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.
The neutron field of various irradiation positions of the TRIGA Mark II reactor of the Jožef Stefan Institute (JSI) has been thoroughly characterized by neutron activation dosimetry and miniature ...fission chambers techniques. In order to have a fully validated calculation scheme to analyze and plan experiments, the gamma field also has to be experimentally validated. The 10-year long collaboration between CEA and JSI is a perfect framework to carry out such a study, and measurements of the gamma field started in late 2016. Several measurement techniques were investigated in in-core and ex-core positions. Online measurements were carried out using miniature ionization chambers (ICs) manufactured by the CEA and PTW Farmer. Positional dependence was studied, showing a decrease in the delayed gamma contribution to the total gamma flux with increasing distance from the reactor core center. To characterize the gamma dose in the core, as well as in the periphery, thermoluminescent and optically stimulated luminescent detectors (TLDs and OSLDs, respectively) were tested. These detectors are commonly used at CEA to measure the gamma dose in a given material, in order to study the nuclear heating in various core elements (control rod, baffle, structural material). Different filters were used in order to assess an integrated dose ranging from a few Gy up to several kGy. The comparisons of experimental results against calculations performed with the JSIR2S code package show a very good agreement. The feasibility of such measurements demonstrates the complementarity between measurements with dosimetry and ICs from low to very high gamma-dose environment, such as in material testing reactors.
The National Center for Energy, Sciences and Nuclear Techniques (CNESTEN)'s Training Research and Isotope Production General Atomics (TRIGA) Mark II is a pool-type light water moderated and cooled ...research reactor operating at a maximum steady state thermal power of 2 MW. The reactor was designed to be used as a training facility for reactor operators, neutron activation analysis, isotopes production, and for implementing different reactor physics experiments. This article deals with the numerical and experimental characterization of reaction rates (RRs) in different irradiation channels inside the CNESTEN's TRIGA Mark II research reactor, located in Rabat/Morocco. The main objective of this study is to validate the calculated neutron RRs against the measured ones and to prove that the new TRIPOLI-4 model of the reactor is capable to reproduce the measured quantities. Therefore, the measurements were carried out using the neutron activation technique and gamma spectrometry measurements. Preliminary simulations were performed with TRIPOLI-4 transport Monte Carlo code to establish the experimental design and set up for activation foils experiments. The selected set of foils with known characteristics were irradiated, at different power levels, inside the irradiation facilities of the TRIGA reactor. The resulting activities were evaluated via <inline-formula> <tex-math notation="LaTeX">\gamma </tex-math></inline-formula> spectrometry measurements. Normalized calculated and measured RRs were compared, and a good agreement was shown for most nuclides, which indicates that the new detailed TRIPOLI-4 model of the TRIGA reactor can accurately predict the relative experimental RRs values. Further work is ongoing to analyze absolute RR values, as well as to carry measurements in other irradiation channels.
Well characterized neutron and gamma fields inside a nuclear reactor are of key importance for its safe operation and for successful utilization of various research reactor irradiation facilities. In ...case of high-flux research reactors such as BR2 in Belgium, Maria in Poland and the future Jules Horowitz Reactor in France, the gamma energy deposition rate in reactor structural components and irradiated samples reaches values well over 10W/g. To assure safe reactor operation, the gamma field and associated heating must therefore be thoroughly characterized in order to provide adequate component and sample cooling.
The gamma field can be divided into two contributions: prompt gamma rays are emitted almost instantly after neutron interaction with nuclei, while the delayed gamma rays are emitted from nuclei, which become radioactive by neutron absorption, generated from fission and other processes. Most modern Monte-Carlo particle transport codes enable the transport of prompt gamma rays; a few support delayed gamma ray generation and transport as well. The latter have mostly been applied to fusion devices, where detailed shutdown dose-rate measurements have been performed. Although the delayed gamma field can also be simulated in fission devices, significant inaccuracy in the result is to be expected due to the computational complexity arising from the large number of radioactive fission products and incompleteness of nuclear data. Furthermore, the unavailability of experimental delayed gamma measurements in fission systems presents an important challenge for the validation of the experimental results.
Previous measurements in several research reactors show that the delayed gamma flux amounts to around 30% of the total gamma flux. However, these evaluations were performed with measurement data obtained during rapid reactor shutdowns (SCRAMs), using a single measurement point per SCRAM.
In this paper we propose a new technique to accurately determine the magnitude of the delayed gamma component and its time evolution, based on synchronous acquisition of fission and ionization chamber signals. The measurements were performed at the JSI TRIGA reactor, using fission and ionization chambers placed in several in-core measurement positions. Their signal was acquired synchronously and at the highest possible acquisition rate in order to distinguish between measurement noise and reactor transients. Using the novel delayed gamma extraction technique we were able to estimate the magnitude of the delayed gamma contribution to be: 18.9% ± 2.0% at the reactor core periphery, linearly increasing towards the reactor core center to 31.4% ± 2.8% of the total measured gamma flux signal after 10min of reactor operation.
In this paper, test measurements of a fibered optically stimulated luminescent dosimeter (OSLD) system performed during a dedicated experimental phase in EOLE zeropower reactor are presented. The ...measurement setup consists of an OSLD crystal connected onto the extremity of an optical fiber and a laser stimulation system, manufactured by the CEA/LIST, Saclay, France. The OSL sensor is remotely stimulated via an optical fiber using a diode-pumped solid-state laser. The OSL light is collected and guided back along the same fiber to a photomultiplier tube through appropriate optical filters. OSL signals are corrected for background noise and integrated to provide the absorbed doses thanks to calibration coefficients. The light stimulation also fully bleaches the crystal (reusable for the next measurement). Results obtained using this system are compared to usual gamma-heating measurement protocol using OSLD pellets. Both measurement processes were calibrated to ensure the comparison of absolute doses in gray at the same positions in the core. The presence of induced radioluminescence in the OSLD during the irradiation was also observed and could be used to monitor the gamma flux. The feasibility of remote measurements is achieved, whereas further developments could be conducted to improve this technique since the readout procedure still requires to withdraw the OSLD off the gamma flux (hence from the core) on account of the dose rate (around a few Gy·h -1 ), and the readout time remains quite long for online applications. Several improvements are foreseen and will be tested in the forthcoming years.
The PETALE experimental program in the CROCUS reactor intends to provide integral measurements on reactivity worth and dosimetry measurement to constrain nuclear data relative to stainless steel ...heavy reflectors. The experimental setup consists in eight successive plates of pure iron, pure nickel, pure chromium, or nuclear-grade stainless steel set at the close periphery of the core. The plates are interleaved with up to nine dosimeters that consist of thin activation foils with different possible materials to be sensitive to different ranges of the neutron spectrum. A precise measurement with a good estimation of the uncertainties and correlations is required, especially when comparing reaction rates, e.g. transmission measurement and/or spectral indices.
The present work focuses on the validation of the dosimetry technics developed in preparation of this experimental program. Different aspects are discussed: monitors, efficiency calibration, self-absorption correction, self-shielding and nuclear data uncertainties. The different sources of uncertainties for the experiment-calculation comparisons are characterized, taking into account all the correlation between the different dosimeters. These correlations are a mandatory element for the aimed Bayesian assimilation in order to avoir overfitting when considering dosimeter providing a similar information.
As part of the design studies of future power and research nuclear reactors, the validation of neutron and photon calculation schemes requires the implementation of nuclear heating measurements. Such ...measurements are usually performed in very low-power reactors, whose core dimensions are accurately known and where irradiation conditions are entirely controlled. The use of thermoluminescent dosimeters is particularly well suited to gamma heating measurements, although the TLDs are also sensitive to neutrons depending on their composition. Hence, for gamma heating assessment in a mixed gamma-neutron field, the neutron sensitivity correction factors of TLDs has to be accurately determined to ensure a better interpretation of doses measured in a reactor with reduced uncertainties. This paper presents the CANDELLE experiment, which is dedicated to the implementation of a pure neutron field calibration of lithium-enriched TLDs, by using the Platform for Studies and Research on Nuclear Energy platform of LPSC. This calibration step is part of a promising approach that has been recently developed at INFN for medical purposes, aiming at determining the gamma and neutron contributions to the doses measured by LiF TLDs in the mixed field, from the peak height ratios of the gamma and neutron calibration glow curves of this type of TLDs. The test irradiations conducted at LPSC within the first phase of the CANDELLE project allowed retrieving the peak height of LiF TLDs in fast neutron field, hence providing an early assessment of the neutron components of doses measured in the EOLE reactor at CEA Cadarache with around 10% uncertainty at 1σ. From these preliminary results, the second phase of the project will be devoted, on the one hand, to the use of moderator screens aiming at achieving a thermal neutron field while avoiding a too strong attenuation of the flux and on the other hand, to the determination of the neutron spectrum at different locations within the irradiation area.
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.
The present work details a new benchmark to be produced to the International Community, for dealing with neutronics code validation in the frame of the analysis of severe accidents in Sodium Fast ...Reactors (SFRs) leading to core degradation and material relocation. The benchmark is based on analysis of selected experiments performed at the Schnelle Null-Energie-Anordnung Karlsruhe (SNEAK). The SNEAK-12A experiments have been considered where core distorted configurations have been simulated. In the paper, the experimental results have been analyzed by using TRIPOLI-4, MCNPX2.7, and Serpent-2 Monte Carlo codes as reference tools, and the ERANOS system of codes for deterministic calculations, all based on JEFF-3.1.1 nuclear data libraries. The complete material balance, and the main degradation sequences are provided within the manuscript. Preliminary results on available experimental results (keff) are given, and additional local quantities are calculated, such as axial flux distributions, as well as detector responses as function of the distance to the degraded part. The benchmark offers an excellent opportunity to validate calculation schemes for strongly heterogeneous interfaces, in particular the preparation of homogenized self-shielded neutron cross-sections for deterministic core calculations, as well as leakage treatment in locally very heterogeneous media. This work is done within the frame of new core design capacities and new ways of conducting experiment in Zero Power Reactors, such as the ZEPHYR project led independently by CEA. The present analysis will be completed by a full nuclear data sensitivity and uncertainty analysis of the reactivity coefficients and keff in a companion paper.
In this paper are presented test measurements of a fibered-OSLD system performed during a dedicated experimental phase in EOLE zero-power reactor. The measurement setup consists of an OSLD crystal ...connected onto the extremity of an optical fiber and a laser stimulation system, manufactured by the CEA/LIST in Saclay. The OSL sensor is remotely stimulated via an optical fiber using a diode-pumped solid-state laser. The OSL light is collected and guided back along the same fiber to a photomultiplier tube. Results obtained using this system are compared to usual gamma heating measurement protocol using OSLD pellets. The presence of induced radio-luminescence in the OSLD during the irradiation was also observed and could be used to monitor the gamma flux.
The feasibility of remote measurements is achieved, whereas further developments could be conducted to improve this technique since the readout procedure still requires to withdraw the OSLD off the gamma flux (hence from the core) on account of the dose rate (around a few Gy.h
-1
), and the readout time remains quite long for on-line applications. Several improvements are foreseen, and will be tested in the forthcoming years.
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