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• Measurement of reaction rates in benchmark reference neutron field in LR-0. • Evaluation of 14N(n,p)14C reaction integral cross section in LR-0 field. • Calculation of reaction rates in IRDFF-II. • Effect of gamma on reaction rates of dosimetrical reaction. • Discrepancies in gamma induced cross sections on gold. 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. [ABSTRACT FROM AUTHOR]

The long-term operation of existing nuclear power reactors is a crucial concern due to the complexities and expenses associated with replacing key components, such as the reactor pressure vessel and reactor internals. Gamma radiation, a byproduct of nuclear reactions and radioactive decay, significantly influences the lifetime of these components. This radiation is responsible for various degradation pathways leading to void swelling in steel reactor components and cracking or other radiation damage in concrete structures. A study conducted at a full-scale mock-up of the VVER-1000 reactor at the LR-0 zero-power reactor employed HPGe and stilbene measurements to analyze gamma spectra behind the reactor pressure vessel and within concrete biological shielding. While simulations behind the reactor pressure vessel aligned with measurements, notably, a marked overestimation of stilbene spectrum calculations occurred deep in concrete, suggesting potential inaccuracies in radiation predictions for power plant structures. [ABSTRACT FROM AUTHOR]

Silicon, especially silicon in the form of SiO2, is a major component of rocks. Final spent fuel storages, which are being designed, are located in suitable rock formations in the Earth's crust. Reduction of the uncertainty of silicon neutron scattering and capture is needed; improved silicon evaluations have been recently produced by the ORNL/IAEA collaboration within the INDEN project. This paper deals with the nuclear data validation of that evaluation performed at the LR-0 reactor by means of critical experiments and measurement of reaction rates. Large amounts of silicon were used both as pure crystalline silicon and SiO2 sand. The critical moderator level was measured for various core configurations. Reaction rates were determined in the largest core configuration. Simulations of the experimental setup were performed using the MCNP6.2 code. The obtained results show the improvement in silicon cross-sections in the INDEN evaluations compared to existing evaluations in major libraries. The new Thermal Scattering Law for SiO2 published in ENDF/B-VIII.0 additionally reduces the discrepancy between calculation and experiments. However, an unphysical peak is visible in the neutron spectrum in SiO2 obtained by calculation with the new Thermal Scattering Law.

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.

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.

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]

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. [ABSTRACT FROM AUTHOR]

Critical experiments using an insertion zone with FLiBe salt performed at the LR-0 reactor at Research Centre Řež (RC Řež) have been compared to application models for solid-fueled, fluoride-salt-cooled high-temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts using sensitivity and uncertainty (S/U) analysis techniques. These experiments support FHR and MSR advanced reactor concepts by informing on neutron spectral effects and nuclear data uncertainties related to fluoride salts. The FLiBe salt in the LR-0 experiments is from the Oak Ridge National Laboratory Molten Salt Reactor Experiment and is enriched to greater than 99.99% 7 Li. This work is part of a broader collaboration between the United States and the Czech Republic on MSR and FHR technology. Results from the S/U analyses for FHR and MSR models indicated the most significant potential source of eigenvalue bias due to nuclear data within the FLiBe salt is radiative capture in 7 Li. Other smaller but potentially significant contributions come from 19 F, 6 Li, and other 7 Li reactions. Similarity comparisons of the RC Řež LR-0 experiments and FHR and MSR application models indicate that the LR-0 FLiBe experiments could be useful as candidate benchmarks for low-enriched uranium–fueled application models. However, sensitivity differences observed included both spectral effects, due to the LR-0 reactor being moderated by water instead of graphite and fluoride salt, and sensitivity magnitude effects driven by the amount of FLiBe inserted and its location. New experiments with refined designs to increase the volume and importance of salt in the system could provide improved data. Results also demonstrate that salt systems using thorium and/or 233 U fuel will require additional experiments with relevant driver fuels. Increased contributions from the graphite moderator and 19 F and covariance between reactions such as 233 U fission and 233 U radiative capture indicate higher uncertainty contributions from the salt and significant differences in contributions from the fuel species due to underlying uncertainties in their nuclear data. [ABSTRACT FROM AUTHOR]

Validation of neutronic parameters for lead-cooled systems, by integral and local experiments, became of primary interest with advances in ALFRED lead cooled reactor demonstrator design. Notably, aiming at advancing the core design from the conceptual to the basic stage, the assessment of uncertainties coming from numerical simulation, with respect to real experiments, was of paramount importance to support the claims of outstanding safety demonstration meant for ALFRED. Among the required evidences, the assessment of the spatial distribution of the neutron flux and power in fuel pins required conceiving ad-hoc experiments, and disposing of state-of-the-art post-irradiation examination capabilities. For all these reasons, these experiments were one of the main focuses of the collaboration between the Research Centre Rez and ENEA. This paper presents the conception phase of the test and discusses some of the main results of the first phase of the experimental campaign, dealing with neutrons propagation, collected during its execution at the LR-0 research reactor (hence before the post-irradiation examination of the experimental pins placed in the lead test rig). The experimental work involved at first neutron spectrum measurements in the energy range from 0.1 to 10 MeV. Additionally, measurement of basic neutronic parameters of lead were performed, such as its reactivity worth, its effect on the neutron spectrum and its slowing-down properties. The comparison of calculations and experimental results shows good agreement. In case of calculation in benchmark model with different nuclear data libraries, the criticality is systematically over-predicted by approximately 150 pcm, which is, however, in the 1σ of the uncertainty interval. Neutron spectrum measurement shows only slight variations being around 10% most of the time. [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 delayed neutron fraction and generation time of the research reactor LR-0 were measured. • Fission chambers were placed in various locations to assess for their response. Signals were analyzed in current mode using a pile noise technique. • The experimental results were found in very good agreement with calculations performed with MCNP associated to ENDF-B.VII nuclear data library. For safety reasons, it is important to simulate precisely the kinetic behavior of light water nuclear reactors (LWR) in nominal or incidental situations. In the point kinetic approximation, a few integral parameters allow describing the reactor's behavior. It is of interest to measure those parameters, such as the delayed neutron fraction and delayed neutron abundances, in low power research reactors loaded with benchmarked fuel configurations to compare with simulations results. In this context, Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA) and Centrum Výzkumu Řež (CVŘ) have been collaborating towards measuring the delayed neutron fraction of LR-0, a VVER-type research reactor operated by CVŘ. This paper presents the first estimation of the delayed neutron fraction of the reactor loaded with six fuel assemblies, arranged in a small hexagonal lattice with a center cavity accessible for instrumentation. This configuration was chosen because it was very close to the so-called reference configuration with seven fuel assemblies, the one that was benchmarked in the NEA IRPheP Handbook. The measurement was performed in July 2021 using the CEA-developed acquisition system SPECTRON dedicated to pile noise experiments. Pile noise refer to a set of techniques which focus on the correlations induced by the fission process in the signals of neutron detectors. The experimental campaign was designed to meet two objectives: first to provide an estimation of the delayed neutron fraction; second to study the impact of the location of the neutron detectors on the quality of the results. Two fission chambers were used, one was fixed and located in the core center and the second was placed in various locations in the reflector. Several configurations were tested with the mobile detector progressively moved up to 20 cm from the closest fuel element. The reactor fission rate was obtained thanks to gamma spectrometry on metal foils (gold, nickel and iron). Simulations with TRIPOLI-4® and JEFF33 libraries were performed to convert the saturated activities into fission rates. The delayed neutron fraction was estimated at 782 ± 10 pcm and the generation time was measured at 39.1 ± 0.5 μs. The uncertainty balance is as follows: 2 % associated to the Diven factor and 1.5 % associated to the core's integral fission rate and 0.5 % for the statistical uncertainty. The reproducibility of the experiments was found very good, with a negligible dispersion amongst experiments (0.2 %). These results are valuable complements to the characterization of LR-0 reference neutron benchmark field. [ABSTRACT FROM AUTHOR]

The present paper describes reaction rate measurement of 89 Y(n,2n) 88 Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88 Y using gamma-ray spectrometry of irradiated Y 2 O 3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 µb, averaged in the 235 U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed. [ABSTRACT FROM AUTHOR]

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, 12.1 MeV, 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 ± 0.005 mb. Notable discrepancies were reported in both 94 Zr(n,g) and 96 Zr(n,g). [ABSTRACT FROM AUTHOR]

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 8 MeV), where more pronounced variations could be observed. [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 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). [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]

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±0.02 µb. 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. [ABSTRACT FROM AUTHOR]

Measurement and calculation of the axial power profile near the boundary of a moderated and non-moderated core is used to analyze the suitability of the neutron-physical process description, mainly the angular cross-section of a water-moderated uranium system. This is also an important issue because it affects the radiation situation above the partly flooded core of a water-moderated reactor. Axial power profiles of various fuel pins irradiated on reactor LR-0 were measured and the results were compared with MCNP6 code calculations using the ENDF/B-VII.0 nuclear data library. The calculated power profile in positions above the moderator level significantly underestimates experimental results. This might be caused by an improper description of the angular distribution of scattered neutrons in a water-moderated uranium system. [ABSTRACT FROM AUTHOR]

Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The present paper aims to compare the calculated and measured spectra after insertion of candidate materials for the Molten salt reactor/Fluoride cooled high temperature reactor system concept into the LR-0 reactor. The calculation is realized with MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-4, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Additionally, comparisons between the slowing down power of each media were performed. The slowing down properties are important parameters affecting the thickness of moderator media in a reactor. [ABSTRACT FROM AUTHOR]

The criticality measurements are the crucial point in design verification of all reactor concepts. The research on fluoride salts used in nuclear reactor technology has been carried out in Řež since 2002 and in the year 2006 also reactor experiments were performed. Due to the lack in the fluoride salt reactor experimental data, series of critical experiments (EROS) were performed on the LR-0 reactor during the SPHINX project. The work was focused on determination of reactivity worth and on evaluation of the calculation to experiment agreement of various nuclear data libraries in case of integral critical experiments in various core configurations. The core consisted of driver zone and different insertions with fluoride salts or graphite and their combinations. These materials were selected as they are perspective in MSR and FHR projects; discrepancies between experiment and calculations were discovered, which may be caused by nuclear data used for fluoride salts. [ABSTRACT FROM AUTHOR]

The study of the reactivity effects, which are in many cases negligible, is not a simple task. Consequently, the special core configuration with high sensitivity to low introduced reactivity changes was developed at LR-0 reactor. It proved to be suitable for precise definition of any material, in terms of its impact on reactivity, being inserted into a reactor core. The results also imply this kind of core configuration to be applicable for the planed precise measurement of the MSR FLIBE salt temperature reactivity coefficients. This paper also examines the agreement between calculations and experiments for some cases of the low reactivity introduction into the core. Besides that, the similar cores can be used for purposes of neutron dosimetry, therefore the neutron spectrum in the experimental region was determined and compared with the results of MCNP calculations with the standard nuclear data libraries. [ABSTRACT FROM AUTHOR]

Abstract: The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100h irradiation on a reactor with power of ~630W and several days cooling. Specifically the nuclides studied are 140Ba, 103Ru, 131I, 141Ce, 95Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived 92Sr, 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 &y& Elsevier]

Highlights: [•] Repairs in fuel assembly. [•] Effect of control rod. [•] Calculations and experiment comparison of inhomogeneities effect. [•] Influence of nuclear data libraries on inhomogeneities effect. [Copyright &y& Elsevier]

Abstract: 97Zr is a relatively high-yield fission product that can be used for zero reactor power determination. The technique is not widely used because in the case of reactors that use zirconium metal in the fuel cladding, it is not only a fission product but is also produced by activation. In an appropriately chosen time interval, results obtained using 97Zr can be compared to those of power determination performed using 92Sr. The knowledge of the ratio between fission-induced 97Zr and the portion of 97Zr activated in the cladding can be used not only for power-density determination but also as an important indication of fuel failures. [Copyright &y& Elsevier]

Highlights: [•] Mixed water moderator was used in an MCNP5 model of LR-0 reactor. [•] 11 Different heavy water percentages were used ranging from 0% to 100%. [•] Neutron destiny and flux were explored at different moderator compositions. [•] Dramatic changes in reactor neutronics were found between 90% and 100% D2O. [Copyright &y& Elsevier]

• Measurement of critical parameters of variously reflected core arrangements. • Measurement of fission density across cores with various neutron reflectors. • Calculation of criticality and spatial distribution of neutron flux in studied cores. • Discrepancies between calculation and experiment. 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 k eff calculated using the ENDF/B-VIII.0 library significantly underpredicts the experiment. The calculations using a new evaluation of 56Fe by the IAEA (INDEN collaboration) do not improve the situation significantly. Generally, the presented results are suitable for the validation of iron cross sections. [ABSTRACT FROM AUTHOR]

Abstract: 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. [Copyright &y& Elsevier]

Highlights: [•] VVER-1000 mock-up criticality coefficients determination. [•] Calculations of criticality using various libraries. [•] Calculations to experiment comparison of criticality coefficients. [•] Influence of thermal scattering law sublibraries on resulted k eff. [Copyright &y& Elsevier]

Abstract: The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5h irradiation on the power level of 9.5W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density. [Copyright &y& Elsevier]

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]

Abstract: The neutron fluence load on a reactor pressure vessel is an important physical quantity affecting material degradation and consequently reliable assessment of pressure vessel integrity and lifetime prolongation beyond designed conditions. This degradation is influenced by mixed neutron–photon source from the core periphery as well as by material parameters of the reactor pressure vessel. Computational procedures and experimental determination of neutron fluxes in the VVER-1000 mock-up internal structures as well as the results achieved are described in this paper. The calculation were performed with the MCNPX code with different nuclear data libraries. Nuclear data were processed using NJOY code. The neutron spectra measurements were performed with a two-parameter stilbene spectrometer. [Copyright &y& Elsevier]

Well-defined neutron spectrum is crucial for calibration and testing of detectors for spectrometry and dosimetry purposes. As a possible source of neutrons nuclear reactors can be utilized. In reactor core most of the neutrons are originated from fission and neutron spectra is usually some form of moderated spectra of fast neutrons. The reactor LR-0 is an experimental light-water zero-power pool-type reactor originally designed for research of the VVER type reactor cores, spent-fuel storage lattices and benchmark experiments. The main reactor feature that influences the performance of experiments is the flexible arrangement of the core. Special types of the possible core arrangements on the reactor LR-0 can provide different neutron spectra in special experimental channels. These neutron spectra are modified by inserting different materials around the channel and whole core is driven by standard fuel assemblies. Fast, epithermal or thermal spectra can be simulated using graphite, H2O, D2O insertions, air, Cd foils or fuel with different enrichment. [ABSTRACT FROM AUTHOR]