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We have investigated following capture reactions: 197 Au(n,g) 198 Au, 63 Cu(n,g) 64 Cu, 45 Sc(n,g) 46 Sc, 181 Ta(n,g) 182 Ta, 140 Ce(n,g) 141 Ce, 139 La(n,g) 140 La, 176 Yb(n,g) 177 Yb, 55 Mn(n,g) 56 Mn, and 141 Pr(n,g) 142 Pr in a standard 252 Cf(s.f.) neutron field. The experimentally derived reaction rates were compared with calculations using state-of-the-art libraries IRDFF-II, ENDF/B-VIII.0, JENDL-5 and JEFF-3.3 performed in MCNP6.2 transport code. The experiment was focused on capture reactions in high energy neutron range. To suppress scattered low energy neutrons, all activation materials were enclosed by a spherical cadmium surface of 30 cm in diameter. Experimentally derived reaction rates of 197 Au(n,g) 198 Au, 63 Cu(n,g) 64 Cu reactions agree with calculations using all state-of-the-art nuclear data libraries. The highest discrepancies were found when evaluating 139 La(n,g) 140 La and 55 Mn(n,g) 56 Mn reactions utilizing all libraries. Experimental reaction rates applying data from the JENDL-5 library agree very well with the results of other reactions except for the 45 Sc(n,g) 46 Sc reaction. 181 Ta(n,g) 182 Ta reaction rate does not agree within uncertainties using any library. The 176 Yb(n,g) 177 Yb reaction rate agrees reasonably well only when performing calculation using JENDL-5 and ENDF/B-VIII.0 libraries, unlike JEFF-3.3 library. 140 Ce(n,g) 141 Ce and 141 Pr(n,g) 142 Pr reaction rates applying JEFF-3.3 and ENDF/B-VIII.0 libraries do not agree within uncertainties., Competing Interests: Declaration of competing interest The authors declare the following financial interests/personal relationships which may be considered as potential competing interests:Michal Kostal reports financial support was provided by Research Center Rez. If there are other authors, they declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 Elsevier Ltd. All rights reserved.)

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 50 Cr (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 23 Na (n,γ), 58 Fe (n,γ), and 59 Co (n,γ) agreed well within 10 % with calculations in thermal and epithermal region, but larger discrepancy was found for the isotopes of tin., Competing Interests: Declaration of competing interest The authors declare the following financial interests/personal relationships which may be considered as potential competing interests: Some of the authors took part on several reviews for the journal. If there are other authors, they declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 Elsevier Ltd. All rights reserved.)

The spectrum averaged cross sections (SACS) in standard neutron field, e.g. 252 Cf(s.f.), is a preferable tool for cross section evaluation and validation. A set of reaction measurements with high energy thresholds was previously performed. The presented work focuses on lower energy threshold reactions, namely on the inelastic scattering of the tin foil, more specifically the reaction 117 Sn(n,n') 117m Sn, and the zinc foil reaction, namely 67 Zn(n,p) 67 Cu. These reactions are of special interest due to their intermediate energy range, which is essential in classical reactor dosimetry and fast reactor dosimetry. The experiments were carried out in a standard neutron field formed by 252 Cf(s.f.) source in Řež. The experimental results were compared with calculations using MCNP6.2, ENDF/B-VII.1 transport library, and ENDF/B-VIII.0 and IRDFF-II cross section data library. Additionally, the calculations using CEA code DARWIN/PEPIN2 using JEFF-3.0/A were executed. The obtained experimental SACS of previously measured reactions were in good agreement with the SACS calculated using the IRDFF-II library. Additionally, the calculational reaction rate of 67 Zn(n,p) 67 Cu was in accordance with the experimental data in case of ENDF/B-VIII.0 nuclear data library. Moreover, the calculational results of 117 Sn(n,n') 117m Sn obtained by DARWIN/PEPIN2 code (using JEFF-3.0/A nuclear data library) are closest to the experimental results., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 Elsevier Ltd. All rights reserved.)

Photonuclear reactions are gaining importance due to their influence on the shielding and activation of components of widely spread accelerators. Therefore, there is a need for accurate data describing photonuclear reactions. The MT25 microtron operated at NPI was used for validation of cross section in the bremsstrahlung. The gold and copper activation foils with known cross sections were used as flux monitors. The bremsstrahlung spectra was simulated with Geant4 code and the spectral averaged cross section above 10MeV for reaction 23 Na(γ, x) 22 Na was calculated. The experimental result was compared with data from the evaluated data libraries and EXFOR database. It was proved that the facility can be used for validation of cross sections and it was showed that all libraries overestimate the photon cross section on 23 Na with the best results with data from JENDL-5 library., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024. Published by Elsevier Ltd.)

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., Competing Interests: Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2023 Elsevier Ltd. All rights reserved.)

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., 252 Cf(sf) reaction neutron field, for validation of lutetium threshold cross sections. SACS were inferred from gamma spectrometry derived reaction rates. The SACS which were derived include 175 Lu (n,2n) 174 Lu, 175 Lu (n,3n) 173 Lu, 175 Lu (n,p) 175 Yb, and 176 Lu (n,n') 176m1 Lu reactions. All these reactions SACS were measured for the first time. MCNP6.2 calculations using JEFF-3.3 or ENDF/B-VIII.0 libraries for lutetium cross sections were compared with experimental data. The agreement was found very poor for all reactions under study. Thus there is a need for their improvement. The presented data can be also used for validation of the various theoretical models., (Copyright © 2022 Elsevier Ltd. All rights reserved.)

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 55 Mn(n,γ) reaction. Measured values of that cross section in the given filtered reference spectra are reported., (Copyright © 2020 Elsevier Ltd. All rights reserved.)

There is a lack of reliable experiments aiming at the prompt fission neutron spectrum of 235 U 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 55 Mn (n,2n) 54 Mn, 197 Au (n,2n) 196 Au, 197 Au (n,3n) 195 Au, 209 Bi(n,3n) 207 Bi, 209 Bi(n,4n) 206 Bi. 209 Bi(n,3n) 207 Bi and 209 Bi(n,4n) 206 Bi reactions were measured for the first time. 58 Ni(n,p) 58 Co 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-3.3 235 U 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 198 Au (n,g) 199 Au reaction rate. The difference of -55.3% for double capture reaction was found in comparison with ROSFOND-2010 calculations., (Copyright © 2020 Elsevier Ltd. All rights reserved.)

Only neutron spectrum standard is 252 Cf spontaneous fission neutron spectrum. However, the high energy tail of this spectrum is loaded with high uncertainty. To reduce this uncertainty, it is crucial to use validated cross sections with low uncertainty. The explored set of reactions covers 58 Ni(n,X) 57 Co, 169 Tm(n,3n) 167 Tm, 197 Au(n,3n) 195 Au, 209 Bi(n,3n) 207 Bi and 209 Bi(n,4n) 206 Bi threshold reactions. Measurement of dosimetric 169 Tm(n,3n) 167 Tm, 209 Bi(n,3n) 207 Bi and 209 Bi(n,4n) 206 Bi reactions spectral averaged cross sections (SACS) is included in NEA's High Priority Nuclear Data Request List. All these reactions SACS were measured for the first time. All SACS were derived from experimentally determined reaction rates by gamma spectrometry using the same high-purity germanium detector. In the case of 169 Tm(n,3n) 167 Tm reaction, the difference between experimental and calculated value using the IRDFF-II library is only 1.45%. Concerning 197 Au(n,3n) 195 Au reaction, the reasonable agreement is achieved only using the TENDL-2017 library. In the case of 209 Bi(n,3n) 207 Bi reaction, agreement within uncertainty is not achieved with any library unlike 209 Bi(n,4n) 206 Bi reaction where the agreement within uncertainty is achieved with IRDFF-II library. The best agreement for 58 Ni(n,X) 57 Co reaction is achieved using ENDF/B-VIII library., (Copyright © 2020 Elsevier Ltd. All rights reserved.)

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 252 Cf(sf) neutron field for neutron dosimetry reactions to validate recently updated IRDFF-II library intended mainly for neutron metrology applications. It covers 46 Ti(n,p) 46 Sc, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc, 58 Ni(n,p) 58 Co, 60 Ni(n,p) 60 Co, 58 Ni(n,2n) 57 Ni, 24 Mg(n,p) 24 Na, 27 Al(n,α) 24 Na and 63 Cu(n,α) 60 Co threshold reactions. Measurement of 60 Ni(n,p) 60 Co SACS is included in NEA's High Priority Nuclear Data Request List since existing data are discrepant. Spectral averaged cross sections were derived from experimentally determined reaction rates by gamma spectrometry using a same high-purity germanium detector to measure all irradiated samples. Measured spectrum-averaged cross sections agree very well within quoted uncertainties with those calculated using the IRDFF-II library. No other library achieves such good performance. Thus, the presented data support use of the cross sections of the mentioned reactions from IRDFF-II library., (Copyright © 2019 Elsevier Ltd. All rights reserved.)

Purpose of this paper is to provide extensive information helpful for anyone performing any experiment involving 252 Cf neutron source and aiming for high precision experiments. The paper summarizes basic characteristics and fields of study using 252 Cf 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 27 Al(n,2n) 26 Al reaction estimation and the way of handling of the 252 Cf 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 252 Cf source validates our MCNP model for studying reaction rates and leakage fluxes in the region above 1 MeV., (Copyright © 2019 Elsevier Ltd. All rights reserved.)

Spectrum-averaged cross sections (SACS) have been measured in the reference 252 Cf(sf) neutron field for the following high-threshold (n,2n) neutron dosimetry reactions since they are especially important due to the high threshold which allows validation of upper parts of prompt fission neutron spectrum. This work includes 59Co(n,2n)58Co, 197Au(n,2n)196Au, 169 Tm(n,2n) 168 Tm, 55 Mn(n,2n) 54 Mn, 93 Nb(n,2n) 92 m Nb and 89 Y(n,2n) 88 Y and for the 59 Co(n,p) 59 Fe threshold reactions. SACS were inferred from experimentally determined reaction rates by gamma spectrometry using a semiconductor high-purity germanium detector to measure irradiated samples. Measured reaction rates agree within quoted uncertainties with those calculated from the IRDFF-1.05 library, except for the reaction 55 Mn(n,2n) 54 Mn, for which the measured value is underestimated by 16%. For this reaction the ENDF-B/VII.1 evaluation agrees with measured reaction rate within uncertainties., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

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., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

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., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

The fast leakage neutron spectra have been measured on spherical nickel benchmark assembly of diameter 50 cm. The 252 Cf neutron source with approximate emission of 5.0·10 8 n/s was placed into the centre of the sphere. Fast neutron spectrum in the range of 1-10 MeV was measured in the distance of 1 m from the sphere centre by means of proton recoil method using scintillation stilbene crystal. The experimental data were compared to transport calculations based on several evaluated nuclear data libraries using MCNP6. MCNP6 was compared with SCALE/MONACO program using the same ENDF/B-VII.1 library which leads to different differential neutron flux results. Best experimental agreement with calculation in MCNP6 is achieved with ENDF/B-VII.1 library. Contrary, worst agreement is achieved with JENDL-4.0 and CENDL-3.1 libraries. Furthermore, cross section sensitivity analysis for elastic and inelastic scattering for both main nickel isotopes ( 58 Ni, 60 Ni) was performed. It was shown that 58 Ni isotope has higher influence on the result than 60 Ni isotope in the entire energy range under study. The highest influence has the elastic XS of 58 Ni around energy of 1.5 MeV. The inelastic cross section (XS) of 58 Ni dominates in the energies above 2 MeV where two percent rise due to the inelastic XS leads up to 3.3% decrease in the neutron flux., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

A well-defined neutron spectrum is an essential tool not only for calibration and testing of neutron detectors used in dosimetry and spectroscopy but also for validation and verification of evaluated cross sections. A new evaluation of thermal-neutron induced 235 U PFNS was performed by the International Atomic Energy Agency (IAEA) in the CIELO (Collaborative International Evaluated Library Organisation Project) project; new measurements of Spectral Averaged Cross sections averaged in the evaluated spectrum are to be obtained. In general, a neutron spectrum in the core is not identical to the pure fission one because fission neutrons undergo many scattering reactions, but it can be shown that PFNS and reactor spectra become undistinguishable from a certain energy boundary. This limit is important for experiments, because when the studied reaction threshold is over this limit, the spectral averaged cross sections in PFNS can be derived from the measured reactions in the reactor core. The evaluation of the neutron spectrum measurements in three different thermal-reactor cores shows that this lower limit is around the energy of 5.5 - 6 MeV. Above this energy the reactor spectra becomes identical with the 235 U PFNS. IAEA CIELO PFNS is within 5% of the measured PFNS from 10 to 14 MeV in a LR-0 reactor, while ENDF/B-VII evaluated PFNS underestimated measured neutron spectra., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

Fast neutron leakage spectra from the light and heavy water sphere of 30cm in diameter with neutron source in its centre were measured by a stilbene scintillation detector in the region of 1-10MeV in the distance of 85cm 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 3-4MeV. Furthermore, JENDL-4 library overestimates experiment in the region of 4-6.5MeV. 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., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

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., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

Zirconium is an important material used in most of reactor concepts for fuel cladding. Thus the knowledge of its cross section is important for reliable prediction of fuel operation. Also 90 Zr(n,2n) reaction, is included in IRDFF files as dosimetry cross section standard. Due to its very high threshold, 12.1MeV, it is suitable for measurement of high energy neutrons. One of possible interesting applications is also evaluation of prompt fission neutron spectra in 235 U and 238 U what is under auspices of the International Atomic Energy Agency in CIELO project. The experimental values - obtained with the LR-0 nuclear reactor - of various zirconium cross sections were compared with calculations with the MCNP6 code using IAEA CIELO, ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND- 2010, and CENDL-3.1 transport libraries combined with the dosimetry cross sections extracted from the IRDFF library. Generally, the best C/E agreement for 90 Zr(n,2n) cross section, was found with the IAEA CIELO 235 U evaluation that includes an updated prompt fission neutron spectra in the evaluated data file. The cross section of this reaction averaged over LR-0 spectra was determined being 28.9 ± 1.2 µb, corrected to spectral shift, spectral averaged cross section in 235 U was determined to be 0.107 ± 0.005mb. Notable discrepancies were reported in both 94 Zr(n,g) and 96 Zr(n,g)., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

A well-defined neutron spectrum is an essential tool for calibration and tests of spectrometry and dosimetry detectors, and evaluation methods for spectra processing. Many of the nowadays used neutron standards are calibrated against a fission spectrum which has a rather smooth energy dependence. In recent time, at the LVR-15 research reactor in Rez, an alternative approach was tested for the needs of fast neutron spectrometry detector calibration. This process comprises detector tests in a neutron beam, filtered by one meter of single-crystalline silicon, which contains several significant peaks in the fast neutron energy range. Tests in such neutron field can possibly reveal specific problems in the deconvolution matrix of the detection system, which may stay hidden in fields with a smooth structure and can provide a tool for a proper energy calibration. Test with several stilbene scintillator crystals in two different beam configurations supplemented by Monte-Carlo transport calculations have been carried out. The results have shown a high level of agreement between the experimental data and simulation, proving thus the accuracy of used deconvolution matrix. The chosen approach can, thus, provide a well-defined neutron reference field with a peaked structure for further tests of spectra evaluation methods and scintillation detector energy calibration., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

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 100cm 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 3.5-7.0MeV 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., (Copyright © 2017 Elsevier Ltd. All rights reserved.)

A well-defined neutron spectrum is essential for many types of experimental topics and is also important for both calibration and testing of spectrometric and dosimetric detectors. Provided it is well described, such a spectrum can also be employed as a reference neutron field that is suitable for validating selected cross sections. The present paper aims to compare calculations and measurements of such a well-defined spectra in geometrically similar cores of the LR-0 reactor with fuel containing slightly different enrichments (2%, 3.3% and 3.6%). The common feature to all cores is a centrally located dry channel which can be used for the insertion of studied materials. The calculation of neutron and gamma spectra was realized with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Only minor differences in neutron and gamma spectra were found in the comparison of the presented reactor cores with different fuel enrichments. One exception is the gamma spectrum in the higher energy region (above 8MeV), where more pronounced variations could be observed., (Copyright © 2016 Elsevier Ltd. All rights reserved.)

The present paper aims to compare the calculated and experimental reaction rates of (23)Na(n,2n)(22)Na in a well-defined reactor spectra of a special core assembled in the LR-0 reactor. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination. The resulting value averaged in spectra is 0.91±0.02µb. This cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of long-term activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Generally the best C/E agreement, within 2%, was found using the ROSFOND-2010 data set, whereas the worst, as high as 40%, was found using the ENDF/B-VII.0., (Copyright © 2016 Elsevier Ltd. All rights reserved.)

Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins., (Copyright © 2015 Elsevier Ltd. All rights reserved.)

The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well., (Copyright © 2014 Elsevier Ltd. All rights reserved.)

Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up., (© 2013 Elsevier Ltd. All rights reserved.)

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 45×45 mm(2) and a 10×10 mm(2) cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6-10 MeV and the photon spectrum in 0.2-9 MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries., (Copyright © 2013 Elsevier Ltd. All rights reserved.)