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This paper is focused on the development of the experimental environment connected to reactor graphite. Regarding its very good neutronic and mechanical properties, graphite will be very important in some new reactor designs, such as high-temperature or molten salt SMR reactors. These new reactor concepts require a new experimental environment as support for further research. In the laboratories of the Research Centre Řež and at the LR-0 reactor, the new experimentally validated graphite environment was created. This large graphite insertion is the largest graphite mono-block, which is possible to assemble at the LR-0 reactor. Sets of experiments for measuring reaction rates of different activation detectors for neutron field mapping were performed. This approach was used for thermal and epithermal region descriptions. For the fast neutron spectrum evaluation, the stilbene scintillation detector was used. All parameters, such as criticality height of moderator level, neutron spectrum, and other parameters for all experiments, were performed using Monte Carlo neutronic codes Serpent and MCNP. The obtained results were finally compared to the measurement of neutron leakage spectra from the graphite cube and graphite cylinder. These specially developed graphite-shaped neutron fields, reactor insertions, and external cube and cylinder with Cf neutron source can be used in the future for validation of not only materials used in SMR reactors but for arbitrary cross-section verification. [ABSTRACT FROM AUTHOR]

The leakage neutron spectrum measurements have been done on benchmark spherical assembly-nickel sphere with a diameter of 50 cm and a copper cube (block) with dimensions of 49.5 x 49.5 x 48 cm3 in Research Centre Rez (RC Rez). The 252Cf neutron source was placed into the centre of nickel sphere and copper cube. The proton recoil method was used for the neutron spectrum measurement using spherical hydrogen proportional detectors (HPD) with pressure of 400 and 1000 kPa (diameter of detectors is 4 cm) and scintillation stilbene (ST) detector (diam. of 1 x 1 cm). The neutron energy range of spectrometer is from 0.04 MeV to 1.3 MeV for HPD and from 1 MeV to 12 MeV for ST. The adequate MCNP neutron spectrum calculations based on data libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, BROND-3.1 were done and compared with the experiment, i.e., calculation to experiment ratio C/E was determined. [ABSTRACT FROM AUTHOR]

The systematic evaluations of spectrum averaged cross sections of dosimetric reactions over a broad range of energies were performed in 252Cf (spontaneous fission) and 235U(nth,f) neutron fields. The neutron sources used in this study were LR-0, VR-1 zero power research light water reactors, LVR-15 10 MW research light water reactor, and 252Cf neutron source with emission specified precisely by the manganese sulphate bath. All spectral averaged cross sections were inferred from measured reaction rates which were derived from gamma spectrometry. These gamma spectrometry measurements were performed using a single detector in all cases. The ratios of 252Cf and 235U spectral averaged cross sections can be used to specify the high energy tail of the 235U prompt fission neutron spectrum as the 252Cf spontaneous fission spectrum is considered as a standard. Furthermore, ratios are independent of cross section uncertainties since uncertainties in the cross sections are eliminated. Theoretical models of fission can be tested based on our measurements. The calculations were performed in MCNP6.2 transport code using different prompt fission neutron spectra and IRDFF-II cross sections for threshold reactions. The ratios are in good agreement using only ENDF/B-VIII.0 235U prompt fission neutron spectrum suggesting it to be harder than in other evaluations. [ABSTRACT FROM AUTHOR]

This article focuses on description of two different neutron fields from linear and cylindrical Inertial Electrostatic Confinement (IEC) neutron generators. Both of these generators are well defined and commonly used. They use a deuterium-tritium reaction that produces neutrons with energies in the range 13 – 16 MeV, depending on the direction and the energy of the incoming deuterium nucleus. Two-parametric spectrometric system for neutron/gamma mixed fields NGA-01 was used to characterize neutron spectra in the proximity of generators. The cylindrical 45x45 mm stilbene scintillator was connected to this device using an active voltage divider. This way, we were able to measure neutron energies in the range 1 - 15 MeV while filtering out gamma radiation, even when counts per second is high. For the neutron spectrum calculation recoil spectra using deconvolution through maximum likelihood estimation was used. Measured neutron spectra have been compared with simulations realized via MCNP6. According to the theoretical prediction, these two types of generators produce different neutron fields. In case of the linear generator the target is very close to point located tritium bombarded by deuterons. Thus the neutron spectrum varies depending on the angle between the detector axis and the axis of the generator. Both experimental results and simulation show a shift of the neutron energy peak in pulse height histogram. For IEC type generators the neutron spectrum is more complicated. The shape and the position of the neutron energy peak depend heavily on the position of the detector. The most prominent effect is in the position in the plane perpendicular to the generator axis. In this case, the peak splits into two peaks that can be measured and distinguished. These results were verified by the diamond detector which was also used for characterization of the IEC type generator. [ABSTRACT FROM AUTHOR]

The cross section is a fundamental quantity which affects the accuracy of Monte Carlo simulations widely used in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations that include full uncertainty quantification is being developed by the International Atomic Energy Agency and expected to be released in January 2020; a preliminary version IRDFF-1.05 was released in 2014 and is being tested in this work. Validation of the cross-section evaluations proposed for this library is a high priority task. The validation can be realized using integral cross sections measured in standard and/or reference neutron benchmark fields. Integral quantities feature significantly lower uncertainties than differential nuclear data. If the neutron spectrum where the cross section is measured is well characterized, then the Spectrum Averaged Cross Section can be used for validating of existing evaluations. [ABSTRACT FROM AUTHOR]

This paper presents some results of the development of two new scintillation detectors usable for the detection and spectrometry of neutron and photon components of the mixed field. It is a series of liquid organic scintillators containing carbon, hydrogen, nitrogen and oxygen. Elements of individual scintillator designs are arranged in various chemical bonds defining physical and chemical properties of the scintillators. These scintillators have advantageous useful properties. They are non-hygroscopic, not poisonous and are not sensitive to contact with the outside atmosphere and can be stored in conventional laboratory containers. [ABSTRACT FROM AUTHOR]