TRIGA type research reactors, even relatively new ones, are originally equipped with rather obsolete irradiation pneumatic transfer systems. Therefore, the irradiation system of Slovenian TRIGA Mark ...II system was renewed in 2015 to improve the overall quality of irradiations. This year, the system was upgraded to allow for automated short irradiations under more thermalized neutron flux. The modernization includes also a so-called “triangular” channel, allowing for in-core irradiation of samples up to 5 cm in diameter and a horizontal channel allowing for irradiations of objects under homogeneous neutron flux in the length of over 60 cm.
Despite offering low-carbon and reliable energy, the utilization of nuclear energy is declining globally due to high upfront capital costs and longer returns on investments. Nuclear cogeneration of ...valuable chemicals from waste biomass-derived feedstocks could have beneficial impacts while harnessing the underutilized resource of ionizing energy. Here, we demonstrate selective methanol or acetaldehyde production from ethylene glycol, a feedstock derived from glycerol, a byproduct of biodiesel, using irradiations from a nuclear fission reactor. The influence of radiation quality, dose rate, and the absorbed dose of irradiations on radiochemical yields (G-value) has been studied. Under low-dose-rate, γ-only radiolysis during reactor shutdown rate (<0.018 kGy min–1), acetaldehyde is produced at a maximum G-value of 8.28 ± 1.05 μmol J–1 and a mass productivity of 0.73 ± 0.06% from the 20 kGy irradiation of neat ethylene glycol. When exposed to a high-dose-rate (6.5 kGy min–1), 100 kGy mixed-field of neutron + γ-ray radiations, the radiolytic selectivity is adjusted from acetaldehyde to generate methanol at a G-value of 2.91 ± 0.78 μmol J–1 and a mass productivity of 0.93 ± 0.23%. Notably, utilizing 422 theoretical systems could contribute to 4.96% of worldwide acetaldehyde production using a spent fuel pool γ-ray scheme. This research reports G-values and production capacities for acetaldehyde for high-dose scenarios and shows the potential selectivity of a nuclear cogeneration process to synthesize chemicals based on their irradiation conditions from the same reagent.
Abstract
Collection and interpolation of radiation observations is of vital importance to support routine operations in the nuclear sector globally, as well as for completing surveys during crisis ...response. To reduce exposure to ionizing radiation that human workers can be subjected to during such surveys, there is a strong desire to utilise robotic systems. Previous approaches to interpolate measurements taken from nuclear facilities to reconstruct radiological maps of an environment cannot be applied accurately to data collected from a robotic survey as they are unable to cope well with irregularly spaced, noisy, low count data. In this work, a novel approach to interpolating radiation measurements collected from a robot is proposed that overcomes the problems associated with sparse and noisy measurements. The proposed method integrates an appropriate kernel, benchmarked against the radiation transport code MCNP6, into the Gaussian Process Regression technique. The suitability of the proposed technique is demonstrated through its application to data collected from a bespoke robotic system used to conduct a survey of the Joz̆ef Stefan Institute TRIGA Mark II nuclear reactor during steady state operation, where it is shown to successfully reconstruct gamma dosimetry estimates in the reactor hall and aid in identifying sources of ionizing radiation.
•For the first time, delayed gamma ray transport outside TRIGA Mark II research reactor biological shield was simulated using Monte Carlo Methods.•Developed method was validated on a simple ...experiment where dose rate field was calculated and measured around transport cask contacting irradiated TRIGA fuel element.•It is demonstrated that major contribution to the dose in the reactor hall in case of loss of water event is from backscattered gamma rays.•In case of loss of water event for reactor pool, reactor platform has to be evacuated for 10 days due to the increased dose rated. Later, corrective actions can be done.
Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate the transport of delayed gamma particles emitted from irradiated TRIGA fuel. At the first step, this method was verified on a simple experiment where irradiated fuel was packed inside a transport cask. The calculated dose rate field in the vicinity of the transport cask was compared to the measurements. Due to an acceptable agreement, the method was later used to analyse Loss of Water Event (LOWE) for the JSI TRIGA research reactor main and spent fuel pool.
Main results of the analysis show that in the case of LOWE for the reactor pool, the reactor building has to be evacuated since dose rates exceed 2 mSv/h. The highest dose rates are expected at the reactor platform. The radiation from the core is backscattered from the ceiling and affects the dose rates inside the whole reactor hall and control room. In the worst case scenario, the operating personnel can start with corrective actions 10 days after the last reactor operation, when dose rates at the reactor platform drop below 20 mSv/h.
If LOWE occurs for the spent fuel pool, the reactor building has to be evacuated. However, the corrective actions by the operating personnel can be started immediately after the accident. This is due to the fact, that the fuel can only be transported into the spent fuel pool after a three-month cooling period inside the main reactor pool, resulting in a lower fuel activity.
Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate neutron and prompt gamma particles emitted from TRIGA research ...reactor during operation. Firstly, the method was validated by measuring dose rates around open beam port number 5 was unplugged. Neutron and gamma dose rates inside the reactor hall in the vicinity of the beam port were calculated and compared to the measurements. Due to the satisfactory agreement, the method was later used to design external shielding for the same beam port when it was upgraded – special mechanism was installed that allows irradiation of larger samples. Computational analysis of the proposed shielding configuration provided acceptable dose rate levels inside the reactor hall. When the shield was constructed, calculated dose rates were confirmed by the actual measurements. No modifications were needed.
•For the first time neutron and gamma ray transport outside JSI TRIGA shielding body was simulated using Monte Carlo Method•Method was validated by comparing calculated and measured dose rates around beam port no. 5 during reactor operation.•Removable shield was placed in front of the beam tube opening in order to study backscattered radiation.•It is demonstrated that developed method can be used in designing appropriate shielding for the future experiments.
High radiation tolerance of functional materials in harsh environments is the key requirement for the operation of particle accelerators, medical devices, nuclear power plants, satellites, and ...spacecraft. Neutron and gamma (γ) radiation can seriously affect the functional properties of the irradiated materials and thus the performance of the entire device. In this work, the feasibility of using (1-x)Pb(Mg1/3Nb2/3)O3–xPbTiO3 (PMN–100xPT) electrocaloric materials in applications where the material is exposed to high neutron and γ-radiation is investigated. For this purpose, three different compositions of PMN–100xPT ceramics (x = 0, 0.1, and 0.35) were prepared and their dielectric, ferroelectric and electrocaloric properties were investigated before and after neutron and γ-irradiation. The samples were irradiated with a neutron fluence of 1015 to 1017 neutrons cm−2 with an energy of 1 MeV, which exceeds the largest expected neutron irradiation in the European Council for nuclear Research (CERN) and simultaneously exposed to γ-irradiation. The neutron and γ-radiation partially affect the functional properties of the PMN–35PT, the ceramic with distinct ferroelectric and weakened relaxor features, with some differences observed in the domain switching behavior, measured by conventional polarization versus electric field (P–E) hysteresis, at the highest radiation dose of 1017 neutrons cm−2. In contrast, the functional properties of the irradiated PMN and PMN–10PT samples with relaxor behavior are quite similar to those of the pristine samples, therefore, we conclude that these materials can be used as working materials in EC coolers exposed to such harsh environments.
CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core ...are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring.
•Neutron and gamma measurements with fission and ionization chambers were performed.•Axial in-core profiles were measured at the JSI TRIGA reactor in pulse and current mode.•Measurements were compared with the Monte Carlo calculation using MCNP model.•Mostly, the C/M agreement is very satisfactory – within a few percent.
Non-intermittent, low-carbon energy from nuclear or biofuels is integral to many strategies to achieve Carbon Budget Reduction targets. However, nuclear plants have high, upfront costs and biodiesel ...manufacture produces waste glycerol with few secondary uses. Combining these technologies, to precipitate valuable feedstocks from waste glycerol using ionizing radiation, could diversify nuclear energy use whilst valorizing biodiesel waste. Here, we demonstrate solketal (2,2-dimethyl-1,3-dioxolane-4-yl) and acetol (1-hydroxypropan-2-one) production is enhanced in selected aqueous glycerol-acetone mixtures with γ radiation with yields of 1.5 ± 0.2 µmol J
and 1.8 ± 0.2 µmol J
, respectively. This is consistent with the generation of either the stabilized, protonated glycerol cation (CH
OH-CHOH-CH
OH
) from the direct action of glycerol, or the hydronium species, H
O
, via water radiolysis, and their role in the subsequent acid-catalyzed mechanisms for acetol and solketal production. Scaled to a hypothetically compatible range of nuclear facilities in Europe (i.e., contemporary Pressurised Water Reactor designs or spent nuclear fuel stores), we estimate annual solketal production at approximately (1.0 ± 0.1) × 10
t year
. Given a forecast increase of 5% to 20% v/v% in the renewable proportion of commercial petroleum blends by 2030, nuclear-driven, biomass-derived solketal could contribute towards net-zero emissions targets, combining low-carbon co-generation and co-production.
The design and test of a robot-compatible, radiation detection instrument providing simultaneous <inline-formula> <tex-math notation="LaTeX">\gamma </tex-math></inline-formula>-ray imaging and ...<inline-formula> <tex-math notation="LaTeX">\gamma </tex-math></inline-formula>-ray spectroscopy is described. The sensing system comprises a cerium bromide inorganic scintillation detector and a cylindrical, lead slot collimator that is configured with a robot-compatible, on-board data acquisition system. The mount for the sensor is a lightweight, bespoke 3-axis gimbal actuated by servos for pan, tilt and rotation of the collimator for imaging capability. This paper discusses the integration of this relatively low-cost radiation detection apparatus with a commercially available, Robot-Operating-System-controlled robotic platform (a Clearpath Robotics™ Jackal). The detection system is compliant with the power and mass payload constraints of the robot. Its performance has been evaluated by means of two practical examples: a) measurements in a laboratory environment to assess the ability of the system to resolve two caesium-137 point sources, and b) deployment at the Jožef Stefan Institute TRIGA Mark II research reactor to assess the ability of the system to characterise the <inline-formula> <tex-math notation="LaTeX">\gamma </tex-math></inline-formula>-ray emission at 1 kW from a horizontal tangential beam port in the reactor hall and from the reactor sample pool above the core.
•Summary of the current state of E&T and VET in EU.•The history and vision of the European Nuclear Experimental Educational Platform.•Development and application of alearning model consisting of ...knowledge skills and attitudes in the learning process.•Teaching methods developed to measure the progress of participants.
This paper presents the educational methods of the European Nuclear Experimental Educational Platform (ENEEP), an open platform continuing the legacy of the H2020 ENEEP project. It offers a brief overview of the current state of education, training, and vocational education in the European Union, emphasising the imperative for a skilled and competent workforce in the future. Subsequent sections outline the objectives, timeline, and structure of the ENEEP project. The most important part is the description of educational methods developed within the project to deliver high-quality tailored courses. Specific attention is given to a series of demonstration courses, wherein the efficacy of these methods was systematically measured and evaluated. The final part of the paper outlines the vision of the ENEEP association its standing in the international educational and training community and highlights key activities planned for the upcoming period.
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