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ITER is of key importance in the European fusion roadmap as it aims to prove the scientific and technological feasibility of fusion as a future energy source. The EUROfusion consortium of labs within Europe is contributing to the preparation of ITER scientific exploitation and operation and aspires to exploit ITER outcomes in view of DEMO. The paper provides an overview of the major progress obtained recently, carried out in the frame of the new (initiated in 2021) EUROfusion work-package called ' Pr eparation of I TER O peration' (PrIO). The overview paper is directly supported by the eleven EUROfusion PrIO contributions given at the 29th Fusion Energy Conference (16–21 October 2023) London, UK [www.iaea.org/events/fec2023]. The paper covers the following topics: (i) development and validation of tools in support to ITER operation (plasma breakdown/burn-through with evolving plasma volume, new infra-red synthetic diagnostic for off-line analysis and wall monitoring using Artificial Intelligence techniques, synthetic diagnostics development, development and exploitation of multi-machine databases); (ii) R&D for the radio-frequency ITER neutral beam sources leading to long duration of negative deuterium/hydrogen ions current extraction at ELISE and participation in the neutral beam test facility with progress on the ITER source SPIDER, and, the commissioning of the 1 MV high voltage accelerator (MITICA) with lessons learned for ITER; (iii) validation of neutronic tools for ITER nuclear operation following the second JET deuterium–tritium experimental campaigns carried out in 2021 and in 2023 (neutron streaming and shutdown dose rate calculation, water activation and activated corrosion products with advanced fluid dynamic simulation; irradiation of several materials under 14.1 MeV neutron flux etc). [ABSTRACT FROM AUTHOR]

As nuclear facilities are dismantled in decommissioning, various and large amounts of waste are generated. Even more inconveniently, they are radioactive waste due to neutron activation. Thus, the neutron capture cross-sections of nuclides targeted for decommissioning are required to evaluate the radioactivity produced. The present study selected 45Sc, 63Cu, 64Zn, 109Ag and 113In among targeted nuclides, and performed the measurements of thermal-neutron capture cross-sections by an activation method at the Kyoto University Research Reactor. The thermal-neutron capture cross-sections were derived on the basis of Westcott's convention. The present results were obtained as follows: 27.18 ± 0.28 barn for 45Sc(n,γ)46Sc, 4.34 ± 0.06 barn for 63Cu(n,γ)64Cu, 0.719 ± 0.011 barn for 64Zn(n,γ)65Zn, 4.05 ± 0.05 barn for 109Ag(n,γ)110mAg and 8.53 ± 0.27 barn for 113In(n,γ)114Inm1+m2 reactions. The results for 45Sc and 64Zn support evaluated cross-section data within the limits of uncertainties, while those for the other nuclides were slightly different from evaluated ones. The obtained results are useful not only for the evaluation of isotope production, but also for the consideration of those nuclides as flux monitors, as an alternative to Au and Co. [ABSTRACT FROM AUTHOR]

AbstractWith an interest in reducing carbon emissions, Canada has committed to deploying a variety of small modular reactor (SMR) designs. This review covers Canada’s research reactor capabilities relevant to SMR development and highlights gaps in these capabilities. Following the commissioning of the subcritical facility at Ontario Tech University, Canada’s capabilities in reactor physics, education, and training will be relatively sufficient. In contrast, materials/fuel irradiation capabilities and neutron scattering techniques are insufficient, with a notable gap in in-reactor test loops needed to develop new fuels. Recommendations are made for a new multipurpose research reactor (MPRR) development and deployment of SMRs in the long term while increasing accessibility to Canada’s existing research reactors in the near term, particularly in Western Canada. [ABSTRACT FROM AUTHOR]

AbstractThis paper presents an overview of ongoing work to qualify the Advanced Test Reactor (ATR) driver fuel elements that have been affected by irradiation-degraded polymethyl methacrylate (PMMA) flux wands. Irradiation testing was performed on PMMA material in contact with aluminum clad material. The cladding was prefilmed with a boehmite oxide layer, an important feature of the ATR driver fuel. The effects on the boehmite layer due to gamma irradiation of the PMMA-aluminum clad system were investigated. PMMA embrittlement, followed by softening and degradation, occurred at high radiation levels. Adhesion between the cladding and irradiated PMMA was observed. Flow testing at prototypic ATR flow rates demonstrated the effective removal of the adhered material. Measurements of the boehmite layer thickness were performed, and Raman spectroscopy was utilized to detect the presence of boehmite in the irradiated PMMA material. [ABSTRACT FROM AUTHOR]

In recent years, various drug delivery systems circumventing the blood–brain barrier have emerged for treating brain tumors. This study aimed to improve the efficacy of brain tumor treatment in boron neutron capture therapy (BNCT) using cerebrospinal fluid (CSF) circulation to deliver boronophenylalanine (BPA) to targeted tumors. Previous experiments have demonstrated that boron accumulation in the brain cells of normal rats remains comparable to that after intravenous (IV) administration, despite BPA being administered via CSF at significantly lower doses (approximately 1/90 of IV doses). Based on these findings, BNCT was conducted on glioma model rats at the Kyoto University Research Reactor Institute (KUR), with BPA administered via CSF. This method involved implanting C6 cells into the brains of 8-week-old Wistar rats, followed by administering BPA and neutron irradiation after a 10-day period. In this study, the rats were divided into four groups: one receiving CSF administration, another receiving IV administration, and two control groups without BPA administration, with one subjected to neutron irradiation and the other not. In the CSF administration group, BPA was infused from the cisterna magna at 8 mg/kg/h for 2 h, while in the IV administration group, BPA was intravenously administered at 350 mg/kg via the tail vein over 1.5 h. Thermal neutron irradiation (5 MW) for 20 min, with an average fluence of 3.8 × 1012/cm2, was conducted at KUR's heavy water neutron irradiation facility. Subsequently, all of the rats were monitored under identical conditions for 7 days, with pre- and post-irradiation tumor size assessed through MRI and pathological examination. The results indicate a remarkable therapeutic efficacy in both BPA-administered groups (CSF and IV). Notably, the rats treated with CSF administration exhibited diminished BPA accumulation in normal tissue compared to those treated with IV administration, alongside maintaining excellent overall health. Thus, CSF-based BPA administration holds promise as a novel drug delivery mechanism in BNCT. [ABSTRACT FROM AUTHOR]

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 the research of new reactors, lithium has become a research direction as a new coolant after sodium. Lithium working fluids introduce various metallic and non-metallic impurities such as oxygen, calcium, carbon, nitrogen, etc. to a greater or lesser extent during production, filling, and even operation. In addition, the processing, welding, and cleaning steps of the pipeline and equipment in the circuit during manufacturing inevitably result in residual dirt, grease, surface oxides, and moisture. These will have a certain impact on the performance of heat transfer and circuit safety through accelerated corrosion or plugging. Therefore, in high-temperature lithium circuits, the impurity concentration of lithium working fluid must be strictly controlled and purified before use. At present, the detection methods for impurity content in lithium working fluids can be divided into two categories: sampling analysis method and online measurement method. Sampling analysis is the most direct detection and analysis method, which can directly obtain the content of various impurities in lithium working fluids, and the results are relatively accurate; The online detection method can generally only detect non-metallic impurities, but it is relatively fast and simple, and is currently a commonly used measurement method on lithium working fluid devices. The plugging meter method has been studied and used as a rather convenient online detection method. This article introduces a small, self-heating, and mobile high-temperature lithium plugging meter system to meet the online detection of different points in the lithium circuit of system engineering. [ABSTRACT FROM AUTHOR]

Aluminium alloys are used for the fabrication of the fuel clad of research-grade nuclear reactors as well as for several types of core components of high-flux research reactors. In order to carry out design and safety analysis of these components, their mechanical and fracture properties are required by the designer. In this work, experiments have been conducted on tensile specimens machined from an aluminium alloy block to evaluate the material stress-strain curve. Experiments have also been conducted on disc-shaped compact tension specimens in order to determine the fracture toughness of aluminium alloy. Numerical simulations of both tensile and fracture specimens have been carried out using the crystal plasticity model. Initially, the slip system level parameters of the crystal plasticity material model have been calibrated using experimental stress-strain data for single as well as polycrystalline aluminium. For the prediction of crack initiation toughness, Rice and Tracey's damage model has been used. The critical damage parameter has been evaluated for a fractured specimen with a crack length-to-width (a/W) ratio of 0.6. The attainment of the critical damage parameter in the analysis corresponds to the instance of experimentally observed ductile crack initiation in the specimen. Later, this model was applied to other fracture specimens with different a/W ratios with values ranging from 0.39 to 0.59. It was observed that the critical damage parameter corresponding to crack initiation in the material has a very small variation, even if the specimens have different crack lengths. It is well-known in the literature that Rice and Tracey's critical damage parameter is a material constant. Hence, we have applied the same model to predict crack initiation for single crystal fracture specimens with two different orientations of the crack plane. It was observed that the <111> orientation is more susceptible to crack initiation and propagation compared with the <100> orientation, as the damage parameter is high in the ligament of the specimen ahead of the crack tip for the same level of applied loading. As the [111] crack plane is more closely packed compared with the [100] plane, the distance between atomic planes is greater for the former, and hence, it is more susceptible to ductile damage. The results of the experiments and the material damage parameter are helpful for the integrity analysis of the fuel clad of research reactors as well as components of high-flux research reactors. [ABSTRACT FROM AUTHOR]

This article examines the impact of temperature on the steel collimator cap and the primary beam shutter. These components will be used to implement the Prompt Gamma Activation Analysis (PGAA) technique in the Moroccan TRIGA Mark-II research reactor. The steel collimator plug is essential for forming the neutron beam, while the main role of the shutter is to stop the beam when the channel is inactive. This study analyzes the effect of temperature on the collimator and shutter system, particularly focusing on the variation in maximum temperature over a month of operation with 8-h cycles per day, the behavior of temperature over 24 h, the total heat flux as a function of the length of the experimental device, the temperature distribution in mild steel (E235) and 304L stainless steel materials, and the total displacement and strain gradient as a function of temperature. All calculations were performed using COMSOL Multiphysics simulation software, based on the finite element method. [ABSTRACT FROM AUTHOR]

The recent advancements in neutron scattering technologies in China, with the development of China's first Spin-Echo Small-Angle Neutron Scattering (SESANS) spectrometer at China Mianyang Research Reactor (CMRR) and the integration of a Very Small Angle Neutron Scattering (VSANS) instrument at China Spallation Neutron Source (CSNS), have significantly bolstered the nation's scientific capabilities. This review aims to highlight the distinctive features and applications of SESANS and SANS, and to offer a valuable contribution by demonstrating how SESANS and SANS can be leveraged for the study of soft matters and solid-state materials, with a particular emphasis on the benefits of techniques such as chain labeling, contrast variation, and contrast matching. We also aim to illustrate what types of information can be gleaned from these methods. The review is structured to first introduce the general concepts of SANS and SESANS, followed by a discussion on the information these techniques can provide. Then the applications of these techniques, in combination with other techniques, on various material investigations will be demonstrated. The review concludes with a summary and future perspectives, aiming to inspire further interdisciplinary research and collaboration, and beneficial to a broader audience. [ABSTRACT FROM AUTHOR]

Several preliminary conceptual designs of nuclear thermal rocket reactor cores are presented that use tin-bonded monolithic ceramic [mononitride (UN), monocarbide (UC), and uranium dioxide (UO2)] fuel plates or pins with molybdenum-tungsten alloy clad. Neutron moderation is provided by a block of Be metal or composite materials using metal hydrides such as ZrH1.6 or YH1.6 with different matrices (MgO or Be). Mainly high-assay low-enriched uranium is considered, but highly enriched uranium is also assessed for a few configurations. Nominal core thermal power is 300 MW corresponding to about 66 kN (15 klbf) of thrust, and with minimal modifications, 500 MW may be possible (25 klbf of thrust). Depending on the configurations, the amount of 235U needed for criticality is 30 to 90 kg, and reactor weight is 2.5 to 3.8 tonnes. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Due to the diverse uses of research reactors compared to power reactors, a variety of safety aspects must be considered in their design and operation. On the other hand, due to the high age of a large percentage of these reactors, the need to update them in order to respond to the growing needs of today’s society is inevitable. One of the items that increase the features in terms of proliferation, economy, usability, security and nuclear to meet the requirements of today’s society is core conversion. The advantages and necessities of using new fuel in research reactors include the possibility of forming a more compact core, reducing operating costs, reducing security challenges, protection, environmental effects, transportation and end-of-cycle processes. In this study, the roadmap for research reactors core conversion is drawn considering all aspects of this issue which would be very useful for research reactor roadmap of any country especially our country. [ABSTRACT FROM AUTHOR]

Glomerular filtration rate (GFR) could be determined more accurately using renal positron emission tomography (PET) than conventional gamma imaging. Copper-64 [T½ = 12.7 h, Eβ+ (max) = 653 keV, β+ branching ratio = 17.8 %, 1346 keV γ-photon (0.54 %), EC (43.8 %), β− emission (38.4 %)] in the form of its hydrophilic complex with 1,4,7-triazacyclononane-1,4,7-triacetic acid (NOTA) is proposed as a potent formulation for renal PET imaging. A lyophilized kit was developed for formulation of ∼370 MBq dose of [64Cu]Cu–NOTA complex in a facile single step process using 64Cu produced by thermal neutron activation in a research reactor. The complex could be synthesized with >99 % yield and retained its integrity even when challenged by apoferritin. The rapid accumulation of [64Cu]Cu–NOTA in the kidney and clearance through urinary path was demonstrated using PET/CT imaging and ex vivo biodistribution study carried out in healthy Wistar rats to elucidate its effectiveness as a renal PET-imaging agent. [ABSTRACT FROM AUTHOR]

LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for modelling the progression of a severe accident (SA) in light-water nuclear power plants and estimating the behaviour of the reactor under SA conditions. The model was built by collecting information about the LVR-15. Since the research reactor can have different core configurations according to the location of the core components, the core configuration with the most fuel (hottest campaign K221) was selected. Then, to create the radial nodalisation, the details of the core components were obtained and grouped in five radial rings and 27 axial levels. The simulation was run with the boundary conditions collected from campaign K221, and the results were compared with the reference values of the campaign with a negligible percentage of error. For the coolant inlet and outlet temperature, the reference values were 318.18 K and 323.5 K, respectively, while for the simulation, the steady state reached 319 K for the inlet temperature and 324 K for the outlet temperature. Additionally, the cladding temperature of the hottest assembly was compared with the reference value (353.72 K) and the steady-state simulation results (362 K). In future work, different transients leading to severe accidents will be simulated. When simulating the LVR-15 reactor with MELCOR, specific attention is required for the aluminium-cladded fuel assemblies, as the model requires some assumptions to cope with the phenomenological limitations. [ABSTRACT FROM AUTHOR]

This paper presents research on the assessment of the radioecological state of plant cover surrounding two research reactor facilities located within the Semipalatinsk Test Site (STS) as examples of nuclear fuel cycle facilities (NFC). Source data on the concentrations of artificial radionuclides in the plant cover were obtained. Quantitative values for 137Cs, 241Am, and 239+240Pu activity concentrations were determined in plants across the perimeters of the facilities, indicating that these compounds may be present in the associated media from the perspective of accumulative bioindication. The values determined for artificial radionuclides in the 'soil‒plant' system around the researched NFC facilities were attributed to radioactive contamination of the STS territory. [ABSTRACT FROM AUTHOR]

One of the most comprehensive resources for thermal neutron-capture data is the Evaluated Gamma-ray Activation File (EGAF), containing data from prompt gamma activation analysis measurements carried out at the Budapest Research Reactor for 245 isotopes. Although these valuable datasets have been freely available for many years, the outdated and cryptic adopted format makes it difficult to utilize the data and it is not generally suitable for modern computational technologies. To help overcome these challenges, we have converted the datasets into an open standard JSON format. Additionally, we have developed a Python implementation of an open-source software package, pyEGAF, designed to interact with the JSON data structures for general purpose access, manipulation, and rapid assessment of the capture-gamma data in EGAF. [ABSTRACT FROM AUTHOR]

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275–300 μSv/h and 175–200 μSv/h, respectively, when the reactor is operating at 10 kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178 keV. Apart from energy reduction, 10 kW normalised neutron flux of south beam pit is also attenuated by the shield by 25 000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated. [ABSTRACT FROM AUTHOR]

The study of aluminum alloys (Al-Fe-Ni-Mg - Aluminum - Iron - Nickel - Magnesium) was carried out to develop high-density research reactor fuel cladding materials. Aluminum alloy as a fuel cladding undergoes a deformation process and heat treatment above the recrystallization temperature. Deformation and heat treatment will change in structure, microstructure, and mechanical properties. This study aims to determine aluminum alloy's phase structure, microstructure, and hardness after deformation and heat treatment. Phase structure analysis was carried out based on the X-ray diffraction pattern method. Observation of microstructural changes was carried out by metallographic-optical Scanning Electron Microscopy (SEM), and elemental analysis of alloying phase compounds was carried out by Energy Dispersive Spectroscopy - Scanning Electron Microscopy (EDS-SEM). The Vickers hardness tester tested the hardness properties of the material. The analysis of the x-ray diffraction patterns showed a tendency to form α and θ (Fe-Al³) phases in the alloy. The metallographic-optical observations showed that the microstructure changed with an increasing heating time of 3 hours. The microstructure tends to form an enlarged equiaxial grain structure with a long heating time. The results of testing the hardness properties of the Al-Fe-Ni-Mg alloy showed an increase of around 163 HV after undergoing the deformation process and heating treatment at 500 oC to 3 hours of heating time decreased around 86 HV. The changes in hardness properties of the Al-Fe-Ni-Mg alloy appear to align with changes in the microstructure. The deformation process and heat treatment impact the mechanical characteristics and microstructure of Al-Fe-Ni-Mg. [ABSTRACT FROM AUTHOR]

With the development of energy technology, hydrogen is gradually becoming widely used. China is also rapidly developing its own hydrogen energy industry in an effort to accomplish the "dual carbon target'. Hydrogen from methane is one of the mainstream sources, of which hydrogen from methane cracking is the decomposition of methane into hydrogen and solid carbon, with the solid carbon also having a high economic value. In particular, this technology does not emit carbon dioxide, which is very friendly to the environment, but it cannot avoid larger energy consumption. In contrast, hydrogen production from methane cracking driven by solar energy not only combines the characteristics of no CO 2 emission, but also greatly reduces the consumption of primary energy, which is the future development trend of hydrogen production from methane. This paper provides a brief overview of the various technological pathways for methane to hydrogen production in the context of China's actual development, focusing on the current status of the development and application of solar methane cracking to hydrogen production technologies at home and abroad. Among them, this paper mainly discusses the characteristics of solar reactors of different sizes and structures, and summarizes the results of related optimization studies. Meanwhile, the application of catalysts in solar reactors and the problem of carbon deposition are analyzed, and the practical application of solar methane cracking hydrogen production system is outlined. Finally, the future development of solar methane cracking hydrogen production technology is prospected from the national situation of China. [Display omitted] • Solar methane cracking for hydrogen can reduce emissions and energy consumption. • The current state of research on solar reactors is summarized. • Solar methane cracking requires higher performing catalysts. • Providing insights into the future development of solar methane cracking in China. [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]

Self-powered neutron detectors (SPNDs) have been utilized within in-core instrumentation to measure neutron flux for control and core flux mapping in nuclear reactors. To estimate neutron flux with SPNDs, a mathematical dynamic model correlating neutron flux and SPND material has been established. Estimation and signal compensation for neutron flux have primarily been developed using transfer-function-based methods or state-space-based methods. Particularly for the rhodium SPND, to compensate for its delayed response to incident neutron flux input, both groups of compensation methods have been widely applied. This Review details the signal compensation methods of neutron flux using transfer-function-based methods, such as those employing analog circuits, dynamic modeling of neutron flux, compensation of neutron flux, and direct inversion. In addition, signal compensation and estimation of neutron flux using state-space-based methods, such as the Kalman filter and H-infinity filter, are reviewed, along with basic calculations based on certain assumptions. Since there are differences in the characteristics of the two groups of methods for the same type of SPND, review comments are also included regarding the stability of compensation methods, based on results obtained from calculations using certain assumptions. [ABSTRACT FROM AUTHOR]

Activation of impurities in topazes happens due to irradiation with thermal neutrons (induced radioactivity occurs), which complicates their further handling. This leads to the need for their long cooling for safe use. The Institute of Nuclear Physics (Kazakhstan) is conducting R&D to develop a method for the effective formation of color centers in topaz during their irradiation in the WWR-K reactor. An irradiation capsule design has been developed in which optimized conditions for irradiating stones in the neutron field of the reactor are formed. The capsule uses shielding materials made of boron carbide and tantalum to cut off thermal neutrons, resulting in a reduction in induced radioactivity in topaz. The effectiveness of the irradiation capsule was tested in the core of the critical facility. As a result, thermal neutron flux is reduced by 5.7 times and the induced activity of the tantalum is reduced by 2.2 times. [ABSTRACT FROM AUTHOR]

This paper presents a methodology for the thermal analysis of our design of a nuclear micro-reactor, addressing the challenges of efficient heat management and safety in compact reactor designs. We introduce an approach that integrates computational fluid dynamics and thermal-hydraulic simulation, for the constraints and operational parameters of micro-reactors. Our methodology highlights the factors that influence micro-reactor thermal performance, including coolant flow rate, temperature, geometric configuration, and heat transfer. We further explore the implications of our findings for reactor design, emphasizing the importance of thermal efficiency and safety margins in the development of advanced nuclear micro-reactors. Significantly, our work not only contributes to the optimization of micro-reactor designs but also provides a framework for future research in nuclear reactor thermal analysis. The implications of our study extend to the broader field of nuclear engineering, offering valuable insights into the safe, efficient, and sustainable deployment of nuclear micro-reactors. [ABSTRACT FROM AUTHOR]

Copyright of Chinese Journal of High Pressure Physics is the property of Chinese Journal of High Pressure Physics Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

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]

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident-tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron-based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy-4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of -12551 pcm, while 310SS demonstrated the highest with a value of -17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials. [ABSTRACT FROM AUTHOR]

Thermochemical processes used to produce clean hydrogen (H 2) on a large scale by utilizing renewable energy and nuclear power sources are considered one of the most environmentally friendly strategies. In this study, the neutronic performance and hydrogen production potential of the PACER fusion blanket have been investigated during the operation period. Firstly, neutronic calculations have been performed to produce an electrical energy of 1000 MW from fusion explosions of 8.36 TJ to be repeated every 40 min with Monte Carlo Calculation Method (MCNP). Minor actinide flouride (MA) fuel out of the nuclear waste of LWRs with volume fractions of 2% have been mixed homogenously in the Flibe coolant. Tritium breeding ratio (TBR) at startup is calculated as 1.27. TBR and energy multiplication factor (M) decrease gradually. TBR >1.05 can be kept over 7 years for a self-sustained reactor. The M due to higher fissionable fuel content in the molten salt computes as 3.3 at initial. The burn up exceeds ∼800 GWd/tM at the end of operation period. Secondly, the potential of hydrogen production for four step iron chloride (Fe–Cl) and three step magnesium chloride (Mg–Cl) thermochemical cycles of the hydrogen unit PACER fusion blanket. The ratio of thermal power (1 − ψ), total thermal power (P hpf) and rate of mass flow of hydrogen ( m ˙ H 2 ) depends on M for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) thermochemical cycle have been presented. The ratio of thermal power (1 − ψ) for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) have been obtained as 0.6494, 0.131 and 0.095 whereas, total thermal power (P hpf) have been computed as 5298.8, 3608.43 and 3488.925 MW, respectively, at start up. The produced of hydrogen amount for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) have been calculated as ∼ 534.65; ∼1236.82 and ∼806.03 kg/year, respectively, at startup. • Burning value of the PACER computed as ∼800 GWd/tM. • Hydrogen production obtained as 12.82 (Fe–Cl) and 39.22 (Mg–Cl) kg/s. • Energy efficiency calculated as 45% (Fe–Cl) and 68% (Mg–Cl). [ABSTRACT FROM AUTHOR]

Canada has operated 17 research reactors at 11 different locations. The spent fuel from these research reactors differs significantly from CANDU fuel, which makes up the vast majority of spent fuel in Canada, and will eventually require disposal. The focus of this paper is to identify properties specific to Canadian research reactor fuel designs that would impact their suitability for disposal. The radionuclide inventory, hazardous chemical inventory, decay heat, residual enrichment, radiation decay rates, and gas generation of several Canadian research reactor fuel designs were simulated using the SCALE 6.2.4 software suite. The National Research Universal U3Si/Al dispersion rod, the National Research Experimental uranium metal X-rod, the Royal Military College UO2 SLOWPOKE-2 core, and the Whiteshell Reactor 1 uranium carbide bundle were investigated. Fuel burnup is the primary driver for the concentration of most radionuclides, hazardous chemicals, decay heat, and radiation decay rates. Carbon-14, chlorine-36, and mercury are driven by initial impurities in the fuel, which vary by fuel design. Low burnup, enriched fuels constitute a reasonable bounding case for the evaluation of criticality safety and proliferation risks. Canadian research reactor fuels are unlikely to present a greater risk of over pressurization from helium generation than CANDU fuel. Overall, the small volume of Canadian research reactor fuels requiring disposal is an important factor in the evaluation of disposal requirements. [ABSTRACT FROM AUTHOR]

This paper introduces a novel concept of a TRISO-fueled salt-cooled reactor (TFSCR). The core contains circulating pipes filled with molten salt carrying TRISO particles. The reactor achieves online refueling by slowly circulating the molten salt through the pipeline. The reactor utilizes the same molten salt as the coolant and graphite as the moderator. The reactor design has the characteristics of safety, economy, and nonproliferation. TRISO particles exhibit greater resistance to neutron irradiation, corrosion, oxidation, and high temperatures compared to conventional fuels. A molten salt-cooled reactor can also operate at higher temperatures, consequently enhancing power generation efficiency. Furthermore, lower operating pressures can mitigate the risk of significant damages and loss of coolant caused by accidents, thereby enhancing reactor safety. This paper presents the basic nuclear design of TFSCR under the assumptions concerning the average fuel power density, the volumetric core power density, and the core temperature. At the same time, the feasibility of online refueling and long-life operation was evaluated by fuel burnup calculation. [ABSTRACT FROM AUTHOR]

Currently, over 100 nuclear power units globally have been in operation for more than 40 years. Hindered by the limitations of computer technology at the time, these nuclear facilities lack detailed electronic drawings. Activities such as equipment replacement and process circuit system modifications during operation result in discrepancies between paper drawings and actual conditions. Given the complexity and irreversibility of nuclear facility decommissioning activities, virtual simulation technology is often employed before the decommissioning process begins to assist in designing and validating decommissioning plans. Consequently, the creation of high-precision 3D models is crucial for subsequent decommissioning designs. Through innovatively utilizing laser-scanning 3D model reconstruction technology in the reconstruction of the model of China's first heavy water research reactor undergoing decommissioning, this paper provides an overview of the process of laser-scanning 3D model reconstruction and its application in reconstructing the heavy water research reactor model. Using a 3D laser scanner, four decommissioning areas of the heavy water research reactor, including the reactor building, secondary water pump room, ventilation center, and low-level radioactive wastewater storage tank area, were subjected to 3D laser scanning. The acquired point cloud data from 572 scanning stations were processed using point cloud processing software for denoising, stitching, and triangulation. The triangulated model was then imported into modeling software for 3D reconstruction, ultimately establishing a digitalized model of the heavy water research reactor suitable for subsequent decommissioning simulation and design. [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]

This article describes the preparation of 99Mo isotope using the direct nuclear reaction 98Mo(n,γ)99Mo by neutrons of the WWR-K reactor using molybdenum (VI) oxide of natural and enriched composition, according to the isotope 98Mo (98.66%) as a target. The possibility of using enriched 98Mo oxide to produce 99Mo was considered. The estimation of the maximum specific activity of 99Mo was obtained in a comparative manner under the same irradiation conditions at the reactor. [ABSTRACT FROM AUTHOR]

Rock physics plays an important role in the oil, gas, and water industries by providing essential data for reservoir management. This study focuses on predicting Shear Wave Velocity (Vs), a key geophysical parameter, using the Deep Belief Network (DBN) and Random Forest (RF) algorithms. The dataset is divided for training, validation, and testing, revealing that the DBN model surpasses the RF and empirical models in accuracy. With a low RMSE value of 0.0398 km/s in testing and an R2 of 0.9775, the DBN model demonstrates precision in Vs prediction. The analysis identifies specific parameters—Measure Depth (MD), Caliper (CALL), Gamma Ray (GR), Shallow Electrical Resistivity (RES-S), Medium Electrical Resistivity (RES-M), Deep Electrical Resistivity (RES-D), Compressional Wave Velocity (DCT), Neutron Porosity (NPHI), and Photoelectric Coefficient Index (PEF)—as significantly influencing Vs, while CALI and GR have a relatively lesser effect. The results of the Vs prediction indicated that the DBN algorithm excels in deep feature extraction, unsupervised pretraining, and efficient handling of complex data. Ultimately, the DBN algorithm proves to be a superior and time-saving alternative for Vs prediction and analysis in subsurface reservoirs. • Rock physics aids oil, gas, and water industries through crucial reservoir data. • Deep Belief Network (DBN) excels in Vs prediction with superior performance. • DBN model achieves remarkable precision with a low RMSE of 0.0398 km/s. • Correlation analysis identifies key parameters influencing shear wave velocity (Vs). • DBN algorithm proves a time-saving, reliable alternative for Vs prediction in subsurface reservoirs. [ABSTRACT FROM AUTHOR]

The CO2 electrochemical reduction reaction (CO2ERR) is heralded for carbon dioxide and renewable energy utilization. However, complexities in catalyst optimization and reactor structure research hinder its practical application. Herein, rather than traditional catalyst optimization, emphasis is placed on refining the reactor structure to enhance gas–liquid mixing. The goal is to raise the CO2 concentration in the reaction zone and extend its residence refining CO2ERR. Building on prior reactor designs, this work introduces the N‐reactor (the nozzle‐type reactor) with a ring electrode suited for gas–liquid mixing and reaction interplay. Through finite element simulations using the gas–liquid flow model, compared with the S‐reactors (the wide‐straight‐type reactor and the narrow‐straight‐type reactor), the N‐reactor's ring electrode shows a 12.99% CO2 concentration rise in the electrode zone and a 67.11% surge at the cathode exit. As a consequence, the total concentration of the product methanol is increased by 6.37%, with a maximum concentration increase of 26.96% and a concentration increase of 49.08% at the cathode outlet. These results validate the feasibility of optimizing the reaction from the perspective of gas–liquid mixing flow and provide novel methods and ideas for further optimization of CO2ERR, contributing to the practical application of the technology. [ABSTRACT FROM AUTHOR]

Determination of Rare Earth Elements (REEs) using spectroscopic methods is challenging. Conventional Instrumental Neutron Activation Analysis (INAA) method utilizing Pneumatic Carrier Facility (PCF) of Dhruva research reactor has been optimized with the short‐lived isotopes such as 161Gd, 165mDy and 153/155Sm for quantification of Gd, Dy and Sm respectively. Such isotopes are difficult to observe by conventional INAA with longer cooling periods owing to their very short half‐lives (1–3 mins). In this work, such isotopes are not only observed but also utilized as activation product for rapid determination of these REEs. The standardized method has been utilized for the first time in liquid samples involving nuclear fuel like U3O8 wherein TBP/n‐dodecane has been applied for chemical separation of Uranium prior to analysis. The INAA method has also been applied for validating the stoichiometry of Gd in Gd3+ co‐doped Zn2P2O7:Tb3+ green emitting solid‐state phosphors, which plays a very pivotal role in sensitizing the terbium ion in its performance for solid state lighting. We believe with huge current demand of REEs in research, science and technology; INAA can be a game changer wherever needed for quantification with high precision and accuracy and can also be extended to other REEs of interest. [ABSTRACT FROM AUTHOR]

Featured Application: Proposed research could be useful in condition monitoring activities for neutron detectors and neutron instrumentation systems. Nuclear power plants (NPPs) are globally important sources of clean energy. However, high operation and maintenance (O&M) costs are one of the factors hindering the development of NPPs. In this study, we focused on the development of a digital algorithm for ex-core instrumentation systems (EISs) to reduce O&M costs, given that the digitalization of EISs is not progressing compared with other instrumentation and control systems in NPPs. Specifically, we developed a digital algorithm for pulse signal processing in EISs, which traditionally require a significant amount of hardware and maintenance efforts. Our purposes were to simplify the configuration and reduce the O&M costs associated with pulse signal processing. To validate the algorithm, we used a detector emulator and a research reactor. Additionally, we explored a new approach to reduce the workload associated with a discrimination characteristics test (DCT) in EISs. The results of this study demonstrated that the proposed algorithm is effective in achieving good linearity within ±5% of the span and response performance with a delay time of 5 ms, which are required for EISs. Furthermore, the proposed method for the DCT shows promising results in reducing the time required for the test compared with conventional methods. [ABSTRACT FROM AUTHOR]

The PIK research reactor includes circuits of heavy-water reflector and liquid regulation representing systems important for nuclear safety. Therefore, the issue of organizing the optimal water-chemical mode and chemical control of coolants for these two systems is relevant. The volume of chemical control for coolants of heavy water circuits was developed taking into account existing experience and the physicochemical characteristics of heavy water. In order to return heavy water to the heavy-water reflector circuit and reduce the dose load on laboratory personnel, a unique automated system of sampling and analysis was created for online monitoring of heavy water in the heavy-water reflector of the PIK reactor. [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]

Production processes become more and more complex in terms of engineering implementation. During the operation of equipment for nuclear power plants or research reactor facilities, it is advisable to consider big data volumes, their change dynamics and deviations from the normal state. The IRT-T reactor facility at the School of Nuclear Science and Technology in National Research Tomsk Polytechnic University, is used for machine learning. [ABSTRACT FROM AUTHOR]

In consultation with The Department of State and the USAEC, and other state and non-state actors – including the Atomic Industrial Forum Inc. and the Fund for Peaceful Atomic Development – organized the first international missions to investigate if and how nuclear technologies could be mobilized as commodities and symbols of possible futures to the rest of the world, with a special focus on underdeveloped countries. Nevertheless, they soon realized that selling atomic development required a set of skills, tools, and knowledge not found in the usual and tested aid programs for agriculture and public health. Based on John Krige's reflections on the co-production of knowledge and the writing of transnational history, we will explore how knowledge for nuclear development was co-produced in countries like Egypt, Turkey, Mexico, and the United States. [ABSTRACT FROM AUTHOR]

Understanding the behaviour of a neutron spectrum with burnup is important for describing various phenomena associated with reactor operation. The quest to understand the neutron spectrum comes with a lot of questions. One question that is usually asked by students is: Does the neutron spectrum harden or soften with burnup? Most textbooks used by students do not provide a definite answer to this question. This paper seeks to answer this question using a 3D model of a standard 17 × 17 pressurised water reactor fuel assembly. Two cases were studied using the Serpent Monte Carlo code: the first considered the fuel assembly with constant boron concentration (traditionally found in many published papers), and the second considered boron iteration (where the boron concentration was reduced with burnup). Neutron spectra for the two cases at beginning of life and end of life were compared for spectral shifts. In addition, thermal spectral indices were used to assess spectrum hardening or softening with burnup. Spectral shifts to lower energies were observed in the thermal region of the neutron spectrum, whereas the fast region experienced no spectral shift. There was an increase in thermal spectral indices indicating that the spectrum became soft with burnup. [ABSTRACT FROM AUTHOR]

Control rod calibration experiment results for the Oregon State TRIGA® Reactor (OSTR) immediately following LEU conversion in 2008, and MCNP® 5 predicted rod worths from the 2008 LEU Conversion Safety Analysis Report (CSAR) are discussed. The reactivity worth of the four OSTR control rods is measured using the rod-pull method. Reactor power and period measurements in this method rely on the fission chamber power detector on the north side of the reflector. It is proposed that the location of the fission chamber and the neutron flux distribution in the core may result in an inaccurate reactor period measurement due to the asymmetry of the neutron flux distribution in the OSTR core. The asymmetry of the flux is believed to be more pronounced during super-criticality, resulting in errors in the time-of-power-rise measurements. As a result, control rod calibration experiments may under-predict or over-predict the reactivity worth of certain control rods. A time-independent Monte–Carlo method for the quantification of these effects is presented. Thermal flux maps at the core axial mid-plane are obtained from the model to inform discrepancies between predicted and observed results. [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]

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Providing sufficient energy in emergency situations in nuclear power plants is of great importance due to their crucial role in improving power plant safety and eliminating disasters. Solar power is capable of providing a secure energy source which could be used for emergency power loads. The main goal of this study is application of solar energy for providing emergency loads of the Tehran research reactor. A 100 kW solar power plant connected to the grid and 400 kWh battery pack which is capable of providing 50 kw loads for 8 h is designed. As the first priority, the plant charges the battery pack and discharges excess energy to the power grid. Therefore, the proposed scheme, in addition to increasing safety, also has economic benefits. Design procedure of the power plant was carried out using PVsyst which is international well known software for designing solar power plants and results were investigated. Economic evaluation of the project is carried out using RETSCREEN software and different schemes are evaluated. Due to the critical nature of the power loads and their sensitivity, application of this scheme for providing emergency energy loads of the Tehran research reactor is suggested. Considering the expansion of nuclear safety and the protection of the environment and humanity, we request the respected editors and reviewers of the journal to have positive opinions about the publication of the article. [ABSTRACT FROM AUTHOR]