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 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.

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 measured and evaluated excitation functions are fundamental quantities that affect the accuracy of all calculations in nuclear applications. Some cross sections, such as 14N(n,p)14C, have added value for special applications, as these reactions may be responsible for possible contamination in industrial processes such as spent fuel reprocessing. For the validation of the evaluated cross sections, we can rely on the comparison of the calculated spectrum averaged cross sections (SACS) for the given neutron spectrum with the measured SACS value. The benchmark reference neutron field has been identified, characterized, and well validated in the LR-0 special core. A very large set of SACS measurements in the LR-0 reference field is measured with low uncertainty, making it an excellent set for deconvolution codes' validation. The impact of the gamma-induced reaction on the production yield of neutron-induced reactions was investigated for most of the benchmarked reactions. Gamma competition was found to contribute at most 1 % for the 197Au(n,2n) reaction, while being substantially lower for other target isotopes and neutron-induced activation reactions.

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.

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 Cf-252 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 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 be-tween 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.

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.

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.

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).

Cross section data are fundamental quantities which affect the accuracy of all calculations in nuclear applications. A new dosimetry library IRDFF-II that contains cross section evaluations with full uncertainty quantification was developed by the International Atomic Energy Agency and released in January 2020 (https://www-nds.iea.org/IRDFF). A previous version, IRDFF-1.05, was released in 2014 and experimental validation of the newly released cross section by spectrum averaged cross section (SACS) measurements is a high priority task. For such purpose, a neutron dosimeter set containing 5 target foils was activated in 2 independent experiments at the VR-1 reactor of the Czech Technical University in Prague. Care was taken to derive SACS with low uncertainties. New experimental evaluation method is in good agreement with previous approaches based on relative measurements using monitor foils. Good agreement of measured SACS and evaluated IRDFF-II cross sections is observed. Slight overestimation of evaluated ENDF/B-VIII.0 235U(nth,f) PFNS above 10 MeV is discussed.

Experimental work at the LR-0 reactor was carried out to determine the possibility of the mock-up neutron field creation for the fluoride salt-based reactors. Previous calculations and experiments have shown that the fast part of the molten salt reactor (MSR) spectrum is shaped by fluorine and even the Teflon material is suitable for neutronics in that energy range. Properly selected spectrum indices can describe the neutron field of the MSR in fast thermal and intermediate parts of the spectrum. Current research has focused on a deeper study of the possibility of using the filtered thermal and intermediate neutron spectrum of the experimental light water reactor for the physics of fluoride salt-based reactor. LiF-BeF2 (FLIBE) capsule and teflon cylinders are used as a spectrum filters in the LR-0 reactor. Measured results show acceptable C/E-1 agreement in the reaction rates and satisfactory agreement for usage of the FLIBE filtered neutron spectrum determined by the 181Ta(n, γ) monitor as a mock-up in thermal to intermediate energy in the fluoride high-temperature reactor (FHR). Concerning MSR, intermediate spectrum can be reproduced to some extent but not as good as in the case of FHR.

Silicon cross-sections, being important for a criticality safety of final spent fuel disposals, were recently reevaluated within the IAEA INDEN project. Similarly, the thermal Scattering Law matrix for silicon dioxide, which is also important for criticality safety, was also reevaluated in the ENDF/B-VIII.0 library. Due to these reasons, a series of validation experiments with silica sand were performed at the LR-0 reactor. This paper describes these validation experiments, which used two different amounts of silica sand placed in the core. The first part of validation was carried out as critical experiments in order to benefit from the suitability of integral experiments for validation. The second part -fast neutron spectrum measurement in the sand -was performed in order to obtain knowledge of its characteristics and its agreement with a calculation. The results showed significant improvement of the Thermal Scattering Law matrix for silicon dioxide available in the ENDF/B-VIII.0 library. They also showed that the new INDEN evaluation of silicon cross-sections, together with its description in the ENDF/B-VIII.0 gives disagreement rate closest to the experiments carried out without silica sand insertions. The spectrum measurement showed that the calculations of fast neutron spectra in the sand show only slight differences between different evaluations of silicon cross-sections. However, the fast neutron spectrum is not dependent on the Thermal Scattering Law. The calculated spectra show relatively good agreement with the measurement.

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.

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.

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 highenergy 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,alpha)48Sc, confirming that the current neutron production reaction models are not suitable for the characterization of the secondary neutrons leaking from the target.

This paper shows a new merthod for characterization of the secondary neutron field quantities, specifically neutron spectrum leaking from 18O enriched H2O XL cylindrical target in IBA Cyclone 18/9 in the energy range of 1-15 MeV. Spectrum is measured by stilbene scintillation detector in different places. The neutron spectra are evaluated from the measured proton recoil spectra using deconvolution through maximum likelihood estimation. A leakage neutron field is an interesting option for irradiation experiments due to quite high flux, but also to the validation of high energy threshold reactions due to relatively high average energy. Measured neutron spectra are compared with calculations in MCNP6 model using TENDL-2017, FENDL-3, and default MCNP6 model calculations. TENDL-2017 and FENDL-3 libraries results differ significantly in the shape of the neutron spectrum for energies above 10 MeV while MCNP6 gives incorrect angular distributions. Activation measurements of different neutron induced reactions support characterization. The 18F production yield is in a good agreement with TENDL-2017 proton library calculation within respective uncertainties.

The spectral averaged cross section is an important quantity used in a validation of nuclear cross section. When the cross sections are averaged over the neutron standard field (Cf-252(s,f) or U-235(n,f) neutron spectrum), they can be used for tuning of evaluations. This kind of quantities is very useful because the data in integral measurements can be determined with a significantly smaller uncertainties than the standard differential data. The experiment was aimed at the spectral average cross sections measurement and was performed in a radial channel of VR-1 reactor (with fuel enrichment 19.75 wt %). The results are in a good agreement within the uncertainties with a previous measurements in LR-0 reactor (with fuel enrichment 3.3 wt %), thus it supports the hypothesis that even significant amount of U-238(n,f) neutrons in the LR-0 reactor spectrum does not have a significant influence. The derived spectral averaged cross sections are as follows: 0.1709 +/- 0.0115 mb for Y-89(n,2n), 10.738 +/- 0.719 mb for Ti-46(n,p), 17.896 +/- 1.181 mb for Ti-47(n,p), 0.294 +/- 0.02 mb for Ti-48(n,p), 72.994 +/- 4.964 mb for Fe-54(n,p), 0.528 +/- 0.036 mb for Cu-63(n,a), 0.444 +/- 0.029 mb for Nb-93(n,2n)Nb-92* and 0.239 +/- 0.016 mb for Ni-58(n,x)Co-57.

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 U-235 prompt fission neutron spectrum (PFNS). Thus, when a reaction has threshold over this limit, the SACS in the thermal-neutron induced U-235 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. (C) 2019 Elsevier Ltd. All rights reserved.