Low-enriched (LEU) U-ZrH fuel, with a
235
U content less than 20% of the total uranium, is being evaluated for possible use in different types of reactors, including space nuclear systems, light ...water reactors (LWRs) and micro-reactors. As a result, it is beneficial to better understand the macrostructural and microstructural changes that occur in this fuel during irradiation. This paper reports the results of the post irradiation examination of an LEU U-ZrH fuel element (30 wt.% U, <20%
235
U) using neutron radiography, precision gamma scanning, chemical analysis, optical metallography and scanning electron microscopy combined with energy dispersive spectroscopy and wavelength dispersive spectroscopy, where the fuel element was irradiated in a Training, Research, Isotope, General Atomics (TRIGA) reactor. Results of microstructural characterization indicated some dehydriding and cracking of the U-ZrH fuel occurred during irradiation; an axial and radial burnup gradient existed in the fuel during irradiation, as measured by gamma scanning and chemical analysis; negligible microstructural changes transpired during irradiation, based on comparison of irradiated and as-fabricated U-ZrH fuel microstructures; and, negligible, fission product-rich, phases could be resolved in a U-ZrH fuel that was irradiated to a calculated 20% depletion of
235
U.
The Washington State University TRIGA reactor was modeled in the SCALE system’s KENO neutron transport code. The model includes 119 fuel rods, a central boral control rod, 4 control blades, 5 ...irradiation ports, and 20 graphite reflector blocks within a rectangular aluminum grid box. The model accurately simulated the irradiation of depleted uranium and highly enriched uranium foils and predicted reaction rate ratios for
238
U(n,2n)
237
U /
238
U(n,
γ
)
239
U and
99
Mo / fission, which compared favorably to empirical values previously measured at Los Alamos. Finally, a sensitivity/uncertainty analysis was performed using the continuous-energy TSUNAMI-3D module of SCALE.
•An improved TRIGA research reactor thermal power calibration method with electrical heaters is presented.•The heat capacity of the reactor pool is determined with various methods.•The heat losses ...from the reactor tank are assessed experimentally and computationally.•The total experimental uncertainty of the calibration method is evaluated.
A fission rate profile benchmark experiment has been performed at the Jožef Stefan Institute TRIGA Mark II reactor. The measurements were made using absolutely calibrated miniature fission chambers developed and manufactured by the Commissariat à l’Énergie Atomique et aux Énergies Alternatives. The aim of the paper is to describe the experimental set-up, fission rate measurements and to present the detailed Monte Carlo computational model of the TRIGA reactor, which was constructed with as used to compute absolute fission rate distributions in the core at a fixed control rod position, taking into account the detailed description of the experimental configuration. The paper focuses on the extensive evaluation of experimental and calculational uncertainties and biases following the International Reactor Physics Experiment Evaluation Project methodology. A comparison between the measured and computed absolute reaction rates concludes the paper, with the agreement being within one sigma standard uncertainty.
•A 3D/0D domain coupling tool was implemented and used to simulate the overall coolant circuit of the TRIGA MARK I IPRR1 openpool reactor.•The tool aim to reduce the computational cost in large ...thermal hydraulic systems.•Start-up, steady state and accidental scenarios were assessed in the TRIGA MARK I IPR-R1 reactor.•The model results provided valuable information from the reactor’s safety point of view.
The computational simulation of large-scale reactors is currently limited by the high computational cost. The system codes allow addressing these problems, although with the well-known loss of local information. The use of coupling domains to reduce the problems looks like a proper alternative to settle this issue. In the present paper, a multi-domain coupling 3-dimensional/0-dimensional method to solve the thermal hydraulics of the TRIGA Mark I IPR-R1 reactor was implemented into a Finite Volume suite. Despite of the broadly literature about coupling methods, even in the nuclear engineering community, most of them manage with different codes in a fully explicit way. In the other hand, the benefit of solve different domain approaches inside the same software is in the use of monolithic algorithms. The proposed method consists on using 3-dimensional full CFD to simulate the reactor pool and 0-dimensional modelling for the external cooling loop. This is made by implementing a set of ad hoc dynamics boundary conditions to model the momentum and energy balances along the pipeline. This strategy was used to perform long-time steady state simulations of the reactor at the design power of 100 kW as well as for the repowering up to 265 kW. The results demonstrated that the core is efficiently cooled at the higher power without need to increase the coolant mass flow rate of the external system. Moreover, two accidental events were simulated: the first case was the Station Black Out at full power of 265 kW. The results indicated that the loss of the external heat sink led to a slow pool heating, but the core remains being cooled by the natural circulation in the pool. In fact, the mass flow rate through the core is only reduced in 15% by the loss of the external loop circulation. Finally, a large-Loss of Coolant Accident for the operational power of 100 kW and keeping the pump running is performed. In this case, the pool is quickly empty if safety systems do not take action and the core is uncovered after 450 s completely losing the core cooling capacity.
Since radioactive waste inventory plays an important role in decommissioning scenario development, we developed an estimation method of the amount of radioactive wastes for a research reactor as a ...case study of the 100 KWth KARTINI reactor. In order to address geometrical complexity of the research reactor with experimental equipment such as beam ports, we calculated the distribution of neutron flux by using Monte Carlo particle transport code, and the radionuclide inventories generated by neutron capture in the reactor components and structures. The amount of radioactive wastes arising from the decommissioning was estimated by assigning radioactive parts according to the clearance criterion of each radionuclide. In addition, the size of a segment of the dismantled biological shielding was considered to estimate the amount of radioactive wastes addressing the actual dismantling activity. It is also found that the amount of radioactive wastes increases with the size of the segment because it includes partially clearance materials.
•To evaluate the TRIGA kinetics parameters.•To model the TRIGA Mark-II reactor using the SRAC2006 code system.•To compare the calculated results with SAR as well as MCNP5 values.•To demonstrate a ...good agreement with the SAR and the MCNP results.•To improve reactor kinetic parameters data for safe operation of TRIGA reactor.
The aim of this study is to evaluate the kinetic parameters of 3MW TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety. The most important kinetic parameters of nuclear reactors are the effective delayed neutron fraction (βeff), the effective decay constant for ith family of delayed neutron precursor (λeff,i), the prompt neutron lifetime (lp) and the mean neutron generation time (Λ). These parameters are calculated using the 3-D diffusion code SRAC-CITATION of the comprehensive neutronics calculation code system SRAC2006 based on the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 in both cases. The calculated results of reactor kinetic parameters are compared to the available safety analysis report (SAR) values of 3MW TRIGA Mark-II reactor by General Atomic as well as the MCNP5 values (numerically benchmark) based on the evaluated nuclear data library ENDL/B-VII.0. It was found that in most cases, the calculated results of kinetic parameters demonstrate a good agreement between the JENDL-3.3 and the ENDF/B-VII libraries as well as the SAR and the MCNP5 values respectively. Therefore, this study will be essential to improve the basic nuclear data of reactor kinetic parameters for safe operation of 3MW TRIGA Mark-II research reactor.
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.
•Different measures for homogeniety of 2D/3D field are described.•The measures were applied for calculation of gamma fields around TRIGA fuel elements.•There is no absolute or best measure of field homogeniety.•Due to simplicity the ratio between the maximum and mean is a very practical measure.
Control rod calibration experiment results for the Oregon State TRIGA® Reactor (OSTR) immediately following LEU conversion in 2008, and MCNP® 5 predicted rod worths from the 2008 LEU Conversion ...Safety Analysis Report (CSAR) are discussed. The reactivity worth of the four OSTR control rods is measured using the rod-pull method. Reactor power and period measurements in this method rely on the fission chamber power detector on the north side of the reflector. It is proposed that the location of the fission chamber and the neutron flux distribution in the core may result in an inaccurate reactor period measurement due to the asymmetry of the neutron flux distribution in the OSTR core. The asymmetry of the flux is believed to be more pronounced during super-criticality, resulting in errors in the time-of-power-rise measurements. As a result, control rod calibration experiments may under-predict or over-predict the reactivity worth of certain control rods. A time-independent Monte–Carlo method for the quantification of these effects is presented. Thermal flux maps at the core axial mid-plane are obtained from the model to inform discrepancies between predicted and observed results.
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.
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