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The inelastic neutron scattering is often followed by the emission of gamma photon. As the prompt gammas have a discrete level character they can be used for the identification of nuclides. Because of this fact, a good knowledge of photon production from inelastic scattering is important. Described research deals with the measurement of gamma originated from inelastic scattering of neutrons on 16O. The 241Am–Be was used as a neutron source because of its high average neutron energy. The oxygen in form of heavy water was used for maximization of neutron flux on oxygen and minimization of background gammas’ production, namely 2223 keV gammas accompanying capture on hydrogen 1H. The gamma spectrum was measured by HPGe and the stilbene detector. The HPGe measured quantities are comparted with calculation and discrepancies between measured and calculated gamma fluxes are reported. Stilbene measurement shows indistinguishability of gamma peaks above 6 MeV.

Copper is an important material for the nuclear industry; therefore, the correct copper cross section are essential. This paper deals with benchmarking the fast neutron leakage spectrum from a copper block with an intense 252Cf source. The spectrum was measured by the proton recoil method using a stilbene scintillator and spectrometer with pulse shape discrimination. The room effect was subtracted experimentally by shielding cone and confirmed computationally. Simulations were performed using the MCNP6.2 Monte Carlo code. A detailed study on uncertainties has been made. Consequently, as the main source of uncertainties, the measurement apparatus was determined (namely, calibration and number of iterations—up to 15%). From the parameter uncertainty, the most important effects are connected with the mass of cube material—the thickness of the assembly (up to 2%). The effect of copper density uncertainty is low because density was determined experimentally with low uncertainty. The uncertainty related to parameter uncertainties in source structural components is negligible because even neglection of all components leads to a shift in leakage spectra below 7%. Several copper nuclear data libraries were tested as well, and it was found that the calculation with JEFF-3.3 gives the most discrepant results with a discrepancy of up to 60%. The closest results were obtained with ENDF/B-VIII.0 (below 5 MeV, discrepancy within 10%) and JENDL-4.0 (above 5 MeV, discrepancy within 15%). ENDF/B-VII.1 library is relatively close in the region above 5 MeV (discrepancy within 15%), below 5 MeV the agreement is worse (up to 40% discrepancy).

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.

The spectrum averaged cross section (SACS) in a standard neutron field is a preferable tool for cross section validation. The presented work uses only neutron standard, i.e.

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.

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.

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.

Purpose of this paper is to provide extensive information helpful for anyone performing any experiment involving 252Cf neutron source and aiming for high precision experiments. The paper summarizes basic characteristics and fields of study using 252Cf neutron source. We show the basic characteristics of our source, precise geometry in MCNP6, isotopic content, distribution of the source in the encapsulation and possible use of encapsulation for 27Al(n,2n)26Al reaction estimation and the way of handling of the 252Cf neutron source in our laboratory. Furthermore, we prove that our source is volumetric, i.e. non-point and can be considered as an isotropic in our experimental settings. Influence of the palladium matrix density on the reaction rates is also investigated. Concerning nuclear data, we are measuring fast neutron leakage spectra in various benchmark sets including spheres of different materials. Our work extends benchmark sets to possible use of cubes, in our case graphite cube of side 30 cm. The neutron spectra are measured by stilbene scintilation detector in the energy range of 1–10 MeV in the steps of 100 keV in the distance of 100 cm from the centre of the cube. There is no significant difference between measurements performed using cube or sphere except lower cost for the cube production. The agreement between calculation using ENDF/B-VII.1 library and experimental data is not satisfactory especially in the regions around 3.5 MeV and 8 MeV. The agreement within 4% of calculation to experiment ratio for neutron leakage fluxes of pure 252Cf source validates our MCNP model for studying reaction rates and leakage fluxes in the region above 1 MeV.

The presented paper summarizes the methodology of fast neutron leakage fluxes measurement using a stilbene scintillation detector on spherical benchmark assemblies. The experiments are realized in a spherically simple geometry, which does not need validation of the neutron transport model. Hence, this type of experiment is suitable for validation of the nuclear cross sections of various isotopes. The 252 Cf neutron source is always placed in the center of a sphere of different pure materials. The selection is focused on structural materials used in the nuclear industry. In the presented set, the materials include light water, heavy water and nickel. For example, the light and heavy water spheres can be used to eliminate the effect of hydrogen and estimate the effect of oxygen. The neutron spectra in the region of 1–10 MeV are measured in a room where the background could substantially influence the final result. For this reason, the estimation of the scattered neutrons is crucial for the correct evaluation of the experiment. The comparison of the calculated and measured neutron fluxes and background signal share are presented. Calculations were performed using the MCNP6 program with the ENDF/B-VII.1 nuclear data library.

Dosimetry of high doses of ionizing radiation is a method widely used not only in the field of basic and applied research but also in various industries. High-dose irradiation is used in the sterilization of foodstuffs and prolongation of their shelf life, sterilization of tools, materials and wastewater, treatment of antiquities attacked by woodworm organisms, biological and genetic research, inhibition of seed germination, and retardation of fruit ripening. Furthermore, it has been applied in radiation resistance testing of materials, defectoscopy, industrial irradiation of materials (e.g., crosslinking and grafting of polymers or gemstone coloring by ionizing radiation), and last but not least, of course, nuclear energetics. In this work, the dosimetric properties of polycarbonate are studied. They are compared with a standardized and certified method using alanine dosimeters. In the case of gammas, the testing was realized for 60Co and neutron radiation was studied in the light water research nuclear reactor (LVR-15).

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 252Cf(s.f) and 235U(nth;fiss).

Lead is an important material, both for fusion or fission reactors. The cross sections of natural lead should be validated because lead is a main component of lithium-lead modules suggested for fusion power plants and it directly affects the crucial variable, tritium breeding ratio. The presented study discusses a validation of the lead transport libraries by dint of the activation of carefully selected activation samples. The high emission standard 252Cf neutron source was used as a neutron source for the presented validation experiment. In the irradiation setup, the samples were placed behind 5 and 10 cm of the lead material. Samples were measured using a gamma spectrometry to infer the reaction rate and compared with MCNP6 calculations using ENDF/B-VIII.0 lead cross sections. The experiment used validated IRDFF-II dosimetric reactions to validate lead cross sections, namely 197Au(n, 2n)196Au, 58Ni(n,p)58Co, 93Nb(n, 2n)92mNb, 115In(n,n')115mIn, 115In(n,γ)116mIn, 197Au(n,γ)198Au and 63Cu(n,γ)64Cu reactions. The threshold reactions agree reasonably with calculations; however, the experimental data suggests a higher thermal neutron flux behind lead bricks. The paper also suggests 252Cf isotropic source as a valuable tool for validation of some cross-sections important for fusion applications, i.e. reactions on structural materials, e.g. Cu, Pb, etc.

This paper presents the measurement of the spectrum-averaged cross section (SACS) of 63Cu(n,2n)62Cu reaction in 252Cf spontaneous fission neutron spectrum. The SACS in the 252Cf spectrum was chosen as a validation tool since 252Cf is the only standard neutron field and 62Cu isotope is not easy to measure by gamma spectroscopy since the gamma line of interest is an annihilation peak, which is also produced by 64Cu isotope. Fortunately, contributions to the annihilation peak from these isotopes can be distinguished due to the very different half-lives. SACS was inferred from the experimental reaction rate. The SACS in the 252Cf spontaneous fission neutron field for the 63Cu(n,2n)62Cu reaction was determined as equal to (0.1763 ± 0.0077 mb). This value agrees with value (0.183 ± 0.007) × 10−3 b within uncertainty presented by W. Mannhart. However, it differs by 12.7% from IRDFF-II value, which is equal to (0.19874 ± 8.954 10−3) × 10−3 b. Furthermore, reasonable agreement is not achieved with ENDF/B-VIII.0, JEFF-3.3, CENDL-3.1, ROSFOND-2010, nor JENDL-4.0 nuclear data libraries.

Only neutron spectrum standard is

There is a lack of reliable experiments aiming at the prompt fission neutron spectrum of

Physical quantities derived from integral experiments can usually be measured much more accurately than that from differential nuclear data. The accurate knowledge of integral parameters provide excellent grounds for testing and tuning differential data such as, for instance, cross sections. Measurement of neutron leakage spectra with 252Cf neutron source located at sphere center is often used for integral experiments. While this type of experiments provide information for cross section tuning, however, care must be taken to avoid misleading interpretation, namely, at high energies due to the very low portion of high energy neutrons in 252Cf spectrum. This issue can be alleviated by the use of point source with different spectra shape. For that purpose one suitable candidate seems to be the AmBe neutron source which has a relatively high average energy and peak character of emitted neutrons. Indeed, AmBe seems an interesting option because the calculated leakage neutron spectra are not very sensitive to the input shape of the neutron spectra. Thus the neutron leakage spectra calculated using tabulated of International Organization for Standardization spectra is nearly the same as stilbene measured AmBe spectra as an input.

Oxygen belongs to the group of the most important isotopes in the nuclear data field. The aim of this paper is validate various oxygen nuclear data libraries in different scenarios with high content of oxygen. For this purpose, fast neutron spectra were measured by a stilbene scintillation detector in the region of 1–10 MeV in the three model cases involving 252Cf neutron source and light water reactor. The cases include measurements of leakage spectra using 252Cf neutron source placed in the centers of the light water and heavy water spheres of 0.30 m diameter. Following measurements were carried out inside the concrete biological shielding of the VVER-1000 mock-up simulator in the LR-0 reactor and in the dry channel located in the center of the special core placed in the LR-0 reactor. In the case of the special core, symmetric active core consisted of six standard fuel assemblies which surround the experimental dry module, where the fast neutron spectrum was measured. The measured neutron spectra were compared with MCNP6 transport code calculations in ENDF/B-VII.1, ENDF/B-VIII.0, JENDL-4.0, and IRSN 16O nuclear data evaluations. Experimental results for all cases follow similar trend. All considered libraries underestimate experimental measurement in the region of 3–4 MeV in all cases.

High quality nuclear data is the most fundamental underpinning for all neutron metrology applications. This paper describes the release of version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) that contains a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60~MeV neutron energy. The library is intended to support: a) applications in research reactors; b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new \mbox{IRDFF-II} library includes 119 metrology reactions, four cover material reactions to support self-shielding corrections, five metrology metrics used by the dosimetry community, and cumulative fission products yields for seven fission products in three different neutron energy regions. In support of characterizing the measurement of the residual nuclei from the dosimetry reactions and the fission product decay modes, the present document lists the recommended decay data, particle emission energies and probabilities for 68 activation products. It also includes neutron spectral characterization data for 29 neutron benchmark fields for the validation of the library contents. The IRDFF-II library and comprehensive documentation is available online at https://www-nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at https://www-nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications., Submitted to Nuclear Data Sheets, 109 pages, figures revised

The results of systematic evaluations of spectrum averaged cross section (SACS) measurements in the fission neutron fields of 252Cf and 235U are presented. The data form a complete database of high-threshold experimental SACS measured in the same installation under the same conditions and using the same high purity germanium gamma spectrometer. This is crucial to reduce the uncertainty of the ratio and the data scattering and therefore, to minimize discrepancies compared to cross section measured under different conditions in different laboratories. This new dataset complements and extends earlier experimental evaluations. The total emission of the 252Cf neutron source during the experiments varied from 9.5E8 to 4.5E8 neutrons per second. The emission was derived in accordance to the data in the Certificate of Calibration involving absolute flux measurements in a manganese sulphate bath. Concerning 235U fission neutron field, the irradiations were carried out in a specifically designed core assembled in the zero power light water LR-0 reactor. This special core has a well described neutron field. After the irradiation, the low volume irradiated samples to be measured by gamma spectrometry were placed directly on the upper cap of a coaxial high purity germanium (HPGe) detector in a vertical configuration (ORTEC GEM35P4). High volume samples were homogenized and strewn into the Marinelli beaker. The HPGe detector is surrounded by the lead shielding box with a thin inner copper cladding and covered with rubber for suppression of background signal and bremsstrahlung. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas (NPA) measured using the semiconductor HPGe detector. The measured reaction rates are used to derive the spectrum-averaged cross sections. Furthermore, measured reaction rates are also compared with MCNP6 calculations using various nuclear data libraries, in particular IRDFF evaluations.

The presented paper focuses on the measurement of fast neutron leakage spectra in the copper assembly using an intense standard 252Cf(sf) source. The neutron spectra were measured by a stilbene organic scintillation detector and compared with MCNP6.2 calculations using copper nuclear data from the ENDF/B-VIII.0, ENDF/B-VII.1, JEFF-3.3, JEFF-3.2, JENDL-4.0 and CENDL-3.1 libraries in the energy range of 1–10 MeV. Discrepancies appear in the intermediate neutron energy range between experiment and calculation. The best agreement with an experiment was achieved using ENDF/B-VIII.0 library in the region of 1–4.5 MeV. In the region of 4.5–10 MeV, the best agreement was achieved using JENDL-4.0 or CENDL-3.1 library. The largest discrepancies in the whole energy range were found for calculations involving JEFF-3.3 copper evaluations.

VR-1 is light water zero-power reactor using IRT-4 M fuel with nominal enrichment of 19.7 wt%, equipped with the radial channel being directly attached to the fuel. Enrichment at the top level of the low enriched uranium limit guarantees that the larger part of non-scattered neutrons in the spectrum is coming from 235U than in case of experimental reactors using enrichments between 2 and 5 wt% (e.g., LR-0). Thus, the measurement of the VR-1 leakage spectrum can be a valuable contribution to the validation process of predicted 235U prompt fission neutron spectrum. Presented second set of measurements is confirming the beam stability at VR-1 and compares the agreement rate in neutron spectrum shape with two benchmarks; light- and heavy water filtered neutron spectrum obtained from spheres. Additionally, gamma spectrum was measured and compared with spherical iron benchmark and VVER-1000 core simulator. Presented measurements have the character of comprehensive characterization of beamline experiments developed at reactor cores with IRT-4 M fuel. Reported good agreement between the experiment and calculation confirm the methodology of using VR-1 as a Mock-Up experiment for benchmarking of high power reactors with IRTM fuel in the neutron transport experiments.

The correct description of neutron transport in lead is an essential task for correct description of tritium production in the DEMO (DEMOnstration Power Station) breeding blanket because some concepts deal with lead as a major component: namely the WCLL (water cooled lithium lead blanket), HCLL (helium cooled lithium lead blanket), and DCLL (dual cooled lithium lead blanket). Concerning the improvement of the knowledge about the transport of fast neutrons in lead, a set of experiments and calculations was carried out to study this problem with a well-defined neutron beam. The neutron flux behind various lead arrangements positioned along the beam axis was measured using a stilbene scintillation crystal (10 mm × 10 mm) with neutron and gamma pulse shape discrimination. The measurement was performed along the beam axis and in the case of the thick target also above the axis, to estimate the neutron angular scatter in lead. The calculations were realized using MCNP6 with various nuclear data libraries. Discrepancies in the angular distribution description in the energy region of about 1 MeV were discovered by these experiments.

The neutron flux distribution in the internal parts of a reactor is an important parameter which affects the absorbed dose, temperature, and gas production in the materials. Therefore, the knowledge of neutron flux distribution, for example in the baffle, is important for estimation of the radiation-induced swelling and activation of this internal component. The power distribution in the periphery fuel assemblies, mostly the lateral 2–3 pin rows, has considerable influence on neutron and gamma flux in the internal reactor elements. The power distribution in the core is mostly calculated by the reactor operator using diffusion codes. These data are used as an input for transport calculations, e.g., with Monte Carlo for the fluences, displacements per atoms, gamma heating and other important values for radiation damage estimation. This paper aims to compare the experimental and calculated reaction rate distribution in a mock-up of VVER-1000 (VVER-1000 Mock-Up) internals placed in the LR- 0 reactor and compares these quantities in case of a partial change of enrichment in one fuel assembly adjacent to the baffle simulator.

Fast neutron leakage spectra from the light and heavy water sphere of 30 cm in diameter with neutron source in its centre were measured by a stilbene scintillation detector in the region of 1–10 MeV in the distance of 85 cm from the spheres surface. We use the light and heavy water to eliminate the effect of hydrogen. 252Cf with the approximate emission rate of 5.5E8 n/s was used as a neutron source for all measurements involved and was placed in the centres of the spheres. The measured neutron spectra are compared with MCNP transport code calculations in ENDF/B-VII.0, ENDF/B-VIII.b4 and JENDL-4 nuclear data libraries. Experimental results for both cases follows similar trend. The best agreement is achieved with ENDF/B-VIII.b4 library in both cases. All libraries underestimate experimental measurement in the region of 3–4 MeV. Furthermore, JENDL-4 library overestimates experiment in the region of 4–6.5 MeV. In addition, we performed cross section sensitivity analysis for elastic, inelastic and (n,α) reaction in JENDL-4 and ENDF/B-VIII.b4 libraries since they have almost independent evaluations of 16O.

The results of systematic evaluations of the spectrum-averaged cross section measurements performed in the spontaneous fission 252Cf neutron field are presented. The Following threshold reactions were investigated: 23Na(n,2n)22Na, 54Fe(n,p)54Mn, 54Fe(n,α) 51Cr, 27Al(n,p)27Mg, 27Al(n,α)24Na, 19F(n,2n)18F, 90Zr(n,2n)89Zr and 89Y(n,2n)88Y. The spectrum-averaged cross sections for 23Na(n,2n)22Na, 54Fe(n,α)51Cr and 89Y(n,2n)88Y reactions were measured for the first time. This quantity is compared with calculations carried with the IRDFF-v1.05 library. There is a notable disagreement exceeding uncertainties only for 54Fe(n,p)54Mn and 54Fe(n,α) 51Cr reactions. The spectrum-averaged cross sections were inferred from experimentally determined reaction rates. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas measured using the semiconductor high purity germanium spectroscopy. The presented experimental data can be used to validate nuclear data libraries and reactions used in the practical reactor dosimetry and to specify high energy tail of the 252Cf neutron spectrum.

Validation of the selected materials cross sections is an important task for their use as flux monitors within experimental measurements. As the reaction rate is a function of neutron spectrum and material cross section, the neutron spectrum must be well defined for the correct cross section validation. This criterion is met in the special core of the LR-0 reactor, where the reaction rates of 181 Ta(n,γ), 54 Fe(n,p), 54 Fe(n,α) and 89 Y(n,2n) reactions were measured. The 197 Au(n,γ) and 58 Ni(n,p) reactions were used as flux monitors during measurements. Results of calculation show that in case of 181 Ta(n,γ) reaction, the values reported in JEFF-3.2 provide a good agreement, while ENDF/B-VII.1 is underpredicting the experiment. In case of 54 Fe(n,p) reaction, JEFF-3.2 is in satisfactory agreement, while ENDF/B-VII.1 overpredicts the result by about 10%. The 54 Fe(n,α) shows significant discrepancies, JEFF-3.2 under predicts by about 20%, while ENDF/B-VII.1 overpredicts result by 40%. The high threshold 89 Y(n,2n) reaction is interesting from the point of view of spectral averaged cross section. It is worth noting that the resulted spectral averaged cross section, being 0.169 with 1σ uncertainty of 0.007 mb, is in very good agreement with previous result, being 0.172 ± 0.006 mb.

18F is an important radioisotope used in nuclear medicine for diagnostic purposes. It is produced most often via 18O (p,n)18F reaction. During its production, the oxygen contained in the H218O molecules placed in the cooled target is irradiated by accelerated protons. The production of each 18F atom is accompanied by anisotropic high-energy neutron emission, which is a problem for ensuring radiation safety. These neutrons are responsible for the high radiation dose in a cyclotron pit during 18F production as well as for activation of cyclotron components or might be responsible even for target material changes due to their high penetration properties. A precise characterization of a neutron field on the surface of the liquid water target, used for 18F production, is presented. The neutron flux characterization is performed both by using large-area activation detectors and stilbene detectors at 1 m from the target in a forward direction. The experiments show significant discrepancies between calculated and measured reaction rates in higher threshold reactions 58Ni(n,x)57Co and 51V (n,α)48Sc, confirming that the current neutron production reaction models are not suitable for the characterization of the secondary neutrons leaking from the target.

As an iron is the main structural component of nuclear power plants as well as future fusion power plants, the validation of neutron incident data libraries of iron is a must. Presented paper fits into ongoing validation activities and presents measuring neutron leakage spectra in the 0.1-1.0MeV region from iron sphere of 100cm in diameter by hydrogen proportional detectors. The experimental result is compared with ENDF/B-VII.1, JEFF-3.2 and CIELO nuclear data libraries. No library reasonably well describes whole region under study. Furthermore, elastic and inelastic XS sensitivity analysis for all iron isotopes was carried out. 54Fe isotope elastic XS influence is comparable with 56Fe XS influence up to 0.8MeV. 57Fe isotope elastic XS is significant in the region of 0.14-0.15MeV. Additionally, there are large differences among libraries in both elastic and inelastic XS for 57Fe. Furthermore, it was found that 58Fe isotope XS has negligible influence on the results. As a neutron source, 252Cf with initial emission rate of 9.53E8 n/s was used in this experiment.

This article describes a method for testing of the efficiency of the biological shielding of the hot cell facility, which were constructed as a part of the project SUSEN. Ten hot cells and one semi-hot cell are present in the facility Radiochemistry II. The shielding is made from steel plates. In order to demonstrate sufficient efficiency of the biological shielding of the hot cells and a correspondence between measured and contractual values at selected points. The test was done using sealed high activity 60 Co sources. The results are also used as a proof of the optimization of radiation protection for the workplace of this type. The results confirm significant optimization of radiation protection at the workplace. The dose received by a staff do not exceed one tens of annual limit during active service. Obtained results fulfill general requirements of radiation protection and will be used for further active service of hot cells facility.

The presented paper aims to evaluate the importance of 54 Fe XS in iron by means of measuring the reaction rates of the selected reactions on 54 Fe and measuring a fast neutron leakage spectra from the iron sphere of 100 cm in diameter by a stilbene scintillation detector with subsequent XS sensitivity analysis. The reactions involved in the study were 54 Fe(n,p) and 54 Fe(n,α). Measured neutron induced reaction rates in 54 Fe are compared with calculated ones in different nuclear data libraries. We show that there are notable discrepancies in 54 Fe(n,α) reaction. The results of the leakage spectra differ significantly in various libraries, library ENDF/B-VII.1 in region 3.5–7.0 MeV gives relatively good agreement. CIELO library underestimates the result; however JEFF-3.2 overestimates results., 252 Cf with the emission rate of 9.53E8 n/s was used as a neutron source for all experiments involved.

The first study of $\phi$-meson production in p–Pb collisions at forward and backward rapidity, at a nucleon– nucleon centre-of-mass energy $\sqrt{S_{NN}}$= 5.02 TeV, has been performed with the ALICE apparatus at the LHC. The $\phi$-mesons have been identified in the dimuon decay channel in the transverse momentum ($p_T$) range 1 < $p_T$ < 7 GeV/c, both in the p-going (2.03 < y < 3.53) and the Pb-going (−4.46 < y < −2.96) directions — where $y$ stands for the rapidity in the nucleon–nucleon centre-of-mass — the integrated luminosity amounting to 5.01 ± 0.19 nb$^{−1}$ and 5.81 ± 0.20 nb$^{−1}$ , respectively, for the two data samples. Differential cross sections as a function of transverse momentum and rapidity are presented. The forward–backward ratio for $\phi$-meson production is measured for 2.96 < |y| < 3.53, resulting in a ratio ∼ 0.5 with no significant $p_T$ dependence within the uncertainties. The $p_T$ dependence of the $\phi$ nuclear modification factor R$_{pPb}$ exhibits an enhancement up to a factor 1.6 at $\phi$= 3–4 GeV/c in the Pb-going direction. The $p_T$ dependence of the $\phi$-meson cross section in pp collisions at $\sqrt s$ = 2.76 TeV, which is used to determine a reference for the p–Pb results, is also presented here for 1 < $p_T$ < 5 GeV/c and 2.5 < y < 4, for a 78 ± 3 nb$^{−1}$ integrated luminosity sample.

The article describes measurement of the 75As(n,2n)74As reaction rate in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor. It presents a comparison of this value with results of calculations. The reaction rate is derived from activity of 74As which is determined using gamma spectrometry of an irradiated As2O3 sample. The resulting value averaged in reactor spectrum is 77.0 ± 3.9 μb. This cross section is important as it is included as neutron flux monitor in the International Reactor Dosimetry and Fusion File. The calculations were carried out with the MCNP6 code using ENDF/B-VII.0, ENDF/B-VII.1 JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. The comparison between calculation and experiment shows very good agreement. The relative difference between calculation and experiment is lower than the related uncertainty in all cases. The discrepancy is even below 1% with ENDF/B-VII.1 and JEFF-3.2.

Spectrum-averaged cross sections (SACS) is an important quantity usable in validation of nuclear cross sections. Especially in case of dosimetrical reactions there is a request on precise validation. This paper presents SACS measured in the reference

With increasing needs of neutron detection in energy and homeland security sectors, there is an aim in the development of new suitable detection materials with possible n/g separation and also satisfactory resolution. This issue is connected with a well-defined neutron field. The neutron standards 252Cf(spontaneous fission) and 235U(nth fission) have smooth spectra; thus, the tests in these fields can be used for testing of neutron gamma pulse shape discrimination (PSD), but cannot reveal possible problems in the deconvolution. New neutron field which is formed from fission spectrum filtered by 1 m thick silicon block was developed in Research Center Rez. Due to the course of the silicon cross section, the well-distinguished neutron peaks occur in filtered neutron spectrum. In this field, various scintillation materials, HIDEX Aqualight, and EJ299-33A were tested.

This paper describes the measurement of 55Mn(n,2n) and 127I(n,2n) reaction rates in a well-defined reactor field in a special core of LR-0 reactor. The reaction rates were derived using gamma-spectrometry by measuring gamma activities of irradiated MnO2 and NaI samples at a high purity germanium (HPGe) detector. The spectral average cross section (SACS) in 235U prompt fission neutron spectrum (PFNS) was experimentally determined to be 0.2393 ± 0.015 × 10−3 b for 55Mn and 1.2087 ± 0.052 × 10−3 b for 127I. These obtained results were compared with calculations by MCNP6 code using ENDF/B VII.1, ENDF/B VII, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND- 2010, CENDL-3.1, and IRDFF nuclear data libraries. In a case of 55Mn, a good agreement with ENDF/B VII.1, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND, and CENDL 3.1 nuclear data libraries was found, where C/E−1 is 0.1%, while IRDFF underestimated by about 15.8%. In the case of 127I, more significant discrepancies were found, where JENDL 3.3 and JENDL 4 overestimate the result by about 31.3%.

This work deals with 23Na(n,2n)22Na and 127I(n,2n)126I reactions in the 252Cf spontaneous fission neutron source. 252Cf neutron source with approximate emission of 6·× 108 n/s was employed for the irradiation of sodium iodide. The spectrum-averaged cross sections (SACS) were then inferred from experimentally determined reaction rates and compared with calculations in MCNP6 using various nuclear data libraries. The experimental reaction rates were derived from the net peak areas (NPAs) measured using the high purity germanium spectroscopy. The measured SACS for the 23Na(n,2n)22Na reaction in the 252Cf spectrum was determined as equal to (8.98±0.32)·× 10−6 b. The resulting SACS for the 127I(n,2n)126I reaction in the 252Cf spectrum was derived as (2.044±0.0072)·× 10−3 b. These experimental data can be used for nuclear data libraries validation and to specify high energy tail of the 252Cf neutron spectrum.

Purpose of this paper is to provide extensive information helpful for anyone performing any experiment involving

We argue that energy and momentum deposition from hard partons into quark-gluon plasma induces an important contribution to the final state hadron anisotropies. We also advocate a novel method of Event Shape Sorting which allow to analyse the azimuthal anisotropies of the fireball dynamics in more detail. A use of the method in femtoscopy is demonstrated., Comment: contribution to Proceedings of the XXIV Baldin seminar on high energy physics problems (Relativistic Nuclear Physics and Quantum Chromodynamics), Sept. 17-22, 2018, Dubna, Russia, 6 pages

Neutron activation analysis is the reference method used for offline determination of the neutron flux density in defined positions. It can be used in the nuclear energy industry-as well as in medical- or space applications. For accurate neutron flux evaluation, well-known and reliable cross sections are needed. In the thermal and fast energy region, many reliable monitoring reactions exists, however, in case of the epithermal and intermediate energy region, there are practically no dosimetry nuclear reactions sensitive specifically in this energy range. Due to this fact, both new data are being measured and methodologies are under development to describe and test this energy region. It was found that various neutron filters can be used to cut parts of neutron spectra and thus methodology based on spectrum filtering could potentially be employed to survey cross sections of interest. It this paper, the use of 3 different filters – B4C, Cd, and In is studied, on the case of the 55Mn(n,γ) reaction. Measured values of that cross section in the given filtered reference spectra are reported.

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.

The presented paper aims to compare various measured neutron induced reaction rates in Aluminium with computed ones in different nuclear data libraries. A 252Cf neutron source with emission rate of 9.53E8 n/s was used. Reactions involved in the study were 27Al(n,g), 27Al (n,p) and 27Al (n,α).

Evaluation of neutron fluence in a reactor pressure vessel (RPV) together with surveillance specimen programs for RPV materials are one of the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of RPV residual lifetime. This paper covers transport of neutrons through the RPV of a VVER-1000 nuclear reactor. This problem is of increased importance as it concerns issues around VVER NPP life extension. With regards to the construction (reduced thickness of the lateral reflector), this issue plays greater role in VVER reactors than in Western types of PWR reactors. RPV material degradation depends mainly on neutron flux and spectra. Both quantities can be calculated or measured. This paper compares MCNP calculations and measurements on zero-power experimental reactor LR-0 with TORT calculations for VVER-1000. The goal is to find a reasonable method for precise estimation of neutron fluence and attenuation factor through the RPV. The calculations were performed with MCNP stochastic code and TORT deterministic transport code. The measurements were performed in a VVER-1000 mock-up placed in reactor LR-0 (Research Center Řež).