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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. [ABSTRACT FROM AUTHOR]

Nuclear data evaluations are periodically updated to render state of the art of knowledge, and the independent validation experiments are still of interest of the community. Current work describes validation of selected capture reactions used for monitoring of thermal neutrons in mixed fields, as well as reactions responsible for activation of components exposed to neutron flux. The validation was performed also for reaction 50Cr (n,γ), which is very important reaction directly affecting criticality safety, because chromium is essential component of stainless steel used as structural component of the core baffle closely adjoining to nuclear fuel. Described experiments were performed in reference neutron field of the LR-0 reactor. The activation rate was derived by gamma spectrometry using well-characterized HPGe detector. It was found that the capture cross sections for the dosimetry reactions 23Na (n,γ), 58Fe (n,γ), and 59Co (n,γ) agreed well within 10 % with calculations in thermal and epithermal region, but larger discrepancy was found for the isotopes of tin. • Experimental validation of capture cross sections was performed in the reference neutron field of the LR-0 reactor. • Homogeneity of the neutron field in the irradiation position has been proven by calculation and experiment. • Main dosimetry reactions as 23Na (n,γ), 58Fe (n,γ), and 59Co (n,γ) agree well with calculation predictions using different nuclear data libraries, except of the capture reactions on isotopes of tin. [ABSTRACT FROM AUTHOR]

Abstract 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 252Cf 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. [ABSTRACT FROM AUTHOR]

Abstract The neutron flux distribution behind a reactor pressure vessel (RPV) is an important parameter that is monitored to determine neutron fluence in the RPV. Together with mechanical testing of surveillance specimens, these are the most important parts of in-service inspection programs that are essential for a realistic and reliable assessment of the RPV residual lifetime. The fast neutron fluence values are determined by a calculation. These calculation results are accompanied by measurements of induced activities of the activation foils placed in the capsules behind the RPV at selected locations, namely in azimuthal profile. In case of discrepancies between the measured and calculated activities of the activation foils placed behind the pressure vessel, it is difficult to determine the source of the deviation. During such analysis, there arises a question on the influence of power peaking near core boundary on neutron profile behind the RPV. This paper compares the calculated and measured increase of the neutron flux density distribution behind the reactor pressure vessel in the azimuthal profile that has arisen from the replacement of 164 fuel pins located close to reactor internals by pins with the higher enrichment. This work can be understood as the first step in the characterization of the effect of incorrectly calculated pin power or burn-up in the fuel assembly at the core boundary relative to the neutron flux distribution behind reactor pressure vessel. Based on a good agreement between the calculated and experimental values, it can be concluded that the mathematical model used to evaluate the power increase is correct. Highlights • Measurement of neutron flux density profile behind RPV in azimuthal profile. • Comparison of calculated and measured neutron flux density profile behind RPV. • Measurement of local power increase on neutron flux density behind RPV simulator. • Measurement and calculation of neutron spectra behind RPV simulator of VVER-1000 reactor. [ABSTRACT FROM AUTHOR]

Abstract 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. Highlights • Neutron spectra behind various Pb targets. • Comparison of the experimental data with calculation. • Study of angular distribution of neutron scatter in Pb. • Discrepancies between calculational and experimental neutron spectra. [ABSTRACT FROM AUTHOR]

During production of radiopharmaceuticals, the radiation situation in cyclotron pit is an important parameter, which is being monitored to ensure fulfilment of the limits and conditions of safe operation. The neutron flux in the structural components of the accelerator is also an important parameter, because the secondary neutrons are responsible for activation of cyclotron structural components and may even affect structural changes in it. This paper aims to characterize the neutron field in inner positions of medical accelerator IBA 18/9 by activation detectors and by means of scintillation spectrometry. The backward angle measurement was realized also in special liquid water target (H 2 18 O) at U120M cyclotron to confirm the data obtained in IBA 18/9 cyclotron. [Display omitted] • Neutron spectrum in backward angle (almost 180°) was measured with scintillation spectroscopy. • Neutron spectrum inside the production cyclotron was characterized. • Neutron spectra around a smaller target in open geometry were measured and compared with previous measurement. [ABSTRACT FROM AUTHOR]

The article describes measurement of the 75 As(n,2n) 74 As 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 74 As which is determined using gamma spectrometry of an irradiated As 2 O 3 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. [ABSTRACT FROM AUTHOR]

Evaluation of the neutron fluence values in a reactor pressure vessel together with surveillance specimens programs from reactor pressure vessel materials are among the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of the RPV’s residual lifetime. The neutron fluence values are determined by means of calculation. These calculation results are accompanied by measurements of induced activity of the activation foil placed in capsules behind the reactor pressure vessel at selected locations. These results can be used for neutron flux density adjustment in the corresponding parts of the reactor system. Such information is not complete, because for determination of the radiation damage, the flux on the inner side of reactor pressure vessel has to be known. If the flux attenuation ratio is known, then the neutron flux density on the outer side can be easily recalculated for the inner side. These attenuation factors, applicable to the power reactors, can be determined experimentally in the VVER-1000 Mock-Up placed in the LR-0 reactor. The comparison of experimental results with calculated ones increases the reliability of their estimations. This paper aims to evaluate the measured neutron flux densities at various VVER-1000 Mock-Up positions by means of comparison with the Monte Carlo simulation. The effect of the various nuclear data libraries on the calculation results is also studied. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Abstract: The paper is intended to show the effect of a biological shielding simulator on fast neutron and photon transport in its vicinity. The fast neutron and photon fluxes were measured by means of scintillation spectroscopy using a 45×45mm2 and a 10×10mm2 cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6–10MeV and the photon spectrum in 0.2–9MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries. [Copyright &y& Elsevier]

Abstract: This work describes the influence of biological concrete shielding model on neutron flux in reactor pressure vessel simulator placed in VVER-1000 transport benchmark. It includes an analysis and a comparison of experimental and calculated values of neutron flux. The thermal neutron fluxes in RPV simulator were measured when the reactor vessel or biological shielding was cladded in a cadmium foil absorbing thermal neutrons. The calculations have been performed using both CENDL 3.1 and ENDF/B VII nuclear data libraries. [Copyright &y& Elsevier]

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 H 2 18O 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. • Measurement of neutron leakage spectra in forward direction. • Measurement of reaction rates in various positions on IBA target. • Calculation of neutron field in all measured positions. • Discrepancies between calculation and experiment. [ABSTRACT FROM AUTHOR]

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 – B 4 C, 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. • Measurement of reaction rates in various neutron filters in reference neutron field. • Calculation of reaction rates in various neutron filters in reference neutron field. • Discrepancies between calculation and experiment. • Testing of 55Mn(n,γ) cross section in various energy regions. [ABSTRACT FROM AUTHOR]

There is a lack of reliable experiments aiming at the prompt fission neutron spectrum of 235U for energies higher than 10 MeV. The presented experiment performed at the LVR-15 light water reactor aimed at the measurement of very high threshold reactions spectral averaged cross sections such as 55Mn (n,2n)54Mn, 197Au (n,2n)196Au, 197Au (n,3n)195Au, 209Bi(n,3n)207Bi, 209Bi(n,4n)206Bi. 209Bi(n,3n)207Bi and 209Bi(n,4n)206Bi reactions were measured for the first time. 58Ni(n,p)58Co reaction was used as a monitor reaction. The experimental spectral averaged cross sections are derived from reaction rates measured by means of high purity Germanium spectroscopy at the well defined detector. The experimental spectral averaged cross sections are compared with calculations using either ENDF/B-VIII.0 or JEFF-3.3 235U prompt fission neutron spectrum and IRDFF-II cross sections. The discrepancy is higher with higher mean response energy for JEFF-3.3 unlike ENDF/B-VIII.0, where the agreement is good within broad range of mean response energy. Moreover, due to the high thermal neutron flux in the reactor, the experimental reaction rate is compared with calculated 198Au (n,g)199Au reaction rate. The difference of −55.3% for double capture reaction was found in comparison with ROSFOND-2010 calculations. • Measurement of high threshold SACS in 235U (n th ,f) field using LVR-15 reactor. • Measurement of 198Au (n,g) 199Au reaction rate. • Comparison of JEFF-3.3 and ENDF/B-VIII 235U PFNS. [ABSTRACT FROM AUTHOR]

• Setting of reference neutron field in LR-0 reactor. • Calculation and measurement of selected reactions reaction rates in LR-0. • Discrepancies between calculation and experiment. • Measurement of (n,2n) reaction spectral averaged cross sections of 75As; 90Zr; 23Na; 89Y; 55Mn; 127I; 19F; 197Au. • Measurement of spectral averaged cross sections of lower threshold 27Al(n,α); 24Mg(n,p); 48Ti(n,p) reactions. A new reactor neutron reference field is proposed, which can be used for measurements of spectrum averaged cross sections (SACS). The reference field is identified in the center of a dry irradiation channel surrounded by six 3.3% enriched fuel assemblies assembled in the zero-power LR-0 reactor with negligible burn-up. The neutron field can be defined as a reference one, because of the well-defined neutron spectrum and neutron flux, well-characterized criticality and core power distribution, and reproducibility of results. Extensive experimental validation was undertaken. A new reference field can be used for the measurement of reaction rates and SACS used in nuclear data validation. It is shown that above 6 MeV the spectrum is undistinguishable from the thermal-neutron induced 235U prompt fission neutron spectrum (PFNS). Thus, when a reaction has threshold over this limit, the SACS in the thermal-neutron induced 235U fission neutron field can be derived from reaction rates obtained in this reference neutron field. A comprehensive set of SACS measurements with uncertainties below 5% in the reference field is presented. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]