•Temperature monitoring was carried out in a TRIGA® reactor in the fuel, core, and pool.•The experimental results were compared with the theorical results.•The tests provided an understanding of the ...thermo-fluid dynamics of the TRIGA® reactor.•The experiments confirm the efficiency of natural circulation in removing heat.•A power measurement channel based on fuel temperature was developed.
The natural convection and forced cooling experiments were conducted to confirm the cooling capability in the IPR-R1 TRIGA® nuclear research reactor. Temperature measurements were carried out in several locations such as: at the coolant inlet and outlet in the core channels, at various pool heights, and at the centre of an instrumented fuel element. The temperature monitored by the instrumented fuel element was performed with it positioned in all the core rings and at different operating powers. A new way of measuring the power released by nuclear fissions in the core using thermal processes has been developed. To monitor temperatures and power, a Data Acquisition System (DAS) was used. The information acquired during the experiments provides an exceptional representation of the IPR-R1 TRIGA® reactor’s thermal performance. The experiments confirm the efficiency of natural circulation in removing the heat produced in the reactor core by the nuclear fission.
Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth ...of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient ΔρBU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.
Fluence profiling at JSI TRIGA reactor irradiation facility Sola, V.; Mandić, I.; Ambrožič, K. ...
Nuclear instruments & methods in physics research. Section A, Accelerators, spectrometers, detectors and associated equipment,
June 2024, Letnik:
1063
Journal Article
Recenzirano
Odprti dostop
We present an analysis of the fluence profile at the JSI TRIGA neutron reactor facility in Ljubljana. For the study, multi-pad Low-Gain Avalanche Diodes (LGADs) are used. The deactivation of acceptor ...doping in the gain layer implant due to the irradiation, typical of LGAD devices, is exploited to map the fluence profile inside the irradiation channels. The amount of active doping of the LGAD gain layer is extracted via capacitance–voltage measurements for each pad before and after irradiation to a fluence of 1.5 × 1015neq/cm2, where neq stands for 1 MeV equivalent neutron count, providing a precise and prompt measurement of the fluence distribution over the LGAD sensor. Experimental results are compared to neutron fluence expectations calculated with Monte Carlo techniques.
The Training, Research, Isotopes, General Atomics (TRIGA) research nuclear reactor is utilized in a wide range nuclear applications and has three versions, Mark I, Mark II, and Mark III. The TRIGA ...reactor that is available in Malaysia is a Mark II type, Reaktor TRIGA PUSPATI (RTP). The core power control system in the RTP is designed based on many components including the Feedback Control Algorithm (FCA), and the control rod selection algorithm (CRSA). The CRSA is used in the RTP operation to provide external control to the reactor core reactivity. Even though the CRSA provide stability to the reactor operation, it has relatively complex software structure which produce delayed response time for the control rods selection. Thus, the Single Control Absorbing Rod (SCAR) algorithm is designed to improve the reactor performance by minimizing the time response for the rods selection. However, the SCAR yields a non-constant tracking performance due to the highly non-linear dynamic of the control rods. Thus, to compensate the drawbacks of the SCAR with the existing FCA controller, a hybrid controller is proposed based on the integration of a linear model predictive control (MPC) and Proportional (P) controller named as Hybrid MPC-P-SCAR. The MPC application in the SCAR enable the controller to handle multiple constraints through built-in prediction ability to select the best sequence of future control rod velocity during transient state. Whereas at the steady-state, the MPC is optimized by a scheduled P control switches within the allowable range of power variation. The performance of the proposed Hybrid MPC-P-SCAR, the MPC-SCAR and the existing Feedback Control Algorithm with SCAR (FCA-SCAR) is compared via simulation and evaluated in terms of tracking performance, workload of the control rod drive mechanism (CRDM), and the offset elimination in the unmeasured disturbance. Overall, the Hybrid MPC-P-SCAR controller reduces the settling time by 45% and the steady-state error by 61% of the nominal value compared to the FCA-SCAR.
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•The Hybrid MPC-P core power control is developed via System ID to closely represent the actual plant dynamics.•The experimental work is presented to demonstrate the capability of the SCAR in improving the reactor control performance.•The Hybrid MPC-P is designed to handle multiple safety parameter constraints with direct switching model.•The Hybrid MPC-P can provide constant tracking performance and guarantee plant stability during operation.
In 2016, the NATO Science for Peace and Security Programme funded research project ”Engineering Silicon Carbide for Border and Port Security” — E-SiCure was launched, its objective being the ...development of radiation-hard silicon carbide (SiC) based detectors of special nuclear materials (SNM), with the aim to enhance border and port security barriers. Detector prototypes based on SiC Schottky Barrier Diodes (SBDs) and neutron converter films were developed. This paper presents the results of a dedicated experimental testing campaign performed at the Jožef Stefan Institute (JSI) TRIGA reactor in which several SiC detector prototypes equipped with 10B and 6LiF converter films were irradiated in the Dry Chamber of the reactor. The obtained results demonstrate a clearly measurable neutron response, which varies linearly with the neutron flux. The measured particle spectra from the SiC detectors exhibit a clear structure, attributable to the nature and energy of secondary particles originating as reaction products from nuclear reactions involving 10B and 6Li isotopes. The determined sensitivity of the detectors, their active volume being 1 mm × 1 mm × 25 μm, 1 mm × 1 mm × 69 μm and 1 mm × 1 mm × 170 μm, was of the order of 2 × 10−5 counts per second, per unit of neutron flux counts s−1 per n cm−2s−1 (for neutron energies between 0 and 5 eV). Scaling the detection sensitivity by a factor of 105, i.e. to an array with a surface of around 20 cm × 2 m, comparable to large BF3 or 3He detectors, would theoretically enable an overall sensitivity of around 2 counts s−1 per n cm−2s−1, which is already comparable to typical neutron sensitivity values of gas detectors, in the range from several to over 100 counts s−1 per n cm−2s−1. Due to its outstanding tolerance to harsh environments (including high temperatures and radiation fields) and superior electronic properties when compared to other semiconductors, SiC is a promising base material for the fabrication of solid-state detectors with stable and long life-time. Improvements in sensitivity combined with the capability of fabricating large modules (SiC arrays), could make SiC an important detection technology, applicable also in the context of border and port security barrier monitoring.
Education, training and isotopes production are the most important uses of the Moroccan 2 MW TRIGA Mark II reactor situated at the National Center for Energy Sciences and Nuclear Techniques (CNESTEN, ...Morocco). To develop new R&D projects in research reactors, the particular and advanced knowledge of neutron and photon flux distribution, within and around the reactor core, is crucial. In order to precisely preparing the experiments in the CNESTEN’s TRIGA reactor, a detailed model was developed using the 3D continuous energy Monte Carlo code TRIPOLI-4 and the continuous energy cross-section data from the JEFF3.1.1 nuclear data library. This new model was used to carry out preliminary neutron and photon calculations to estimate flux levels in the irradiation channels as well as to calculate kinetic parameters of the reactor, core excess reactivity, integral control rods worth and power peaking factors. As a first step of the validation of the model, the obtained results were compared with the experimental ones available in the Final Safety Analysis Report (FSAR) of the TRIGA reactor. A study is being carried out at the end of which the results will be published as an evaluated benchmark. Furthermore, this work aims at experimentally characterize the reaction rates in various irradiation channels inside and outside the reactor core. The measurements are carried out using the neutron activation technique. To set up the experimental design for the activation experiments a series of preliminary calculations were performed using the TRIPOLI-4 model to calculate the expected gamma flux/intensity levels of various materials after irradiations in different positions in the irradiation facilities. Different activation foils with known characteristics are then irradiated and the activity of several isotopes is measured with the Gamma Spectrometry Method. The measured relative reaction rates are then compared with the calculated ones evaluated through the new TRIPOLI-4 reactor model. Fairly good agreement was found, which indicates that the new computational model is accurate enough to reproduce experiments.
The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation ...of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the keff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.
The National Center for Nuclear Energy, Sciences and Technology (CNESTEN) located in Rabat, Morocco, operates a 2 MW TRIGA Mark II research reactor. This type of reactor is specially designed to ...effectively implement the various fields of nuclear research such as neutron activation analysis, neutron radiography, detectors testing, radioisotopes production as well as education and training. In the last few years, a collaboration between the French Atomic Energy and Alternative Energies Commission (CEA) and the CNESTEN was established to expand the utilization of the TRIGA computational model by carrying out new in-situ measurements in order to characterize neutron and photon fields within and beyond the TRIGA reactor core. These new measurements will consolidate the knowledge of neutron and photon fluxes in different irradiation and instrumentation channels. The results of these experiments will also be used to extend the experimental validation of the new developed TRIPOLI-4® computational model of the reactor and to quantify the uncertainties and biases. This paper focuses on the neutron and photon flux characterization of two irradiation channels inside the TRIGA reactor core. Neutron measurements are ensured by activation dosimetry whereas photon measurements are ensured by thermo-luminescent detectors (TLD400 – CaF 2 : Mn). Based on these techniques, the experiments were carried out during an experimental campaign conducted in June 2022. Preceding the implementation of these measurements, two experimental devices were specifically designed and manufactured to ensure a reproducible positioning of the detectors in the selected irradiation channels. This paper presents the experimental results analysis and the associated uncertainty quantification. These results will be then compared to the calculation ones obtained by the computational models of the TRIGA reactor.
This paper focuses on the development of a new computational model of the CNESTEN’s TRIGA Mark II researchreactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was ...developed toassess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor,control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used toaccurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivityand uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were usedto build the T4 model. In this regard, the differences between the two models are mainly due to mathematicalapproaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part dealswith the development and the validation of the T4 model. The results obtained with the T4 model were comparedto the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report(FSAR). Different core configurations were investigated via simulations to test the computational model reliabilityin predicting the physical parameters of the reactor. As a fairly good agreement among the results wasdeduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values.The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out toquantify the nuclear data uncertainty in the multiplication factor keff . For that purpose, the T4 model was used tocalculate the sensitivity profiles of the keff to the nuclear data. The integrated-sensitivities were compared to theresults obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools anddifferences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities.Moreover, the nuclear data uncertainties in the keff were derived using the COMAC-V2.1 covariance matriceslibrary and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the keff isaround 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopesto the nuclear data uncertainty in the keff is not negligible and should be taken into account when performingS/U analysis.
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