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Conventional fluids used in fission‐based water‐cooled nuclear reactors have lower heat transfer coefficients (HTCs) and thermal conductivity, which has led researchers to explore high‐performance fluids that can enhance heat transfer in routine operation and prevent core meltdown in the case of accidents. It is important to investigate a wide range of fluids that can help designers improve thermal hydraulic characteristics, such as HTC, critical heat flux, and minimum departure from nucleate boiling ratio (MDNBR). In this study, the effectiveness of nanofluids in enhancing heat transfer parameters, including thermal conductivity and heat capacity, was investigated. Four different nanofluids (Al2O3–H2O, ZrO2–H2O, Ag–H2O, and Si–H2O) with pure water as the primary coolant in an HPR‐1000 nuclear reactor were compared using computational methods. Due to computational limitations, only the flow channel among four fuel rods with the highest power density in the core was simulated using Eulerian computational fluid dynamics. The results of this study show that silver water (Ag–H2O) nanofluid outperformed other nanofluids and pure water. It had a higher average HTC and MDNBR, with a 67.15 % and 45.23 % improvement, respectively, compared to pure water. The fuel rod wall temperature was also reduced by 28.5 K with Ag–H2O compared to water. Comparison of current simulated results with literature data shows a good agreement. [ABSTRACT FROM AUTHOR]

A heated junction thermocouple (HJTC) is a device designed to monitor the water level of nuclear reactors to prevent explosions due to overheating. This device consists of heated and non-heated thermocouples. When heat is applied, the temperature rise in HJTC varies depending on whether the device is submerged in water or exposed to air. In this study, we analyzed the thermal characteristics of HJTC through 3D COMSOL simulation. The temperature profile of the cross-section was extracted based on three design parameters: applied heat, heating length of the wound coil, and diameter of the HJTC. The temperature difference between the two thermocouples was more pronounced upon exposure to air. Meanwhile, high applied heat, short coil heating length, and small diameter of the HJTC contributed to a more pronounced temperature difference based on the surrounding environment. These results can aid in the appropriate HJTC design to achieve the desired detection sensitivity. [ABSTRACT FROM AUTHOR]

Currently, the properties of molten lithium, sodium and potassium fluoride eutectic mixtures with different additions are immensely important for the development of molten salt nuclear reactors. In the present work, the density of molten FLiNaK mixtures with additions of neodymium fluoride was studied by the Archimedean method. The neodymium fluoride addition increased the density of the 46.5 mol% LiF–11.5 mol % NaF–42.0 mol % KF (FLiNaK) and FLiNaK + 25 mol% NdF3 mixture from 2.00 g⋅cm−3 to 3.25 g⋅cm−3, respectively. Thermal diffusivity was measured by the laser flash method. It was found to decrease abruptly as the NdF3 concentration increased. Thermal conductivity of the FLiNaK–NdF3 system, which was calculated using thermal diffusivity, density and heat capacity values, was lower than that of molten FLiNaK at the same temperature. The composition with 25 mol % NdF3 (0.69 W⋅m−1⋅K−1) had a lower value of thermal conductivity than molten FLiNaK without additions (0.74 W⋅m−1⋅K−1) at the same temperature of at 973 K. It can be concluded that neodymium fluoride additions resulted in the density growth and decrease in the thermal diffusivity, heat capacity and thermal conductivity of molten FLiNaK. The change in the neodymium fluoride concentration can affect the technological process in nuclear reactor. [ABSTRACT FROM AUTHOR]

The main features of the loading pattern optimization (LPO) problem, such as high-dimensionality, multi-modality, and non-linearity, make it difficult to achieve a truly optimal configuration. In recent years, metaheuristic methods have been successfully used to solve this problem. In this research, a discrete golden eagle optimization (DGEO) algorithm has been developed to solve the LPO problem in the first cycle of VVER-1000 reactor core. To evaluate the proposed algorithm, a linear multi-purpose fitness function has been used to improve the safety parameters of the reactor core by obtaining a flatter power distribution during the first cycle, and also to enhance the economic parameters by increasing the cycle length and reducing the cost of fuel recycling. For this purpose, a FORTRAN program has been written to map the DGEO algorithm for the LPO problem using the PMAX and PARCS core calculation code to compute the fitness function in each iteration. To speed up the calculations, parallel computing has been applied in the written program. The results demonstrated that the loading pattern, which is suggested by the DGEO algorithm, enhances all the safety and economic parameters in the fitness function. Thus, the DGEO algorithm is highly reliable for the LPO problems in the VVER 1000 reactor core.

Nuclear reactors, as a class, exhibit nonlinear and higher-order system characteristics, posing a consistent challenge for researchers in the design of effective controllers. The specific focus of this study is on the Pressurized Heavy Water Reactor (PHWR), a representative example of such intricate systems. Given the inherent complexity of higher-order system dynamics, this work tackles the challenge by employing a reduced-order modeling approach, capturing the essence of the original system's behavior. In the context of this research, a novel approach is taken in the design of a controller for the PHWR system. The method involves the use of an optimization-based Fractional Order Proportional Integral Derivative (FOPID) controller tailored for the lower-order model of the PHWR. The reduced-order model is derived through the application of the Balanced Truncation method, which enables the creation of a simplified yet representative model that faithfully emulates the behavior of the original higher-order system. The optimization of the FOPID controller parameters is achieved through the adoption of the Grey Wolf Optimization (GWO) algorithm. To substantiate the efficacy of the proposed controller, a comprehensive performance analysis is conducted. Various performance indices are employed to evaluate the controller's robustness and overall effectiveness in regulating the PHWR system.

Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors. [ABSTRACT FROM AUTHOR]

AbstractA virtual reality learning module to train nuclear engineering students in reactor operations to understand reactor power excursions has been developed. The learning module was taught with an Oculus-2 headset and controllers (now called Meta Quest 2). The class was comprised of 71 undergraduate students, mostly in their fourth year of the nuclear engineering curriculum at Texas A&M University. The learning module simulation of power excursion, called pulsing the reactor, was modeled after the Texas A&M Engineering Experiment Station TRIGA reactor. First, the students visited the TRIGA reactor for pulsing and answered a technical quiz on the subject. Next, the students performed pulsing in the equivalent virtual reality module developed in this work.One of the primary learning objectives in the laboratory exercise was the role of passive and active safety mechanisms in a rapid reactivity insertion and power excursion. Data from the actual reactor visit showed that most students did not understand a key passive safety mechanism during the reactor visit. However, the students showed a notable improvement in their understanding of the safety mechanisms after the virtual reality reactor visit.When asked if the virtual reality learning module would have made the quiz at the reactor easier, 96% of the students reported that at least one of the quiz questions would be have been better answerable with the virtual reality module. Students also noted that the virtual reality module needed to expand its scope to include more details and teaching components. Although most students were reluctant to completely replace the pulsing reactor visit with its virtual reality module version available at the time of the study, they appreciated it as a learning reinforcement tool. Student opinion may change more favorably in the future with continued improvements and enhancements of the module. [ABSTRACT FROM AUTHOR]

Autonomous control systems can enhance the economic viability of complex systems such as nuclear microreactors. Advanced forms of these control systems benefit from techniques for high-resolution field data reconstruction from sensor data. Basis projection methods present a useful set of methodologies for nonparametric reconstruction of these field data. In this work, high fidelity multiphysics simulations are used to generate microreactor temperature distributions that are then used to evaluate the viability of a basis projection method which employs a generalized set of basis functions, or dictionaries. Additionally, metrics for evaluating the accuracy of the temperature distribution are presented that are motivated both from the fields of traditional signal processing and statistics. The performance of a method based on matching pursuits is compared to a method that uses the sequential ordering of basis functions. The matching pursuits-based method gives improved performance for highly sparse representations of temperature distributions. Additionally, accuracy is analyzed for multiple temperature distributions which represent different potential reactor operation configurations. Overall, the results from this work should be considered as an upper bound on the performance of basis projection methods which employ generalized basis functions for real time reactor control. • Autonomous systems enhance nuclear microreactor economic viability. • Basis projection methods aid in nonparametric field data reconstruction. • Matching pursuits-based method outperforms sequential basis function method. • Theoretical foundation provided for evaluating alternative bases. • Generalized bases evaluated for nuclear reactor field reconstruction. [ABSTRACT FROM AUTHOR]

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131I is a fission product produced in a nuclear reactor by irradiating tellurium dioxide, with a half-life of 8.02 day. The most important and widely used method for making 131I is irradiation using a nuclear reactor and post-irradiation followed by dry distillation. The advantage of the dry distillation process is that the process and the equipment are relatively simple, namely TeO2 (m.p. 750 °C), which can withstand heating during reactor irradiation. Based on TeO2 irradiation by neutron following the technique of dry distillation was explained for production of 131I on a large scale. A dry distillation followed the radioisotope production operation using the 30 MW GA Siwabessy nuclear reactor to meet national demand. TeO2 targets are 25 and 50 g irradiated for 87–100 h. The resulting 131I activity is 20.29339–368.50335GBq. According to the requirements imposed on the radionuclide purity of the preparation, the contribution of 131I training in the resulting preparation was not less than 99.9 %

This paper explores the vibrational, thermophysical, and structural properties of thorium nitride using density functional theory. An agreement is observed between the calculated and experimental properties of the lattice constant and bulk modulus. In the density functional perturbation theory, diagrams of phonon spectrums and vibrational densities of states along high symmetry paths are calculated. Based on the phonon dispersion diagram, no imaginary frequencies are found, indicating that the crystalline structure is dynamically stable. The compound also exhibits a phonon gap in the range 154-300 cm-1. Under high temperature and pressure, quasi-harmonic Debye models are used to evaluate thermodynamic properties such as Debye temperature, thermal expansion coefficient, entropy isothermal bulk modulus, and vibrational specific heat capacity. As the temperature increases, the volume of the system at a constant pressure decreases, while it increases for all pressures at a constant temperature. As temperature increases at a constant pressure, the coefficient of thermal expansion increases, indicating that the crystalline lattice is transferring more heat.

Abstract This study introduces a novel method for enhancing Boiling Water Reactor (BWR) operation simulations by integrating machine learning (ML) models with conventional simulation techniques. The ML model is trained to identify and correct errors in low-fidelity simulation outputs, traditionally derived from core physics computations. These corrections aim to align the low-fidelity results closely with high-fidelity data. Precise predictions of nuclear reactor parameters like core eigenvalue and power distribution are crucial for efficient fuel management and adherence to technical specifications. Current high-fidelity transport calculations, while accurate, are impractical for real-time predictions due to extensive computational demands. Our approach, therefore, utilizes the standard two-step simulation process-assembly-level lattice physics calculations followed by whole-core nodal diffusion computations-to generate initial results, which are then refined using the ML-based error correction model. The methodology focuses on improving simulation accuracy in regular BWR operations rather than developing a universal ML predictor for reactor physics. By training an advanced neural network model on the difference in high-fidelity and low-fidelity simulations, the model can reduce the nodal power error from low-fidelity simulations to around 1% on average and the core eigenvalue down to under 100 pcm. This result is under the condition of the normal variations of control rod pattern and core flow rate changes in standard BWR operations used in the training and evaluation of the machine learning model. This work suggests a promising approach for achieving more accurate, computationally feasible simulation solutions in nuclear reactor operation and management.

Nuclear reactors, as a class, exhibit nonlinear and higher-order system characteristics, posing a consistent challenge for researchers in the design of effective controllers. The specific focus of this study is on the Pressurized Heavy Water Reactor (PHWR), a representative example of such intricate systems. Given the inherent complexity of higher-order system dynamics, this work tackles the challenge by employing a reduced-order modeling approach, capturing the essence of the original system's behavior. In the context of this research, a novel approach is taken in the design of a controller for the PHWR system. The method involves the use of an optimization-based Fractional Order Proportional Integral Derivative (FOPID) controller tailored for the lower-order model of the PHWR. The reduced-order model is derived through the application of the Balanced Truncation method, which enables the creation of a simplified yet representative model that faithfully emulates the behavior of the original higher-order system. The optimization of the FOPID controller parameters is achieved through the adoption of the Grey Wolf Optimization (GWO) algorithm. To substantiate the efficacy of the proposed controller, a comprehensive performance analysis is conducted. Various performance indices are employed to evaluate the controller's robustness and overall effectiveness in regulating the PHWR system. [ABSTRACT FROM AUTHOR]

The results of experimental studies on attenuating neutron and photon radiation in a nuclear reactor using a protective metal hydride composite are presented. The distribution profile of the dose and spatial energy is obtained for primary and secondary gamma radiation. It is shown that the dose of gamma radiation behind the protection is associated with capturing gamma rays generated in the initial layer of the material. Based on the results obtained, the calculation model of the experiment is verified in the two-dimensional geometry for the properties of the material protecting the reactor. The verification involves the discrete ordinate method based on the DORT package. Deviations between the calculated and experimental values of the fast neutron relaxation lengths do not exceed 5%. The same deviations obtained for gamma radiation are under 7%, which confirms the validity of the calculation technique and the potential to apply the data obtained to proceed with designing radiation protection based on the metal hydride composite. [ABSTRACT FROM AUTHOR]

This study introduces a novel method for enhancing Boiling Water Reactor (BWR) operation simulations by integrating machine learning (ML) models with conventional simulation techniques. The ML model is trained to identify and correct errors in low-fidelity simulation outputs, traditionally derived from core physics computations. These corrections aim to align the low-fidelity results closely with high-fidelity data. Precise predictions of nuclear reactor parameters like core eigenvalue and power distribution are crucial for efficient fuel management and adherence to technical specifications. Current high-fidelity transport calculations, while accurate, are impractical for real-time predictions due to extensive computational demands. Our approach, therefore, utilizes the standard two-step simulation process-assembly-level lattice physics calculations followed by whole-core nodal diffusion computations-to generate initial results, which are then refined using the ML-based error correction model. The methodology focuses on improving simulation accuracy in regular BWR operations rather than developing a universal ML predictor for reactor physics. By training an advanced neural network model on the difference in high-fidelity and low-fidelity simulations, the model can reduce the nodal power error from low-fidelity simulations to around 1% on average and the core eigenvalue down to under 100 pcm. This result is under the condition of the normal variations of control rod pattern and core flow rate changes in standard BWR operations used in the training and evaluation of the machine learning model. This work suggests a promising approach for achieving more accurate, computationally feasible simulation solutions in nuclear reactor operation and management. [ABSTRACT FROM AUTHOR]

This paper proposed a modeling method of the thermo-fluid-solid coupling simulation for the hot water pipeline with the insulation layer outside in the nuclear reactor. The temperature and stress distributions are analyzed and compared with the tests; the effects of the insulation layer thickness and support interval are investigated. The maximum stress occurs at the support points. The maximum stress increases approximately linearly with the increase of the support interval. The maximum deflection occurs at the middle of the two supports. The maximum deflection increases rapidly with the increase of the support interval, which is approximately in a parabolic relationship. The axial elongation ratio increases with the insulation layer thickness. The simulation results agree with the test results very well with regard of both stress and temperature distributions. The simulation method can also be used for the stress and temperature analysis in various cases, such as chemical-vessel and gas-oil pipeline with various thermal fluids inside. [ABSTRACT FROM AUTHOR]

This essay analyzes the historiography of the German attempts during the Second World War to develop and harness the economic and military applications of uranium fission as well as the postwar "politics of the past." This literature, which began immediately after the end of the war and continues to the present day, includes contributions from scientists, historians, journalists, playwrights, and other authors. The essay is divided into mostly chronological sections. [ABSTRACT FROM AUTHOR]

Abstract Background Silver-111 is a promising β −-emitting radioisotope with ideal characteristics for targeted radionuclide therapy and associated single photon emission tomography imaging. Its decay properties closely resemble the clinically established lutetium-177, making it an attractive candidate for therapeutic applications. In addition, the clinical value of silver-111 is further enhanced by the existence of the positron-emitting counterpart silver-103, thus imparting a truly theranostic potential to this element. A so-fitting matching pair could potentially overcome the current limitations associated with the forced use of chemically different isotopes as imaging surrogates of lutetium-177, leading to more accurate and efficient diagnosis and treatment. However, the use of silver-111-based radiopharmaceuticals in vivo has faced obstacles due to the challenges related to its production and radiochemical separation from the target material. To address these issues, this study aims to implement a chromatographic separation methodology for the purification of reactor-produced silver-111. The ultimate goal is to achieve a ready-to-use formulation for the direct radiolabeling of tumour-seeking biomolecules. Results A two-step sequence chromatographic process was validated for cold Ag-Pd separation and then translated to the radioactive counterpart. Silver-111 was produced via the 110Pd(n,γ)111Pd nuclear reaction on a natural palladium target and the subsequent β −-decay of palladium-111. Silver-111 was chemically separated from the metallic target via the implemented chromatographic process by using commercially available LN and TK200 resins. The effectiveness of the separations was assessed by inductively coupled plasma optical emission spectroscopy and γ-spectrometry, respectively, and the Ag+ retrieval was afforded in pure water. Recovery of silver-111 was > 90% with a radionuclidic purity > 99% and a separation factor of around 4.21·10−4. Conclusions The developed separation method was suitable to obtain silver-111 with high molar activity in a ready-to-use water-based formulation that can be directly employed for the labeling of radiotracers. By successfully establishing a robust and efficient production and purification method for silver-111, this research paves the way for its wider application in targeted radionuclide therapy and precision imaging.

We conducted absorption experiments on Ti using hydrogen isotopes, protium and deuterium, to gain insights into the kinetics of hydrogen atoms, particularly across the Ti surface. Rather than relying on various surface pretreatments, we identified the initial temperature of Ti in vacuum as the determining factor in achieving a high absorption rate and content. Specifically, we found that the initial Ti temperatures of approximately 900 and 980 °C are optimal for protium and deuterium, respectively, likely due to sufficient removal of the native surface oxide on Ti. The dependence of the absorption rate and amount of hydrogen isotopes on the temperature during absorption was partially explained by the phase transformation and exothermicity of the hydrogenation reaction. Moreover, this dependence strongly indicated the existence of a tunneling transport process. Additionally, we developed a kinetic model for hydrogen isotope transport in Ti to conduct numerical simulations. The absorption curves, calculated using the developed model, closely matched a series of experimental curves obtained at different temperatures. These curves were from an identical set of equations and kinetic parameters for both protium and deuterium. Our numerical kinetic model has the potential to serve as a valuable simulation tool for various applications, such as the design of high-performance hydrogen storage systems and maintenance strategies against hydrogen embrittlement of structural materials. • Protium and deuterium absorption experiments in Ti performed. • The initial Ti temperature in vacuum determines high absorption rate and content. • The initial Ti temperatures of 900 (protium) and 980 °C (deuterium) are optimal. • Temperature dependence indicates the existence of tunneling transport process. • A kinetic model of hydrogen isotope transport for numerical simulations presented. [ABSTRACT FROM AUTHOR]

The results of experimental studies of the hydrodynamics of a coolant in the outlet section of a cassette fuel assembly of the RITM reactor of a low-power nuclear power plant are presented. The purpose of the work was to study the redistribution of axial velocity and coolant flow rate at the outlet of the fuel rod bundle, behind the fuel assembly cap, near the coolant extraction pipe, and in the holes of the upper support plate. To achieve the goal, experiments were carried out on a research stand with an air working medium on a model of the outlet section of the fuel assembly, including the outlet fragment of the fuel rod bundle, the cap, the upper support plate with the coolant extraction pipe. When conducting research, the pneumometric method and the method of injection of a contrast impurity were used. An area covering the entire cross section of the model was chosen as the study area. The picture of the coolant flow is represented by cartograms of the distribution of axial velocity and coolant flow rate, as well as cartograms of the distribution of the contrast impurity. The experimental results can be used in the design of new active sections of RITM reactors. The resulting experimental base can be used for local validation of CFD programs and onedimensional thermal-hydraulic codes used to substantiate the thermal reliability of cores. [ABSTRACT FROM AUTHOR]

Background: Silver-111 is a promising β−-emitting radioisotope with ideal characteristics for targeted radionuclide therapy and associated single photon emission tomography imaging. Its decay properties closely resemble the clinically established lutetium-177, making it an attractive candidate for therapeutic applications. In addition, the clinical value of silver-111 is further enhanced by the existence of the positron-emitting counterpart silver-103, thus imparting a truly theranostic potential to this element. A so-fitting matching pair could potentially overcome the current limitations associated with the forced use of chemically different isotopes as imaging surrogates of lutetium-177, leading to more accurate and efficient diagnosis and treatment. However, the use of silver-111-based radiopharmaceuticals in vivo has faced obstacles due to the challenges related to its production and radiochemical separation from the target material. To address these issues, this study aims to implement a chromatographic separation methodology for the purification of reactor-produced silver-111. The ultimate goal is to achieve a ready-to-use formulation for the direct radiolabeling of tumour-seeking biomolecules. Results: A two-step sequence chromatographic process was validated for cold Ag-Pd separation and then translated to the radioactive counterpart. Silver-111 was produced via the 110Pd(n,γ)111Pd nuclear reaction on a natural palladium target and the subsequent β−-decay of palladium-111. Silver-111 was chemically separated from the metallic target via the implemented chromatographic process by using commercially available LN and TK200 resins. The effectiveness of the separations was assessed by inductively coupled plasma optical emission spectroscopy and γ-spectrometry, respectively, and the Ag+ retrieval was afforded in pure water. Recovery of silver-111 was > 90% with a radionuclidic purity > 99% and a separation factor of around 4.21·10−4. Conclusions: The developed separation method was suitable to obtain silver-111 with high molar activity in a ready-to-use water-based formulation that can be directly employed for the labeling of radiotracers. By successfully establishing a robust and efficient production and purification method for silver-111, this research paves the way for its wider application in targeted radionuclide therapy and precision imaging. [ABSTRACT FROM AUTHOR]

Mono-energetic γ -beams (Δ ∼ 1 0 eV) based on thermal neutron capture, in a nuclear reactor, using the Mn(n , γ) reaction were utilized for generating a fast neutron source from Zinc, via the 6 7 Zn(γ , n) reaction. One of the incident γ -lines of the Mn source at E γ = 7 2 4 4 keV, photoexcites by chance a resonance level in 6 7 Zn, with subsequent emission of neutrons at an energy of 191 keV. The cross-section for this process was measured and found to be σ (γ , n) = 2 5 2 ± 4 1 mb with an intensity of the order of 104 n/s. The angular distribution of the 191 keV neutron group was also measured. [ABSTRACT FROM AUTHOR]

The DANSS detector, i.e., a scintillation detector with a sensitive volume of 1 m3, is installed at the Kalinin nuclear power plant under the reactor core on a mobile stage. The distance from the detector center to the reactor center is varied from 10.9 to 12.9 m. The detector records about 5000 events of inverse beta decay per day a background level from cosmic muons of the order of two percent. The analysis on the search for sterile neutrino taking into account the count rate of antineutrino events is described. The final ratio of the observed rate of absolute DANSS counting to that predicted based on the Huber and Mueller model was 0.98 ± 0.04. New results on the search for sterile neutrino based on 7 million of antineutrino events are presented. When using the conservative estimate of 7% for the total systematic uncertainty in the absolute count rate, the limits in the parameter space of the hypothetical sterile neutrino are determined. In particular, for high (≳10 eV2) parameters , the values of are excluded at the 90% confidence level. [ABSTRACT FROM AUTHOR]

Summary: Radiopharmaceutical research and bioscience or medical applications have changed during the decades in the function of available technologies for radioisotope production and imaging detection technologies. In the 1940s to 1960s year, radioisotopes were supplied by reactor production, such as 131I for thyroid uptake study, 197Hg, 203Hg-chlormerodrin for brain scans, 198Au-colloid for liver scans, and 85Sr-chloride for the bone scans. In the late 1960s, with the development of the gamma camera by Hal Anger, and the 99Mo/99mTc generators by researchers of the Brookhaven National Laboratory, the focus was moved to research and diagnostic application of 99mTc radiopharmaceuticals, and these compounds have been in use until now. In that age, the chemistry of technetium, including the production of different chelators and radiolabeling processes, were developed. Cyclotron-producing radioisotopes for medical application emerged in 1955, and for a long time, it was used to produce gamma-emitting radionuclides such as 67Ga, 111In, and 201Tl. After the development of PET câmera around 1975, positron-emitting radioisotopes started to be produced routinely in cyclotrons, such as 11C, 13N, and 18F, gaining market status for diagnostic applications after Alfred Wolf and colleagues developed the 18F-FDG, around 1978. In the following year, positron-emitting radiometals 68Ga, 64Cu, and 89Zr, began to be produced on a large scale to label molecules with short or long circulation times in the body. The results in the specificity of the diagnostic opened the possibility of using 90Y and 177Lu, both beta-emitting radioisotopes, to label the same molecules for therapeutic purposes; thus, theranostics pair 68Ga or 64Cu/90Y or 177Lu started to be used in nuclear medicine. From the 2010s, true theranostic pairs gain interest, such as 44gSc/47Sc, 64Cu/67Cu, 83Sr/89Sr, 86Y/90Y, 124I/131I, 152Tb/161Tb, and 152Tb/149Tb, stated to be produced and evaluated for clinical applications. All these developments in radioisotope production and imaging systems also boosted the development of molecule radiolabeling processes, mainly through automated systems, to ensure speed, reproducibility, and less exposure to professionals involved in preparing these molecules. Furthermore, several chelating agents have been developed to complex different metals in a simple, effective, and stable way, such as the DOTA, NOTA, HBED molecules, among others. Conclusão: Trends in radioisotopes and radiolabeling methods result from scientific and technological developments in different areas but with congruent purposes. The result is a large number of products that can be chosen by their physical, chemical, biological, or economic characteristics.

Radiation events such as nuclear war, nuclear reactor incidents, and the deployment of a radioactive dispersal device (dirty bomb) are all significant threats in today’s world. Each of these events would bring significant challenges to clinicians caring for patients with burns and traumatic injuries who are also contaminated or irradiated. The result of a nuclear exchange in a densely populated area could result in thousands of patients presenting with trauma, burns, and combined injury (trauma and burn in an irradiated patient). In this review, we will discuss the three major types of ionizing radiation: alpha, beta, and gamma, and their respective health hazards and biological effects. Additionally, we will discuss the types of burn injuries in a nuclear disaster, caring for the contaminated patient, and managing the combined injury of burn trauma with acute radiation syndrome. The reader will also be left with an understanding of how to prioritize lifesaving interventions, estimate the absorbed dose of radiation, and predict the onset of acute radiation syndrome. While some animal models for morbidity and mortality exist, there is limited modern day human data for patients with combined injury and burns associated with a nuclear disaster due to the infrequent nature of these events. It is extremely important to continue multidisciplinary research on the prevention of, preparedness for, and the response to nuclear events. Furthermore, continued exploration of novel treatments for radiation induced burns and the management of combined injury is necessary.

In nuclear reactors, fine aerosol particles are one of the important carriers of radioactive substances. In the case of a serious accident in a nuclear reactor, the resuspension behavior of particles under the effect of transient airflow may cause a radioactive leak, which is an important content of the radioactive source term analysis. According to the deposition patterns of particles on the wall, particle resuspension can be divided into monolayer resuspension and multilayer resuspension. For the monolayer resuspension, the particle resuspension motion can be divided into lift-off motion determined by lift and adhesion, slide motion determined by drag and friction, and roll motion determined by aerodynamic torque and adhesion torque. For the multilayer resuspension, particle deposition structures affect particle resuspension and produce additional effects including cohesion, shielding effects, coverage effects, restructuration and saltation, which complicates the resuspension process. The resuspension fraction, resuspension rate, and resuspension rate constant are used to describe the resuspension process quantitatively. In terms of physical force, the resuspension process usually depends on the forces from flow including drag, lift, Brownian force, and thermophoresis force, as well as the forces from walls or other particles including cohesion, van der Waals adhesion and friction. For a nuclear reactor system, wall forces acting on particles during the resuspension are more complicated due to the potential capillary force of condensate droplets (pressurized water reactor (PWR)) and the potential sintering force under the high-temperature condition (high-temperature gas-cooled reactor (HTGR)). To predict the resuspension process, scholars proposed various prediction methods including empirical formula, mechanism model (force/torque equilibrium model and energy accumulation model), and numerical simulation method. The empirical formula is only applicable to limited working conditions, the mechanism model may lack key parameters and result in uncertain accuracy, and numerical methods including computational fluid dynamics and the Monte Carlo method can consider more realistic physical processes, but the computational cost in engineering applications is enormous. Based on above principles and methods, the particle resuspension study in the field of nuclear reactors was reviewed in this paper. For PWRs, aerosol resuspension can be induced by continuous fluid flow under convection, transient fluid flow from hydrogen explosion, and bubbling of bubbles in the liquid phase. The resuspension caused by continuous fluid flow possesses the characteristics of long duration and wide occurrence region, which accounts for the main proportion of aerosol resuspension in PWR. For HTGRs, micron-scale graphite dust is the main radioactive carrier. Available reactor experimental data are only obtained from the AVR reactor. Existing research usually uses a combination of resuspension mechanism models and modeling tests to study its resuspension fraction and evaluate dust leakage under accidents. For fusion reactors, dust is mainly generated by the fragmentation of the co-deposited layers formed by plasma and walls. The experiment is usually conducted through a simplified device of the vacuum chamber. Existing experimental research mainly focuses on the influence of geometric parameters of the breach under LOVA (loss of vacuum accidents).

Aerosols in advanced reactors are considered the primary carriers of radioactive nuclides, and the motion behavior of aerosol particles in the near-wall region of reactors is of significant importance for reactor safety analysis. For example, for a high-temperature gas-cooled reactor (HTGR), the graphite dust generated due to the operation of spherical fuel elements may move in the primary circuit under the drag action of the coolant, which will result in deposition, resuspension and then affects the reactor performance. The aerodynamic drag produced by the interaction between fluid and particles during the near-wall motion of aerosols was studied, and the computational fluid dynamics method for near-wall drag numerical calculation was employed in this paper. In order to construct the model for non-spherical particles, several spherical particles (3-5) were combined into non-spherical particles according to the rod-shaped dust particle samples obtained from samples in high-temperature gas-cooled reactor. Based on the geometric model of the particle, the Realizable k-ε turbulence model was used to simulate the Reynolds stress of the mainstream fluid, and the enhanced wall treatment model was used to calculate the turbulent viscosity in the near-wall and around particle region. The three-dimensional incompressible steady SIMPLE algorithm was employed for the computations. The results show that the drag force exerted by the airflow on the particles can be divided into two components, namely, the pressure difference drag force and the shear drag force. Compared to the drag force on the particle under uniform inflow in an infinite space, the near-wall particles withstand a greater aerodynamic drag (enhancement factor) due to non-uniform velocity distribution, offset pressure distribution, and asymmetrical flow vortices. As particles get closer to the wall, the effect of aerodynamic drag enhancement becomes more pronounced. Moreover, the enhancement effect of aerodynamic drag on the near-wall particles becomes more evident as the non-spherical particles elongate and the flow-facing area (angle of attack) increases. Finally, considering the shape of non-spherical rod-shaped particles and the drag enhancement due to near-wall shear flows, the least squares method was applied to fit the expression using an exponential format for the enhancement factor of near-wall aerodynamic drag for different particles. It shows that the difference of drag enhancement factors at the same δ (dimensionless particle-wall distance) among different particles in the near-wall region usually does not exceed 10%. Thus, in most cases, the enhancement factor of spherical particles can be used to estimate non-spherical particles, roughly. This study provides a basis for investigating particle motion characteristics in the near-wall region of advanced reactors.

The supercritical carbon dioxide (S-CO2) nuclear energy and power system uses S-CO2 as the working fluid to convert nuclear heat directly or indirectly into electricity or mechanical energy. S-CO2 is considered one of the most promising working fluids for applications in conventional and high-temperature scenarios. The unique properties of S-CO2, such as its non-phase-change behavior, high density, compressibility near the pseudo-critical region, low thermal cycle compression ratio, minimal thermal inertia equipment requirements, and suitability for direct cycle and power generation propulsion technologies, simplify the system, reduce volume, lower weight, improve maneuverability, reduce noise, and enhance thermal efficiency. The advantages of S-CO2 power conversion technology not only meet the development needs of the next generation of advanced energy conversion technology but also align with the future trends of nuclear power systems. Leading energy technology countries, including the United States, conducted extensive research on S-CO2 power conversion technology in areas such as fossil fuel power generation, nuclear energy, waste heat recovery, and solar thermal power. Some of these projects have already entered the engineering demonstration phase. In China, S-CO2 power conversion technology was included in the “14th Five-Year Plan” for scientific and technological innovation in the energy sector. Numerous universities and industrial sectors in China are actively conducting research and development in this field. The development of S-CO2 nuclear power system in three historical stages for nearly 60 years, including the concept creation, research restart, and collaborative innovation, was summarized in this paper. The research status of S-CO2 nuclear energy and power system in reactor design, material technology, thermohydraulics, heat exchangers, turbomachines, system operation, control, and security policies was also introduced. The technical challenges and key research directions faced by S-CO2 nuclear energy and power system in basic research and engineering were pointed out. Currently, energy superpowers represented by China and the United States have completed laboratory tests of S-CO2 power conversion systems, and are expected to move towards medium-scale engineering demonstration or even large-scale application in the next 5 to 10 years. S-CO2 nuclear energy and power system has the prerequisites for engineering development, and is expected to lead the transformation of advanced nuclear energy technologies. Nevertheless, numerous technical challenges are still existing in core design, materials, thermal-fluid mechanics, and safety analysis. Additionally, engineering efforts are required to tackle challenges related to high-efficiency heat exchangers, highpower high-speed electric motors, heavy-duty electromagnetic bearings, and turbine power generation systems. Simultaneously, coordinated efforts in both fundamental and applied research are essential to enhance the advanced nature and engineering feasibility of S-CO2 nuclear energy and power systems.

This document is an editorial published in the journal Frontiers in Energy Research. It discusses the importance of multi-physics and multi-scale modeling and simulation methods for nuclear reactor applications. The editorial highlights the need for comprehensive analysis of various factors such as neutronics, fuel properties, thermal and hydraulic engineering, and chemical kinetics in order to achieve safe and cost-effective operation of nuclear reactors. The editorial also mentions several research papers that have explored the coupling of different physics and scales in nuclear reactor simulations, as well as advancements in modeling and simulation tools. The document concludes by emphasizing the potential benefits of incorporating modeling and simulation in the design, operation, and maintenance of nuclear reactors. [Extracted from the article]

Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors.

The nuclear heating rate is a critical parameter for reactor core design and irradiation tests, it is typically determined via experimental measurements in a reactor experimental channel. In-pile calibration of calorimeter is an important method for measuring the nuclear heating rate in a fission reactor. This paper summarizes the working principle of the common in-pile calorimeter and reviews the current situation and research progress with regard to in-pile calorimetry employed worldwide. The structural design and performance characteristics of single-cell and multi-cell (differential) calorimeters are compared, and the design method of multimodal integrated measurement device, which represents one of development directions of calorimeters, is introduced. Moreover, the in- and out-pile calibration principles and application methods are described. The advantages and disadvantages of the calibration methods are analyzed, and the developing trend and direction for next-generation in-pile calorimeters are prospected.

In order to ensure the effective supervision of safety criticality during the initial loading and refueling of the reactor and prevent critical accidents during the core loading, the primary loading system of the nuclear reactor is developed. The system detector assembly, monitoring cabinet, measuring channel cabinet and lifting scheme are designed in detail, and the system is fully tested and verified. The detector assembly with boron coated proportional counter tube as the core is designed with pressure resistance and sealing structure, and the integrated high shielding anti-interference cable design is adopted to ensure the stability of signal transmission. The signal processing device adopts the form of cabinet and plug-in box to realize the functions of signal conditioning, pulse counting, alarm display, cycle value calculation and historical record. The isolation design of the signal processing and display plug-in of the measurement channel box further ensures the reliability of the system. Through test and verification, the system’s water pressure tightness, radiation characteristics and stability can meet the requirements of neutron flux rate monitoring during the initial loading process. It has been applied to units 3 and 4 of Fangchenggang nuclear power station.

Uranyl ammonium carbonate (AUC), with the chemical formula UO2CO3·2(NH4)2CO3, plays a crucial role in the wet conversion of uranium hexafluoride (UF6) into uranium dioxide (UO2) or triuranium octaoxide (U3O8) for nuclear fuel production, and is used in commercial and research reactors. In this study, the precipitation of AUC from uranyl fluoride (UO2F2) solution and its subsequent conversion into U3O8 powder were investigated. AUC precipitation was performed at uranium concentrations in UO2F2 solution of 80–120 gL-1, ammonium carbonate (NH4)2CO3 concentrations of 200–400 gL-1, and (NH4)2CO3 to U (C/U) ratios of 5–9. The conversion of AUC into U3O8 powder was studied and sintering of the U3O8 nuclear material derived from ammonium uranyl carbonate (ex-AUC U3O8) was conducted at temperatures of 1000–1800 °C. The kinetics of AUC precipitation from the UO2F2 solution were studied using fundamental kinetic equations, and the kinetics of AUC conversion into UO3 were examined using an isoconversion method based on the thermogravimetric analysis of AUC. The final product of U3O8 nuclear material was characterized using typical techniques, such as thermogravimetric analysis, X-ray diffraction, and scanning electron microscopy. This study provides valuable insights into the production and characterization of AUC and U3O8 nuclear materials, which are key materials in the nuclear fuel industry.

Heat-pipe cooled reactors belong to the group of nuclear reactors using heat pipes filled with liquid metals (such as sodium or potassium) to cool the core. Due to the passive heat removal system, there is no need to use closed loops with pumps, and the reactor can be operated with reduced operational requirements. Consequently, this system can be used in remote locations without access to an electrical grid, or it can be used for space applications. This paper deals with a neutronic study of the Special Purpose Reactor Design-B concept from the Idaho National Laboratory. Using Monte Carlo code Serpent 2 and ENDF/B VIII.0 library, a fixed-temperature model was created to calculate the safety characteristics of the system. This included reactivity coefficients, power distribution, neutron flux spectrum, and criticality safety. In a simplified depletion calculation, an effect of fuel depletion on safety systems was determined as well as a decay heat.

Heat-pipe cooled reactors belong to the group of nuclear reactors using heat pipes filled with liquid metals (such as sodium or potassium) to cool the core. Due to the passive heat removal system, there is no need to use closed loops with pumps, and the reactor can be operated with reduced operational requirements. Consequently, this system can be used in remote locations without access to an electrical grid, or it can be used for space applications. This paper deals with a neutronic study of the Special Purpose Reactor Design-B concept from the Idaho National Laboratory. Using Monte Carlo code Serpent 2 and ENDF/B VIII.0 library, a fixed-temperature model was created to calculate the safety characteristics of the system. This included reactivity coefficients, power distribution, neutron flux spectrum, and criticality safety. In a simplified depletion calculation, an effect of fuel depletion on safety systems was determined as well as a decay heat. [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.)

For heat extraction in nuclear systems, interest in the design of nuclear reactors with heat pipes has increased. The determination of heat limitations is one of the remarkable factors for safety when heat pipes are used for nuclear systems. In this study, capillary and entrainment limit values for the heat pipe were calculated in a heat pipe reactor with potassium working fluid operating at 650 K. Five different effective porous radii (10.1x10-6, 10.225x10-6, 10.35x10-6, 10.425x10-6 and 10.6x10-6 m) and five different wick permeability (4.75x10-12, 5x10-12, 5.25x10-11, 5.5x10-12 and 5.75x10-12 m²) is considered for sintered copper wick heat pipe. While the effects of effective porosity radius, wick permeability, and wick radius on the capillary barrier were studied, only the effects of effective porosity radius were studied. While the effects of effective porosity radius, wick permeability, and wick radius on the capillary barrier are studied, only the effects of effective porosity radius are studied. The highest values of the capillary and entrainment limits are obtained when the porosity radius is 10.1x10-6 m. Besides, maximum capillary limits are achieved when the wick permeability is 5.75x10-12 m² and the effective porosity radius is 10.1x10-6. This study aims to determine the optimum effective porous radius and wick permeability for this reactor and investigate the effect of effective porous radius and wick permeability on the heat pipe limitations. [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.)

The attenuation of neutron radiation and the accompanying gamma radiation from a research reactor by means of shielding filled with metal hydride (using cement binder) is studied experimentally. The energy distribution of the neutrons and gamma radiation in the material is studied. The contribution of secondary gamma quanta to the total dose rate beyond the shielding is determined. [ABSTRACT FROM AUTHOR]