The characterization of experimental locations of a research nuclear reactor implies the determination of neutron and photon flux levels within, with the best achievable accuracy. In nuclear ...reactors, photon fluxes are commonly calculated by Monte Carlo simulations but rarely measured on-line. In this context, experiments were conducted with a miniature gas ionization chamber (MIC) based on miniature fission chamber mechanical parts, recently developed by the CEA (French Atomic Energy and Alternative Energies Commission) irradiated in the core of the Jožef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia. The aim of the study was to compare the measured MIC currents with calculated currents based on simulations with the MCNP6 code. A discrepancy of around 50% was observed between the measured and the calculated currents; in the latter taking into consideration only the prompt photon field. Further experimental measurements of MIC currents following reactor SCRAMs (reactor shutdown with rapid insertions of control rods) provide evidence that over 30% of the total measured signal is due to the delayed photon field, originating from fission and activation products, which are untreated in the calculations. In the comparison between the measured and calculated values, these findings imply an overall discrepancy of less than 20% of the total signal which is still unexplained.
•Among the kinetic parameters, the most important ones are βeff and Λ.•Several methods including the Rossi-α and Feynman-α techniques, slope fit and MCNPX code have been investigated.•The Monte Carlo ...MCNPX code was used to simulate a geometrical model of the TRIGA core.•The results of the methods have been validated.
In this study, noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) and dynamic method such as slope fit method have been used to calculate effective delayed neutron fraction (βeff) and neutron reproduction time (Λ) in Accelerator Driven Subcritical TRIGA reactor. The obtained results have been compared with MCNPX code results. The relative difference between MCNPX code with Feynman-α and Rossi-α techniques and slope fit method for βeff are approximately −5.4%, 1.2%, and −10.6%, −14.8%, respectively, and also for Λ is approximately 2.1%. According to results, the noise methods can been considered ideal for detection with high efficiency and zero dead time and in the slope fit method, the decay of the delayed neutrons has been neglected and only the prompt neutrons have been taken into account. In addition, quantities simulated in the current study are validated against both the reference data and the results of MCNPX code. Therefore, the purpose of this study is to simulate the commonly used experimental methods by MCNPX code and investigate the convergence as well as accuracy of the computational results for different analysis methods in calculation of the kinetic parameters in an Accelerator Driven Subcritical TRIGA reactor.
In this work we report some preliminary results regarding the analysis of electron paramagnetic resonance (EPR) response of alanine pellets and alanine pellets added with gadolinium used for ...dosimetry at the TRIGA research reactor in Mainz, Germany. Two set-ups were evaluated: irradiation inside PMMA phantom and irradiation inside boric acid phantom. We observed that the presence of Gd2O3 inside alanine pellets increases the EPR signal by a factor of 3.45 and 1.24 in case of PMMA and boric acid phantoms, respectively. We can conclude that in the case of neutron beam with a predominant thermal neutron component the addition of gadolinium oxide can significantly improve neutron sensitivity of alanine pellets. Monte Carlo (MC) simulations of both response of alanine and Gd-added alanine pellets with FLUKA code were performed and a good agreement was achieved for pure alanine dosimeters. For Gd2O3-alanine deviations between MC simulations and experimental data were observed and discussed.
•ESR analysis of alanine pellets with and without Gd irradiated at Mainz neutron beam.•Improvement of neutron sensitivity of alanine with gadolinium inside PMMA phantom.•Good agreement between ESR results and MC simulation for pure alanine pellets.
The primary aim of this work is to improve the analysis of the dynamic behaviour of the TRIGA Mark II reactor at the University of Pavia through a zero-dimensional approach. Besides the coupling ...between neutronics (point-reactor kinetics with six delayed neutron precursors group) and thermal-hydraulics (two-region model, with fuel and coolant) implemented in earlier works, the new model considers also the time behaviour of the mass flow rate due to natural circulation, of the neutron poisons and of the primary and secondary pool temperature. The system of coupled first-order differential equations is non-linear, as some state variables, such as the mass flow rate and the coolant temperature, multiply each other. The Simulink™ programming environment for dynamic analysis and control purposes is used to solve the system. A comparison with experimental data collected on-site for different reactor power transients and with measurements of the poison anti-reactivity during reactor shut-down and of the pool temperature allows the validation of the model. The model results and the experimental data reach a remarkable agreement. In addition, a linear stability analysis of the reactor is performed through the root locus and the stability map in terms of the thermal feedback coefficients. This analysis shows how the power level influences the dynamic of the system, and that, for certain values (always negative) of the fuel thermal feedback coefficient, positive values of the one for the moderator still ensures the system stability.
•Thermal hydraulic model with time evolution of poisons, mass flow rate, pool temperature.•The presented model is validated using experimental data.•Dynamic behaviour of the TRIGA reactor on the entire operative power range.•Both short-term and long-term reactor response is considered.•Stability analysis of the TRIGA reactor by means of the root locus method and stability maps.
Fifteen tile samples from archaeological digs in central Italy were analyzed for trace element content using comparative neutron activation analysis. To accommodate the simultaneous irradiation of a ...large number of samples and standard reference materials, we designed a custom sample holder for the Penn State Breazeale Reactor 2″ × 6″ tube and characterized the thermal and epithermal neutron flux across all sample locations. Analysis of sample irradiation data produced concentration values for 14 elements. Statistical analyses of the data set using agglomerative hierarchical clustering indicated that the samples are from two distinct sites, which was confirmed by provenance information received after analysis.
Neutronics analysis of TRIGA Mark II research reactor Rehman, Haseebur; Ahmad, Siraj-ul-Islam
Nuclear engineering and technology,
February 2018, 2018-02-00, 2018-02-01, 2018-02, Letnik:
50, Številka:
1
Journal Article
Recenzirano
Odprti dostop
This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark ...cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, θ) and 3-D (r, θ, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.
•Among the kinetic parameters, the most important ones are βeff and Λ.•The kinetic parameters at the end of cycle for 20days operation at 1MW reactor power have been calculated using MCNPX code.•The ...system sub-criticality levels, the required accelerator current and power have been calculated for each burnup step.•Burnup effects on neutronic parameters and accelerator requirements are analysed for a life cycle of the reactor core.
In this paper, the kinetic parameters at the end of cycle for 20days operation at 1MW reactor power have been calculated using MCNPX code for source multiplication factors (Ks) of critical and 0.97 levels. Accelerator Driven Subcritical TRIGA reactor has been considered as the case study of the problem. The system sub-criticality levels, the required accelerator current (Ip) and power (Pacc) have been calculated for each burn up step. According to the results, the required Ip and Pacc increased as system sub-criticality level decreased. Also, the results show that the effective delayed neutron fraction (βeff) does not depend on sub-criticality levels significantly but the neutron reproduction time (Λ) strongly depends on the number of fuel elements in the core and sub-criticality levels. In addition, it was observed that in comparison with the beginning-of-cycle values, at end-of-cycle, the Λ increases by 112.6% and 108.4% respectively but the βeff decreases by 6.1% and 1.8% respectively in critical and 0.97 levels.
The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the 197Au(n,γ)198Au reaction rate. The ...calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor.
•Measurements of the axial distributions of 197Au(n,γ)198Au reaction rates were made.•The axial distributions were calculated with the MCNP code.•The calculated and the measured distributions are in very good agreement.•The calculation of axial neutron flux distributions with MCNP is validated.
•We develop a detailed MCNP model of the Geological Survey TRIGA Reactor.•We present a simplified approach to considering burnup.•The model is validated against available reactor data.•We present ...evidence of inaccuracies in the ENDF B/VII zirconium libraries.
The United States Geological Survey TRIGA Reactor (GSTR) is a 1MW reactor located in Lakewood, Colorado. In support of the GSTR’s relicensing efforts, this project developed and validated a Monte Carlo N-Particle Version 5 (MCNP5) model of the GSTR reactor. The model provided estimates of the excess reactivity, power distribution and the fuel temperature, water temperature, void, and power reactivity coefficients for the current and limiting core. The MCNP5 model predicts a limiting core excess reactivity of $6.48 with a peak rod power of 22.2kW. The fuel and void reactivity coefficients for the limiting core are strongly negative, and the core water reactivity coefficient is slightly positive, consistent with other TRIGA analyses. The average fuel temperature reactivity coefficient of the full power limiting core is −0.0135$/K while the average core void coefficient is −0.069$/K from 0% to 20% void. The core water temperature reactivity coefficient is+0.012$/K.
Aim of this work is to reproduce the dynamic behavior of the TRIGA Mark II reactor of the University of Pavia on the entire operative power range (i.e. 0–250 kW) using a zero dimensional approach. In ...this work the coupling between neutronics and thermal-hydraulics in natural circulation has been considered. In specific, a point reactor kinetics model with one energy group and six delayed neutron precursors groups has been adopted while for thermal-hydraulics modeling two regions have been defined (i.e. the fuel and the coolant). The nonlinear system of coupled Ordinary Differential Equations has been solved by means of MATLAB Simulink®, which represents a reliable tool for dynamic and control analysis. The model has then been validated through the comparison with a set of experimental data collected in four different reactor power transients, obtaining a very satisfying agreement. Finally, the linear stability analysis of the TRIGA reactor has been performed by means of the root locus, finding out that the power level at which reactor is operating deeply influences the position of the poles of the transfer function between control rod height and neutron density. These results can then be employed as a reliable starting point in designing an automatic device for reactor power control.
•Dynamic behavior of the TRIGA Mark II reactor on the entire operative power range.•Stability analysis of the TRIGA reactor by means of the root locus method.•Thermal-hydraulic model with natural circulation.•Validation of the model using experimental data.
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