Targeted Alpha Therapy (TAT) is considered the best method of treatment of malignant tissues. Alpha particles have interesting properties of destroying the cancerous cells due to their short penetration and high linear energy transfer. Hence, the use of alpha emitting radionuclides allows for targeting specific cancerous cells and killing of the individual cells while minimizing the toxicity to the surrounding healthy cells. The sources of the alpha for such application are the major concern. Investigation shows that Actinium-225 and Bismuth-213 are potential candidates for the production of alpha for TAT. Because of its short halve-life, the best method of production of Bismuth-213 is through 225 Ac/ 213 Bi generator hence the study of the production of Actinium-225. Critical literature review reveals that proton bombardment of thorium-232 is the best method of producing Actinuim-225. Therefore, the theoretical calculations of the production cross-sections were conducted using a nuclear modular code Empire. The calculated cross-sections were compared with one another and with the measured production cross-sections obtained from the experimental nuclear reaction data exchange format library contained in the nuclear data section of the International Atomic Energy Agency (IAEA), to investigate the effect of optical model’s configurations and the effect of the dispersive and relativistic optical model potentials on the production cross-section. The optical models’ configurations were designated by D1, D2 and D3 respectively for models that used to calculate coupled-channel transmission coefficients for the incident channel in addition to Distorted-Wave Born Approximation uncoupled, used for coupled-channel transmission coefficients for the incident and outgoing channel in addition to Distorted-Wave Born Approximation uncoupled, and used for Distorted-Wave Born Approximation calculation for all collective levels. The potentials are dispersive and relative potential, non-dispersive and relative potential, and the non-dispersive and non-relativistic potentials. The result obtained shows that the optical model configuration used to calculate coupled-channel transmission coefficients for the incident channel in addition to Distorted-Wave Born Approximation uncoupled produces the better effect on the production cross-section. Similarly, the model with Non-Dispesive and Non-Relativistic potentials produces netter yield while the optical model with Dispersive and Relativistic potential is more physically conservative. Hence any of the optical models can be used for the calculation of the production cross-section, regardless of the dispersive and relativistic potentials of the optical model.
The Interacting Boson Models, IBM1 and IBM2, analyzed in their O(6) limit are used to describe the properties of the 190 Hg isotope. A fitting procedure is used to determine the most accurate parameters to the calculated energy levels of this nucleus. The calculated ground, gamma, and beta bands energy of each model are compared to the previously measured data. Quadrupole electromagnetic transition strengths for 190 Hg, obtained from both IBM1 and IBM2, are presented and associated with previously experimental results. Finally, potential energy surfaces (PES) of 190 Hg in the dynamical symmetry O(6) of the IBM1 as an occupation of the restriction of the deformation are determined and discussed. All calculations are satisfied with experimental data.
This paper presents a multiple linear regression framework for reconstructing the historical helium-to-proton (He/p) flux ratio of galactic cosmic rays (GCRs) leveraging long-term, ground-based neutron monitor datasets. Optimized via a Stochastic Gradient Descent (SGD) algorithm, the model was calibrated and trained on direct satellite measurements from PAMELA and AMS-02 covering the period 2006–2019 and was subsequently applied to estimate the He/p ratio for the extended period of 1980–2019. The reconstruction results show good agreement when validated against independent historical data from balloon-borne e xperiments. Given that the He/p ratio is a crucial indicator of cosmic ray modulation, utilizing the extensive time coverage of the neutron monitor network provides deeper insights into solar modulation dynamics over past solar cycles.
The double-differential cross-sections (DDCS) and angular distributions for the neutron-induced reactions on the 107 Ag isotope were calculated at incident energies of 3, 5, 10, and 15 MeV within the pre-equilibrium region using the Kalbach model implemented in the PRECO-6 code and in conjunction with Feshbach-Kerman-Konnin statistical theory. Emitted nucleons’ total cross-sections were decomposed into direct (MSD) and compound (MSC) multistep components following Kalbach systematics expressed as d 2 σ/dΩ.dE . Results show that the d 2 σ/dΩ.dE a☺nd angular distribution for 107 Ag decrease systematically as the incident neutron energy increases, with markedly clearer structures for light ejectiles (neutrons and protons) than for heavy composite particles (deuterons, tritons, 3 He, and α particles). The latter also requires higher incident energies and exhibits spectral gaps under the present conditions. For two channels (n, n) and (n, p), calculations indicate that existing experimental data are not yet sufficient to firmly establish the energy dependence of the cross-sections, underscoring the need for precise measurements to constrain pre-equilibrium model parameters.
A numerical model for direct numerical simulation of compressible two-phase media is applied to the problem of gas-bubble rise in liquid lead. Because liquid metals are opaque and exhibit a very large density contrast between the gas and liquid phases, direct numerical simulation is an important tool for investigating interfacial dynamics and bubble-shape evolution. The paper presents 2D and 3D calculations performed with a two-phase flow code under development at IBRAE RAN using a previously developed surface-tension model integrated into the numerical algorithm. The numerical algorithm employs an explicit HLLC scheme with MUSCL reconstruction and equations of state to model compressible two-component media. The code is designed for the direct numerical simulation of thermal fluid dynamics in a two-phase compressible medium, including two-component mixtures, accounting for interfacial heat and mass transfer and using equations of state for Stiffened Gas or the Noble-Abel equation for weakly compressible media. The computed evolution of bubble shape and the time-dependent rise velocity are compared against experimental data for bubble rise in narrow rectangular channels. The simulated value of the rise velocity of a gas bubble with initial diameter of 3 mm agrees with the experimental average value of the rise velocity of moderate bubbles, entering the experimental channel through a nozzle of 3 mm in size, within 40%. This discrepancy is attributed to the insufficiently fine computational mesh used in the simulations.
The variation of nuclear reaction cross-sections with the variation in projectile energies is called excitation functions has been a subject of great interest since last few decades. They beautifully display the pre-equilibrium as well as equilibrium emission of particles. Nucleon-induced reactions on medium-mass nuclei are important for nuclear-structure studies and isotope production. Selenium-80 is a target for producing bromine, arsenic, and germanium isotopes. The aim of this study is to compute and validate cross sections for key proton- and neutron-induced reactions on 80 Se that produce 76,77 Br, 76 As, 77 Ge, and 80 As. Excitation functions for 80 Se(p, 4n) 77 Br, 80 Se(p, 5n) 76 Br, 80 Se(p, α) 76 As, 80 Se(n, α) 77 Ge, and 80 Se(n, p) 80 As were calculated with the COMPLET code and compared to literature data. COMPLET reproduces the main features of the measured excitation functions, matching peak energies and magnitudes for most channels. The results of such calculations were compared to predictions of a well-established optical potential and with experimental data, reaching very good agreement. COMPLET reliably models the investigated reactions on 80 Se and yields cross sections useful for isotope production and reaction studies. The results aid beam-energy selection and experimental planning for producing 76,77 Br, 76 As, 77 Ge, and 80 As and support further model refinement.
This paper provides experimental justification for selecting cement-based materials for engineered safety barriers (ESB) in geological disposal facilities (GDF) for radioactive waste. The long-term performance of three materials — Ordinary Portland Cement Concrete (OPCC), Calcium Aluminate Cement Concrete (CACC), and Nirex Reference Vault Backfill (NRVB) —was comparatively analyzed under conditions simulating groundwater impact in a crystalline rock massif. A one-year filtration experiment was conducted using synthetic groundwater at elevated pressure. The evaluation included comprehensive assessment of phase composition evolution, mechanical strength, hydraulic permeability, and chemical/electrochemical parameters of the interacting liquid phase. Results demonstrated that OPCC maintained increased content of strengthening phases (C-S-H, ettringite), enhanced mechanical strength exceeding the required minimum of 7.5 MPa, and consistently low hydraulic permeability (~10 -12 m/s), meeting ESB criteria for contact with bentonite buffer. The pH of its filtrate remained below 11.5, minimizing alkaline degradation risks for bentonite. In contrast, NRVB exhibited complete dissolution of C-S-H and portlandite, resulting in significant strength reduction and high hydraulic permeability, making it unsuitable for bentonite buffer contact. For aluminate concrete, cross-reaction of metastable phase CAH 10 into less dense C 3 AH 6 caused substantial strength decrease and increased hydraulic permeability exceeding maximum permissible values (10 -10 m/s), limiting its GDF application. Based on these findings, Ordinary Portland Cement Concrete emerges as the preferred material for ESB construction in GDF. These results establish foundation for subsequent geochemical modeling of long-term cement barrier evolution.
Background . The radioisotope lutetium-177 ( 177 Lu) has been suggested for utilizing in radio-immunotherapy. Now, 177 Lu is mainly produced by neutron activation through nuclear reactors, although cyclotron-based radioisotope generation may also be assessed. Methods . In this investigation, the small cyclotron located in Karaj, Iran, with a maximum deuteron energy of 14 MeV, was simulated to generate 177 Lu using a 176 Yb target with the TALYS and EMPIRE codes. Here, the generation yield of the 176 Yb(d,p) 177 Yb→ 177 Lu reaction was computed via different incident energies. Results . The estimated cross-section amounts of the reaction (d,p) and its rival reaction (d,x) were compared with other reports. The maximum cross-section amount for the (d,p) interaction was evaluated to be 125 mb at 10.5 MeV and for the (d,x) interaction to be 121 mb at 9.5 MeV. Through Pade fitting, the computed production yields were 315 and 319 MBq/µA.h at 14 MeV deuteron energy from TALYS and EMIRE simulators, respectively. Conclusion . For a one-hour irradiation at 100 μA beam current, the activity was approximately estimated 31.5 GBq of 177 Lu at 14 MeV. The EMPIRE code’s Fermi Gas Model offers computational efficiency and predictive capability for unknown nuclei, though its limitations at low excitation energies and for deformed nuclei introduce systematic uncertainties. Further experimental validation of the calculated cross-sections and development of efficient targetry and separation chemistry are recommended for practical implementation of this cyclotron-based production route.
Laser-induced nuclear fusion offers the tantalizing prospect of a clean, virtually limitless energy source. Traditional inertial confinement fusion (ICF) approaches struggle with challenges in laser efficiency, beam uniformity, and target design. The present research paper proposes a novel method of laser-induced nuclear fusion utilizing a ring laser configuration. The ring laser’s unique geometry enables highly focused, omnidirectional energy delivery to a fusion fuel target. This design aims to overcome the limitations of traditional ICF systems by enhancing energy deposition efficiency and increasing compression symmetry.Numerical simulations using specialized codes such as HYADES,LILAC and VORPAL has been used to investigate the potential advantages of this concept.
This study presents a comprehensive neutronic analysis of dual-cooled annular duplex fuel assemblies incorporating UO2-ThO2 composition for application in advanced pressurized water reactors (PWRs). The investigation employs a 13 × 13 fuel assembly configuration to evaluate the operational enhancement potential of modern reactor systems. The research methodology focuses on the comparative assessment of burnup characteristics across varying uranium enrichment levels, benchmarking the neutronic performance of dual-cooled duplex fuel against conventional solid 17 × 17 and dual-cooled 13 × 13 UO2 assemblies. The results demonstrate that the proposed UO2-ThO2 dual-cooled duplex fuel configuration with 7 wt.% U-235 achieves discharge burnup equivalent to that of conventional solid UO2 assemblies. Safety analysis encompasses the quantification of plutonium isotope production and minor actinide generation, revealing that dual-cooled duplex assemblies produce significantly reduced quantities of plutonium isotopes and lower concentrations of minor actinides, including neptunium (Np), americium (Am), and curium (Cm), relative to conventional all-UO₂ assemblies. Reactivity coefficient analysis confirms that both the fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) maintain consistently negative values throughout the operational cycle. These coefficients not only satisfy but exceed the established safety criteria for PWR operations, thereby demonstrating the enhanced safety margins and operational performance characteristics inherent to the dual-cooled duplex fuel assembly design.
Research reactors (RRs) are widely used for research in fundamental physics, materials science, and for radioisotope production. One of the problems in the field of mathematical modeling of research reactors is the verification of computational codes used for the calculation of neutron-physical parameters. To pass the verification procedure, it is necessary to justify the methods of mathematical modeling and the accuracy of the RR geometric model. Currently, precision programs allow creating a RR geometric model of any degree of detail. However, the degree of this detail has not been formulated. The paper considers the justification of the parameters of spatio-temporal discretization of IVV-2M-type fuel assemblies when calculating the burnup distribution. Determining the frequency of dividing the fuel assembly model into layers by the height of its active part significantly affects such parameters as the effective neutron multiplication factor (Keff), energy release at the maximum stress point. Moreover, the degree of detail and the set of calculation statistics affect the calculation duration, which can be reduced by applying a certain approach to modeling. It should also be noted that the issue of the need for a fuel pin (or more detailed, in the horizontal plane) degree of burnup calculation was not considered due to the lack of control over the orientation of the IVV-2M fuel assemblies in the horizontal plane during the operation of the IVV-2M research reactor.
In gas-bonded fuel pins the interior space is filled with helium in order to provide heat removal from the fuel. Under irradiation, as a result of gaseous fission products release, the initial gas composition under the fuel pin cladding changes, which leads to a change in its thermophysical characteristics. Post-irradiation examinations have shown that the partial pressure of nitrogen under the fuel pin cladding increases with fuel burn-up increase, since nitrogen atoms also release from uranium-plutonium nitride fuel under the cladding, in addition to the krypton, xenon and helium inert gases. The mechanisms of release of inert gases and nitrogen from nitride are different: the diffusion mechanism for inert gases and the knocking out of fission fragments for nitrogen. The difference in the gas release mechanisms leads to a significant quantitative difference in the gases release under the cladding. The specific nitrogen yield under the fuel pin cladding is significantly lower than the specific yield of other gases, but its presence is an important factor, since the dissociation temperature of the mixed nitride, as well as the nitriding of the inner surface of the cladding, depends on the nitrogen pressure under the fuel pin cladding. A generalization and analysis of the results of post-irradiation examinations of the nitrogen content in the gas mixture under the claddings of 87 investigated fuel pins after irradiation in the BN-600 reactor as part of 17 experimental fuel assemblies with mixed uranium-plutonium nitride fuel, irradiated to maximum fuel burn-up from 3 at. % to 9 at. % is carried out in the paper.
Nucleon-induced nuclear reactions are a significant field in nuclear physics with numerous applications like as in the production of medically important radioisotopes. The primary objective of this study is to analyze the excitation function of nucleon-induced nuclear reactions on the arsenic-75 isotope across projectile energies from 10 MeV to 100 MeV using COMPLET code. The excitation functions of the seven reaction channels: 75As(p, 3n)73Se, 75As(p, pn)74As, 75As(p, p5n)70As, 75As(p, p2p)73As, 75As(p, n)75Se, 75As(n, 2n)74As, and 75As(n, p)75mGe were investigated, analyzed and compared with experimental data within the energies from 10 MeV to 100 MeV. The calculated excitation functions showed strong agreement with experimental data obtained from the EXFOR data base, as assessed using Pearson’s correlation coefficient. Both pre-equilibrium and equilibrium nuclear excitation functions for all nucleon-induced reaction channels displayed a strong correlation with experimental results, except for the neutron-induced reaction channel, 75As(n, p)75mGe, which exhibited a moderate correlation. Studies have indicated that the pre-equilibrium reaction mechanism primarily governs the high-energy segment of the excitation function, whereas the low-energy segment is dominated by the equilibrium reaction mechanism for both neutron and proton-induced nuclear reactions on arsenic-75. Thus, utilization of the COMPLET code and the EXFOR data base has facilitated a detailed analysis of induced nuclear reactions in producing radionuclides with diverse applications.
The paper presents the results of the computational analysis of an accident scenario in the spent fuel storage pool of the multipurpose fast research reactor. The simultaneous failure of the spent fuel cooling system and the ventilation systems is being considered as beyond design basis accident (BDBA) conditions. The computational analysis of the BDBA scenario was performed with the computer program KUPOL-MT designed to simulate thermohydraulic processes in the room atmosphere of the reactor facility containment systems. During the computational analysis with KUPOL-MT, a nodalization scheme for the spent fuel storage pool rooms was developed. The scheduled loading of spent fuel assemblies in the storage pool was considered in the accident simulation, together with the emergency unloading of the entire reactor core. It was shown that the water temperature in the spent fuel storage pool remained below the boiling point and there was no violation of the integrity of fuel cladding for three days after the start of the accident. As the fuel cladding did not collapse, there were no radiation consequences in the considered accident scenario. The calculation results demonstrated insignificant hydrogen stratification in the room above the spent fuel storage pool. It was shown that hydrogen content did not exceed the maximum concentration limit for three days after the start of the BDBA.
Cr-Mn Austenitic Stainless Steel (ASS) are utilized in nuclear reprocessing plants, fast-breeder reactors, pressurised water reactors and boiling water reactors because of their ease of fabrication and welding. In this study two fillers were used, ER308L and ER308L-16, to weld ultra-low nickel Cr-Mn ASS using gas metal arc welding (GMAW) and shielded metal arc welding (SMAW) process. The simulation of the welding processes was carried out for the evaluation of stresses induced in the welding. The heat input is employed at the weld bead for modelling and analysis. From analysis, it was found that the weld bead width is less in GMAW as compared to SMAW. The transient thermal and structural analysis was carried out for evaluation of stresses in the weld using ANSYS. The weld for both welding methods was analysed using varying welding speeds. It was revealed that as the welding speed increases there is a decrease in Von-Mises stresses in the weld.
The article considers the possibility of using the mixed (Am, Cm)-fraction extracted from minor actinides (MA) for large-scale production of powerful heat sources for space radioisotope thermoelectric generators (RTG). Such a choice of the MA fraction removes the limitation on the proportion of 236 Pu in plutonium and leaves only a limitation on the proportion of 238 Pu (above 80%). The irradiation of the (Am, Cm) mixture and the separated Am-fraction of MA are considered and compared. The irradiated material was placed in the central assembly of the VVER-1000 light-water power reactor core. It is proposed to remove only fission products (FP) from the irradiated material and to use the remaining three-component (Pu, Am, Cm) mixture as a heat source in the RTG. It has been shown that the mixed (Am, Cm)-fraction of MA is a preferable starting material compared to the Am-fraction since it provides the higher specific heat release of the produced three-component (Pu, Am, Cm) mixture at comparable rates of plutonium production and its isotopic compositions.
This study applies the COMPLET nuclear reaction code to calculate excitation functions for eleven alpha-induced reactions on stable copper ( 63 Cu, 65 Cu) and antimony ( 121 Sb, 123 Sb) isotopes, aiming to predict production cross-sections for medically significant radionuclides such as 68 Ga, 67 Ga, 66 Ga, 65 Zn, 124 I, and 123 I. Reactions are simulated across an alpha energy range of 10–80 MeV to evaluate excitation functions. Fixed nuclear model parameters, such as an initial exciton number n 0 = 4 (2p+2n+0h) and level density parameters ACN/K (K = 10) expressions tied to compound nucleus mass, were used to compute theoretical cross-sections. Model outputs were systematically compared with experimental data obtained from the EXFOR database. Statistical and graphical analyses demonstrated an excellent agreement, with Pearson correlation coefficients ranging from 0.74 to 0.94. Sensitivity analyses confirmed that variations in exciton numbers and level density parameters significantly influenced the shape and peak positions of the excitation functions, highlighting the importance of accurate parameter selection. These findings validate the COMPLET code as a reliable tool for modeling alpha-induced nuclear reactions, especially when experimental data are scarce. The results contribute to improved nuclear data evaluations and provide critical support for the planning of radionuclide production in medical applications. The study also includes a detailed covariance analysis to minimize discrepancies between model predictions and experimental data, emphasizing the importance of theoretical methods in contemporary nuclear research.
The challenging geopolitical landscape and the imposition of sanctions necessitate a comprehensive overhaul of economic processes within our nation. In response, the Government of the Russian Federation has delineated a list of products and technologies earmarked for substitution with domestic equivalents in the immediate future. A pivotal focus lies on the import substitution of software. Pursuant to Presidential Decree No. 166 dated March 30, 2022, titled “Measures to Ensure Technological Independence and Security of the Critical Information Infrastructure of the Russian Federation”, effective January 1, 2025, state authorities and procurement entities are prohibited from deploying foreign software on their critical information infrastructure assets. The spearhead of the IT import substitution initiative in Russia has been State Corporation “Rosatom”, which, within the framework of a unified digital strategy aimed at bolstering IT import independence, is focused on fostering collaboration with Russian small and medium enterprises. A noteworthy initiative within Rosatom’s import substitution program, involving SMEs, entails the modernization of existing GEFEST and GEFEST800 PCs. These systems are designed for operational neutronic calculations for fast reactors utilizing sodium coolant at the Beloyarsk Nuclear Power Plant, the world’s only station with industrial-scale fast neutron reactors, namely, BN-600 and BN-800. The overarching objective of this project is to establish a unified cross-platform PC, known as GEFEST-M, based on the current GEFEST and GEFEST800 models. This unified system will incorporate specially developed calculation modules and methodologies, with the future applicability envisaged for reactors like BN-1200 and other BN-type reactor units. A pivotal attribute of the GEFEST-M project is the shift away from proprietary closed software development tools. Additionally, it underscores the capability to operate seamlessly on computing equipment and operating systems developed domestically. The developer spearheading the GEFEST-M PC project is an accredited organization in the realm of information technology – LLC “ASU-LEADER”.
Sodium-cooled fast neutron reactors are one of the most environmentally friendly reactor types and play an important role in the development of Russia’s nuclear power industry and the long-term supply of nuclear fuel for the industry. The long-term successful operation of the industrial-scale fast neutron reactor BN-600 and the commissioning of a new power unit with a more powerful BN-800 reactor have made Russia a world leader in the construction and operation of fast neutron reactors. This paper describes the stages of mastering the capacity of the sodium-cooled fast neutron power reactors BN-600 and BN-800 at the Beloyarsk NPP. Based on the experience of BN-600 and BN-800 operation, a power unit with a BN-1200M reactor is being created, which can be considered as the head in a series of power units with BN-1200M reactors. The creation of the series will significantly expand the fuel base of nuclear power by reusing spent nuclear fuel from other NPPs, and minimizing radioactive waste by burning the longest-lived isotopes from spent nuclear fuel from other reactors. The main task in creating the BN-1200M is to achieve economic characteristics that ensure its competitiveness with a serial thermal neutron reactor VVER of a similar power level. According to the General Scheme for the Placement of Electric Power Facilities approved by the Government of the Russian Federation until 2042, the lead power unit with the BN-1200 reactor will be built at the Beloyarsk NPP as power unit No. 5.
The possibility of obtaining 238 Pu in the BN-1200M (about 1,200 MWe) fast neutron power reactor is considered. To obtain products with a purity of 80–85% in terms of 238 Pu, options were proposed with installation of irradiation assemblies (IA) with breeding and moderating elements in the radial blanket. Neptunium and americium oxides and their mixtures were considered as target materials of the breeding elements for producing target nuclide. Zirconium hydride was considered as a moderator. The effective density of targets, IA arrangement, target/moderator materials content ratio, and irradiation time were varied. Calculations were made to determine mass and composition of produced plutonium and IA decay heat, as well as IA effect on the global and local pin-to-pin power profile in the standard fuel assemblies of the reactor core. When using targets made of neptunium oxide, options were effective only at an effective density of approximately 1.1 g/cm³ were only effective because of self-shielding effect. Irradiation during 4–5 years ensures required 80% purity of 238 Pu with target purity in terms of 236 Pu within 2 ppm. When using Am, 236 Pu purity is also ensured, but 238 Pu fraction is somewhat lower than 80% due to the higher fraction of 242 Pu. The use of Am and Np mixture makes it possible to reduce the fraction of 236 Pu and prevent high accumulation of 242 Pu, but in this case 242 Cm accumulation should be taken into account, and appropriate time period for its decay heat reduction should be chosen.
