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In order to validate the neutron capture cross section in 241Am, comparisons of calculations with activation experiments at the TRIGA reactor at JSI and pile oscillation measurements at the MINERVE reactor at CEA Cadarache were done. The results suggest that the evaluation in the new test library JEFF-4T1 on average provide an improvement of the C/E results compared to the existing nuclear data library releases. However, due to some discrepancies, additional experiments at JSI as well as further development of the computational model of the JSI TRIGA reactor are planned to improve the accuracy and reliability of the neutron activation analysis and the nuclear data inferred from them.

The problem of calculating sensitivity coefficients to nuclear data in burnup problems is addressed. A non-intrusive method based on the random sampling of nuclear data and on linear regression is proposed to overcome the limitations of dealing with a complex coupled problem. Sensitivity plots are reported for the actinide concentrations that build up in irradiated fuel and that have the largest impact for the safe and economical management of spent nuclear fuel. Despite using a simplified pincell model, the range of applicability of the calculated sensitivities is further discussed. The sensitivity study is corroborated by an analysis of how different nuclear data libraries perform in the prediction of the neutron multiplication factor in a burnup problem. The origin of the discrepancies between the results obtained using the JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries are investigated.

The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides U , U and Pu , on Am and Na , Ni , Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy., Acknowledgements The authors would like to thank all contributors to the JEFF project that shared their insights and provided their specific input for the present JEFF-3.3 evaluation and its immediate predecessors JEFF-3.2 and JEFF-3.1.2.

The JSI TRIGA reactor features several in-core and ex-core irradiation facilities, each having different properties, such as neutron/gamma flux intensity, spectra and irradiation volume. A series of experiments and calculations was performed in order to characterise radiation fields in irradiation channel thus allowing users to perform irradiations in a well characterised environment. Since 2001 the reactor has been heavily used for radiation hardness studies for components used at accelerators such as the Large Hadron Collider (LHC) at CERN. Since 2010 it has been extensively used for testing of new detectors and innovative data acquisition systems and methods developed and used by the CEA. Recently, several campaigns were initiated to characterise the gamma field in the reactor and use the experimental data for improvement of the treatment of delayed gammas in Monte Carlo particle transport codes. In the future it is planned to extend the testing options by employing pulse mode operation, installation of a high energy gamma ray irradiation facility and allow irradiation of larger samples at elevated temperature.

Methodologies to derive cross section data from spectrum integrated reaction rates were studied. The Westcott convention and some of its approximations were considered. Mostly measurements without and with transmission filter are combined to determine the reaction cross section at thermal energy together with the resonance integral. The accuracy of the results strongly depends on the assumptions that are made about the neutron energy distribution, which is mostly parameterised as a sum of a thermal and an epi-thermal component. Resonance integrals derived from such data can be strongly biased and should only be used in case no other data are available. The cross section at thermal energy can be biased for reaction cross sections which are dominated by low energy resonances. The amplitude of the effect is related to the lower energy limit that is used for the epi-thermal component of the neutron energy distribution. It is less affected by the assumptions on the shape of the energy distribution. When the energy dependence of the cross section is known and information about the neutron energy distribution is available, a method to correct for a bias on the cross section at thermal energy is proposed. Reactor activation measurements to determine the thermal 241 Am(n, γ ) cross section reported in the literature were reviewed. In case enough information was available, the results were corrected to account for possible biases and included in a least squares fit. These data combined with results of time-of-flight measurements give a capture cross section 720 (14) b for 241 Am(n, γ ) at thermal energy.

For many practical applications, homogeneity of a field, whether neutron or electromagnetic, is of significance. Definitions of homogeneity or its calculation are diverse in the existing literature, however. This paper is an attempt to explore the concept of homogeneity and the ways it can be measured or calculated. The list of the measures that were considered includes: the ratio between the maximum and the minimum value, the ratio between the maximum difference and the mean value, the variance of the field and its generalization, and the mean and the variance of the field's gradient. These were applied to a realistic gamma-ray irradiation scenario, both by deriving a semi-analytical approximation and in a Monte Carlo numerical simulation. The irradiation arrangement included a hollow cylindrical irradiation channel and different numbers of spent fuel elements taken from a TRIGA reactor. By manipulating the geometry of this setup, fields of varying degrees of homogeneity were simulated and the homogeneity measures were calculated. By applying principal component analysis the maximum value to the weighted mean ratio was put forth as the most desirable measure. Finally, the application of this measure was demonstrated by choosing one of the irradiation configurations previously considered that produced a homogeneous gamma-ray field.

An overview of the TRIGLAV package, a tool for the computational analysis of TRIGA Mark II reactors based on lattice transport code WIMSD for unit cell cross section homogenization and four-group diffusion theory in 2D cylindrical coordinates, is presented. Possible unit cells include four types of fuel elements and different non-fuel elements, such as water, graphite, control rod, etc. Apart from the calculation of the main parameters, i.e. the multiplication factor and neutron flux distribution, secondary parameters, such as fuel power and temperature distribution, xenon defect, etc., are also calculated. The program has been compared with the Monte Carlo code MCNP5. Some possible applications of the TRIGLAV program are also outlined in the paper.

SANDY is a nuclear data sampling code compatible with nuclear data files in ENDF-6 format. Exploiting the basic theory of stochastic sampling, SANDY generates random nuclear data samples that reproduce the covariance information stored in the ENDF-6 files. Such random data are rewritten in perturbed ENDF-6 or PENDF files and can be used as nuclear code inputs to produce perturbed responses. After the statistical analysis of the sampled responses, the sample mean and variance are quantified. Not only can SANDY be used for the study of the response sample variance, but it can also estimate global sensitivity indices for correlated and uncorrelated input parameters using a variance-based decomposition method. SANDY was tested for the generation of random ENDF-6 files with perturbed cross sections and resonance parameters. These files were used for the uncertainty quantification of integral responses, such as the one-group integrated cross section and the resonance integral of several isotopes. Then, the response variance was apportioned to the several inputs and the parameters producing the largest impact were identified.

The equivalence between the neutron source and flux spectra is often implicitly used in practice although many times the users are not even aware of this. This work identifies two conditions under which the equivalence holds. Namely, if the neutron interaction between the source region and the volume where flux is observed is negligible, and the neutron mean track length in the observed volume does not depend on their energy, source and flux spectra are equivalent. Consequently, a flux determined on a closed surface from a full system calculation can be replaced by an equivalent source for a simplified model including only the region contained by the surface.

An evaluation of the criticality and relative fission rate radial distribution experiments in an 69.2-cm diameter aluminium sphere filled with intermediately enriched UO2F2 aqueous solution is presented. An evaluation of the total experimental uncertainty has been performed within the framework of the International Criticality Benchmark Evaluation Project (ICSBEP) and International Reactor Physics Benchmark Evaluation Project (IRPhEP). In addition, the uncertainty due to the uncertainties in the nuclear data is evaluated in this paper. It has been determined that the highest contribution to the overall uncertainty in the effective multiplication factor keff is due to the uncertainty in the uranium enrichment ( ∼ 550 pcm in keff due to the 4% uncertainty in 235U enrichment). The highest experimental uncertainties in the relative fission rate profile in terms of the relative standard uncertainty (Lrs) are 0.032, 0.016, and 0.020 due to the uncertainty in guide tube diameter, filling tubes position, and detector position, respectively. It is estimated that the uncertainty in the nuclear data contributes approximately 400 pcm–1200 pcm to the uncertainty in keff depending on the covariance data library used. keff and radial fission rates have been calculated with MCNP and COG Monte Carlo neutron transport codes and have been compared to the experimental benchmark values. In general, the various Monte Carlo codes have given similar results with deviations within uncertainties.

Constant improvements of the computational power and methods as well as demands of accurate and reliable measurements for reactor operation and safety require a continuous upgrade of the instrumentation. In particular, nuclear sensors used in nuclear fission reactors (research or power reactors) or in nuclear fusion facilities are operated under intense mixed neutron and gamma-ray fields, and need to be calibrated and modeled to provide selective and accurate neutron and gamma-ray measurements. The French Atomic Energy and Alternative Energies Commission (CEA) and the Jožef Stefan Institute (JSI) have started an experimental program dedicated to a detailed experimental benchmark with analysis using Monte Carlo particle transport calculations and a series of neutron and gamma-ray sensor types used in the JSI TRIGA Mark II reactor. CEA has setup a simplified TRIPOLI-4® modeling scheme of the JSI TRIGA reactor based on the information available in the IRPhEP benchmark in order to facilitate analysis of future neutron and gamma-ray measurements. These allow the CEA to perform a TRIPOLI-4 instrumentation calculation scheme benchmarked with the JSI MCNP model. This paper presents the main results of this CEA calculation scheme application and the analysis of their comparison to the JSI results obtained in 2012 with the MCNP5 & ENDF/B-VII.0 calculation scheme. This paper will conclude with some information about the new experimental program to be carried out in 2022 in the TRIGA reactor core.

In this paper, two main exercises have been carried out to describe the effect that varying an albedo boundary condition has in the computation of observables such as decay heat, neutron emission rate and nuclide inventory from a PWR fuel assembly (or a configuration of assemblies) during a depletion scenario. The SERPENT2 code was then employed to emphasize the importance of modeling a proper boundary condition for such purposes. Moreover, the effect of taking into account more than a single fuel-pin region for depletion studies while varying the type of boundary condition, was also accounted for. The first exercise has the main objective of comparing in a single fuel assembly the albedo variations ranging from 1.1 up to full vacuum conditions. By comparing to the reference assembly (considered to be the case of full reflective conditions), relative differences up to +17% were observed in decay heat and up to almost -30% in neutron emissions. Also, a clear dependence on the albedo was detected if more than one depletable zone was considered while computing the integral value of observables of interest. Regarding the second exercise, where a 3 × 3 configuration of fuel assemblies is being now considered with a reflector section in the middle, a negligible effect on the observables was observed for the single fuel pin zone case; instead, an effect in the 244Cm computation when analyzing two fuel pin-zones produced a change in the neutron emission rate during cooling time up to 2.5% (while comparing it to the reference single assembly case).

Methods for random sampling of correlated parameters are presented. The methods are implemented for sampling of resonance parameters in ENDF-6 format and a link to the open-source code ENDSAM is given. The code has been validated on realistic data. Additionally, consistency of covariances of resonance parameters of three major contemporary nuclear data libraries (JEFF-3.2, ENDF/B-VII.1 and JENDL-4.0u2) has been checked.

Within the bilateral project between the CEA Cadarache and the Jožef Stefan Institute (JSI) a wide variety of measurements using multiple fission chambers simultaneously inside the reactor core were performed. The fission rate axial profiles were measured at different positions in the reactor core and at different control rod configurations. A relative comparison of the calculated fission rates using the MCNP code and the measured fission rates was performed. In general the agreement between the measurements and calculations is good, with the deviations within the uncertainties. For better observation and understanding of the neutron flux redistribution due to the control rod movement, the neutron flux and fission rate had been calculated through the entire reactor core for different control rod configurations. The detector position with minimum signal variations due to the regulating and compensating control rod movement during normal operation was determined. The minimum variation is optimal in case we want to reliably determine the reactor power without influence of the regulating and compensating control rod positions.

MYRRHA is a multi-purpose research reactor able to operate in sub-critical and critical modes and currently in the design phase at SCK•CEN. The choice of LBE was driven by its chemical stability, low melting temperature, high boiling point, low chemical reactivity with water and air and a good neutronic performance. As a drawback, the neutron capture in 209Bi results in the production of 210Po, a highly radiotoxic alpha emitter with relatively short half-life (≈138 days). The 210Po production represents a major safety concern that has to be addressed for the reactor licensing. In this work we used the ALEPH-2 burnup code to accurately calculate the 210Po production in a MYRRHA operating cycle. The impact of using different nuclear data libraries was evaluated and the reliability of the results was determined by quantifying the uncertainty of the 210Po concentration. The uncertainty quantification was carried out sampling the currently available nuclear data covariance matrices with the SANDY code. Also, estimates of the sensitivity profiles were obtained with a linear regression approach. The activation yield of the 209Bi neutron capture reaction was assessed as the largest nuclear data source of uncertainty, however the lack of covariances for such data represent a capital drawback for the 210Po content prediction.

The methodology to derive cross-section data from measurements in a cold neutron beam was studied. Mostly, capture cross-sections at thermal energy are derived relative to a standard cross-section, e.g. the cross-section of the 1 H(n,γ), 14 N(n,γ), or 197 Au(n,γ) reaction, and proportionality between the standard and the measured cross-section, evaluated at different energies in the sub-thermal region, is often assumed. Due to this assumption the derived capture cross-section at thermal energy can be biased by more than 10%. Evidently the bias depends on how much the energy dependence of the cross-section deviates from a direct proportionality with the inverse of the neutron speed. The effect is reduced in case the cross-section is not derived at thermal energy but at an energy close to the average energy of the cold neutron beam. Nevertheless, it is demonstrated that the bias can only be avoided in case the energy dependence of the cross-section is known and proper correction factors are applied. In some cases the results are also biased when the attenuation of the neutron beam within the sample is neglected in the analysis. Some of the cross-section data reported in the literature suffer from such bias effects. Hence, the results have to be corrected using the correction factors presented in this paper.

The EURACOS experiment series (sodium, iron, and no shielding) has been re-evaluated in the scope of the OECD/NEA SINBAD project to determine how useful these benchmarks can be for the validation of the modern nuclear data evaluations. The measured activation rates of different foils as well as unfolded proton recoil spectra at different depths in sodium and iron have been compared to the Monte Carlo calculated parameters. The results suggest good agreement with the experiment only for the high-threshold 32S(n,p) reaction detectors. The comparison of the lower threshold 103Rh(n,n′) and 115In(n,n′) reactions, and the (n,γ) reaction on 197Au under cadmium cover, reveals large discrepancy between the experimental and calculated spatial distributions for these reaction rates which is expected to be due to the uncertainties in the source and background description rather than due to the cross section uncertainties. On the other hand, the unfolded neutron spectra shapes are in relatively good agreement with the calculated ones. The overall results show that the EURACOS Na and Fe experiments cannot be considered of benchmark quality mainly due to unsatisfactory information on the neutron source. Nevertheless, some activation measurements from the EURACOS sodium and iron experiments such as the high threshold reactions and the spectra measurements are potentially useful for modern nuclear data and computer code benchmarking.

Two methods for random sampling according to a multivariate lognormal distribution – the correlated sampling method and the method of transformation of correlation coefficients – are briefly presented. The methods are mathematically exact and enable consistent sampling of correlated inherently positive parameters with given information on the first two distribution moments. Furthermore, a weighted sampling method to accelerate the convergence of parameters with extremely large relative uncertainties is described. However, the method is efficient only for a limited number of correlated parameters.

The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×103Svh-1 to 6×106Svh-1, silicon dose equivalent from 6×102Gy/hsi to 3×105Gy/hsi, and neutron air kerma from 4.3×103Gyh-1 to 2×105Gyh-1. Ratio of fast (1MeV

The SINBAD database currently contains compilations and evaluations of over 100 shielding benchmark experiments. The SINBAD database is widely used for code and data validation. Materials covered include: Air, N. O, H2 O, Al, Be, Cu, graphite, concrete, Fe, stainless steel, Pb, Li, Ni, Nb, SiC, Na, W, V and mixtures thereof. Over 40 organisations from 14 countries and 2 international organisations have contributed data and work in support of SINBAD. Examples of the use of the database in the scope of different international projects, such as the Working Party on Evaluation Cooperation of the OECD and the European Fusion Programme demonstrate the merit and possible usage of the database for the validation of modern nuclear data evaluations and new computer codes.

A sensitivity and uncertainty analysis was carried out to estimate the uncertainty in the neutron multiplication factor k eff and to identify the most important nuclear data for neutron induced reactions for criticality calculations of the latest MYRRHA designs. Sensitivity profiles, i.e. sensitivity to the nuclear data as a function of incoming neutron energy, were derived for both a critical and sub-critical core. They were calculated using codes that are based on different methodologies including stochastic and deterministic calculations (i.e. SCALE, MCNP and XSUN). The neutron induced nuclear data sensitivity analysis outlined the following quantities to be of special importance for the MYRRHA reactor concept: 239 Pu(n,γ) both in resonance and fast energy region, (n,f) fast, χ and ν fast; 238 U(n,n′) fast, (n,γ) resonance and fast, (n,n) resonance and fast; 240 Pu ν fast; 238 Pu(n,f) both resonance and fast; 56 Fe(n,γ) both resonance and fast. Differences of less than 4% between codes were obtained for these quantities, with few exceptions ( 238 Pu(n,f), 238 U(n,n) and 56 Fe(n,γ) reactions). Nuclear data covariance matrices of different libraries (SCALE-6, COMMARA-2 and JENDL-4.0m) were used to derive the uncertainty in k eff based on the calculated sensitivities. This study reveals that the largest contributions to k eff uncertainty result from the uncertainty in the average prompt neutron fission multiplicity of 239 Pu, in the 238 U inelastic scattering cross section and 239 Pu fission cross section, using the covariances from SCALE-6, COMMARA-2 and JENDL-4.0m, respectively.

The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) concept is a flexible experimental lead-bismuth cooled and mixed-oxide (MOX) fueled fast spectrum facility designed to operate both in sub-critical (accelerator driven) and critical modes. One of the key issues for the safe operation of the reactor is the uncertainty assessment during the design works. The main objective of the European project CHANDA (solving CHAllenges in Nuclear DAta) Work Package 10 is to improve MYRRHA relevant nuclear data in order to reduce the reactor parameter uncertainties derived from them. In order to achieve this goal, several tasks have been undertaken. First, a sensitivity study of MYRRHA integral parameters, such as energy dependent cross sections, fission spectra and neutron multiplicities, to nuclear data has been conducted resulting in a list of MYRRHA relevant quantities (nuclides and reactions). On the second task, an analysis of the existing experimental data and evaluations for the quantities included in the list has been carried out. In this framework, the impact on the multiplication factor of quantities from different nuclear data libraries for different nuclides, reactions and energy regions has been investigated on the MYRRHA MOX critical core model. As the next step, new experiments and evaluations will be performed in order to improve existing nuclear data libraries.

We generated a preliminary set of resonance parameters for 182,183,184,186W in the neutron energy range of thermal up to several keV. The evaluation methodology uses the Reich-Moore approximation to fit with the R-matrix code SAMMY, the high-resolution measurements performed in 2010 and 2012 at the Geel linear accelerator facility. For 183W, the transmission data and capture cross sections calculated with the set of resonance parameters are compared with the experimental values, and some of the average properties of the resonance parameters are discussed. In the analyzed energy range, this work almost doubles the existing resolved resonance evaluations in the ENDF/BVII. 1 library. A preliminary analysis of the performance of the calculated cross sections based on Lead slowing-down benchmarks is presented and briefly discussed., JRC.D.4-Standards for Nuclear Safety, Security and Safeguards

Covariance data in the existing evaluated nuclear data libraries often include large relative uncertainties and mathematical inconsistencies, which arise especially in combination with random sampling. The 232 Th evaluation from the ENDF/B-VII.1 library has been taken as an example. Possible solutions for mathematically impossible correlation matrices with negative eigenvalues and too low correlation coefficients between inherently positive parameters with large relative uncertainties are proposed. Convergence of the random sampling for lognormal distribution with extremely high relative standard deviations is slow by nature. Using weighted sampling, single parameters or a limited number of correlated parameters with large uncertainties can be sampled. Efficient sampling of a large number of correlated parameters with extremely large relative uncertainties remains unsolved.

Inherently positive parameters with large relative uncertainties (typically ≳ 30 % ) are often considered to be governed by the lognormal distribution. This assumption has the practical benefit of avoiding the possibility of sampling negative values in stochastic applications. Furthermore, it is typically assumed that the correlation coefficients for comparable multivariate normal and lognormal distributions are equivalent. However, this ideal situation is approached only in the linear approximation which happens to be applicable just for small uncertainties. This paper derives and discusses the proper transformation of correlation coefficients between both distributions for the most general case which is applicable for arbitrary uncertainties. It is seen that for lognormal distributions with large relative uncertainties strong anti-correlations (negative correlations) are mathematically forbidden. This is due to the asymmetry that is an inherent feature of these distributions. Some implications of these results for practical nuclear applications are discussed and they are illustrated with examples in this paper. Finally, modifications to the ENDF-6 format used for representing uncertainties in evaluated nuclear data libraries are suggested, as needed to deal with this issue.

Three different methods for multivariate random sampling of correlated resonance parameters are proposed: the diagonalization method, the Metropolis method, and the correlated sampling method. For small relative uncertainties (typical for s-wave resonances) and weak correlations all methods are equivalent. Differences arise under difficult conditions: large relative uncertainties of inherently positive parameters (typical for widths of higher-l-wave resonances) and/or strong correlations between a large number of parameters. The methods are tested on realistic examples; advantages and disadvantages of each method are pointed out. The correlated sampling method is the only method which produces consistent samples under any conditions. In the field of reactor physics, these methods are mostly used for the sampling of nuclear data, however, they may be used for any data with given uncertainties and correlations.

Conference of 2016 International Conference on Nuclear Data for Science and Technology, ND 2016 ; Conference Date: 11 September 2016 Through 16 September 2016; Conference Code:130583; International audience; In the framework of the OECD/NEA WPEC subgroup 41, ways to improve neutron induced capture cross sections for 241Am and 237Np are being sought. Decay data, energy dependent cross section data and neutron spectrum averaged data are important for that purpose and were investigated. New time-of-flight measurements were performed and analyzed, and considerable effort was put into development of methods for analysis of spectrum averaged data and re-analysis of existing experimental data.

A method for correlated random sampling is presented. Representative samples for multivariate normal or log-normal distribution can be produced. Furthermore, any combination of normally and log-normally distributed correlated variables may be sampled to any requested accuracy. Possible applications of the method include sampling of resonance parameters which are used for reactor calculations.

Activation measurements can be used for neutron spectrum determination in irradiation facilities. It is very important that self-shielding is taken correctly into account. Activation measurements of monitor materials were performed at the CEA EOLE facility. Self-shielding factors for the monitor materials were calculated at the Commissariat a l’Energie Atomique with the TRIPOLI 4.6 code using a detailed whole-core model, and at the Jožef Stefan Institute with the MCNP code using a simplified model. The results were compared to assess the relative merits and the achievable accuracy of the calculations. In the end, the results were used to determine the neutron spectrum of the studied location in the EOLE irradiation facility.

There is a growing interest in analyzing the possibility for current nuclear power plants operation extension. In that sense, life management programs, considering safety components ageing, are being developed and employed. On the other side, the large uncertainties of the ageing parameters as well as the uncertainties associated with most of the reliability data collections are widely acknowledged. This paper deals with uncertainty analysis associated with specific ageing rates database. The analysis is conducted using an analytical unavailability model applied for a selected safety system in a nuclear power plant. The most important problem is the immense uncertainty associated to the component ageing data sets as well as the lack of the very data in general, which would correspond to the more detailed modelling of ageing. New probability distributions, encompassing the ageing rates available in the considered data set, are suggested. The obtained results indicate the extent to which the uncertainty of the considered ageing data set, given the inherently assigned probability distribution, influences the performed unavailability calculations. Additionally, comparative analysis regarding the insights gained out of the application of the suggested probability distributions is conducted.

In this paper we investigate the influence of various up-to-date nuclear data libraries, such as ENDF/B-VI.6, ENDF/B-VII.0 and JEFF 3.1, on the multiplication factor of the TRIGA benchmark with fuel made of enriched uranium and zirconium hydride and SB light-water reactor benchmarks with fuel made of fissile material in zirconium matrix. The calculations are performed with the Monte Carlo computer code MCNP. Differences of ∼600 pcm in k eff are observed for the benchmark model of the TRIGA reactor, while there are practically no differences in the k inf of the fuel. Therefore, an investigation is performed also for hypothetical homogeneous and heterogeneous systems with different leakage. The uncertainty analysis shows that the most important contributors to the difference in k eff are the Zr isotopes (especially 90 Zr and 91 Zr) and thermal scattering data for H and Zr in ZrH. As the differences in k eff due to the use of different cross section libraries are relatively large, there is certainly a need for a review of the evaluated cross section data of the zirconium isotopes.

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

For nuclides with a large number of resonances the covariance matrix of resonance parameters can become very large and expensive to process in terms of the computation time. By converting covariance matrix of resonance parameters into covariance matrices of background cross-section in a more or less coarse group structure a considerable amount of computer time and memory can be saved. The question is how important is the information that is discarded in the process. First, the uncertainty of the 55Mn resonance integral was estimated in narrow resonance approximation for different levels of self-shielding using Bondarenko method by random sampling of resonance parameters according to their covariance matrices from two different 55Mn evaluations: one from Nuclear Research and Consultancy Group NRG (with large uncertainties but no correlations between resonances), the other from Oak Ridge National Laboratory (with smaller uncertainties but full covariance matrix). We have found out that if all (or at least significant part of the) resonance parameters are correlated, the resonance integral uncertainty greatly depends on the level of self-shielding. Second, it was shown that the commonly used 640-group SAND-II representation cannot describe the increase of the resonance integral uncertainty. A much finer energy mesh for the background covariance matrix would have to be used to take the resonance structure into account explicitly, but then the objective of a more compact data representation is lost.

In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)–Gold (0.1 wt%) foils (disks of 5 mm diameter and 0.2 mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization.

Neutron fluxes and reaction rates in MCNP output tallies are normalized per source neutron. Since the absolute normalization for the fixed source (NPS) calculations is straightforward, the absolute normalization for the criticality (KCODE) calculations is addressed. The correspondence between the MCNP calculated and measured flux is sought. The normalization constant at a certain thermal power of a neutron multiplication system is derived and discussed. This normalization constant is the same for both neutron flux and reaction rate tallies.

Rhodium foils of about 5 mm diameter were obtained from IRMM. One foil had thickness of 0.006 mm and three were 0.112 mm thick. They were irradiated in the pneumatic transfer system and in the carousel facility of the TRIGA reactor at the Jozef Stefan Institute. The foils were irradiated bare and enclosed in small cadmium boxes (about 2 g weight) of 1 mm thickness to minimise the perturbation of the local neutron flux. They were co-irradiated with 5 mm diameter and 0.2 mm thick Al-Au (0.1%) alloy monitor foils. The resonance self-shielding corrections for the 0.006 and 0.112 mm thick samples were calculated by the Monte Carlo simulation and amount to about 10% and 60%, respectively. The consistency of measurements confirmed the validity of self-shielding factors. Trial estimates of Q0 and k0 factors for the 555.8 keV gamma line of 104 Rh were made and amount to 6.657 0.18 and (6.617 0.12) � 10 -2 , respectively.

The canister filling problem arises when optimizing the spent nuclear fuel repository in hard rock. It is shown the problem is NP-hard. Constructive heuristics followed by remove and reinsert local optimization are considered. Additionally, some theoretical insight on the algorithm operation is provided. Several variants of the algorithm are compared on random and realistic datasets. The obtained results have shown that constructive heuristics give satisfactory results for both random and realistic input data, although there is still some room for improvements.

Modern Monte Carlo transport codes in combination with fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, β eff , and mean neutron generation time, Λ . We calculate β eff and Λ for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It is observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. β eff varies for 12 wt.% uranium standard fuel with 20% enrichment from 0.0080 for a small core (43 fuel rods) to 0.0070 for a full core (90 fuel rods). It is found that calculated value of β eff strongly depends also on the nuclear data set used in calculations. The prompt neutron lifetime mainly depends on the amount (due to either content or enrichment) of 235 U in the fuel as it is approximately inversely proportional to the average absorption cross-section. It varies from 28 μs for 30 wt.% uranium content fuelled core to 48 μs for 8.5 wt.% uranium content LEU fuelled core. Description of the calculation method and detailed results are presented in the paper.

Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI) and Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA Cadarache). The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu) covered with thermal neutron filters (Cd and Gd). The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter) on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

Whenever the sample size in neutron activation analysis cannot be made small enough, self-shielding effects need to be taken into account. When several resonance absorbers are present in the sample, resonance interference must also be considered. Estimation of the self-shielding factors by the Monte Carlo technique is too cumbersome for routine application. Various simplified approaches were compared to rigorous Monte Carlo calculations, pointing out their potential limitations. Good results are obtained using self-shielding factors calculated from evaluated nuclear data libraries and tabulated as a function of the dilution cross-section. The dilution cross-section depends on the material composition and the sample dimensions through the equivalence principle, which is well known in reactor physics. Resonance interference is calculated by solving the neutron spectrum slowing-down equation from cross-sections in 640-group representation. The MATSSF code was written for the purpose and is available on request.

The main aim of this paper is the development and validation of a 3D computational model of TRIGA research reactor using Serpent 2 code. The calculated parameters were compared to the experimental results and to calculations performed with the MCNP code. The results show that the calculated normalized reaction rates and flux distribution within the core are in good agreement with MCNP and experiment, while in the reflector the flux distribution differ up to 3% from the measurements.

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system.

Absolute average capture cross sections of gold, thorium, tantalum, molybdenum, copper and strontium in 252 Cf spontaneous fission neutron spectrum were simulated for two types of experiment setups preformed by Z. Dezso and J. Csikai and by L. Green. The experiments were simulated with MCNP5 using cross section data from the ENDF/B-VII.0 library. The determination of neutron backscattering was calculated with the use of neutron flagging. Correction factors to experimentally measured values were determined to obtain average cross sections in a pure 252 Cf spontaneous fission spectrum. Influence of concrete wall thickness, air moisture and room size on the average cross section was analyzed. Correction factors amounted to about 30%. Corrected values corresponding to average cross sections in a pure 252 Cf spectrum were calculated for 197 Au, 232 Th, 181 Ta, 98 Mo, 65 Cu and 84 Sr. Average cross sections were also calculated with the RR_UNC software using IRDFF-v.1.05 and ENDF/B-VII.0 libraries. The revised average radiative capture cross sections are 75.5±0.1 mb for 197 Au, 87.0±1.6 mb for 232 Th , 98.0±4.5 mb for 181 Ta, 21.2±0.5 mb for 98 Mo, 10.3±0.3 mb for 63 Cu, and 34.9±6.5 mb for 84 Sr.

Following a series of recent cross section evaluations on tungsten isotopes performed at Oak Ridge National Laboratory (ORNL), this paper presents the validation work carried out to test the performance of the evaluated cross sections based on lead-slowing-down (LSD) benchmarks conducted in Grenoble. ORNL completed the resonance parameter evaluation of four tungsten isotopes – 182,183,184,186 W – in August 2014 and submitted it as an ENDF-compatible file to be part of the next release of the ENDF/B-VIII.0 nuclear data library. The evaluations were performed with support from the US Nuclear Criticality Safety Program in an effort to provide improved tungsten cross section and covariance data for criticality safety sensitivity analyses. The validation analysis based on the LSD benchmarks showed an improved agreement with the experimental response when the ORNL tungsten evaluations were included in the ENDF/B-VII.1 library. Comparison with the results obtained with the JEFF-3.2 nuclear data library are also discussed.

CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring.

A review of the experimental data for natC(n, γ) and 12C(n, γ) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK•CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Revay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.