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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. 252 Cf 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 16 O. [ABSTRACT FROM AUTHOR]

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 0.1–1.0 MeV region from iron sphere of 100 cm 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. 54 Fe isotope elastic XS influence is comparable with 56 Fe XS influence up to 0.8 MeV. 57 Fe isotope elastic XS is significant in the region of 0.14–0.15 MeV. Additionally, there are large differences among libraries in both elastic and inelastic XS for 57 Fe. Furthermore, it was found that 58 Fe isotope XS has negligible influence on the results. As a neutron source, 252 Cf with initial emission rate of 9.53E8 n/s was used in this experiment. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Highlights • Validation of the neutron spectra in VR-1 channel. • Measurement of gamma spectra in VR-1 channel. • Comparison of the experimental data with calculation. • Discrepancies between calculational and experimental spectra. Abstract 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. [ABSTRACT FROM AUTHOR]

The results of systematic evaluations of the spectrum-averaged cross section measurements performed in the spontaneous fission 252 Cf neutron field are presented. The Following threshold reactions were investigated: 23 Na(n,2n) 22 Na, 54 Fe(n,p) 54 Mn, 54 Fe(n,α) 51 Cr, 27 Al(n,p) 27 Mg, 27 Al(n,α) 24 Na, 19 F(n,2n) 18 F, 90 Zr(n,2n) 89 Zr and 89 Y(n,2n) 88 Y. The spectrum-averaged cross sections for 23 Na(n,2n) 22 Na, 54 Fe(n,α) 51 Cr and 89 Y(n,2n) 88 Y 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 54 Fe(n,p) 54 Mn and 54 Fe(n,α) 51 Cr 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 252 Cf neutron spectrum. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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