The MInes NEutron Radiography facility (MINER facility) installed at the United States Geological Survey TRIGA Reactor provides new capabilities for both researchers and students at the Colorado ...School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a microchannel plate imaging detector, and the associated electronics.
Radiographs of a sensitivity indicator-a resolution indicator developed by the American Society for Testing and Materials-taken using both the digital detector and the transfer method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a comparison between digital and film radiography at the new facility. The calibration pins contain simulated defects of known dimensions, including pellet-clad gaps, gaps between pellets, and central voids within the pellets. Comparison of the radiographs taken by the two methods reveals that the digital detector does not produce high-quality images when compared to film radiography. Additionally, there are a number of artifacts in the digital images produced by the image acquisition system. The quality of the film images demonstrates that the problems with the digital images are a product of the digital imaging system and not the neutron beam.
In 2016, the ”E-SiCure” project (standing for Engineering Silicon Carbide for Border and Port Security), funded by the NATO Science for Peace and Security Programme, was launched. The main objective ...is to combine theoretical, experimental and applied research towards the development of radiation-hard SiC-based detectors of special nuclear materials (SNM), and by that way, to enhance border and port security barriers. Along the plan, material modification processes are employed firstly to study, and secondly to manipulate the most severe electrically active defects (which trap or annihilate free charge carriers), by specific ion implantation and defect engineering. This paper gives an overview of the experimental activities performed at the JSI TRIGA reactor in the framework of the E-SiCure project. Initial activities were aimed at obtaining information on the radiation hardness of SiC and at the study of the energy levels of the defects induced by neutron irradiation. Several Schottky barrier diodes were fabricated out of nitrogen-doped epitaxial grown 4H-SiC, and irradiated under Cd filters in the PT irradiation channel in the JSI TRIGA reactor with varying neutron fluence levels. Neutron-induced defects in the material were studied using temperature dependent current-voltage (I-V), capacitance-voltage (C-V) and Deep-Level Transient Spectroscopy (DLTS) measurements. Our prototype neutron detectors are configured as 4H-SiC-based Schottky barrier diodes for detection of secondary charged particles (tritons, alphas and lithium atoms) which are result of thermal neutron conversion process in
10
B and
6
LiF layers above the surface of the 4H-SiC diodes. For field testing of neutron detectors using a broad beam of reactor neutrons we designed a standalone prototype detection system consisting of a preamplifier, shaping amplifier and a multichannel analyser operated by a laptop computer. The reverse bias for the detector diode and the power to electronic system are provided by a standalone battery-powered voltage source. The detector functionality was established through measurements using an
241
Am alpha particle source. Two dedicated experimental campaigns were performed at the JSI TRIGA reactor. The registered pulse height spectra from the detectors, using both
10
B and
6
LiF neutron converting layers, clearly demonstrated the neutron detection abilities of the SiC detector prototypes.
•One of the key components in ADSR is spallation target.•The spallation source parameters for beam spatial distributions have been investigated.•The Monte Carlo MCNPX code was used to simulate a ...model of the TRIGA core.•The results of the methods have been validated.
In this paper, effect of the spallation target neutronic parameters such as beam profile, proton beam energy (Ep), target thicknesses and for different materials in an Accelerator Driven Subcritical TRIGA reactor has been investigated. Monte Carlo code MCNPX has been used to calculate neutronic parameters such as: spallation neutron yield (Yn/p), neutron spectrum, angular distribution of the spallation neutrons coming out of the target and spallation neutron production spectra in the target region for two cases of uniform and parabolic beam spatial distributions. According to the results, the relative difference of spallation neutron yield increases by 3.45% using a parabolic spatial distribution instead of a uniform spatial distribution and also this parameter increases by 7.91% using a thick target instead of a thin target, respectively. Moreover, with increasing Ep from 115MeV up to 2GeV, the relative difference of spallation neutron yield improves by 4813.58%. Additionally, in comparison to uniform spatial distribution of the proton beam, having parabolic spatial distribution has the advantage of more homogeneity in the spatial distribution of generated spallation neutrons as well as heat generation in the target. Also, Yn/p/Ep parameter increases with target atomic number and proton beam energy. Therefore, our results are indicative of the fact that investigating sensitivity of the target spallation neutronic parameters is necessary in order to optimally design a cost-efficient Accelerator Driven Subcritical Reactors (ADSRs).
The fleet of research and training reactors is aging, and no new research reactors are planned in the United States; thus, there is a need to expand the capabilities of existing reactors to meet ...users' needs. To address these needs, the Colorado School of Mines added a neutron beamline facility to the U.S. Geological Survey TRIGA Reactor (GSTR), a 1-MW(thermal) Mark-I TRIGA reactor located at the Denver Federal Center in Lakewood, Colorado. The original GSTR design did not include any beam ports, and future research efforts will benefit from a neutron beam at the GSTR. Adding new beamline facilities to existing research reactors is both rare and challenging, and this paper describes the design and installation of a new neutron beamline facility at a Mark-I TRIGA reactor with no existing beamline facilities. The design and construction of a radiation beamstop for the new beamline is described in detail. A neutronics model of the neutron beam provides researchers with a useful tool for experiment design. The new neutron beam has a measured length-to-diameter ratio of 200 ± 10, a neutron flux of 2.2×10
6
± 6.4×10
5
n/cm
2
-s, and an average cadmium ratio of 7.4 using copper, gold, manganese, and indium foils.
Using the Westcott convention, the Westcott flux,
φ
w
; modified spectral index,
r
T
n
/
T
0
; neutron temperature,
T
n
; and gold-based cadmium ratios were determined for various sampling positions ...in the U.S. Geological Survey (USGS) Training, Research, Isotopes, General Atomic (TRIGA) Mark I reactor. Westcott parameters were determined by a bare multi-monitor method. Thermal-neutron temperature measurements were made using lutetium foils. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. Measurement of the neutron spectrum has resulted in a better knowledge of the reactor core and will improve predictive radioisotope production calculations necessary for neutron activation analysis and medical isotope production.
•Accuracy assessment of a new home developed burnup code called BUCAL1.•BUCAL1 has the capability of using several depletion calculations.•MCNPX2.7 and BUCAL1 code to code verification through burnup ...analysis.•The 2MW TRIGA MARK II Moroccan research reactor is used as benchmark.•Comparison shows that BUCAL1 is enough accurate to be used in burnup analysis.
The availability of accurate burnup data is an essential first step in any systematic approach to enhancement of economics, safety and performance of a research reactor. This first step requires the utilization of a well verified burnup code system. In this work a newly home-developed burnup code called BUCAL1 is presented. The code provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP (version 5c). BUCAL1 has the capability of using several depletion calculation schemes that do not exist in several other burnup code systems such as: shuffling, refueling and multicycles burnup calculation, in an automatic way.
The accuracy and precision of BUCAL1 were tested for U-Zrh fuels, by a code to code verification with MCNPX2.7, by predicting the burnup parameters of the 2MW TRIGA Mark II Moroccan research reactor. Continuous energy cross section data from the more recent nuclear data evaluation ENDF/B-VII.0 as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. Analysis of the verification results shows that BUCAL1 is enough accurate to be used in burnup calculations.
•We proposed a methodology to calculate importance weighted kinetic parameters.•Kinetic parameters of the TRIGA reactor have been calculated using MCNIC method.•The results of the method have been ...validated.•Parameter Λ is highly sensitive to the reactor geometry and reflector thickness.•MCNIC method can efficiently be used to calculate the kinetic parameters in ADSRs.
The effective delayed neutron fraction (βeff) and neutron reproduction time (Λ) play an important role in nuclear reactor control and safety analysis. In this work, calculation of these parameters in an Accelerator Driven Subcritical TRIGA reactor is investigated using MCNIC method. The result of this method has been compared with those obtained by MCNPX code, 1/v poisoning method and reference data. The relative difference between MCNIC method results with MCNPX code and reference data for βeff is −3.0% and 1.7%, respectively. Our results indicate that, the Λ value from MCNIC method has relative differences of −19.4% and −10.3% with MCNPX code and 1/v poisoning method calculations, respectively. These values are increased by increasing the reflector thickness. Finally, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.
•The control rod worth is measured using the positive period method.•Core excess reactivities are measured at different power levels.•Shut down margin of the reactor has been determined.•Fuel and ...water temperatures are measured at different power levels.•All the safety parameters are compared with the safety analysis report of the reactor.
Measurement and validation of safety parameters of a nuclear reactor are required for reactor start up, normal power operation, experimental research and shutdown. The reactivity of the control rod is one of the important parameters for management of reactor operations, and is used for the prediction of control rod position at startup and the estimation of the core excess reactivity during the reactor operation. In this study, some reactor safety parameters such as control rod worth, core excess reactivity, shutdown margin, reactivity changes by fuel and control rods, and temperature effect on reactivity has been measured using digital instrumentation & control (I&C) system of the BAEC TRIGA Research Reactor (BTRR). All of these safety parameters have significant effects on the reactor control system. The measured total worth of all control rods of BTRR are 14.888 $, 14.672 $, 14.348 $ for 1.5 folding time, doubling time, 5 folding time, respectively. The measured reactivity has also been compared with the previously measured reactivity. The core excess reactivity and shutdown margin were found to be 6.38 $ and 5.20 $, respectively. The measured values were found to be within the safety limit as mentioned in the Safety Analysis Report (SAR) of the BTRR.
Radioactive lanthanides have become an important imaging, diagnostic and therapeutic tool in the medical field. The objective of our research was to investigate the feasibility of producing ...radioactive lanthanides with high specific activity in a small-scale research reactor using the Szilard–Chalmers method. The results indicate that the activated nuclides recoil out of the target after neutron capture and we obtain enrichment of the radionuclide compared to the bulk irradiation. These first attempts, result in enrichment factors and yields that are low but indicates a possibility of using this technique if the method is further optimized.
A 100 kW TRIGA Mark I research reactor has been functioning since 1960 at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil. The reactor was installed for research, manpower ...training and production of radioisotopes to be used in industry, agriculture and medical treatment. The IPR-R1 TRIGA is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactors in the world and they are characterized by inherent safety. The IPR-R1 was designed to operate at 30 kW steady-state power and 100 kW power for limited periods, namely four hours a day, two days a week. The reactor could not meet the increasing demand for operation with natural circulation alone. The forced cooling system was built in the 70s and the power was upgraded to 100 kW. Recently the core configuration and instrumentation were upgraded again to 250 kW at steady state. This paper describes some results of experiments that have been carried out in the IPR-R1 reactor in recent years with the aim of evaluating the behavior of some thermal-hydraulic and neutronic operational parameters. The research project meets the recommendations of the IAEA for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB).
•The core configuration and instrumentation of the IPR-R1 TRIGA reactor were upgraded from 100 kW to 250 kW at steady state.•Thermal hydraulic and neutronic experiments were carried out in the IPR-R1 with the aim of evaluating operational parameters.•A data acquisition and a software were developed to allow on line monitoring of the main reactor operational variables.•The experiments have given better comprehension of the reactor behavior and helped to develop numerical models.
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