This paper presents the preliminary characterization of a new reduced-height CALORRE differential calorimeter designed by Aix Marseille University and fabricated in order to be tested and qualified under irradiation in the MITR experimental reactor within the framework of the CALOR-I program. The paper begins by focusing on the preparation of the irradiation campaign, providing a concise description of the MITR core, the water loop facility, and the newly fabricated differential calorimeter. Then, this work summarizes neutronic calculations performed for the CALORRE differential calorimeter to provide data for the irradiation campaign. MCNP calculation code was used to predict the neutron and gamma flux spectra, and total nuclear heating rate to the components of the calorimeter in a variety of positions. The MCNP model parameters and methodology are detailed and results are used to perform 3-D thermal modelling. Estimation of the response of the calorimeter calculated under real conditions considering local heat sources determined by the NRL of the MIT using MCNP code is given. Next, the paper presents the experimental characterization conducted under laboratory conditions. This section includes a detailed presentation of the updated experimental set-up and the key metrological characteristics of the sensor response obtained from the experimental results. In conclusion, the paper offers some final remarks and prospects to realize the irradiation campaign successfully.

In recent years, radiation mapping has attracted widespread research attention and increased public concerns on environmental monitoring. In terms of both materials and their configurations, radiation detectors have been developed to locate the directions and positions of the radiation sources. In this process, algorithm is essential in converting detector signals to radiation source information. However, due to the complex mechanisms of radiation-matter interaction and the current limitation of data collection, high-performance, low-cost radiation mapping is still challenging. Here we present a computational framework using Tetris-inspired detector pixels and machine learning for radiation mapping. Using inter-pixel padding to increase the contrast between pixels and neural network to analyze the detector readings, a detector with as few as four pixels can achieve high-resolution directional mapping. By further imposing Maximum a Posteriori (MAP) with a moving detector, further radiation position localization is achieved. Non-square, Tetris-shaped detector can further improve performance beyond the conventional grid-shaped detector. Our framework offers a new avenue for high quality radiation mapping with least number of detector pixels possible, and is anticipated to be capable to deploy for real-world radiation detection with moderate validation.

This work summarizes neutronic calculations performed for the CALORRE differential calorimeter specifically designed by Aix Marseille University to inform the irradiation campaign planning for testing in MITR within the framework of the CALOR-I research program. MCNP software was used to predict the neutron and gamma flux spectrum, and total nuclear heating rate to the components of the calorimeter in a variety of positions. Results were compared to evaluate spatial bias, core loading effects, optimization for axial position, and provide input data for thermal multi-physics modelling.

This paper presents the preliminary characterization of a new reduced-height CALORRE differential calorimeter designed and fabricated for the irradiation in the MITR within the framework of the CALOR-I program. The paper begins by focusing on the preparation of the irradiation campaign, providing a brief and concise description of the MITR core and the newly fabricated differential calorimeter assembly. Next, the paper presents the preliminary experimental characterization conducted under laboratory conditions. This section includes a detailed presentation of the updated experimental set-up and the key metrological characteristics of the sensor response obtained from the first experimental results. Furthermore, the paper estimates and highlights the responses of the calorimeter calculated by means of a 3-D numerical thermal model under real conditions considering local heat sources determined by the NRL of the MIT using MCNP code. In conclusion, the paper offers some final remarks and prospects to realize the irradiation campaign successfully.

This article reviews the work to date on the CALOrimeter with Radial thermal transfers for nuclear REactors (CALORRE) differential calorimeter patented by Aix-Marseille University (AMU) and the French Alternative Energies and Atomic Energy Commission (CEA) in 2015. The article presents the results obtained with the first prototype of the CALORRE calorimeter qualified under real conditions during an irradiation campaign in the MARIA reactor in 2015, including previously unpublished details. Then, studies of different CALORRE calorimetric cells characterized by experiments under laboratory conditions are described. Several configurations were studied to determine the influence of cell height, horizontal fin geometry, and structural material composition on calorimeter response. These calculations provide for a calibration protocol by generating a heat source inside each cell, with evaluation of linearity, sensitivity, range, reproducibility, response time, and absolute temperatures. Finally, within the framework of a new research program called Compact-CALORimeter Irradiations inside the MIT research reactor (CALOR-I) and financed by AMU Foundation (A*Midex), a design optimization of the calorimeter assembly was carried out in order to remove contact thermal resistances and provide a new very compact CALORRE calorimeter suited for the in-core water loop of the Massachusetts Institute of Technology (MIT) reactor (2 $\text{W}\cdot \text{g}^{-1}$ peak nuclear heating rate). The response of this new very compact calorimeter is estimated using 3-D numerical thermal simulations under real conditions.

This paper deals with the CALORRE differential calorimeter patented by Aix-Marseille University and the CEA in 2015. Firstly, the paper focuses on the presentation of the first prototype of CALORRE calorimeter qualified under real conditions during the MARIA irradiation campaign in 2015. Then, a review of the studies restricted to one CALORRE calorimetric cell realized thanks to experimental characterizations under laboratory conditions is detailed. Several configurations were studied to determine the influence of the cell height, its horizontal fin geometry and the nature of the material of its structure on its response for a calibration protocol: linearity, sensitivity, range, reproducibility, response time and absolute temperatures. Finally, within the framework of the new CALORI project, an optimization of the calorimeter assembly and its design were carried out in order to remove contact thermal resistances and provide a new configuration of CALORRE calorimeter suited for the in-core water loop of the MIT reactor (2 W.g-1). The response of this new calorimeter is estimated thanks to thermal simulations.

This paper presents an experimental study of the calibration of a non-adiabatic single-cell calorimeter called KAROLINA. This type of sensor is used to quantify the nuclear heating rate inside MTRs (Material Testing Reactors). The calibration of this sensor type is carried out in laboratory conditions without irradiation. The calibration corresponds to a preliminary step, which is crucial to use this sensor in real conditions. In the case of the studied single-cell calorimeter owning no heating element, the calibration is realized with a specific test bench, called BERTAN, based on thermal transient regimes. First of all, the paper describes the methodology of the determination of the calorimeter sensitivity and the dedicated experimental set-up developed recently. Then the paper is devoted to a parametric study carried out on the KAROLINA single-cell calorimeter to determine the influence of several parameters on its thermal time constant (temperature, velocity, phase, method).

This paper gives a short review of sensors dedicated to measuring nuclear heating rate inside fission reactors in France and USA and especially inside Material Testing Reactors. These sensors correspond to heat flow calorimeters composed of a single calorimetric cell or of two calorimetric cells at least with a reference cell to obtain a differential calorimeter. The aim of this paper is to present the common running principle of these sensors and their own special characteristics through their design, calibration methods, and in-pile measurement techniques, and to describe multi-sensor probes including calorimeters.

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or similar to 5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup. (C) 2017 Elsevier B.V. All rights reserved.

The MIT Nuclear Reactor Laboratory (NRL) has irradiated lithium-beryllium fluoride (flibe) salt as part of an on-going U.S. Department of Energy-funded Integrated Research Project to develop a Fluoride Salt High-Temperature Reactor (FHR). As part of this project, the NRL has carried out two irradiations of FHR materials in static flibe at 700 degrees C in the MIT Research Reactor. These irradiations marked the start of a program evaluating the tritium production and release from the fluoride salt system at high temperature; in particular, there is interest in the evolution of tritium from the salt into solid materials and cover gasses. This paper describes the experience gained from the irradiation of flibe with respect to the detection of tritium. It covers the development of techniques for monitoring the evolution of tritium from the salt during irradiation and the factors particular to the FHR system that influence this process, including the radiolytic production and release of volatile fluorine and fluoride products as a function of temperature. In addition, it discusses the measurement of tritium partitioning between the different materials in the experiment due to the confluence of diffusion, adsorption, and chemical and radiolytic reactions.

Herein we report on the characterization of defects formed in polycrystalline Ti3SiC2 and Ti2AlC samples exposed to neutron irradiation up to 0.1 displacements per atom (dpa) at 350 +/- 40 degrees C or 695 +/- 25 degrees C, and up to 0.4 dpa at 350 +/- 40 degrees C. Black spots are observed in both Ti3SiC2 and Ti2AlC after irradiation to both 0.1 and 0.4 dpa at 350 degrees C. After irradiation to 0.1 dpa at 695 degrees C, small basal dislocation loops, with a Burgers vector of b = 1/2 [0001] are observed in both materials. At 9 +/- 3 and 10 +/- 5 nm, the loop diameters in the Ti3SiC2 and Ti2AlC samples, respectively, were comparable. At 1 x 10(23) loops/m(3), the dislocation loop density in Ti2AlC was approximate to 1.5 orders of magnitude greater than in Ti3SiC2, at 3 x 10(21) loops/m(3). After irradiation at 350 degrees C, extensive microcracking was observed in Ti2AlC, but not in Ti3SiC2. The room temperature electrical resistivities increased as a function of neutron dose for all samples tested, and appear to saturate in the case of Ti3SiC2. The MAX phases are unequivocally more neutron radiation tolerant than the impurity phases TiC and Al2O3. Based on these results, Ti3SiC2 appears to be a more promising MAX phase candidate for high temperature nuclear applications than Ti2AlC. (C) 2015 Elsevier B.V. All rights reserved.

We use research reactors first for fuel qualification. Even though this was only one of the items in the table presented by Dr. Ridikas from IAEA, for us it is an important one. We perform load following on the French nuclear fleet and we have a special regulation for pellet cladding interaction (PCI), which we can only fully be qualified in research reactors. We also load MOX fuel in our 24 900 MW units, which are licensed for MOX and we like to be able to perform special measurements for such fuels, even though in an international programme UOX fuel will be the primary interest for most of the other industrial partners. Of course, we also use research reactors for materials irradiation and qualification. It is no surprise that we spoke at the conference with AREVA about the non-occurrence of flakes in our reactor pressure vessels and, therefore, of the absence of embrittlement problems that potentially arise from such hydrogen flakes. We also gave details about ultrasonic measurement techniques and hydrogen degassing techniques in French steel warehouses. If you follow French media, steel fracture toughness in all kinds of forged steels has become a heated topic. I am very glad that I came to Bucharest. I could only fully follow 2 sessions, one chaired by Dr Gerstenberg from Munchen, if I remember correctly, and the other by Edgar Koonen, the Program Committee Chairman on International Issues. They were very good sessions, and I learned many things, for which I thank the speakers and the organizers. Research reactors in Europe are important for the nuclear scientific community, and also for the larger industrial community. This is why we believe that your community should be helped and supported.

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is n...

The Massachusetts Institute of Technology Reactor (MITR) is a 6 MW tank-type reactor with a compact core design. It is moderated and cooled by light-water and has a heavy-water reflector. The MITR is a unique test bed for a wide range of nuclear material and fuel research. In one of the recent fuel cycles, detectors found elevated neutron levels on the reactor top lid during the operation with the Water-loop facility installed. A preliminary investigation theorized that the neutrons were streaming out via a thin CO2 gas gap, which is used as a thermal isolation space for the in-core experiment. The current study employs neutron transport simulations to investigate the physical interpretation of the observed phenomenon. The state-of-the-art general purpose Monte Carlo code MCNP5 is adopted as the computational tool. The calculation results identify the typical characteristics of the streaming effect by quantifying the fraction of upward-streaming neutrons. Efforts have also been made to explore feasible engineering options for mitigating this neutron streaming from the Water-loop facility. It is concluded that a borated guide tube is not sufficient to reduce the neutron level on the reactor top lid; however, a stepped thimble design is effective in reducing the neutron source contribution.