Search

The McMaster nuclear reactor (MNR) is an important research facility that not only provides researchers with neutrons for fundamental science, but also supplies the medical industry with isotopes used for cancer treatment. To ensure the safety and performance of the MNR, modeling of the thermal hydraulics during nominal and accidental conditions is required. For such a task, system codes are customarily used. While system codes can assess the safety aspects of complex thermal–hydraulic systems, the question arises whether such systems can be modeled appropriately in a three-dimensional manner, such as computation fluid dynamics (CFD), while capturing the thermal mixing of the coolant. Modeling the thermal hydraulics of nuclear research reactors using CFD is relatively new. Therefore, validating such methods against experimental data is of utmost importance. The validation of CFD is the main focus of this work. More specifically, the influence of the loss of heat sink on the pool temperature is assessed using the CFD code Ansys CFX. The validation basis was provided by an experiment performed at the MNR in March 2023. In this experiment, the secondary heat removal system was intentionally shut down, and the pool temperature was measured in a few positions. The results obtained by modeling the loss of heat sink in the MNR using Ansys CFX agree well with the experiment. [ABSTRACT FROM AUTHOR]

In view of the complexity and importance of nuclear research reactor (NRR) installations, it is imperative to uphold high standards of reliability and quality in the electricity being supplied to them. In this paper, the performance of low-voltage (LV) distribution feeders integrated with NRRs is improved in terms of reduced distribution loss, improved voltage profile, and reduced greenhouse gas (GHG) emissions by determining the optimal location and size of photovoltaic (PV) systems. In the second stage, the power quality of the feeder is optimized by reducing the total harmonic distortion (THD) by optimally allocating D-STATCOM units. In the third and fourth stages, the reliability and resilience aspects of the feeder are optimized using optimal network reconfiguration (ONR) and by integrating an energy storage system (ESS). To solve the non-linear complex optimization problems at all these stages, an efficient meta-heuristic Chernobyl disaster optimizer (CDO) is proposed. Simulations are performed on a modified IEEE 33-bus feeder considering the non-linear characteristics of NRRs, variability of the feeder loading profile, and PV variability. The study reveals that the proposed methodology can significantly improve the service requirements of NRRs for attaining sustainable research activities. [ABSTRACT FROM AUTHOR]

In this paper, we study the influence of densified microsilica and colloidal nanosilica admixtures on the mechanical strength and the microstructural characteristics of special mortars used for immobilizing radioactive concrete waste. The experimental program focused on the replacement of cement with micro- and/or nanosilica, in different proportions, in the basic composition of a mortar made with recycled aggregates. The technical criteria imposed for such cementitious systems, used for the encapsulation of low-level radioactive waste, imply high fluidity, increased mechanical strength and lack of segregation and of bleeding. We aimed to increase the structural compactness of the mortars by adding micro- and nanosilica, all the while maintaining the technical criteria imposed, to obtain a cement matrix with high durability and increased capacity for immobilizing radionuclides. The samples from all the compositions obtained were analyzed from the point of view of mechanical strength. Also, micro- and nanosilica as well as samples of the optimal mortar compositions were analyzed physically and microstructurally. Experimental data showed that the mortar samples present maximum compressive strength for a content between 6 and 7.5% wt. of microsilica, respectively, for a content of 2.25% wt. nanosilica. The obtained results suggest a synergistic effect of micro- and nanosilica when they are used simultaneously in cementitious compositions. Thus, among the analyzed compositional variants, the mortar composition with 3% wt. microsilica and 2.25% wt. nanosilica showed the best performance, with an increase in compressive strength of 23.5% compared to the control sample (without micro- and nanosilica). Brunauer–Emmett–Teller (BET) analysis and scanning electron microscopy (SEM) images highlighted the decrease in pore diameter and the increase in structural compactness, especially for mortar samples with nanosilica content or a mixture of micro- and nanosilica. This study is useful in the field of recycling radioactive concrete resulting from the decommissioning of nuclear research or nuclear power reactors. [ABSTRACT FROM AUTHOR]

Nuclear reactor control is pivotal for the safe and efficient operation of nuclear power plants, facilitating the regulation of reactor reactivity. This study introduces an optimized fractional-order proportional-integral-derivative (FOPID) controller tailored for maintaining reactivity levels in nuclear power plants, particularly during load-following operations. The controller adjusts the position of control rod to regulate power output effectively. We enhance FOPID controller's performance using a modification of Planet Optimization Algorithm (POA-M), leveraging the strengths of the Arithmetic Optimization Algorithm (AOA) to improve its exploitation capabilities. We evaluate the efficacy of POA-M-FOPID controller against traditional techniques, including POA, AOA, and Particle Swarm Optimization (PSO). We assess its performance using the Egyptian Testing Research Reactor (ETRR-2) as a case study. Our results demonstrate that the POA-M-FOPID controller outperforms alternative algorithms across various control metrics, exhibiting superior resilience and efficiency. Notably, the utilization of the POA-M-FOPID controller yields remarkable improvements in reactor power performance, achieving significantly reduced settling time (25.27 sec) and maximum overshoot (0.67 %) compared to conventional FOPID controllers incorporating POA, AOA, and PSO methods. These findings underscore the effectiveness of POA-M-FOPID in enhancing nuclear reactor control systems, offering potential benefits for broader nuclear power industry in terms of safety, stability, and operational efficiency. [ABSTRACT FROM AUTHOR]

The purpose of this study is to provide a new deterministic model for the SLOWPOKE-2 nuclear research reactor at École Polytechnique de Montréal (EPM) with LEU (Low Enriched Uranium) core. Using the latest release of the code system DRAGON5 and DONJON5 and the cross-section data library ENDFB.VII rel.1 evaluation, the developed model is applied to simulate the neutronic behavior of the SLOWPOKE-2 research reactor. We studied the separate temperature effects of the main components of the core (i.e., fuel, coolant/moderator, beryllium reflector, and water reflector). The contribution of different physical phenomena to the RTC was assessed. The temperature reactivity feedback calculated using the deterministic approach based on the DRAGON5 and DONJON5 code system using the ENDF/B-VII.1 evaluated nuclear data library produced in the WIMD-D4 format is in good agreement. Therefore, this work proves the capability of DRAGON5 and DONJON5 codes, normally used for power reactors, to reliably simulate a low-power research reactor. [ABSTRACT FROM AUTHOR]

A major concern for nuclear research reactors under normal operating conditions is that they may release radioactive elements into the atmosphere, endangering public health and the environment. The present study concentrated on the detailed radiological dose assessment resulting from the atmospheric release of 41Ar activity from the TRIGA Mark‐II research reactor in Bangladesh during its normal operational condition at full power of 3 MW. The total effective dose equivalent (TEDE), ground deposition activity, organ‐committed dose, and pathways dose values have been evaluated under different weather conditions using the HotSpot 3.1.2 code. The weather data have been gathered from the Bangladesh Meteorological Department (BMD), Dhaka. Two significant seasons with various weather stability effects have been considered for dose analysis. Atmospheric dispersion of 41Ar was evaluated using the Gaussian plume model. From the obtained results, the maximum TEDE of 4.06 × 10−9 Sv at 0.19 km distance from the reactor site for stability class B during the summer season is found to be well below the annual effective dose limit of 1 mSv recommended by the ICRP. During the rainy season, the maximum TEDE of 1.76 × 10−9 Sv at 0.92 km distance for stability class E is also found to be negligible compared to the dose limit. From the organ‐committed dose analysis, skin is investigated as the highest dose absorber compared to other organs. The pathways dose analysis concludes that the submersion and ground shine doses are observed to be maximum at 0.20 km and 1.0 km distances for the summer and rainy seasons, respectively. Based on the identified results, dose values have been found to be within the limiting values, ensuring environmental and human health safety. Such comprehensive dose analysis due to the atmospheric release of 41Ar activity is very significant from the perspective of ensuring the radiological and environmental safety of research‐type nuclear reactors under normal operational conditions. [ABSTRACT FROM AUTHOR]

Autonomous nuclear reactor monitoring is a key aspect of the International Atomic Energy Agency's strategy to ensure nonproliferation treaty compliance. From the rise of small modular reactor technology, decentralized nuclear reactor fleets may strain the capacities of such monitoring and requires new approaches. We demonstrate the superior capabilities of a gamma detection system to monitor the criticality of a zero power nuclear reactor from beyond typical vessel boundaries, offering a powerful alternative to neutron-based systems by providing direct information on fission chain propagation. Using the case example of the research reactor CROCUS, we demonstrate how two bismuth germanate scintillators placed outside the reactor vessel can precisely observe reactor criticality using so called noise methods and provide core status information in seconds. Our results indicate a wide range of applications due to the newly gained geometric flexibility that could find use in fields beyond nuclear safety. [ABSTRACT FROM AUTHOR]

One of the wastewater treatment units in the Tehran Research Nuclear Reactor (TRNR) is an evaporator unit, which consists of a semi-batch single-stage kettle-type evaporator with heating coils and auxiliary equipment. The concentrate of this unit is immobilized in cement. So, the precipitate specie, concentrate composition, and its amount are important parameters for further waste treatment steps. In this research, a new bulk mathematical model is developed based on the total mass balance, the partial mass balance of ions in the concentrate and precipitates, as well as the energy balance to investigate the performance of the Tehran Research Nuclear Reactor evaporator. The resulting ordinary differential equations are solved numerically with a proposed algorithm. In this simulation, Pitzer equations are used to calculate the activity coefficient of water and ions, density, and solubility of salts. The temperature and pressure dependence of the developed model parameters are also considered. Moreover, the effect of precipitate formation on the overall heat transfer coefficient with time dependence correlation is also taken into account. Results show a good agreement between the experimental data and the developed model, and the relative error of different parameters is reported between 2% and 13%. • New model developed to estimate performance of Tehran Nuclear Reactor evaporator. • Conservation of ions mass as well as the energy are considered in this model. • New algorithm proposed to solve the ordinary differential equations. • Thermophysical properties of ions and concentrate are derived from pitzer equation. • Mathematical model predict the experimental data by the relative error of 2 - 13%. [ABSTRACT FROM AUTHOR]

Gamma-ray spectroscopy is an effective technique for radioactive material characterization, routine inventory verification, nuclear safeguards, health physics, and source search scenarios. Gamma-ray spectrometers typically cannot be operated in the immediate vicinity of nuclear reactors due to their high flux fields and their resulting inability to resolve individual pulses. Low-power reactor facilities offer the possibility to study reactor gamma-ray fields, a domain of experiments hitherto poorly explored. In this work, we present gamma-ray spectroscopy experiments performed with various detectors in two reactors: The EPFL zero-power research reactor CROCUS, and the neutron beam facility at the Ohio State University Research Reactor (OSURR). We employed inorganic scintillators (CeBr3), organic scintillators (trans-stilbene and organic glass), and high-purity germanium semiconductors (HPGe) to cover a range of typical—and new—instruments used in gamma-ray spectroscopy. The aim of this study is to provide a guideline for reactor users regarding detector performance, observed responses, and therefore available information in the reactor photon fields up to 2 MeV. The results indicate several future prospects, such as the online (at criticality) monitoring of fission products (like Xe, I, and La), dual-particle sensitive experiments, and code validation opportunities. [ABSTRACT FROM AUTHOR]

The Idaho National Laboratory (INL) is a government research facility that has been conducting scientific research on nuclear energy and military applications for over 70 years. INL has made significant contributions to the nuclear energy industry, including generating the first usable electricity from nuclear power and developing nuclear propulsion systems for Navy submarines and aircraft carriers. However, there is also a dark side to the INL, as it has been responsible for worker health issues, pollution, and radioactive waste. The article highlights concerns about worker safety, hazardous conditions, and retaliation against those who report safety concerns. [Extracted from the article]

Abstract—The article considers the efficiency and possibility of introducing into operation of VVER-1000 power units with the lifetime extended owing to the sliding pressure of the second circuit. The history of research by Russian nuclear engineers on VVER lifetime extension is reviewed. It is shown that reduction of fresh steam pressure within the allowed limits will make it possible to release additional reactivity through release of the temperature effect and to generate additional energy during the period of unit operation on the power effect. A mathematical model of changes in the main parameters of the reactor unit, including reactivity balance equations, Stodola–Flugel equation for the turbine, and heat balance and heat transfer equations in the steam generator, is compiled. The expected increase in power generation for modern fuel loadings of VVER-1000 reactors is calculated. The degree of compliance of the proposed mode with the current technological limitations of the large-series VVER-1000 power units is analyzed. The results of the analysis are confirmed by the numerical experiment on the multifunctional analyzer of the VVER-1000 unit modes developed by the National Research Nuclear University MEPhI built on the basis of the ENICAD environment and the certified diffusion neutronic code PROSTOR. Conclusions are drawn about the possibility of introducing the proposed mode into the operational practice of modern VVER-1000 power units. [ABSTRACT FROM AUTHOR]

A novel solid-state neutron and gamma radiation monitor-dosimeter based on biopolymer polylactic acid (PLA) is presented. The resulting detector (PLAD) technology takes advantage of property changes of the renewable PLA resin when subject to ionizing nuclear radiation. A simple yet rapid and accurate (±10%) low-cost (<$0.01/detector) mass loss upon dissolution (MLD) technique was successfully developed; MLD is based on a simple mass balance for discerning neutron and/or gamma doses using small (40 mg, ~4 mm diameter) ultra-low-cost (<$0.01) resin beads via dissolution in acetone. The GammaCellTM Co-60 irradiator, and the PUR-1 12 kW fission nuclear research reactor were utilized, respectively. Irradiation absorbed doses ranged from 1 to 100 kGy. Acetone bath temperature was varied from ~40 °C to ~54 °C. Results revealed a strong dependence of MLD on acetone bath temperature between neutron and gamma photon dose components; this allowed for the unique ability of PLAD to potentially perform as both a neutron-cum-gamma or as a gamma or neutron radiation dosimeter and intensity level detector. A linear trend is found for combined neutron and gamma radiation doses from 0 to 40 kGy when dissolution is conducted above 50 °C. The important potential ability to distinguish neutron from gamma radiation fields was scoped and found to be feasible by determining MLD at 45 °C. The potential was studied for simultaneous use as an in-core neutron and gamma monitor of an operating 3 GWt light-water reactor (LWR). Scoping tests were conducted with the pre-irradiated (@ 20 °C) PLAD resin beads followed by heating to in-core LWR coolant (300 °C) conditions for ~30 s corresponding to the time to reach ~40 kGy total doses in a typical 3 GWt LWR. MLD results were unaffected, indicating the exciting and unique potential for in situ (low-cost, accurate and rapid) simultaneous mapping of neutron and gamma radiation fluxes, related dosimetry, and fission power level monitoring. [ABSTRACT FROM AUTHOR]

New values of neutron fluxes and spectral parameters f and α were determined experimentally in all irradiation devices of the TRIGA Mark I IPR-R1 nuclear research reactor at Nuclear Technology Development Centre (CDTN), Belo Horizonte, Brazil. Sets of monitors Au, Fe, Zn and Zr were irradiated bare and Cd-covered, according to "Cd-ratio for multi-monitor" method. Values were validated by analysing the certified reference material BCR-320R irradiated in chosen channels. The calculations were made based on irradiation channel values and the average values of the Carousel. The results of En-score point out that the k0-method is producing reliable results. From now on, the values of mass fractions in several matrices, the production and studies with radioisotopes will be more accurate and the activities calculated more precisely. [ABSTRACT FROM AUTHOR]

Studying the power quality (PQ) is an essential issue to ensure the safe and accurate operation of sensitive equipment particularly for nuclear installations. Assessment of PQ involves collecting and analysing data resources and then evaluating it with reference to PQ standards. There are many alternatives for PQ and it is difficult to make an appropriate selection among them in the existence of their multiple criteria which are usually conflicted. So this selection subject can be classified as a Multi Criteria Decision Making (MCDM) problem. To do so, a reliable and scientific method for studying and evaluating the overall system PQ is required. This study aims to assess performance of PQ for the electrical power system at a Nuclear Research Reactor (NRR) during a certain period using multiple measures for the most decisive PQ phenomena. It focuses on a number of the most important PQ phenomena namely frequency fluctuation (deviation), unbalances of current and voltage, current and voltage harmonic distortion, flicker and power factor. After combining all results into six samples (alternatives), the criteria weights are determined based on an objective method for weighting which is called CRITIC method. Then, the alternatives are ranked using compromise MCDM method-VIKOR method. The obtained results are analyzed and discussed to evaluate performance of NRR electrical system from the PQ view. It showed that the compromise solution that obtained by CRITIC-VIKOR can be a guide to facilitate the PQ evaluation of nuclear installation electrical system. Also, it can empower the operators with the benefits of benchmarking and monitoring a single index instead of several indices. Moreover, it is very useful for helping stakeholders to understand how the PQ performance changes under a certain operating condition of the facility. Finally, it is can be considered as a good model to weight each PQ phenomena and identify the time intervals for best and worst total PQ in NRR. [ABSTRACT FROM AUTHOR]

Indian nuclear research reactors provide facilities to carry out irradiation experiments on the core structural materials to assess the radiation damage in these materials. The radiation damage in the material strongly depends on the temperature of irradiation. The temperature-controlled instrumented irradiation capsule (TCIIC) is a device which facilitates irradiation of specimens at a constant higher temperature in a reactor. This paper discusses design of PID controller to establish a constant temperature in the TCIIC system. The control parameters are estimated through placement of dominant pole-zero in the "S-plane" and a stable region of the PID controller parameters was defined. The time response result from the proposed method has been compared with other well-known PID controller design methods and also verified experimentally. The proposed method shows that the controller parameters can provide a stable and reliable performance of temperature control needed in the irradiation experiment. [ABSTRACT FROM AUTHOR]

The results of the analysis of the radiation state of environmental objects in the sanitary protection zone (SPZ) and surveillance zone (SZ) of the research nuclear reactor (RNR) WWR-M of the Institute for Nuclear Research of the National Academy of Sciences of Ukraine (INR) for 2021 are analyzed and compared with the data for previous years (from 2016 to 2020). The levels of total beta activity of sedimentary dust and atmospheric precipitation and wastewater from the main reservoirs of INR, the concentration of beta-active aerosols in the surface layer of atmospheric air, as well as the content of 137Cs radionuclide in soil and vegetation samples, were studied. The research was conducted in the laboratory of the Center for Ecological Problems of Atomic Energy of INR. The obtained results show that in 2021 there was no excess of the values of controlled parameters compared to previous years. There was also no significant difference in the specific activity of the radionuclide 137Cs in soil and vegetation samples from SPZ and ZS. Studies conducted in the period from 2016 to 2021, allowed to accumulate significant factual material on the radiation situation in the sanitary protection zone and surveillance zone and to substantiate the conclusion about the absence of man-made radioactive contamination of the environment due to the operation of RNR WWR-M of INR of NAS of Ukraine. [ABSTRACT FROM AUTHOR]

In 1981, Israel launched a preventive military strike against a nuclear reactor that Iraq was constructing at the Tuwaitha Nuclear Research Center. The low fungibility of Iraq's power resources, and especially its nuclear weapons program, shaped Israel's decision-making process. First, it motivated Israel's executives to disaggregate Iraq's capabilities. Though they identified Baghdad's emerging nuclear weapons program as a threat, there was little pressure to act more broadly. Second, it pushed Israel's leaders to favour delaying strategies to stall Iraq's program. Third, it motivated Israel's leaders to favour a preventive military strike to delay Iraq's nuclear program from becoming more fungible. [ABSTRACT FROM AUTHOR]

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. [Display omitted] • 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. [ABSTRACT FROM AUTHOR]

Fault diagnosis of Medium Voltage power Cables (MVCs) research nuclear reactor, incredibly inaccessible/remote ones, has to be carefully identified, located, and fixed within a short time. Therefore, this paper proposes a perfect simultaneous fault diagnosis scheme based on Multiclass Support-Vector Machine (MCSVM) in the fractional Fourier domain. First, the three-phase sending currents are simulated under different conditions then their features are extracted using Discrete Fractional Fourier Transform (DFRFT). Afterward, the features reduction process occurs via the Singular Value Decomposition (SVD) approach. MCSVM scheme is used to diagnose faults (i.e., discover, categorize, and trace) using reduced features outcome from DFRFT and SVD stages. Alternating Transient Program/Electromagnetic Transient Program (ATP/EMTP) simulations have been carried out for 22 kV unreachable MVC. Different kernels of SVM, i.e., linear, quadratic, or polynomial, and diverse factors of DFRFT, i.e., α, are investigated in simulations to obtain the optimum performance (i.e., best α and kernel pair). Hence, performance analysis of the proposed diagnosis method under different conditions (i.e., various fault resistances, locations, and inception angles) concluded two highest accuracy and lowest time settings, which were found at α = 0.5 (for both) quadratic kernel, and linear kernel, respectively. Moreover, the linear kernel achieves 99.8% accuracy rate, the lowest execution time (10 ms), and fault tracing error rate of 0.525789%, which is proper for real-time applications. Besides, our proposed method is more reliable and accurate against variable operating conditions (fault resistances, distances, and inception angles), leading to more reliable power production systems. [ABSTRACT FROM AUTHOR]

The future Ricochet experiment aims at searching for new physics in the electroweak sector by providing a high precision measurement of the Coherent Elastic Neutrino-Nucleus Scattering (CENNS) process down to the sub-100 eV nuclear recoil energy range. The experiment will deploy a kg-scale low-energy-threshold detector array combining Ge and Zn target crystals 8.8 m away from the 58 MW research nuclear reactor core of the Institut Laue Langevin (ILL) in Grenoble, France. Currently, the Ricochet Collaboration is characterizing the backgrounds at its future experimental site in order to optimize the experiment's shielding design. The most threatening background component, which cannot be actively rejected by particle identification, consists of keV-scale neutron-induced nuclear recoils. These initial fast neutrons are generated by the reactor core and surrounding experiments (reactogenics), and by the cosmic rays producing primary neutrons and muon-induced neutrons in the surrounding materials. In this paper, we present the Ricochet neutron background characterization using 3 He proportional counters which exhibit a high sensitivity to thermal, epithermal and fast neutrons. We compare these measurements to the Ricochet Geant4 simulations to validate our reactogenic and cosmogenic neutron background estimations. Eventually, we present our estimated neutron background for the future Ricochet experiment and the resulting CENNS detection significance. Our results show that depending on the effectiveness of the muon veto, we expect a total nuclear recoil background rate between 44 ± 3 and 9 ± 2 events/day/kg in the CENNS region of interest, i.e. between 50 eV and 1 keV. We therefore found that the Ricochet experiment should reach a statistical significance of 4.6 to 13.6 σ for the detection of CENNS after one reactor cycle, when only the limiting neutron background is considered. [ABSTRACT FROM AUTHOR]

Bangladesh Atomic Energy Commission TRIGA Research Reactor (BTRR) is a MK II type nuclear research reactor with a maximum thermal output power of 3 MW. Nuclear research/test reactors are designed for the development of nuclear science and technology including all necessary safety features. Safety issues are the prime concern for nuclear reactors to protect the nuclear accidents. BTRR has several passive and engineered safety features against any accidental failure of its control mechanisms leading to the meltdown of the reactor core or any other nuclear incident/accident. The nuclear percent power channel (NPP-1000) is one of the vital parts of the safety systems of the BTRR. Suddenly, the Instrumentation and Control (I&C) system of BTRR is tripped with a system SCRAM message "NPP HV Low". As a safety system, it is very compulsory to resolve the occurred uncertain issue before going to further reactor operation. To solve that kind of unwanted problem, effective troubleshooting steps were taken very carefully. Circuit analysis showed that a fault transistor was found inside the "HV trip circuit". By replacing this transistor with a good one, the corresponding relay is now functioning successfully and the "NPP HV low trip" condition as well as the "trip alarm" have been removed from the I&C system. Several types of experiments, research and training are now going on using this Bangladesh Atomic Energy Commission (BAEC) TRIGA Research Reactor facility by ensuring proper safety. Different troubleshooting experiences provide various strategies which are helpful for resolving many upcoming problems and also enrich the problem analyzing ability for further known or unknown tasks". [ABSTRACT FROM AUTHOR]

Reactor TRIGA PUSPATI (RTP) is the only research nuclear reactor in Malaysia. Maintenance of RTP is crucial which affects its safety and reliability. Currently, RTP maintenance strategies used corrective and preventative which involved many sensors and equipment conditions. The existing preventive maintenance method takes a longer time to complete the entire system's maintenance inspection. This study has investigated new predictive maintenance techniques for developing RTP predictive maintenance for primary cooling systems using machine learning (ML) and augmented reality (AR). Fifty papers from recent referred publications in the nuclear areas were reviewed and compared. Detailed comparison of ML techniques, parameters involved in the coolant system and AR design techniques were done. Multiclass support vector machines (SVMs), artificial neural network (ANN), long short-term memory (LSTM), feed forward back propagation (FFBP), graph neural networks-feed forward back propagation (GNN-FFBP) and ANN were used for the machine learning techniques for the nuclear reactor. Temperature, water flow, and water pressure were crucial parameters used in monitoring a nuclear reactor. Image marker-based techniques were mainly used by smart glass view and handheld devices. A switch knob with handle switch, pipe valve and machine feature were used for object detection in AR markerless technique. This study is significant and found seven recent papers closely related to the development of predictive maintenance for a research nuclear reactor in Malaysia. [ABSTRACT FROM AUTHOR]

• GAMA MSR is developed by the Department of Nuclear Engineering and Engineering Physics, Universitas Gadjah Mada. • GAMA MSR is in early stage of development. Therefore, simulation is required to obtain important neutronic characteristics. • Reactivity control is done by fuel level adjusting and utilizing control rod. Fuel draining can be used as shutdown system. • OpenMC and SCALE was used to calculate criticality, conversion ratio, and fissile and fertile inventory and reactivity feedback. • The results showed that this reactor is inherently safe characteristics and the fuel draining is proven to be an effective shutdown system. The GAMA molten salt reactor (GAMA-MSR) is a small modular MSR design developed by the Nuclear Reactor Research Team of the Department of Nuclear Engineering and Engineering Physics, Universitas Gadjah Mada. In the GAMA-MSR design, the reactor core is placed above the primary heat exchanger. The primary heat exchanger is partially filled with fuel in normal operation, allowing room to be filled with the fuel drained from the reactor. Thus, in addition to control rods, the design of GAMA-MSR provides unique reactivity control through the balance of fuel salt fuel from the reactor core to the primary heat exchanger vice versa by adjusting fuel pump capacity. This design allows to shutdown the reactor by switching off the fuel pump. The reactor is initially fueled with 7LiF-ThF 4 -UF 4 with 75: 20.1: 4.9 mol composition. The uranium has 19.75 % U-235 mol fraction. The reactor criticality and nuclide composition in the reactor fuel are calculated using OpenMC code during the reactor operation lifetime, with periodic adjustments to the fuel composition to simulate the effects of the fissile and fertile additions to the fuel and the extraction of actinides and fission products from the fuel. The calculated reactor thermal power is 60 MWth in this paper. The reactor is operated until it achieves a burnup value of 34,000 MWd/tHM. At this burnup value, the U-235 inventory is consumed and the amount is reduced from 794 kg to 250 kg. The U-233, Pu-239 and Pu-241 are produced with the amount are 246 kg, 24 kg and 8 kg respectively at the end of reactor operation. The effects of temperature, fuel density change, and fission product neutron poisons are also calculated. The average temperature reactivity coefficient value is approximately - 4.4395 × 10 - 5 Δ k / k per °C. The effect of thermal expansion has been included in the calculation of this temperature reactivity coefficient value. The average value of the void reactivity coefficient is + 1.2606 × 10 - 3 Δ k / k per % of void fraction for the void fraction value range up to 50 %. However, molten salt fuel suffers only small changes in density during operation. The negative value of the temperature reactivity coefficient will be dominant; thus, the GAMA-MSR becomes inherently safe. The GAMA-MSR is designed for a 30-year operation time. The effect of the fission product to the reactor reactivity is - 1.4597 × 10 - 2 Δ k / k. Full draining of the fuel to primary heat exchanger brings the reactor to shutdown condition with the reactivity of −0.20525. Results showed that this reactor has an inherently safe characteristic and has an effective shutdown system, especially due to the fuel draining system. [ABSTRACT FROM AUTHOR]

To predict the thermal hydraulic behavior in nuclear power plants, traditional system codes such as RELAP, MARS, and CATHARE, designed for one-dimensional two-phase flows, have been employed. However, these codes face limitations in simulating unknown phenomena or new geometries, highlighting the need for advanced three-dimensional simulations for the safety analysis of reactors, especially small modular reactors with passive safety systems. Recent efforts in multidimensional thermal hydraulic analysis have employed various codes for specific scenarios, such as pressurized thermal shock and hydrogen distribution in containment vessels. Consequently, the Korea Atomic Energy Research Institute (KAERI) has developed the CUPID code since 2007, which utilizes a two-fluid, three-field model on unstructured grids, enabling efficient large-scale simulations through MPI-based parallelization. Despite its successes, the computational efficiency of pressure-based solvers, which consume a significant portion of simulation time, is identified as a crucial area for improvement. This study introduces and verifies the meshless Geometric Multi Grid (GMG) technique, a novel approach to enhance the efficiency of solving the pressure correction equation on unstructured grids. The GMG technique, aimed at reducing computational time and increasing scalability, is compared to the conventional Bi-Conjugate Gradient method through various numerical simulations, demonstrating its effectiveness and potential as a superior alternative for thermal hydraulic simulations in nuclear power plant analysis. • Introduces the CUPID code, a three-dimensional simulation tool developed by KAERI, utilizing a two-fluid, three-field model on unstructured grids. • Proposes a novel meshless Geometric Multi Grid (GMG) method, designed to significantly improve the computational efficiency. • Demonstrates through comparative studies that the GMG method substantially reduces computational time compared to Bi-Conjugate Gradient methods. • Confirms through extensive numerical simulations that GMG offers stable convergence on structured and unstructured grids. [ABSTRACT FROM AUTHOR]

The TRIGA Mark II research reactor at the Jožef Stefan Institute in Slovenia achieved first criticality in 1966. Since then, the reactor has been playing an important role in developing nuclear technology. The reactor has been mainly used for research, education of university students, training of operators of the Krško nuclear power plant (start of operation in 1983) and other nuclear specialists, isotope production and beam applications. The reactor is experiencing a high level of activity today, engaging in a diverse range of experiments and studies across reactor physics, environmental research, radiation hardness testing as well training and education. The future of nuclear technology in Slovenia is focused on new NPPs, while the research community is looking forward to a possible new nuclear reactor. The basic initiatives are at a very preliminary stage: the primary choice is dual-core pool-type reactor, with a zero-power core and a separate MW-size core, cooled and moderated with light water. Such a dual-core configuration is designed to meet the varied requirements of the European Union member states. Another option would be hosting one or more micro-reactors with electrical and/or heating power producing capability that could offer stronger support toward demonstration of prototype small modular reactors in prototype future electrical grids. [ABSTRACT FROM AUTHOR]

As an alternative to research nuclear reactors, a compact accelerator-driven neutron generator that uses a lithium beam driver could be a promising candidate since it produces almost no undesired radiation. However, providing an intense lithium-ion beam has been difficult, and it has been thought that the practical application of such a device would be impossible. The most critical problem of insufficient ion fluxes has been solved by applying a direct plasma injection scheme. In this scheme, a pulsed high-density plasma from a metallic lithium foil generated by laser ablation is efficiently injected and accelerated by a radio-frequency quadrupole linear accelerator (RFQ linac). We have obtained a peak beam current of 35 mA accelerated to 1.43 MeV, which is two orders of magnitude higher than a conventional injector and accelerator system can deliver. [ABSTRACT FROM AUTHOR]

In this presented study, the cooling problem of the I.T.U. Triga Mark-II reactor has been handled and analyzed, and solutions were proposed. First of all, a thermal model of the reactor, heat exchanger, and cooling tower trio was established in the reactor. With this model, which was obtained with the help of experimental data, the parameters affecting the change of reactor water temperature over time were determined, and significant findings were obtained by investigating the possibilities of increasing the cooling power of the existing system. Then, using these mathematical equations, the effects of parameters that can affect the power of the reactor cooling system are investigated. The parameters affecting the cooling power are the cooling water flow rates in the second cooling circuits and the deposited layer that may exist as a result of numerical calculations. Different models have been created with machine learning algorithms (page regression, decision tree) to estimate the effect of the deposit layer. The mathematical and predictive models obtained with the experimental data for the heat transfer coefficient of the deposit layer, hbd, were compared. The pace regression algorithm modeled the hbd values with the least error rate (RMSE: 1.66) among the models. It has been calculated that the average tank water temperature will decrease by approximately 3.5 °C if the deposits layer is cleared. [ABSTRACT FROM AUTHOR]

The experience of conducting chemical intra-loop decontamination of individual serial and research nuclear propulsion reactors by experimental single-bath high-concentration or multiple-bath low-concentration methods is examined. It is shown that reagent-free methods make it possible to lower the level of contamination of the first loop by fragments of fuel composition remaining on the equipment and pipelines after a leaky reactor core is replaced. The obtained data give grounds for recommending that the regulatory decontamination be adjusted by expanding the collection of regular methods. [ABSTRACT FROM AUTHOR]

This report presents the methods and calculated results of the activity inventory in the structural components of the Dalat Nuclear Research Reactor (DNRR). These components include the shielding concrete, the reactor tank, and its inside irradiated facilities; the thermal and thermalizing columns; and the horizontal channels. The MCNP5 code with a three-dimensional neutron transport model was used to calculate the neutron flux distribution, neutron energy spectrum at different locations, and activation cross sections of long-lived radioactive nuclides in activated major materials, including heavy concrete, reflection graphite, and aluminum of the reactor. The calculated results of the energy spectrum and activation cross sections of MCNP5 were used in the ORIGEN2.1 point depletion code to calculate the neutron-induced activity of activated materials at different time points by modeling the irradiation history and radioactive decay. Radioactivity of long-lived key activation products such as 41Ca, 60Co, 55Fe, 63Ni, and 152Eu was modeled, and volumes of radioactive waste mainly of ordinary concrete, graphite, and aluminum in the structural components of the reactor were estimated. Experimental results of neutron flux and specific activities of some typical nuclides such as 60Co, 152Eu, 55Fe, and 63Ni in activated aluminum samples showed good agreement with the calculated results. As part of the national regulation requirements, the obtained data have been used to develop the decommissioning plan for the operational DNRR, with an estimation of about 10 years before its permanent shutdown. [ABSTRACT FROM AUTHOR]

Changes in the thermal power of a nuclear research reactor will lead to changes in experimental, irradiation, and testing conditions. Consequently, reactor core parameters are inevitably susceptible to changes. One such parameter is gamma heating (GH), which results from gamma interaction with materials. In this work, a gamma thermometer was used to measure GH over the course of 7 operational days and nights. In addition, the Monte Carlo reactor physics code Serpent-2 was used to evaluate the sensitivity of common detection methods for monitoring reactor core parameters such as neutron fluxes, GH, and gamma flux under the following conditions: reactor core power variation, reactor core fuel shuffling, and detector vicinity fuel assembly shuffling. The GH values obtained through measurements and calculations were linearly proportional to the reactor power. In addition, the Serpent-2 code for the McMaster nuclear reactor showed that despite maintaining the reactor power core at the same level, the fuel burnup distribution could alter the studied parameters. [ABSTRACT FROM AUTHOR]

To measure fuel rod swelling in a nuclear research reactor, our research has focused on the development of an optical confocal chromatic sensor, allowing contactless measurements of radial changes in fuel rods. This sensor must be able to operate in the harsh and hot environment of a reactor, especially under high neutron fluences up to $10^{19}~{n}_{\mathrm{ fast}}/\text {cm}^{2}$ and gamma doses of a few GGy. When exposed to such radiation levels, the properties of bulk optical glasses are known to be affected by two main phenomena: radiation-induced attenuation (RIA) and radiation-induced refractive index change (RIRIC). With the objective of testing glasses as candidates for a future confocal chromatic sensor, we created a dedicated setup for the online monitoring of the glasses’ RIRIC in the core reactor with a measurement principle relying on interferometry. Since not only the refractive index (RI) but also the length of the samples changes under neutron radiation through compaction, we measured the variation of the optical path (OP) ($nL$). To assess the performance of our setup, preliminary in-lab tests were performed to evaluate the feasibility of measuring the small RI changes caused by external temperature changes. Due to this study, we were able to obtain data that were not yet available in the literature regarding the temperature-induced RI variation of several bulk glasses, including radiation-hardened glasses, up to 350 °C. This information is crucial for the targeted application, as the confocal chromatic sensor will have to operate at such elevated temperatures. [ABSTRACT FROM AUTHOR]

The article explores the challenges of Micro nuclear reactors before they can be implemented. Topics include vSMR technology is in its infancy and in most cases, the designs have not yet been approved and tested, initial cost of design will be high and government funding will be imperative for the development of vSMRs.

The Institute for Hot Chemistry in Karlsruhe was founded 1959, to support the development of a reprocessing process for spent nuclear fuel. Based on the PUREX process, variants and improvements for the combination of extraction and redox steps were investigated from laboratory to near‐commercial pilot scale. By the end of nuclear reactor research, the re‐named institute for technical chemistry turned to new topics in environmental and energy related research. During the transition period, the institute participated in the GALLEX collaboration, an international basic physics experiment with the construction and operation of a radiochemical solar neutrino detector. [ABSTRACT FROM AUTHOR]

The BR2 nuclear research reactor in Mol, Belgium, runs in successive phases of operation (cycles) and shutdown, whereby a water basin surrounding the reactor vessel undergoes periodic changes in physico-chemical parameters such as flow rate, temperature, and radiation. The aim of this study was to explore the microbial community in this unique environment and to investigate its long-term dynamics using a 16S rRNA amplicon sequencing approach. Results from two sampling campaigns spanning several months showed a clear shift in community profiles: cycles were mostly dominated by two Operational Taxonomic Units (OTUs) assigned to unclassified Gammaproteobacterium and Pelomonas , whereas shutdowns were dominated by an OTU assigned to Methylobacterium. Although 1 year apart, both campaigns showed similar results, indicating that the system remained stable over this 2-year period. The community shifts were linked with changes in physico-chemical parameters by Non-metric Multidimensional Scaling (NMDS) and correlation analyses. In addition, radiation was hypothesized to cause a decrease in cell number, whereas temperature had the opposite effect. Chemoautotrophic use of H2 and dead cell recycling are proposed to be used as a strategies for nutrient retrieval in this extremely oligotrophic environment. [ABSTRACT FROM AUTHOR]

Background: The etiology of brain cancer is poorly understood. The only confirmed environmental risk factor is exposure to ionizing radiation. Because nuclear reactors emit ionizing radiation, we examined brain cancer incidence rates in the USA in relation to the presence of nuclear reactors per state. Methods: Data on brain cancer incidence rates per state for Whites by sex for three age groups (all ages, 50 and older, and under 50) were obtained from cancer registries. The location, number, and type of nuclear reactor, i.e., power or research reactor, was obtained from public sources. We examined the association between these variables using multivariate linear regression and ANOVA. Results: Brain cancer incidence rates were not associated with the number of nuclear power reactors. Conversely, incidence rates per state increased with the number of nuclear research reactors. This was significant for both sexes combined and for males in the 'all ages' category (β = 0.08, p = 0.0319 and β = 0.12, p = 0.0277, respectively), and for both sexes combined in the'50 and older' category (β = 0.18, p = 0.0163). Brain cancer incidence rates for counties with research reactors were significantly higher than the corresponding rates for their states overall (p = 0.0140). These findings were not explicable by known confounders. Conclusions: Brain cancer incidence rates are positively associated with the number of nuclear research reactors per state. These findings merit further exploration and suggest new opportunities for research in brain cancer epidemiology. [ABSTRACT FROM AUTHOR]

To monitor compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT), the International Monitoring System (IMS) is being established which will include 40 sensor systems for atmospheric xenon radioactivity. Radioactive isotopes of the noble gas xenon provide the most likely observable radioactive signatures of underground nuclear explosions. These isotopes are frequently detected by IMS noble gas systems as a result of normal operational releases from different types of nuclear facilities including nuclear power plants (NPPs), medical isotope production facilities (MIPFs), and nuclear research reactors (NRRs). Improved knowledge of the contribution of different emission sources on IMS observations strengthens the screening of radioxenon measurements to exclude observations not relevant to emissions from a nuclear explosion. The contribution of NPPs and MIPFs to the global radioxenon emission inventory is fairly well understood. NRRs have yet to be systematically assessed. This paper is the first attempt to assess the total emission inventory of NRRs expressed as annual total discharges. The results can enhance understanding of those sources most likely to impact IMS background observations and to guide future studies on contributions to IMS station background. [ABSTRACT FROM AUTHOR]