, for validation of lutetium threshold cross sections. SACS were inferred from gamma spectrometry derived reaction rates. The SACS which were derived include 175Lu(n,2n)174Lu, 175Lu(n,3n)173Lu, 175Lu(n,p)175Yb, and 176Lu(n,n)176m1Lu reactions. All these reactions SACS were measured for the first time. MCNP6 calculations using JEFF-3.3 or ENDF/B-VIII.0 libraries for lutetium cross sections were compared with experimental data. The agreement was found very poor for all reactions under study. Thus there is a need for their improvement. The presented data can be also used for validation of the various theoretical models.