This paper presents a new procedure to optimize the geometric parameters of a n-type coaxial HPGe detector. It is based on a statistical technique called “Design of Experiments” (DoE). This technique ...aims to identify the most influential parameters and to determine the optimal configuration. In this work, The effects of each parameter on the detector responses have been investigated by a fractional factorial design. Only the most influential factors contributing to the detector response have been selected. Precise modeling of these factors was then performed using a full factorial design. Based on the results obtained from this design, the full energy peak efficiencies according to the geometric parameters were modeled by a multiple-linear regression. These models have been statistically validated by analysis of variance (ANOVA). The optimal combination of the geometric parameters has been identified using the desirability function approach, which is a useful tool to optimize multi-response problems. A verification test was performed to validate the results obtained. It was observed that the relative deviation found between experimental and simulated values was less than 5%.
•Optimization of geometrical dimensions of an HPGe detector.•Application of Monte Carlo simulation and “Design of Experiments” technique.•Investigating influence of each detector parameter on the FEPE.•Achieving good agreement between the measured and the simulated results.
The uncertainties in nuclear data give inaccurate values of reaction rates in shielding benchmarks. Consequently, the inclusion of the shielding benchmarks in the validation of the nuclear data would ...contribute to the production of general-purpose nuclear data evaluations. The validation of nuclear data goes through an essential step of sensitivity and uncertainty analysis. During this step, the evaluator determines which nuclear data necessitate adjustment. In this work, sensitivity and uncertainty analysis of reaction rates due to two primordial cross sections elastic and capture uncertainties of important isotopes in PCA-REPLICA benchmark are carried out. The reaction rates and their sensitivities are calculated by MCNP6.1 code. The weight window and energy cut-offs are two variance reduction techniques used in this calculation. The covariance matrices of the studied cross sections are processed by NJOY99 software. The sensitivities and uncertainties of the estimated reaction rates are calculated in four energy groups for three detectors 103Rh (n,n’)103Rhm, 115In (n,n’)115Inm and 32S (n,p)32P. The obtained results show that the elastic cross sections of 56Fe, 16O and 1H give large uncertainties on reaction rates, whereas the captures of these isotopes have small uncertainties on reaction rates. The analysis of the sensitivities and the uncertainties of the reaction rates due to elastic and capture cross sections of the 56Fe, 1H and 16O indicate that the elastic cross section of 56Fe needs improvement in the energy range 0.1 MeV–19.60 MeV and the elastic cross sections of 1H and 16O need adjustments in the energy range 1.35 MeV–19.60 MeV.
•Reaction rates calculation in PCA-REPLICA using variance reduction techniques.•56Fe, 1H, 16O cross sections analysis using Rh-103 (n,n’)Rh-103m detector.•56Fe, 1H, 16O cross sections analysis using In-115 (n,n’)In-115m detector.•56Fe, 1H, 16O cross sections analysis using S-32 (n,p)P-32 detector.
In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle ...MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U235(n, f), U235(nν¯) and H1(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.
The main objective of this study is to assess the neutronic modeling and calculations of the nuclear heating reactor NHR-5. To this end, the neutronic parameters of NHR-5 reactor underwent a ...comparative and validation protocols/methods through the analysis of the finite multiplication factor keff and some neutronic parameters including D,νΣf,Σa as well as the power and flux distributions, using the evaluated nuclear data libraries ENDF/B-VII.1, JEFF3.1 and JENDL4.0 based on 172 energy groups.
In this study, the collision probability approach is used to simulate and calculate the neutronic parameters in NHR5 reactor using the deterministic core diffusion code DONJON5 and the lattice transport code DRAGON5. The group constants of the reactor components were produced by DRAGON5. The DONJON5 code was then used to calculate the effective multiplication factor, excess reactivity, and power and flux distributions by introducing these group constants. The results are compared to those of the experiment to validate the calculation scheme. The results of the simulations reveal that the computed neutronic parameters consistently produce reasonable and consistent results of flux and power distributions, excess reactivity and all other parameters when compared to the experiment values. Therefore, the reactor models of NHR5 developed by DRAGON and DONJON codes were good in predicting the effective multiplication factor as well as the studied neutronic parameters. In this study, we being able to show the potential validation of the reactor physics lattice transport code DRAGON5 and the core diffusion code DONJON5, as well as the nuclear data libraries ENDF/B-VII.1, JEFF3.1 and JENDL4.0.
•Neutronic Modeling and calculations of the Nuclear Heating Reactor NHR-5.•Validation of the deterministic transport code DRAGON5 and diffusion code DONJON5.•Calculation and analysis of excess reactivity as well as the power and flux distributions by deterministic codes.•To rely on DRAGON5 and DONJON5 codes for NHR-5 calculations.•The good consistency of the results ensures that a thermal-hydraulic analysis will be performed for NHR reactor.
The purpose of neutronic calculations is to determine many principal integral parameters such as effective multiplication factor (keff), reaction rate, spectrum indices, etc. These parameters, are ...based on several cross sections as well as their uncertainties. However, the uncertainty propagations effect will give, in turn, inaccurate values of these integral parameters. Therefore, the margin of reactor safety can be decreases. In order to minimize these risks, the adjustment of basic nuclear data (Cross section) is required.
Cross sections adjustment techniques consist essentially of using available information from integral experiments to improve the basic nuclear data. In this work, the multi-group cross section adjustment is processed for data which are primordial in neutronic nuclear calculations and their covariance matrices are available in the ENDF files. The adjustment process is based on the uncertainty of keff using the maximum likelihood method (i.e. general least squares method GLLSM). So, several critical benchmarks, their sensitivities matrices and the cross sections covariance matrix are required when using this method. Benchmarks have been taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE), their sensitivity matrices and the covariance matrices of the desired cross sections have been calculated by MCNP6 code and NJOY software respectively. This study investigates the effects of the correlation between reactions on the prior and posterior nuclear data uncertainty, adjusted cross-sections and their standard deviations and the adjusted integral parameters (keff). A significant reduction of the a-priori uncertainties and a good convergence of the C/E ratio are observed after adjustment by using the a-posteriori covariance and cross sections data.
•The group cross section uncertainty estimation to the keff in the ENDF/B-VII.1 evaluation using MCNP6.1 and NJOY99.•The cross section adjustment process by using linear least squares method in two cases (correlation and non-correlation).•The covariance matrix adjustment by using linear least squares method.•The comparison between prior and posterior keff in the two cases.
In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle ...MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U235(n, f), $ U_{235}(n\bar{\nu})$ and H1(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.
•Analyzing uncertainty on keff due to cross section uncertainties in KRITZ-2:19.•Sensitivity vectors are calculated using the Monte Carlo code MCNP6.1 - Ksen card.•Cross sections were processed by ...the recent nuclear data processing system NJOY2016.
In criticality and stability studies of the nuclear reactor, it is important to evaluate the impact of the uncertainties of the basic nuclear data (cross sections) on the different neutron parameters. In this work sensitivity and uncertainly analysis of KRITZ-2:19 experiment at 21.1 °C was carried out with respect to cross-section uncertainties of 1H, 10B, 16O, 235U, 238U, 239Pu, 240Pu, 241Pu and 241Am isotopes on the effective multiplication factor ” keff ” for JENDL-4.0 and ENDF/B-VII.1 evaluated libraries. Sensitivity vectors calculated by card KSEN/MCNP6.1 have been combined with the covariance matrices (from ENDF/B-VII.1, JENDL-4.0, and/or ENDL-3.2) generated by the ERRORJ module of NJOY2016 to produce nuclear data uncertainties. Comparison between JENDL-4.0 and ENDF/B-VII.1 evaluated libraries consistency among the calculated-to-experiment values of keff and the overall computational uncertainties is discussed. The keff was found to have the largest sensitivity to fission and capture cross-sections of 239Pu for the two libraries JENDL-4.0 and ENDF/B-VII.1. Neutron fission reaction of 239Pu for the two evaluated libraries was found to contribute to the major part of the keff uncertainty due to the cross-sections. The discrepancy of the calculated and measured keff values was strongly correlated with the uncertainty in the calculated keff due to the 239Pu neutron fission cross-section uncertainty.
This paper provides a comparison between multiplication factors within the UO2 fuelled cores that were measured during the original KRITZ-1 experiments performed in the early 1970s. Results were ...calculated using the MCNP6.1 code and the most recent cross section libraries, ENDF/B-VII.1 and JENDL-4.0. The results of this study reveal a good level of agreement between calculated and measured outcomes; indeed, the maximum relative error of calculated keff from experimental data does not exceed 0.5% (absolute value) for all cores and for both evaluations. The reactivity temperature coefficient, expressed as the temperature dependence of the reactivity, was examined into details using the seven factors that incorporate the keff parameter rather than just the five that exist in the literature. In order to better investigate the influence of differences in neutron cross sections we also quantified the temperature effect on the five factors of the infinite multiplication factor on each one separately by admitting a pin cell model.
Calculated thermal and fast non-leakage probabilities using JENDL-4.0 are found to overestimate the ones calculated using ENDF/B-VII.1 for all temperatures, possibly because the JENDL-4.0 library overestimates some absorptions, especially in the resonance region of fissionable nuclides. This discrepancy between the libraries decreases with increasing temperature, however, while the inverse tends to occur in the case of standard deviations. Our assessment has shown that the temperature coefficient in the thermal temperature range is linked to thermal spectrum and water density effects, while within the epithermal temperature range this is strongly dependent on the thermal shapes of the cross sections of the uranium isotopes; the resonance escape probability highlights that the error is mainly due to the Doppler broadening effect.
•Seeing that the works on KRITZ-1 are limited over all the world, an analysis of the KRITZ-1 experiments light water moderated lattices with uranium rods, at 20 °C, 90 °C, 160 °C, and 210 °C temperatures; using MCNP6.1 code and the ENDF/B-VII.1 and JENDL-4.0 libraries are made.•The reactivity decomposition to seven factors allowed us to make a detailed analyze to the reactivity temperature coefficient and to investigate better the influence of cross sections differences. A detailed analysis has made to the thermal components contribution.•The main contributions that led to the RTC are linked to the water density effect. The contribution is very weak and appears as constant with the temperature because it corresponds to the Thermal Spectral Shift effect.•An analysis has made for the calculated thermal and fast non-leakage probability.•The overall RTC value becomes slightly negative and this effect appears stronger in the temperature range from 160 °C to 210 °C.
•Evaluation of 235U elastic cross section uncertainty and covariance matrix.•ENDF/B-VII.1 and JENDL-4.0 libraries are studied using MCNP6.1 and NJOY99 codes.•Linear least squares method is applied to ...adjust the cross section and covariance.•Prior and posterior (post-adjusted) of keff are compared.
The main aim of this study is to estimate the uncertainty on the multiplication factor (keff) caused by the elastic cross-section uncertainty of the 235U nucleus, and the adjustment of this cross-section using the generalized linear-least squares method. For this purpose, the sensitivity matrix of keff with respect to 235U elastic cross section in the two nuclear evaluations ENDF/B-VII.1 and JENDL-4.0, has been assessed in several thermal, intermediate and fast benchmarks; which have been taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). The selection of these benchmarks is based on chi-square statistic test (χ2) that measures the statistical weight of deviations between calculus and experience results of the integral parameters caused by uncertainty in differential data. In this statistical test the sensitivity and covariance matrices are used. The sensitivity matrix has been calculated by the adjoint-weighted perturbation technique of the multi-purpose Monte Carlo code MCNP6.1. The 235U elastic covariance matrix has been generated in 44 neutron energy groups by the ERRORJ module of the nuclear data processing system NJOY99. The prior and posterior uncertainties on the keff, the adjusted elastic cross-section of the 235U and covariance matrix have been estimated using FORTRAN 95 code developed in our laboratory based on the generalized linear-least squares method.
The results obtained show that: the 235U elastic cross-sections taken from JENDL-4.0 requires an adjustment which can reach 6% in 0–100 keV, 4.5% in 100 keV–625 keV and 2.5% in 625 keV–0.5 MeV, while those from ENDF/B-VII.1 needs an adjustment which can reach 5% in 0–100 keV, 3% in 100 keV-625 keV and 1% in 625 keV–0.5 MeV interval. The adjusted cross-section in the two evaluations ameliorates the posterior calculated keff with respect to the experimental keff. Also the posterior covariance matrix decreases the nuclear uncertainty of the adjusted keff.
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