Proportional counters filled with hydrogen are suitable for fast neutron detection and spectrometry. As their sensitivity is not limited only to this type of ionizing radiation as desired, an approach suitable for event discrimination was investigated. Several counters of the given construction were subjected to various radiation fields and the effort was directed to the examination of pulse shape discrimination method intended to distinguish neutrons from gamma rays and parasitic events (microdischarges) if these also occurred. Rising edge duration of the output pulses and their amplitudes were mapped into diagrams enabling visual discrimination while the criteria for further computational processes are to be determined. Pulse waveforms from charge-sensitive preamplifier with the proportional counters connected were recorded using a field-programmable gate array (FPGA)-based custom-build acquisition unit and then postprocessed. Moreover, the possibility of operation verification and energy calibration based on the introduction of helium-3 admixture into the gas filling was studied, as well as the energy resolution abilities. In addition, the theoretical basis for particle discrimination is presented in the form of description of events and principles involved.

Copper is an important structural material used in nuclear technology, often used as a cover for spent fuel canisters or planned to be used in fusion devices. Despite its significance, there is a lack of integral experiments useful for validating and improving the evaluations of copper nuclear data. To address this gap, a neutron leakage experiment was conducted a few years ago using a point 252Cf(s.f.) neutron source placed inside a large block of copper. In this work a pencil beam transmission-attenuation experiment (a broomstick) employing various thicknesses (5-20 cm) of copper blocks (cylinders of 6 cm in diameter) was undertaken to expand the dataset of available experiments for copper in the fast neutron energy range (1-10 MeV). This type of experiment has the highest sensitivity to the total cross sections, and sensitivities are different from other integral experiments, making it a complementary measurement to already existing integral data. The measurement was performed using stilbene scintillation spectrometry. Measured transmission shows that the current INDEN evaluation, proposed to be adopted for ENDF/B-VIII.1 and JEFF-4 libraries, exhibits excellent agreement with experimental data. The JEFF-3.3 evaluation displays significant discrepancies, consistent with previous results from integral experiments involving copper. In the case of JENDL-5, discrepancies were found in the energy region 1.7-4.9 MeV.

The long-term operation of existing nuclear power reactors is a crucial concern due to the complexities and expenses associated with replacing key components, such as the reactor pressure vessel and reactor internals. Gamma radiation, a byproduct of nuclear reactions and radioactive decay, significantly influences the lifetime of these components. This radiation is responsible for various degradation pathways leading to void swelling in steel reactor components and cracking or other radiation damage in concrete structures. A study conducted at a full-scale mock-up of the VVER-1000 reactor at the LR-0 zero-power reactor employed HPGe and stilbene measurements to analyze gamma spectra behind the reactor pressure vessel and within concrete biological shielding. While simulations behind the reactor pressure vessel aligned with measurements, notably, a marked overestimation of stilbene spectrum calculations occurred deep in concrete, suggesting potential inaccuracies in radiation predictions for power plant structures.

The assessment of the neutron flux distribution in nuclear power plant components, like reactor internals and the pressure vessel, is one of the most important parts of their residual lifetime evaluation process because the irradiation damage of these components is strongly dependent on it. The fast neutron fluences in power reactors are generally determined using calculations and verified by measurements. Discrepancies between them sometimes occur, which can be caused by be the inaccuracies in the power distribution used in neutron transport calculations. This paper quantifies the effect of uncertainty in power density on the neutron fluences behind the reactor pressure vessel (RPV). An increase in power density was experimentally simulated by implementing a higher uranium enrichment of certain selected pins. The experiment was also simulated using the MCNP code with the ENDF/B-VII.1 library. Both the experimental and calculational data show a significant local increase in neutron flux. Even behind the RPV, the increase is as high as 25%, while the local power increase in the perturbed pins is about 70–80%, which is approximately equal to the 2 σ power density uncertainties. A good agreement between measurement and calculation was found.

Prompt gamma radiation mostly comes from neutron reactions, thus naturally accompanying neutron transport. Therefore, it is an important part of the gamma field in nuclear installations, which accompanies neutrons leaking from the system. This also implies that good knowledge of the prompt gamma description of iron is essential in reactor dosimetry, especially as part of gamma heating and related issues. As the prompt gamma has a discrete character, it can be efficiently used in nuclear forensics for the identification of various isotopes. Prompt gamma rays from neutron capture on iron and chlorine were measured using HPGe and stilbene detector. Iron (as FeSO4) and chlorine (as NaCl) were dissolved in water to maximize the thermalization process of 252Cf(s.f.) neutrons emitted from a centrally positioned neutron source in the solution tank. Not only iron and chlorine peaks were found in the HPGe spectra, but six sulfur peaks from ferrous sulphate were identified as well. The experiments were simulated with the MCNP6.2 code using the ENDF/B-VIII.0, JEFF-3.3 and JENDL-5 libraries. Peak comparisons using HPGe results were generally discrepant in all libraries. A study of prompt gammas from thermal neutron capture reveals significant differences between the libraries studied. The comparison of whole spectra using stilbene results shows discrepancies in the gamma production data for chlorine in all libraries. Calculations with FeSO4 give quite good results for JENDL-5 and ENDF/B-VIII.0 and discrepant results for JEFF-3.3.

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.

The paper presents a first comprehensive measurement of the total fission gamma spectrum (TFGS) of 252Cf(s.f.) spanning an energy range from 0.2 to 14.3 MeV using stilbene scintillator spectroscopy. The measurement was performed with the 252Cf(s.f.) source placed in a flexo-rabbit ending, with a stilbene detector located at a distance of 100 cm from the source. The room effect was determined by separate measurement with a 50 cm long lead shielding cylinder placed between the source and the detector. The obtained spectrum was corrected for the effect of the 252Cf source structural parts and the flexo-rabbit ending, using correction function obtained by simulation with the MCNP6.2 code. The main benefit of this independent measurement is the wide energy range, from 0.2 up to 14.3 MeV, in a single experiment, whereas existing measurements typically cover narrower intervals. The knowledge of the TFGS is crucial for various applications involving 252Cf(s.f.) sources, including the design and analysis of nuclear systems, the validation of nuclear data, and the estimation of radiation in the nuclear medicine. The comparison of TFGS with the recently evaluated 252Cf(s.f.) prompt fission gamma spectrum (PFGS) indicates that delayed gammas may contribute above 2 MeV up to the highest energies; the current assumption was that delayed gammas contribution above 2 MeV was negligible.

Having knowledge of the spectrum of fast neutrons in the irradiation position of a reactor plays an important role in many experimental settings. If well described, it can be used as a reference neutron field for validating and testing measuring devices or detectors. This paper compares model calculations with measurements of the neutron spectrum in the research reactor VR-1. The fast neutron spectrum was measured in the radial channel of VR-1 reactor using a stilbene scintillation detector and the NGA-01 measuring device. The NGA-01 is a twoparameter spectrometric system for neutron and gamma radiation fields. The calculations were performed in a critical mode of the MCNP6 code using ENDF/B-VII.1 nuclear data library. Generally, a good agreement was achieved. However, a slight discrepancy in the neutron spectrum between the calculated and measured values was found in the 2-3 MeV region. In other regions, the differences between calculated and experimental values are comparable with uncertainties. Since the detector is distanced from the core, the neutron beam can be assumed parallel and therefore possible issues with stilbene anisotropy are eliminated. Thus, this experiment can be used for validation of a reactor leakage spectrum and nuclear cross sections in the energy region above 6 MeV, which is of interest for the nuclear data section of International Atomic Energy Agency.

Prompt gamma radiation from neutron interactions is an important issue as it affects the operation of nuclear facilities (radiation protection or gamma heating issues) or has various uses (non-destructive identification of elements). Despite of importance, its production is inaccurately described in present nuclear data libraries. Therefore, a set of experiments have been carried out focusing on this radiation - prompt gamma from neutron inelastic scattering on oxygen and prompt gamma from radiative capture on manganese, both present the form of an aqueous solution. The gammas were induced by the Am-Be neutron source (for oxygen) and 252Cf spontaneous fission neutron source (for manganese). In the case of oxygen, the present libraries can be used, but all the libraries tested overestimated the experiment by 20–30%, considering the experimental uncertainty of 6-9%. INDEN-4.0 library is giving the best results. In the case of manganese, none of the libraries tested give acceptable results. The INDEN updated data for gamma production on manganese give usable results, but there is still a discrepancy from -36% to 55% with uncertainties from 6-15%.

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.

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.

Prompt gamma radiation is an important part of radiation fields with neutrons. Its good description is vital for precise predictions in void swelling issues because it is a contributor in gamma heating effect. The energy of prompt gamma rays is unique for each nuclide, thus can be used as its signature. Prompt gamma activation analysis is a technique using these gammas in nuclide identification. Despite the prompt gamma importance, current nuclear data libraries describe its origins very inaccurately. Measurement of gamma leakage spectrum of manganese bath with 252Cf neutron source was performed using well-defined HPGe detector with high-density polyethylene shielding and stilbene scintillator. Stilbene measured gamma/neutron spectrum was separated by a Pulse Shape Discrimination method. MCNP6.2 simulation of the experiment was performed with ENDF/B-VIII.0 and JEFF-3.3 libraries for neutrons (and gamma production from neutron interactions) and MCPLIB04 library for photons. Analysis of results revealed that ENDF/B-VIII.0 and JEFF-3.3 description of prompt gamma radiation production is significantly discrepant. INDEN updated data for gamma production on 55Mn was tested as well and show significant improvement compared to original ENDF/B-VIII.0 data. Nevertheless, significant discrepancies up to 50% were still observed. An alternative approach using prompt gamma production tables provided by IAEA was tested as well. Its results are basically in agreement with the calculation using INDEN updated 55Mn data.

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.

Using D–T generators as a neutron source for precise measurements requests a well-known neutron spectrum and spatial distribution of neutron flux. It is important in cross section measurements as well as in deep penetration issues. Thus, it is very important not only to validate the fluence accurately but also accurately validate the neutron spectrum in various positions. Experimental validation of two different D–T neutron generator experiments was performed using proton recoil method. Measurements performed at different angles and neutron generator voltages show relatively good agreement with the calculations. Despite the poor resolution, it was shown that stilbene is usable for the measurement of neutron energy shift due to changes in deuteron voltage.

Prompt capture gammas are an important part of the fission reactor gamma field. Because some of the structural materials after neutron capture can emit photons with high energies forming the dominant component of the gamma spectrum in the high energy region, the following study of the high energy capture gamma was carried out. High energy gamma radiation may play a major role in areas of the radiation sciences as reactor dosimetry. The HPGe measurements and calculations of the high-energy aluminum capture gamma were performed at two moderator levels in the VR-1 pool-type reactor. The result comparison for nominal levels was within two sigma uncertainties for the major 7.724 MeV peak. A larger discrepancy of 60% was found for the 7.693 MeV peak. The spectra were also measured using a stilbene detector, and a good agreement between HPGe and stilbene was observed. This confirms the validity of stilbene measurements of gamma flux. Additionally, agreement of the wide peak measurement in 7–9.2 MeV by stilbene detector shows the possibility of using the organic scintillators as an independent power monitor. This fact is valid in these reactor types because power is proportional to the thermal neutron flux, which is also proportional to the production of capture gammas forming the wide peak.

Dosimetry cross sections are fundamental quantities for proper determination of the neutron fluences in points of interest under heavy radiation load. One of the critical applications is the Reactor Pressure Vessel aging management, related to the correct estimation of its residual lifetime or for many non-reactor applications including neutron dosimetry of accelerator-based fields or space applications. The neutron flux of neutrons above 12 MeV in reference to fission spectra is below 1%. Therefore, new reference neutron fields with average energy above 5 MeV should be developed for the validation of neutron dosimetry cross sections up to 60 MeV. This paper presents the testing of a new methodology for the use of quasi monoenergetic neutron fields, where different sensitivity allows validations of the dosimetry cross section at energies much higher than the average energy of around 2 MeV typical of fission spectra. The exact shape of the neutron spectrum in the tested fields is measured by stilbene spectrometry. The total flux is determined from Ni and Al flux monitors. The developed methodology was applied to the validation of selected reactions from the IRDFF-II library showing satisfactory agreement.

Liquid organic scintillators are important devices for measurements of neutron radiation. Currently, large-scale liquid organic scintillators have capabilities of detecting neutrons, but the determination of the neutron energy spectra is a challenge. This work aims to measure the responses of two liquid two-component scintillators to mono-energetic neutron radiation and to determine their light output function, which is necessary for proper neutron energy spectra determination. Both scintillators are composed of the solvent di-iso-propyl-naphthalene (DIPN) mixed isomers. The first scintillator, labeled PYR5/DIPN, contains the luminophore 1-phenyl-3-(2,4,6-trimethyl-phenyl)-2-pyrazoline with a concentration of 5 g/L. The second scintillator labeled THIO5/DIPN contains the luminophore 2,5-bis(5-tert-butyl-benzoxazol-2-yl)thiophene also with a concentration of 5 g/L. The responses to neutron energies of 1.5 MeV, 2.5 MeV, and 19 MeV are measured at PTB in Braunschweig. The responses to neutron energies of 2.45 MeV and 14 MeV were measured at CTU in Prague using DD and DT reactions. The responses to a silicon filtered beam were measured at Research Centre Řež. The measurements were processed using a two-parameter spectrometric system NGA-01 to discriminate neutrons from gamma rays. The obtained responses are dominated by recoil protons from elastic collisions of neutrons with hydrogen atoms. The edge of the response of recoil protons gives information about the light output of neutrons, compared to gamma rays for the same radiation energy. The light output function for protons in the PYR5/DIPN scintillator is L(Ep)=0.6294Ep−1.00(1−exp(−0.4933Ep0.95)). The light output function for protons in the THIO5/DIPN scintillator is L(Ep)=0.6323Ep−1.00(1−exp(−0.4986Ep0.9883)). The light output functions well resemble the standard shape, and they are quite similar to each other. That suggests a weak influence of the luminophore on the light output function. The light output functions are ready to be incorporated to the response matrix for the neutron energy spectra determination.

Liquid organic scintillators are important materials for measurements of neutron radiation. This work aims to develop and optimize the composition of liquid organic scintillators so they can be used for fast neutron spectrometry. As neutron radiation is usually accompanied with γ ray radiation, this work is focused on γ/n discrimination. It is shown that Di-iso-propylnaphthalene Mixed Isomers and 1-Methylnaphthalene are capable of good γ/n discrimination. Afterwards, the concentration of luminophores is varied and it is shown that the PYR and THIO luminophores with concentrations above 3g/l have Figure of Merit FOM> 1.27 above the energy threshold of 0.14MeVee and 0.17MeVee, respectively.

This paper summarizes the issue of the validation of the silicon-filtered neutron beam transport in the deep neutron transport penetration experiment in iron. Iron is an essential structural material important for nuclear technology. The use of a silicon-filtered beam is a very interesting method because some significant peaks occur in the spectrum, helping to study selected wide energy regions during the deep neutron transport in the iron. The detailed characterization of the silicon-filtered beam has been performed in the past as well. Therefore, the input spectrum for the penetration experiments is well-known. The character of the input spectrum is reflecting the fine structure of the silicon cross section in region 1-8 MeV. Based on the agreement between calculated and measured attenuation in groups located within the neutron flux peaks, one can reveal possible problems in neutron transport description. The results are confirming satisfactory agreement of neutron transport description in ENDF/B-VII.1 in the majority of energy regions, while in the interval 4.7-6 MeV, underprediction in attenuation can be observed. This seems to be a consequence of discrepancies in the angular distribution of scattered neutrons. These results constitute an advance to previously performed integral experiments characterizing the neutron transport in iron using Cf-252(s.f) and U-235(nth;fiss).