DOI: 10.3897/nucet.12.215113

Authors: Ivan V. Burievsky, Andrey V. Gulevich, Olga S. Gurskaya, Viktor M. Dekusar, Vladimir A. Eliseev, Nikita A. Zakirov, Kirill A. Isanov, Valery V. Korobeynikov, Yulia A. Kuzina, Andrey L. Moseev, Vladimir M. Troyanov, Ruben A. Shaginyan

Abstract: Today, thermal reactors, which are the basis of modern Russian nuclear energy, can consume plutonium, but in limited quantities and with a low content of threshold isotopes, so their fuel supply is actually limited by natural uranium reserves. Fast reactor technology was created for extended production of plutonium and their fuel supply capabilities are limited by the resources of 238U, which is two orders of magnitude more than 235U. The paper presents the following possible functions of fast reactors in the future two-component nuclear energy system, the transition to which in the coming decades is a strategic goal of the Rosatom State Corporation: electricity generation; maintenance the balance of plutonium in the system of fast and thermal reactors; “refinement” of plutonium from spent nuclear fuel in thermal reactors; disposal of long-lived nuclear energy wastes; production of isotopes highly demanded by the industry and medicine. Obviously, the specialized fast reactors creation will be required to increase the efficiency of these functions. As a result, thermal and fast reactors can be used to create a sustainable, multi-functional and waste-free new nuclear power industry with an almost unlimited fuel resource.

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DOI: 10.3897/nucet.12.215111

Authors: Valery P. Smirnov, Aleksandr A. Samsonov, Anton Yu. Leshchenko, Aleksej V. Nosov, Yuriy S. Mochalov, Sergey M. Savchits, Aleksandr A. Legen’kij

Abstract: The paper presents results of testing a mock-up high-temperature facility for the treatment of mixed uranium-plutonium nitride (MUPN) spent nuclear fuel (SNF) from the BREST-OD-300 reactor. The mock-up facility is designed to test selected high-temperature treatment (HTT) process parameters using model nuclear fuel (MNF) and allows design development testing of pyroprocessing equipment for fast neutron reactor SNF at the Reprocessing Module of the Experimental and Demonstration Energy Complex. The performance of the mock-up facility was preliminarily evaluated using MNF dummies; temperature distribution, tightness, and dummy product unloading efficiency under specified HTT conditions were obtained. The mock-up tests on the uranium nitride were carried out to prove the feasibility of the MNF nitriding/denitriding and its oxidation to U3O8, followed by the reduction to UO2. After the tests, the pulverized fuel was characterized for the particle size distribution, the bulk density, the angle of repose, and the phase composition by X-ray diffraction analysis. A material balance of uranium-containing product distributed over the units of the facility was drawn. The nitriding/denitriding processes proved ineffective to pulverize the MNF (25% of the loaded amount remained in claddings); nevertheless, X-ray diffraction showed a finely-pulverized UN-U2N3 mixture in the powdered product. The nitride MNF oxidized in the Ar–10% O2 gas mixture and the U3O8 reduced to UO2 in the Ar–4% H2 atmosphere completely. According to the X-ray diffraction analysis, the produced UO2 did not contain a U3O8 phase and was a coarse powder, i.e. the fraction of particle size less than 100 µm was only 12.5%. The quantity of uranium in the product removed from the mock-up facility was 58% of the total amount used for the tests. The product loss was caused by the product deposits in the retort (13%) and a significant dust carryover from the retort during the tests (31%). The design of the mock-up HTT facility requires improvements to reduce the internal deposits and the dust carryover from the facility together with off-gases.

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DOI: 10.3897/nucet.12.215081

Authors: Sergey M. Dmitriev, Tatiana D. Demkina, Alexander A. Dobrov, Denis V. Doronkov, Alexey N. Pronin, Anton V. Ryazanov, Oleg O. Karpukhov, Oleg B. Samoilov, Dmitry L. Shipov

Abstract: The paper presents the results of an experimental study and comparative analysis of coolant hydrodynamics at the outlet of a fuel assembly (FA) with upper nozzles of three different designs. The FAs are intended for the core of the RITM reactor of a small nuclear power plant (SNPP). The aim of the work was to study the axial flow velocity at the outlet of the fuel rod bundle and behind the upper nozzles of different designs, as well as to identify regions of the fuel rod bundle from which flow could enter the sampling tube leading to a resistance thermometer (RTD). The experiments were conducted on an experimental test bench with an air working fluid using a model of the FA outlet section. The model consisted of a fragment of the fuel rod bundle, replaceable mock-ups of the upper nozzles for three design variants, as well as mock-ups of the mounting plate for the FA and the coolant sampling tube to the RTD. The pneumometric method was used to determine the axial flow velocity. The tracer injection method was employed to identify the regions of the fuel rod bundle from which flow could enter the sampling tube to the RTD. The hydrodynamic flow pattern is presented in the form of maps of axial velocity and tracer concentration. The experimental results allowed detailing the turbulent flow structure and led to the following conclusions. The shape of the flow passage of the upper nozzle confusor has little effect on the axial flow already before the outlet spacer grid. However, a significant effect of the flow path geometry of the upper nozzles is manifested when the flow exits the fuel rod bundle. The shape of the confusor flow passage and the flow path geometry of the upper nozzles have a considerable influence on the axial velocity magnitude and flow structure. The greatest contribution to the change in axial velocity comes from the reduction in the flow passage area of the upper nozzles. The non-uniformity of the flow structure from the upper nozzle windows is determined by the sampling tube, which blocks part of the flow passage. The flow exiting the upper nozzle with three windows is the most non-uniform. During flow through the upper nozzle confusors, a mixing process is observed, expressed as a displacement of the tracer toward the sampling tube. The greatest transverse tracer displacement occurs in the confusor of the three-window upper nozzle. To minimize the influence of flows with different temperatures on the RTD readings, it is necessary to use upper nozzles with three windows. The research results have been used by specialists of JSC “OKBM Afrikantov” to substantiate engineering solutions in core design.

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DOI: 10.3897/nucet.12.203079

Authors: Hassan Alsaleh, Mohammad Albarhoum

Abstract: A benchmark-based neutronic and steady-state thermal-hydraulic assessment was performed to evaluate beryllium reflector configurations for enhancing irradiation capability in the MTR-10 MW research reactor. The reference one-trap core was modified to a symmetric four-trap layout, and a constrained sequence of graphite and beryllium reflector substitutions was assessed with MCNP4C2/MTRMC. The four-trap configuration reduced the initial excess reactivity from 141.71 to 82.85 mk. The selected 22-block beryllium reflector recovered the excess reactivity to 139.38 mk, corresponding to 96.04% recovery of the reactivity loss relative to the shifted four-trap base case. The four-site average thermal neutron flux increased from the reference central-trap mean of 1.055 × 1014 to 1.319 × 1014 n·cm-2·s-1, an average gain of 25.0%, while the maximum site reached 1.904 × 1014 n·cm-2·s-1. After applying a comparable active neutron-history basis, OpenMC reproduced the same recovery trend, giving 96.94% recovered reactivity. The steady-state thermal-hydraulic assessment used CAUDVAP/TERMIC-1H at 550 m3/h, giving a total mass flow rate of 151.607 kg/s and a mean pressure drop of 5.372 kPa. The maximum plate power in the hottest fuel element was 28.93 kW, giving a hot-plate peaking factor of 1.24 and a hot-plate heat-flux input of 49.4 W/cm2. The TERMIC-B2 case at 50 W/cm2 therefore represents the rounded-up hot-plate condition for the final configuration, with redistribution and DNB ratios above unity over the assessed hot-channel velocity interval.

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DOI: 10.3897/nucet.12.201427

Authors: S M Sayeed Anowar, Abid Hossain Khan, Md Mahfuzur Rahman, Md Tanbir Hasan, Md Waliuzzaman Talha, Md Shamim Hassan

Abstract: Neutronic analysis of nuclear fuel assemblies is crucial for optimizing reactor core design. The conventional techniques involve computationally expensive simulations. This work explores the feasibility of developing efficient machine-learning models for neutronic analysis. This was done by leveraging artificial neural networks (ANNs) for evaluating two important neutronic parameters: neutron multiplication factor (k∞) and void coefficient (αv), of 17 × 17 pressurized water reactor (PWR) fuel assemblies. OpenMC neutronics code was employed to generate a training set of 6000 data points for the machine learning models with six input parameters: U-235 enrichment, densities of fuel and moderator, concentration of burnable poisons in fuel and moderator, and fuel pitch. Results reveal that the four-layer ANN model with sigmoid activation functions in the hidden layers exhibited excellent performance. The ANN predicts k∞ with an average predictive error below 106 pcm, well within the acceptable limit of 500 pcm. Among the optimization algorithms employed, the simplex search method outperformed genetic algorithms, simulated annealing and particle swarm optimization. Utilizing the developed ANN model in simplex search, the correlation was derived for fuel pitch separating the under- and over-moderated regions under cold zero power (CZP) conditions. It was observed that the fuel pitch must be kept ≤ 1.434 cm to maintain the reactor’s under-moderated status throughout its operational lifetime. Overall, the accuracy of ANN model was found to be satisfactory, opening doors for further improvements.

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DOI: 10.3897/nucet.12.199104

Authors: Georgy V. Tikhomirov, Mikhail Yu. Ternovykh, Alexander A. Ryzhkov, Aleksandr S. Gerasimov

Abstract: A benchmark for neutron transport simulations of fast reactors with lead coolant and mixed-nitride uranium-plutonium fuel is presented. An approach for developing adapted test problems is proposed for extending an open database of neutron transport simulation test problems with a simple and full description to widen a potential audience. The formulation of the benchmark is based on a conceptual computational model with no excessive detailing of design elements, which reduces the number of potential errors in the preparation of initial data and the model description in the tool. In addition, a methodology for developing a test problem with an equilibrium composition of nuclear fuel with a minimal reactivity margin for burn-up based on existing fast reactor design concepts is described to expand the cross-verification base of fast reactors. The proposed core contains fuel in three zones (regions) with fuel rods having different fuel column heights. This ensures a simplified core design, a negative void effect of reactivity, and effective flattening of the energy deposition distribution along the core radius. The benchmark may also be of use in an initial verification of neutron transport simulation tools for finding the errors related to the methods implemented in the tool. Comparative results of the benchmark calculations with the ENDF/B-VII.1-based data of SCALE 6.2 and the BNAB-RF data are presented, providing a reference point for the potential audience of the benchmark problem. Besides, the work highlights an educational purpose of the problem and prospects of creating a system of computational benchmarks for innovative fast nuclear reactors included in the Generation IV list.

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DOI: 10.3897/nucet.12.196675

Authors: Vincent Yao Agbodemegbe, Edward Shitsi, Savior Anyorgbor

Abstract: The heat removal characteristics of water under supercritical and compressed fluid conditions are of direct relevance to the thermal-hydraulic optimization of nuclear reactor rod bundles. In this study, numerical simulations of upward water flow through a heated 2×2 rod bundle were performed using STAR-CCM+ at pressures of 23 MPa and 25 MPa, with systematically varied heat and mass fluxes. The SST k-ω (Menter) and realizable k-ε turbulence models were applied with All y+ wall treatment and validated against the experimental data of H.Y. Gu et al. The SST k-ω model demonstrated superior near-wall prediction accuracy, establishing its suitability for supercritical flow simulations, while both models performed comparably in resolving bulk flow temperature. At constant heat flux (300 kW/m2), the heat transfer coefficient increased with both mass flux and pressure. At constant mass flux (228 kg/m2s), it decreased with increasing heat flux but increased with increasing pressure, confirming the combined governing role of the heat-to-mass flux ratio and an independent pressure effect attributable to pseudo-critical temperature shifting. Contrary to conventional expectations, elevated turbulent kinetic energy under supercritical conditions did not correspond to enhanced heat transfer. Superior heat transfer performance was instead associated with lower turbulence levels under compressed fluid conditions, a behavior attributed to the severe degradation of thermal conductivity, specific heat, and density near the pseudo-critical temperature, which suppresses turbulent transport effectiveness. TKE variations were more pronounced at 25 MPa than at 23 MPa with increasing mass flux, and overall TKE behavior was governed by a complex interaction of pressure, heat flux, mass flux, and axial position.

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DOI: 10.3897/nucet.12.199627

Authors: Ouadie Kabach, El Mahjoub Chakir

Abstract: This study evaluates transuranic (TRU) management in dual-cooled annular PWR fuel using UO2 and (232Th-233U-235U)O2 matrices, with solid-pin assemblies as references. Pin-cell screening and assembly depletion calculations were performed to assess criticality, cycle performance, fissile-inventory retention, actinide evolution, and temperature feedback. Under a 3% leakage allowance, the best-performing configuration, Case-9 with a 0.3 mm TRUO2 ring in the thorium-based matrix, achieved a discharge burnup of 71.36 GWd/tHM and a cycle length of 594.70 effective full-power days (EFPDs), representing a 29.5% increase over the solid UO2 reference. Its end of the irradiation cycle fissile inventory ratio reached 0.469. In UO2-based assemblies, 238U conversion sustained plutonium production, preventing systematic net depletion of the principal Pu isotopes. Conversely, the absence of 238U from the thorium-based matrix eliminated matrix-driven 239Pu production, while 233U regeneration supported late-cycle reactivity. In Case-9, 239Pu, 240Pu, 241Pu, and 241Am decreased by approximately 97.5%, 73.6%, 59.0%, and 91.9%, respectively; however, 242Pu, 243Am, and curium accumulated through successive capture chains. All configurations exhibited negative fuel and moderator temperature coefficients, ranging from -1.72 to -2.60 pcm/K and -20.90 to -55.15 pcm/K, respectively. These results identify thorium-based annular fuel with localized TRU loading as the most promising configuration from a neutronic point of view examined for combining TRU consumption, extended irradiation potential, and favorable reactivity feedback.

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DOI: 10.3897/nucet.12.191179

Authors: Alyaa A. Abd-Aljbar, Mahdi J. S. Al Musawi, Huda H. Kassim, Asmaa A. Elbndag, Mushtaq A. Al-Jubbori, I. Hossain, Fadhil I. Sharrad, N. Aldahan

Abstract: The Interacting Boson Models, IBM1 and IBM2, analyzed in their O(6) limit are used to describe the properties of the 190Hg 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 190Hg, obtained from both IBM1 and IBM2, are presented and associated with previously experimental results. Finally, potential energy surfaces (PES) of 190Hg 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.

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DOI: 10.3897/nucet.12.191038

Authors: Teshale Ayalew Mekuriaw, Yihunie Hibstie Asres, Senamaw Mequanent Zegeye, Ermias Yitayew Beyene

Abstract: 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 80Se that produce 76,77Br, 76As, 77Ge, and 80As. Excitation functions for 80Se(p, 4n)77Br, 80Se(p, 5n)76Br, 80Se(p, α)76As, 80Se(n, α)77Ge, and 80Se(n, p)80As 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 80Se and yields cross sections useful for isotope production and reaction studies. The results aid beam-energy selection and experimental planning for producing 76,77Br, 76As, 77Ge, and 80As and support further model refinement.

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DOI: 10.3897/nucet.12.182365

Authors: Abdalsatar Khlil Maroof, Akram Mohammed Ali

Abstract: The double-differential cross-sections (DDCS) and angular distributions for the neutron-induced reactions on the 107Ag 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 d2σ/dΩ.dE. Results show that the d2σ/dΩ.dE a☺nd angular distribution for 107Ag 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, 3He, 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.

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DOI: 10.3897/nucet.12.178562

Authors: Olumide Oluwasanmi Ige, Tongshinen S. Longmena, Jonah A. Sunday

Abstract: 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 225Ac/213Bi 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.

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DOI: 10.3897/nucet.12.170769

Authors: Abdollah Khorshidi

Abstract: Background. The radioisotope lutetium-177 (177Lu) has been suggested for utilizing in radio-immunotherapy. Now, 177Lu 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 177Lu using a 176Yb target with the TALYS and EMPIRE codes. Here, the generation yield of the 176Yb(d,p)177Yb→177Lu 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 177Lu 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.

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DOI: 10.3897/nucet.12.155378

Authors: Kseniya A. Shcherbakova, Iliya A. Lagoida

Abstract: 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.

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DOI: 10.3897/nucet.12.191338

Authors: Ekaterina A. Tyupina, Pavel P. Kozlov, Kirill A. Boldyrev, Victoria V. Krupskaya

Abstract: 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 CAH10 into less dense C3AH6 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.

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DOI: 10.3897/nucet.12.196493

Authors: Vladimir V. Chudanov, Anna E. Aksenova, Mikhail V. Bayaskhalanov, Alexey A. Leonov, Artem A. Makarevich

Abstract: 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.

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DOI: 10.3897/nucet.12.191448

Authors: Maksim Chubarov, Mikhail Zakharov, Georgy Tikhomirov

Abstract: This study presents a comparative analysis of two geometric representations of the flow domain for RANS simulations of a high-temperature gas-cooled reactor (HTGR) core segment: a realistic model with explicitly resolved fuel pebbles and a porous-medium representation based on volume-averaged properties. The computations were performed with the LOGOS software package using the k–ω SST turbulence model. The realistic model reproduces local flow and heat-transfer features of helium, including recirculation, local acceleration in interstitial gaps, and temperature hot spots, with peak values of 2.69 m/s and 1310 K. In contrast, the porous-medium approach yields a spatially smoothed solution with markedly lower extrema: the maximum velocity does not exceed 0.5 m/s and the maximum temperature is 727 K. Using precomputed resistance coefficients, the porous model predicts an average helium velocity in the pebble bed that is 76.4% lower, while the mean temperature and pressure drop differ by about 9%. The realistic representation is preferable when local thermal-hydraulic quantities are required, e.g., for transient and accident analyses. The porous-medium approximation, owing to its computational efficiency, is well suited for preliminary calculations and integral assessments in large computational domains, provided that the model is properly parameterized.

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DOI: 10.3897/nucet.12.191457

Authors: Alexandr I. Godes, Daniil V. Fedorov, Oleg V. Kondrashov, Vladimir L. Shablov

Abstract: The article is devoted to the development of an alternative fuel cycle concept for fusion reactors with magnetic plasma confinement using D-3He fuel. Turning to this type of fusion fuel is believed to solve the key problems of thermonuclear fusion installations with traditional D-T fuel: the need to handle radioactive tritium and the so-called first wall problem arising from the intense flux of high-energy neutrons from the fusion reactor. In this connection, the problem arises of improving the accuracy of predicting the characteristics of the low-radioactive D-3He fuel cycle of fusion reactors with magnetic plasma confinement and determining ways to improve the efficiency of their operation, taking into account its peculiarities: higher realization temperatures, as a consequence, higher losses for bremsstrahlung and synchrotron radiation and the use of magnetic configurations with a high value of the parameter β (β – ratio of the plasma thermal pressure to the magnetic pressure). To solve this problem, it is necessary to have a set of computational-theoretical approaches to the modelling of processes in plasma and a detailed description of the plasma energy balance, based on modern data on the rates of thermonuclear reactions and a correct description of energy losses for bremsstrahlung and synchrotron radiation. The paper presents a set of software tools for calculating the rates of the main thermonuclear reactions based on the effective radius approximation, analytical parametrizations of losses for both types of radiation, taking into account relativistic effects, and for modelling the kinetics of processes occurring in the plasma, combined with the determination of the parameters of the Lawson criterion and the fusion triple product for different modes of D-3He fuel use (fully catalyzed or fully uncatalyzed modes, 3He self-supply mode, various power gain factors and various fusion reactor sizes).

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DOI: 10.3897/nucet.12.191451

Authors: Boris A. Vasilyev, Alexandr N. Kryukov, Mansur R. Farakshin, Sergei B. Belov, Vladimir S. Sheryakov, Artem E. Kuznetsov, Ilia A. Filin

Abstract: In accordance with the strategy for the development of two-component nuclear energy with NFC closure adopted in the Russian Federation, fast neutron reactors must ensure the effective use of mixed uranium-plutonium fuel based on plutonium extracted from SNF of thermal reactors, and then from their own SNF. The first domestic BN reactors used fuel developed by that time based on enriched uranium dioxide. The development of mixed uranium-plutonium oxide (MOX) fuel began with the irradiation of experimental fuel assemblies (FAs) in the research reactor BOR-60, as well as in power reactors BN-350 and BN 600. The BN-800 design was already focused on the use of mixed uranium-plutonium fuel. Due to the delay in the launch of industrial production of MOX fuel created at the Mining and Chemical Combine, the BN-800 was launched (2016) with a hybrid core formed mainly from FAs with uranium fuel manufactured at MSZ JSC. The share of FAs with MOX fuel manufactured at the pilot production facilities of JSC SSC RIAR and the Production Association MAYAK was 16%. The hybrid core was operated until the eighth operation cycle. The transition of BN-800 core to a full loading with MOX fuel was completed by the beginning of the 13-th operation cycle (2023). For fabrication of MOX fuel, plutonium of various isotopic compositions is used, which is ensured by the method implemented at the production site for adjusting the mass fraction of plutonium in the fuel. Currently, the core is being transferred to FAs with fuel cladding from more radiation-resistant steel EK164-CW, which will subsequently allow starting work to increase fuel burnup. At that, compared with current core operation mode with the use of ChS68-CW steel for pin cladding, it is planned to increase fuel burnup from 9.5% h.a. up to 12% h.a. by the maximum value and from 66 MW·d/kg to 86 MW·d/kg by the average value. As part of the development of the technology for burning out minor actinides from the 14-th operation cycle (2024), three FAs have been irradiated in the reactor, which include four fuel elements each with the addition of americium (0.9%) and neptunium (0.6%).

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DOI: 10.3897/nucet.12.191438

Authors: Daniil M. Arkhangelsky, Yuliya S. Daichenkova, Mikhail A. Kalugin, Dmitry S. Oleynik, Denis A. Shkarovsky

Abstract: The calculation of neutron kinetics functionals, including the effective fraction of delayed neutrons (βeff) and the prompt neutron generation time (Λ), by the Monte Carlo method is known to be challenging due to the complexities involved in calculating the adjoint function. In the context of the Monte Carlo method, it is optimal to represent the importance function as the asymptotic number of descendants of a neutron placed at a given point in phase space. In practice, it is possible to consider only a finite number of generations. This limitation forms the basis of the Iterated Fission Probability (IFP) method. However, this approach is associated with several disadvantages, including the convergence of the adjoint source through a given number of generations, the appearance of statistical noise, and high memory consumption. In this paper, we present a methodology for calculating the multigroup importance function using the matrix method implemented in the MCU code. The computational model is partitioned into a finite number of tally objects and energy groups. During the modeling process, the elements of the fission matrix are tallied, and the neutron importance function for each energy group of each object is calculated at the post-processing stage. The methodology was validated by calculating βeff and Λ for 6 ICSBEP Handbook experiments. The one- and 14-group approximation of the importance function yielded almost identical results, with a negligible difference (less than 1%), due to the minor change in importance in the energy range where most neutrons are generated.

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DOI: 10.3897/nucet.12.191307

Authors: Evgeny A. Vertikov, Kirill V. Zaporzhin, Dmitry A. Oleksyuk, Vladislav A. Khamaza, Alexey M. Khudykin, Maxim A. Glazov, Oleg N. Morozkin

Abstract: The article presents the main results of the work cycle on modernizing the source code of the SC-INT computer program designed for subchannel thermal-hydraulic calculations of the water-cooled nuclear reactors cores. The mathematical description of the program is briefly provided, including the method of allocation of control volumes in space, the discrete analog of the basic conservation laws forming the system of nonlinear equations, as well as the method of its solution. The path passed on the internal modernization of the program is described in detail: ejection of outdated Fortran programming language constructions, transition to structure-oriented approach of writing source code, development of modular architecture, as well as implementation of the alternative numerical algorithm for solving the main system of nonlinear equations using the PETSc library. As an example of the SC-INT program capabilities, which appeared after the above described modernizations, the results of thermal-hydraulic calculation in fine-mesh subchannel approximation of a full-scale VVER-1000 reactor core are presented. The core under consideration is assembled from fuel assemblies of different designs: with and without installed «Vikhr» and «Progonka» type intensifier grids. It is demonstrated that the residuals on the main coolant parameters achieved in the simulation of the full-scale core match in order with the corresponding values characteristic for calculations of small-scale experimental fuel assembly models. Thermal-hydraulic calculations of full-scale cores in the subchannel approximation opens the possibility for development of coupled program complexes designed for improved estimation of the parameters of multiphysics processes in the cores of water-cooled nuclear reactors.

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DOI: 10.3897/nucet.12.173459

Authors: Shiyang Peng, Dong Wang, Shuang Li, Jiayu Li, Hua Xie

Abstract: The processing of radioactive concentrates generated during the operation of lead bismuth fast reactors is extremely difficult, with complex components including heavy metals such as Pb and Bi, radioactive nuclides such as Cs, Sr, and Co, alpha emitters such as 210-Po, as well as organic matter and suspended particles. This study focuses on the pretreatment challenges of this radioactive liquid waste, systematically screening 13 types of mature commercial flocculants and evaluating their efficiency in the flocculative removal of key heavy metals and simulated radionuclides from the concentrate. The experimental results showed that the flocculant #9 had a removal efficiency of over 99% for Pb, Bi, and Co within 20 minutes, and a removal efficiency of 96% for the 210-Po analog element Te. The effluent pH was close to neutral, which was superior to other reagents. Through optimization via single-factor experiments and Response Surface Methodology, the optimal process parameters for flocculant #9 were determined as follows: pH 7.8, dosage of 1.17 g/L, and flocculation time of 15.9 minutes. This study provides an efficient and low-cost pretreatment process for radioactive concentrate from lead-bismuth reactors, which can significantly reduce the subsequent deep purification load of 21°Po and offers technical support for the optimization of nuclear waste liquid treatment processes.

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DOI: 10.3897/nucet.12.173615

Authors: Muhammad Thowil Afif, Farisy Yogatama Sulistyo, Dmitriy G. Veretennikov

Abstract: In this study, the thermal–hydraulic behavior of the Indonesian Micro Reactor PeLUIt-40, a 40 MWt modular pebble-bed high-temperature gas-cooled reactor (HTGR), was analyzed under reduced coolant flow and loss-of-coolant accident (LOCA) conditions. Three-dimensional computational fluid dynamics (CFD) simulations using a porous-media core model and volumetric heat sources from neutronic analysis were performed. Steady-state simulations were conducted for helium flow rates from 100% to 25% of nominal, and a transient LOCA was simulated by reducing flow to zero over ten seconds. The results show that maximum core temperatures increased nonlinearly with reduced flow, exceeding TRISO fuel limits at 25% flow, while outlet duct temperatures remained well homogenized. During the LOCA, passive buoyancy-driven circulation limited temperature rise, stabilizing around 1026 °C. These findings provide preliminary insight into PeLUIt-40 thermal hydraulics, highlighting the need for further validation and extended transient analysis to confirm safety margins.

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DOI: 10.3897/nucet.12.169538

Authors: Terefa Urgessa

Abstract: Theoretical excitation functions for the alpha particle-induced reactions 59Co(α,2p4n)57Co, 59Co(α,2pn)60Co, and 59Co(α,4p5n)54Mn were analyzed and compared to the experimental excitation functions. Experimental cross sections were calculated using the PACE4 and TALYS-2.0 statistical model codes across an energy range of 33.1–54.5 MeV. PACE4 calculations used different values of level density parameters at k = 8–10, while TALYS-2.0 calculations were performed with multiple level density models (ldmodel 1–5). The analysis also examined the sensitivity and uncertainty of theoretical cross sections to changes in level density parameter. A quantitative comparison of the experimental and theoretical excitation functions was performed using calculations of reduced chi-square (χ2red) and normalized root mean square deviations (NRMSD). The findings indicate that PACE4 calculations more accurately reflect both the size and form of the experimental excitation functions compared to TALYS-2.0 calculations for the reaction channels studied. The quantitative analysis indicates the predominance of the compound nucleus reaction mechanism, suggesting that pre-equilibrium and direct contributions have a minimal impact within the energy ranges studied. This study provides a thorough justification for the nuclear theoretical models and offers consistent nuclear reaction data for alpha particle-induced reactions on a 59Co target.

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DOI: 10.3897/nucet.12.176131

Authors: Raouf M. Elfaramawy, Amr A. Sarhan, Sherif Helmy, Amged S. El-Wakeel

Abstract: The rapid penetration of renewable energy sources (RESs) into modern power systems has introduced unprecedented challenges in frequency regulation, primarily due to the intermittency and variability of wind and solar generation. Traditional large-scale nuclear and fossil-fuel-based power plants, while reliable, lack the flexibility required for rapid response to frequency fluctuations in high-renewable environments. Small modular reactors (SMRs) have emerged as a promising alternative, combining nuclear reliability with enhanced operational flexibility (Elfaramawy et al. 2025). This research presents a comparative study of two control strategies—proportional-integral-derivative (PID) and model predictive control (MPC)—for frequency regulation in a renewable-rich grid, where SMRs act as flexible balancing resources. Using validated Simulink-based dynamic models of SMR dynamics, grid frequency response, and renewable variability, the performance of PID and MPC controllers in mitigating disturbances caused by renewable fluctuations and large-scale contingencies was analyzed. Results indicate that both controllers provide effective frequency regulation, with MPC offering modest advantages in overshoot suppression and settling time. While PID demonstrates reliable performance and ease of deployment, MPC enables predictive optimization and explicit constraint handling. The findings suggest that both control strategies are viable for SMR-based frequency support, with MPC providing incremental improvements in dynamic response for future high-renewable grids.

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DOI: 10.3897/nucet.11.182048

Authors: Alexander M. Boldyrev, Evgeny F. Seleznev, Evgeny P. Lyapin, Andrey M. Tuchkov, Sergey G. Stolyarov, Yan S. Gaitanov, Ilya Yu. Chernyshov, Anton A. Peregudov, Mikhail Yu. Semenov, Ivan V. Tormyshev, Elizaveta P. Averchenkova, Vyacheslav A. Mishin, Daria V. Panova, Vsevolod V. Losev

Abstract: 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”.

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DOI: 10.3897/nucet.11.170677

Authors: Ruslan A. Vnukov, Elena A. Rodina, Yurii S. Khomyakov

Abstract: The possibility of obtaining 238Pu 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 238Pu, 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 238Pu with target purity in terms of 236Pu within 2 ppm. When using Am, 236Pu purity is also ensured, but 238Pu fraction is somewhat lower than 80% due to the higher fraction of 242Pu. The use of Am and Np mixture makes it possible to reduce the fraction of 236Pu and prevent high accumulation of 242Pu, but in this case 242Cm accumulation should be taken into account, and appropriate time period for its decay heat reduction should be chosen.

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DOI: 10.3897/nucet.11.182046

Authors: Yuri V. Nosov, Nikolay N. Oshkanov, Mikhail V. Bakanov, Ivan I. Sidorov

Abstract: 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.

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DOI: 10.3897/nucet.11.168649

Authors: Vladimir A. Apse, Vasily B. Glebov, Evgeny G. Kulikov, Gennady G. Kulikov

Abstract: 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 236Pu in plutonium and leaves only a limitation on the proportion of 238Pu (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.

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DOI: 10.3897/nucet.11.161809

Authors: Sachin Ambade, Chetan Tembhurkar, Devendra Shahare

Abstract: 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.

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DOI: 10.3897/nucet.11.155986

Authors: Cherie Sisay Mekonen, A. V. Mohan Rao, Yohannes Amare Dessie

Abstract: This study applies the COMPLET nuclear reaction code to calculate excitation functions for eleven alpha-induced reactions on stable copper (63Cu, 65Cu) and antimony (121Sb, 123Sb) isotopes, aiming to predict production cross-sections for medically significant radionuclides such as 68Ga, 67Ga, 66Ga, 65Zn, 124I, and 123I. 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 n0 = 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.

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DOI: 10.3897/nucet.11.176910

Authors: Denis G. Zaryugin, Sergey T. Leskin, Stepan A. Mikhin, Vladimir V. Sergeev, Victor I. Slobodchuk

Abstract: For reactors with high decay power (VVER-1000, VVER-1200, VVER-TOI, etc.), no reliable reactor vessel cooling is achieved via the concept of retaining the corium inside the reactor pressure vessel only by means of heat transfer to water in the process of its boiling within the reactor vessel vault since high heat fluxes (density above 106W/m2) cause boiling evolving into film boiling and to a heat transfer crisis. A comparative analysis of different cooling systems shows that the most reliable and effective way to prevent the heat transfer crisis is an intensification of heat transfer on the cooled surface by using an air-droplet spray device (ADSD), which can be used to retain the core melt inside the reactor pressure vessel. The paper provides an overview of studies on gas-droplet cooling of high-temperature surfaces, and presents preliminary calculations using correlations of the non-wetting cooling mode as being of greatest interest for the ADSD. Methods have been considered for simulating gas-droplet cooling in system codes. The qualitative agreement of the calculation results with experimental data is shown. The peculiarities of spray cooling of the large VVER reactor vessel are identified. A diagram of the possible locations of spray nozzles within the reactor vessel vault is demonstrated. Due to the lack of a single verified spray cooling model for high-temperature surfaces, experimental studies are considered to be the key approach in the scientific problem under consideration, that is cooling of the large VVER reactor vessel in the event of a beyond design basis accident.

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DOI: 10.3897/nucet.11.176904

Authors: Aleksey V. Popov, Egor N. Kulakov, Daria R. Danilova, Alexander S. Lisyanskii, Fyodor A. Svyatkin, Sviatoslav O. Kiriyenko, Egor V. Bubenshikov, Irina V. Pavlova, Victor V. Khodakovsky, Valery N. Chigarev

Abstract: During the design of the secondary circuit for a nuclear power plant with the BREST-OD-300 reactor, engineers and designers faced specific challenges related to the properties of the lead coolant. The key issue was its high crystallization temperature, which necessitated fundamentally new approaches to the thermal circuit design of the power unit. Particular attention was paid to developing the feedwater heating system, as its parameters and configuration directly affect the thermal balance and operational stability of the reactor plant. Throughout the design process, the configuration of this system underwent multiple revisions, reflecting the complexity of achieving optimal technical solutions while meeting stringent reliability and safety requirements. Each modification to the feedwater heating system design requires comprehensive analysis of its impact on all key components of the power unit, including both the reactor and turbine circuits. Engineers employed an Integrated Сalculating Mathematical Model (ICMM) to evaluate the system performance in normal and emergency operating modes, incorporating interconnections between all technological processes. This approach not only helps identify potential issues at early design stages but also optimizes equipment operating parameters. This approach achieves significant improvements in the power unit’s economic efficiency while simultaneously meeting all reliability and safety requirements. The research placed special emphasis on analyzing accident scenarios, particularly the feedwater cooldown line failure. Engineers applied mathematical modeling to pinpoint critical parameters and design failure mitigation solutions. These included upgrades to backup heating systems and optimization of control algorithms. Thus, this work highlights the necessity of employing modern modeling methods when developing innovative nuclear reactors. Such approaches enable advance development of technical solutions to ensure safe and efficient operation of these advanced power units.

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DOI: 10.3897/nucet.11.167007

Authors: Sinem Uzun, Yasin Genç, Hanan Rifai, Ouadie Kabach, Bedrettin Uzun, Zouhair Sadoune, El Mahjoub Chakir, Hamid Amsil, Fadi El Banni

Abstract: Nuclear energy sustainability and deployment flexibility can be significantly enhanced through Small Modular Reactors (SMRs) technology. Critical to their operational success is the thorough assessment of thermal-hydraulic characteristics, especially when incorporating advanced fuel design concepts. This research conducts an extensive thermal-hydraulic analysis examining various thorium-based fuel formulations, including (Th-235U)O2, (Th-233U)O2, and an innovative (Th-233U-235U)O2 composition, benchmarked against standard UO2 fuel. The investigation encompasses both solid fuel arrangements and dual-cooled annular assembly designs, focusing on safety optimization and operational efficiency enhancement. The analysis focuses on key safety parameters, including pressure drop, coolant enthalpy, fuel centerline temperature, and Departure from Nucleate Boiling Ratio (DNBR). Results for solid fuel configurations reveal that thorium-based fuels exhibit reduced pressure drop, more efficient enthalpy distribution, lower peak fuel temperatures, and higher DNBR values compared to conventional UO2, highlighting improved thermal stability and safety margins. The (Th-233U-235U)O2 mixture demonstrates a balanced performance by mitigating the limitations of other thorium compositions. In annular configurations, all fuel types benefit from enhanced heat removal due to the dual cooling surfaces, resulting in further reductions in pressure drop and peak temperatures, as well as a significant increase in DNBR values. The highest DNBR, reaching up to 3.051, confirms the annular geometry’s superior safety performance against boiling crises.

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DOI: 10.3897/nucet.11.173059

Authors: Aleksey F. Grachev, Liudmila M. Zabudko, Evgeny E. Marinenko, Sergey I. Porollo, Anna V. Belyaeva, Feodor N. Kryukov, Vadim G. Teplov, Mikhail V. Skupov

Abstract: 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.

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DOI: 10.3897/nucet.11.173061

Authors: Anatoly A. Kazantsev, Olga V. Supotnitskaya, Evgeniya A. Ivanova, Nataliya E. Astakhova

Abstract: 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.

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DOI: 10.3897/nucet.11.159699

Authors: Alexandra A. Zyryanova, Yulia V. Volchikhina, Ivan M. Russkikh, Oleg L. Tashlykov

Abstract: 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.

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DOI: 10.3897/nucet.11.163189

Authors: Hanan Rifai, Ouadie Kabach, El Mahjoub Chakir, Zouhair Sadoune

Abstract: 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.

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DOI: 10.3897/nucet.11.159698

Authors: Senamaw Mequanent Zegeye, Yihunie Hibstie Asres, Guadie Degu Belete, Teshale Ayalew Mekuriyaw, Yesigat Enawugaw Alemu, Kelebie Tamiru Worku, Mekonen Amare Abeje

Abstract: 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.

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DOI: 10.3897/nucet.11.157254

Authors: Ayan Kumar Makar

Abstract: 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.

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DOI: 10.3897/nucet.11.169077

Authors: Aleksei V. Panov, Gennady V. Koz’min

Abstract: The article presents the experience of training foreign students of Rosatom State Corporation’s flagship universities at the MEPhI Resource Center. It is shown that over the period 2018–2024, more than 500 bachelor’s, specialist and master’s students from 33 countries studying at the five largest flagship universities of the state corporation (VSU, MEPhI, SPbPU, TPU, UrFU) underwent practical training in English and Russian at the Resource Center of Obninsk Institute for Nuclear Power Engineering (IATE MEPhI). Students were trained in six educational programs of 13 and 14 specialty groups. In the field of non-energy application of nuclear technologies, practice-oriented training at the Resource Center is carried out to a greater extent on radiation technologies in agriculture and the food industry. High results of cooperation in training foreign students in the field of agro-nuclear technologies between the IATE MEPhI and the National Research Center “Kurchatov Institute” – RIRAE were noted. Practical training of students at RIRAE is carried out on a unique gamma-installation GUR-120 for irradiation of agricultural raw materials and food products for the purposes of phytosanitary and microbiological safety, increase in shelf life, radiation stimulation. The competencies acquired by students in the field of agricultural radiation technologies are in demand by them when applying for jobs in the Centers of Nuclear Science and Technology. The interaction of the Resource Center of the IATE MEPhI and the leading research institutes of Obninsk in training foreign students is the basis for the development of the international scientific and educational center “Obninsk Tech” for training personnel for foreign businesses of the State Corporation “Rosatom” in the nuclear industry and related areas of application of nuclear technologies.

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DOI: 10.3897/nucet.11.169083

Authors: Svetlana V. Kravets, Dmitry A. Klinov, Yury A. Kurachenko

Abstract: The paper presents the results of determining the possibility of using the horizontal experimental channels of the MBIR reactor for neutron capture therapy studies. The collimator configuration for the neutron beam extraction with specified properties was justified computationally. The peculiarities of the reactor give grounds for a positive assessment of this prospect, primarily the hard spectrum and the uniquely high intensity of the beams. The paper considers the capability of channel No.5 as the most suitable for neutron capture therapy due to a combination of characteristics. The simplest possible axisymmetric collimator was selected for the calculations to assess the key functionalities of neutron capture therapy. The configuration and material composition of the collimator are defined by the experience of calculations. Two fundamental characteristics were analyzed to assess the capabilities of the neutron beam of MBIR’s channel No.5 for neutron capture therapy. These are the dose in the target (soft tumor tissue) containing 65 ppm of 10B, and the dose in healthy tissue containing 18 ppm of 10B. The task in the series of calculations was as follows: to determine the dynamics of the key values for neutron capture therapy with a variable thickness of the moderator in the collimator channel – the time for gaining a fixed “therapeutic” dose in the target (tumor) and the time for gaining the maximum “tolerance” dose in healthy tissue when the target moves along the depth of the phantom. The distribution of these characteristics through the depth of the tissue allows us to conclude that the beam extraction configuration under consideration is effective. The obtained results of the spectral neutron distribution at the outlet of channel No.5 and the estimated dose characteristics in healthy tissue and in the tumor confirm that it is technically possible to use this channel for neutron capture therapy.

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DOI: 10.3897/nucet.11.151728

Authors: Afroza Shelley, Md. Abidur Rahman Ishraq

Abstract: The objective of this study is to manage the initial excess reactivity of the VVER-1200 reactor through different approaches, including the use of burnable absorbers and control rods in fuel assemblies with varying enrichments (from 4.95 wt.% to 2.0 wt.%) by using the Monte Carlo code Serpent. It is found that 30 integrated burnable absorbers (IBAs) rods are required to reduce the infinite multiplication factor (kinf) from 1.32408 to 1.06983 in assemblies containing 4.95 wt.% enriched fuel. On the other hand, assemblies with enrichment below 2.5 wt.% do not require any IBA rods. Control rods (CRs) can also effectively reduce excess reactivity at the beginning of life (BOL), provided they are inserted more than 80% of the assembly’s height. When fully inserted, the CRs provide a reactivity worth of 85.5 pcm/cm in 4.95 wt.% enriched assemblies at BOL. Fuel enrichment has a significant impact on delayed neutron fraction as well as the FTC, and MTC. While the effect of IBA rods on these parameters is found to be insignificant; however they do harden the neutron spectrum at the BOL due to the thermal neutron absorption by isotopes 155Gd and 157Gd.

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DOI: 10.3897/nucet.11.161715

Authors: Sergei I. Spiridonov, Rena A. Mikailova, Vladimir K. Kuznetsov

Abstract: The regulations for radioecological monitoring (REM) of agroecosystems in areas surrounding operational nuclear power plants (NPPs) are characterized by a high degree of detail. Obtaining the required empirical data involves significant financial costs. This raises the question: to what extent do monitoring data reflect the impact of NPPs on the agricultural environment? Based on long-term emissions data from the Kursk and Rostov NPPs, the dynamics of radionuclide accumulation in soil originating from NPPs were calculated. It was shown that the concentrations of long-lived 137Cs and 90Sr in soil increase over time during NPP operation by 30–40 times. However, a comparison of the calculations with REM data over many years indicates that the contribution of NPP-derived 137Cs and 90Sr to the specific activity of these radionuclides in soil does not exceed 0.01–0.1%. Thus, REM primarily detects 137Cs and 90Sr of global and Chernobyl origin. Due to the negligible NPP contribution, the spatial distribution of 137Cs within 30-km zones is uniform and follows a lognormal distribution. According to estimates, the activity levels of NPP-derived radionuclides in soil, recommended for REM, are below detection limits. The study concludes that REM guidelines for agroecosystems around NPPs should be optimized. Adjustments may include revising the list of radionuclides analyzed, reducing the types of agricultural products monitored, and increasing the intervals between sampling. The need to focus efforts on developing REM programs for assessing impacts in normal and emergency situations is emphasized.

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DOI: 10.3897/nucet.11.155487

Authors: Cherie Sisay Mekonen

Abstract: This study examines the current status, challenges, and future prospects of nuclear science education and training (E&T) in Ethiopia. Nuclear education began in the 1960s with basic theoretical concepts, and the first structured academic efforts emerged in the 1990s through MSc program at Addis Ababa University. However, there are still no dedicated undergraduate degrees or specialized training programs in nuclear science. To identify a path forward, the study compares Ethiopia’s progress with selected developed countries that have successfully leveraged nuclear technology, emphasizing the importance of collaboration and shared experience. The study focuses on four essential pillars: human resource development, knowledge management, knowledge networks, and long-term sustainability. It aims to serve as a communication tool for both national policymakers and institutions to support the growth of nuclear E&T in alignment with Ethiopia’s development goals. Major challenges include inadequate infrastructure, limited funding, lack of qualified personnel, and insufficient laboratory facilities. If Ethiopia’s present nuclear energy ambitions are to succeed, the government will require competent human resources to develop nuclear facilities, assure safe and secure operations, and conduct nuclear E&T and research. To address these needs, the study recommends developing internationally aligned curricula, investing in skilled faculty and technicians, expanding research infrastructure, and building public trust through awareness and education. The paper concludes by highlighting the critical role of international and national collaboration in strengthening nuclear education in Ethiopia and presents practical recommendations to guide future development in areas such as energy, health, agriculture, industry, and other disciplines.

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DOI: 10.3897/nucet.11.160017

Authors: Anton S. Lapin, Viktor Yu. Blandinsky, Vladimir A. Nevinitsa, Stanislav B. Pustovalov, Alexey A. Sedov, Stanislav A. Subbotin, Pyotr A. Fomichenko

Abstract: Two stages of the MTRR-SCW reactor operation are planned: a test stage and a research stage. This paper considers the research stage of the MTRR-SCW experimental reactor operation, the purpose of which is to investigate current and advanced light-water reactors. The MTRR-SCW driver-type core provides a fast neutron spectrum with the possibility for the local warmup in ampoule devices and independent loop channels. Irradiation channels will be installed in the core center and periphery, as well as instead of the reactor’s changeable reflector cartridges. The MTRR-SCW irradiation channels and independent loops will provide ample opportunities both for undertaking a research on effects from neutron irradiation of different materials, and for testing a variety of fuel assembly designs and operating conditions (temperature, pressure, neutron spectrum), as well as for investigating transients and emergency processes. The MTRR-SCW channels can be used to irradiate different types of fuel, and structural and absorbing materials with different coolant inlet temperatures (from 250 to 450 °C) and, consequently, its inlet density (from 800 to 100 kg/m3 respectively), providing different neutron spectrum options for the experimental fuel assembly in a range from thermal to fast spectrum. The MTRR-SCW allows experiments to increase power and simulate emergency processes, including reactivity accidents (RIA). The strong primary and safeguard vessels of the independent loop channels also make it possible to simulate loss-of-pressure emergencies of the LB LOCA and SB LOCA types. The peripheral independent loop channel will allow undertaking experiments for simulation of alternative reactor concepts with reactors with SKD coolant parameters, such as a single-circuit concept with the pseudophase transition in the core (VVER-SKD-1700), and with natural coolant circulation in the core (SKDI). In addition, the peripheral channel allows accelerated irradiation of fuel rods used in current VVER reactors, taking into account the reproduction of the ratio between damaging dose and burnup rates.

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DOI: 10.3897/nucet.11.152695

Authors: Aleksandr A. Savelev, Valeriy I. Rachkov

Abstract: The paper presents the results of the development of modified solid-phase extractants (SPE) based on N,N,N’,N’-tetraoctyl diglycolamide (TODGA) intended for the selective extraction of americium-241 from nitric acid solutions of liquid radioactive waste. (LRW). The structural and physico-chemical properties of synthesized materials, including granulometric composition, porosity, and stability in nitric acid media, have been studied. The developed SPEs are stable in nitric acid solutions and meet the size requirements for use in industrial sorption columns. The results obtained can be used as a basis for creating technological schemes for the processing of LRW based on synthesized experimental modified samples of SPE TODGA, in particular for the isolation of minor actinides, including americium-241.

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DOI: 10.3897/nucet.11.160012

Authors: Petr B. Baskov, Boris S. Salamakha, Yakov V. Glazyuk, Artur A. Namakshinas, Sergey A. Bondarenko, Ilya M. Mushin, Alexander S. Khudin

Abstract: The proposed solution for the production of new boron-containing radiator materials – composite coatings with neutron conversion material (10B isotope) – for ionization chambers electrodes is based on the technology of chemical and structural modification of the surface using heat-resistant oxide materials (silicon and boron oxides). The paper presents the results of the development of a neutron-sensitive radiator material based on an oxide micro-dimensional composite coating consisting of an intermediate adhesive layer of silicon dioxide (SiO2) and a neutron-sensitive functional layer of boric anhydride (B2O3). The technological basis consists of the sequential processes of thermal destruction of polyorganosiloxane of the PES-5 brand and pyrolytic decomposition of boric acid (H3BO3). Studies using infrared and fiber-optic interferometric spectroscopy have shown that during the formation of a silicon dioxide layer, an amorphous silicate of a linear chain structure with a developed surface is formed, which contributes to the subsequent formation of a functional boric anhydride layer. The specific neutron sensitivity of boron-containing radiator coatings was measured by alpha spectrometry. It was found that at a surface density of boric anhydride of 2.5 mg/cm2, the specific neutron sensitivity is of the order of 3∙10-18Coulomb/neutron. It is shown that the boron-containing radiator coating retains its integrity and nuclear physical properties during thermocyclic tests (4 cycles at 600 °C). The boron-containing radiator coating is characterized by high adhesion properties to the metallic surface of the electrodes (grade 321 steel) of the ionization chamber. The composite coating is resistant to vibration when exposed to high-frequency (200 Hz) and low-frequency (6–35 Hz) vibration loads.

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DOI: 10.3897/nucet.11.160016

Authors: Sergey V. Terekhov

Abstract: Use of liquid-metal coolants in nuclear power plants has been the cause of unfailing interest in thermal-physical properties of metals and their alloys from both experimentalists and theorists. Power polynomials are used by experimentalists to approximate temperature changes in heat capacity, the coefficient of thermal linear (volume) expansion and other quantities. These polynomials have different form in different temperature intervals and need to be joined at the interval ends. This approach creates a number of difficulties in developing a unified methodology for calculating not only the thermal functions of metals, but also for predicting the behavior of their melts and alloys. The following was used in the paper to solve the problem at hand: the author’s model of a two-phase local-equilibrium region (with different order parameters) and the modified rule of component mixing, taking into account the coordinated arrays of experimental data on the initial metals for calculating predictively the thermal-physical performance of potassium and sodium melts. It has been shown that using the model of a two-phase local-equilibrium region, new approximating functions and empirical formulas lead to a sufficiently adequate estimation of heat capacities, thermal linear expansion coefficients, densities, thermal conductivities and thermal diffusivities of melts. A discrepancy has been found between the mathematical description of the thermal conductivity of the K56Na44 melt and its experimental values.

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DOI: 10.3897/nucet.11.149291

Authors: Dmitriy G. Veretennikov, Abdelmeguid Abouellail, Ahmed H. Ali, Sergey V. Bedenko

Abstract: The study focuses on modeling a device designed for conducting experiments under low reactor neutron fluence rate conditions. The relevance of the work is driven by the challenges of accurately measuring low reactor neutron fluence rates, which are critical for the safety and efficiency of nuclear reactors and for conducting research in various scientific fields. To create a directed neutron flux, a horizontal collimation channel of the Prizm-AN irradiation device with a radioisotope capsule neutron source of the internal bounded neutron (IBN) type is being investigated. The paper presents the results of particle transport modeling using the PHITS and SOURCES-4C codes, which allowed for the evaluation of the generated neutron fluence rate parameters and the effective dose rate of radiation. The obtained results confirm the functionality of the proposed device and its compliance with radiation safety requirements, making it suitable for conducting experiments similar to those performed on reactor installations.

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DOI: 10.3897/nucet.11.143957

Authors: Md. Abidur Rahman Ishraq, Sabyasachi Roy, Valery Victorevich Afanasiev

Abstract: In this study, a simplified computational model of the blanket mock-up is created using the SERPENT Monte Carlo Code. The nuclear data is obtained from the enriched ENDF/B.VII.1 data library to conduct this study. The model is validated, as the error percentage for 63Cu(n,2n)62Cu and 65Cu(n,2n)64Cu reactions is less than 10% when compared to experimental results. The computational model is used to calculate the tritium production rate in different lithium zones with various neutron multipliers (U, Pb) and without any multipliers. The results show that the tritium production rate with a uranium multiplier is 86% higher than with a lead multiplier and 238% higher than with no multiplier. The neutron energy spectrum shows a peak in the 0.1 MeV to 10 MeV energy range for every case. This study also examines the effects of fusion neutrons on different isotopes, providing valuable data on how materials behave under high-speed neutron exposure.

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DOI: 10.3897/nucet.11.151566

Authors: Petr B. Baskov, Boris S. Salamakha, Yakov V. Glazyuk, Artur A. Namakshinas, Sergey A. Bondarenko, Ilya M. Mushin, Alexander S. Khudin

Abstract: The paper presents the results of material testing for the purpose of obtaining radiator material – acomposite coating with neutron conversion material – for ionization chambers (IC) which contain the 6Li isotope and convert neutron radiation to a flux of high-energy charged particles through the 6Li(n, α)3H nuclear reaction. The proposed method for forming lithium-containing radiator material allows ensuring a high temperature resistance of up to 600 °C and a mechanical strength at the expense of adhesion to the IC electrode material (grade 321 steel). The advantages of a lithium-containing radiator, compared to a boron radiator, are explained by the smaller cross-section of the 6Li-neutron interaction: the smaller efficiency of the "neutron → charged particle" conversion is made up for by a high power density and a prolonged free path of reaction products in the radiator material, which makes it possible to increase the surface density of 6Li atoms, while reducing the extent of "burnup" in neutron fields. The IC electrode radiator material consists of a two-layer composite coating comprising an adhesive silicate layer and a functional neutron-sensitive lithium fluoride layer. Measurements at an alpha spectrometric facility have shown that the coating has a high energy output (~ 2.8·10–3 MeV/neutron), which remains stable after four thermal cycles of up to 600 °C. The coating is resistant to vibration when exposed to frequencies of 35 to 200 Hz. The paper presents the results of testing the IC mockup with a lithium-containing radiator material. When irradiated with a neutron flux of 6·103 cm–2·s–1, the IC mockup sensitivity value was about 10–15A·s·cm2/neutron, which agrees with the calculated value.

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DOI: 10.3897/nucet.11.151561

Authors: Vladimir I. Folomeev, Svetlana M. Ganina, Alexandr G. Isakov, Liudmila M. Zabudko, Evgeny E. Marinenko

Abstract: In the process of a post-irradiation examination of nitride pins irradiated as part of experimental BN-600 reactor fuel assemblies, a noticeable elongation and a higher ovality of nitride pins cladding were observed as compared with oxidefuel pins. The analysis results suggest that this takes place due to a local mechanical effect fuel pellets have on the cladding. The potential cause for the fuel-cladding thermomechanical interaction is the fuel pellet misalignment, the pellet jamming and cracking as a result of thermal stresses under reactor power rise, and subsequent jamming of large fuel fragments in the cladding. The paper presents the fuel jamming model which assumes the fuel-cladding interaction via an axial force in each calculated radial layer. The axial force is determined from the condition of equilibrium in the radial cladding and fuel cross-section and ensures that the jamming condition (the same growth in axial deformation) is complied with. Jamming is assumed along the entire core height in the presented model version. The presented results of the jamming model verification, based on post-irradiation examination results of the BN-600 EFA-11 (experimental fuel assembly), show that the model allows simulating the fuel pin cladding elongation. The EFA-11 fuel assembly with fuel pins of the BREST-OD-300 reactor type with the cladding made on EP823-Sh ferritic-martensitic steel and nitride uranium-plutonium fuel was irradiated in the BN-600 reactor to the maximum burnup of 9 at.% and the maximum damage dose of 107.6 dpa.

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DOI: 10.3897/nucet.11.150529

Authors: Nikita V. Kovalev, Aleksandr M. Prokoshin, Polina V. Davydova, Vladimir A. Korolev

Abstract: The paper examines the radiation characteristics of noble platinum group metals (PGMs) extracted from spent nuclear fuel (SNF) of the VVER-1000 reactor. These are the so-called reactor-grade ruthenium, rhodium and palladium. PGMs are radioactive when extracted from SNF, but after the decay cooling of ruthenium for about 27 years, and of rhodium for about 13 years, they can be used in unlimited quantities. There is no sense in decay cooling of reactor grade palladium due to its radioactive isotope 107Pd having a half-life of 6.5 million years. As specified by regulatory documents, such palladium can be freely used only in quantities of up to 34 g. Pd is a soft beta emitter with a maximum beta particle energy of 34 keV. The calculation results show that the mean free path of beta particles from 107Pd in palladium metal is 0.8 μm, so reactor-grade palladium emits only from the surface layer, and other electrons are absorbed in the material itself. The mean free path of electrons with an energy of 34 keV in biological tissue is about 20 μm, which does not exceed the thickness of the skin epidermiscorneous layer. Calculations have shown that the equivalent dose rate (EDR) on the surface of reactor-grade palladium is 0.04 μSv/h, which is below the public EDR value. As a result, a conclusion is made that reactor-grade palladium does not pose a danger in the event of external contact.

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DOI: 10.3897/nucet.11.139558

Authors: Nikita Popov, Alexey Suchkov, Mikhail Zharkov, Veronika Kirillova, Alexander Bazhenov, Pavel Dzhumaev, Alexey Vertkov, Oleg Sevryukov

Abstract: Lithium-based capillary porous systems (CPS) made of tungsten mesh are a part of the most prospective approach to plasma facing components. Currently, tungsten mesh as a part of the CPS is mounted directly on the experimental assembly without a proper joining with a substrate. Tungsten mesh filled with steel could be used as a base structure for the CPS. The experiment considers the wetting of tungsten by steel melt and the features of short- and long-term interaction between the two materials. Wetting was studied by improved sessile drop experiment. The results show that an average contact angle is 69° for SS316LN and 83.2° for SS420 melt on tungsten substrate with a temperature of 500–650 °C. Tungsten-steel composite was manufactured by infiltration of tungsten mesh with a steel melt. As a result of an active dissolution of tungsten in steel melt, (Fe,Cr)7W6 interaction layer with a thickness up to 10 µm forms around tungsten. Optimal structure with the thinnest intermetallic layer is obtained in the zones with the lowest temperature and the highest cooling speed.

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DOI: 10.3897/nucet.11.127108

Authors: Anatoly N. Shmelev, Vladimir A. Apse, Evgeny G. Kulikov, Gennady G. Kulikov, Vasily B. Glebov

Abstract: The paper presents the results obtained in numerical evaluations of a possibility to reach self-sufficiency of fissile materials in the joint system of fast and thermal reactors. These studies considered the joint system consisting of thermal light-water reactors of VVER-type and fast lead-cooled reactors of BREST-type, which are operated within the frames of the closed (Th-U-Pu) fuel cycle. It was assumed that fast reactors (FR) of BREST-type used the mixed thorium-plutonium nitride fuel, while thermal reactors (TR) of VVER-type used the mixed oxide fuel of natural uranium and 233U. Uranium isotope 233U was produced in the Th-fraction of FR fuel for further introduction into the fresh composition of TR fuel, while plutonium was produced in the natural uranium fraction of TR fuel for further introduction into the fresh composition of FR fuel. The numerical studies resulted in the determination of the conditions necessary to provide fuel self-sufficiency in the joint TR-FR system. The following key findings of the research may be noted: – It is demonstrated the possibility to create the joint TR-FR system with inherent fuel self-sufficiency; – Involvement of thorium and 233U in the closed NFC of the joint TR-FR system can arrange an optimal regime for production and consumption of main FM; – Purposeful change of thermal power and introduction of natural uranium in the fuel composition of BREST-type FR made it possible to reach fuel self-sufficiency of the joint TR-FR system; – Application of radiogenic lead instead of natural lead allowed us to reduce necessary values of thermal power and content of natural uranium nitride in the fuel of BREST-type FR.

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DOI: 10.3897/nucet.11.144671

Authors: Anton S. Lapin, Viktor Yu. Blandinsky, Vladimir A. Nevinitsa, Stanislav B. Pustovalov, Alexey A. Sedov, Stanislav A. Subbotin, Pyotr A. Fomichenko

Abstract: The nuclear power system has faced a challenging issue of significantly improving the characteristics of nuclear fuel breeding while maximizing the advantages of the technology of vessel-type pressurized water reactors used extensively in nuclear power. This is possible via switching to supercritical coolant parameters. An increase in production of fissile nuclides, as compared with traditional pressurized water reactors, is achieved by switching to a harder neutron spectrum due to reducing greatly the coolant density and using a dense fuel lattice. A necessary condition for the VVER-SKD design development is the establishment of an experimental base. A multipurpose test research reactor, MTRR-SCW, is the potential testing ground for the reactor technology, and for new structural and fuel materials and fuel rods. The paper presents the key characteristics of the MTRR-SCW reactor, as well as the potential MTRR-SCW applications at different stages of its operation (testing and research). A concept is proposed at the initial stage for the reactor phased rise to power, which will make it possible to justify the efficiency of the MTRR-SCW fuel with increased linear loads through experiments in the independent central loop channel. This concept also involves phased validation and study of the joint operation of the reactor plant and the steam turbine plant as part of the MTRR-SCW nuclear power plant. At the research stage of operation, safe operating limits need to be determined, and the choice of normal operating modes for the VVER-SKD power reactor justified, and experimental studies need to be undertaken to investigate the behavior of structural materials and fuel compositions as part of experimental fuel rods for the advanced light-water reactor cores with different neutron spectra. Long-term irradiation of experimental fuel rods is planned to be carried out in the MTRR-SCW’s independent peripheral loop channel, and experimental simulation of emergency processes to be performed in the reactor’s central loop channel. This paper deals with the issues to be addressed prior to starting the VVER-SKD power reactor design. Issues have been identified that can be fully or partially solved at effective facilities, as well as the applications for the MTRR-SCW prototype reactor. The paper presents the key characteristics of the MTRR-SCW reactor, and describes in detail the concept for the phased development of the research reactor capabilities and phased rise to power.

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DOI: 10.3897/nucet.11.138995

Authors: Oleg Konstantinovich Vlasov, Irina Aleksandrovna Zvonova, Nataliia Vladilenovna Shchukina

Abstract: The work was aimed to the study the 137Cs and 131I transport dynamics along the trophic chain, from the atmosphere to soil, vegetation, dairy cattle, milk, and the human body. This was done with the use of the radioecological simulation model based on instrumental data from the “Warsaw” scenario of the International Atomic Energy Agency’s (IAEA) EMRAS project, which were collected after the Chernobyl nuclear accident. The created simulated radioecological model was used for reconstruction of the 131I radioactivity in dairy cattle milk in the Warsaw and Ostroleka areas located in the central part of Mazovia, where the cattle dairy milk radiometry was performed in April-May 1986. The aim of the study was to verify the radioecological model, to assess the reliability and uncertainty of the milk block of the computed model. It was found that the successive account of discrepancies between computed and reconstructed instrumental data of 131I radioactivity dynamics in green fodder on farms in Mazovia areas and the shortage of clean fodder stocks harvest lead to a successive improvement of the agreement between computed and instrumental data of 131I activity in the dairy milk. For all variants of accounting correction the ratio of computed data to instrumental data with recalculated 131I activities in the atmosphere of the Ostroleka Area is on average 2.5–3 times closer to their ideal value of 1 than in calculations with instrumental data for a cloud. The prognostic properties of the computational model at the stage of reconstruction of the specific 131I activity in milk, estimated as the ratio of computed data to instrumental data, can be estimated as 1.05±2.0; the values of these ratios for correction of thyroid radiation doses to the population due to the contaminated milk consumption are estimated as 1.3±2.5.

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DOI: 10.3897/nucet.11.130422

Authors: Md. Abidur Rahman Ishraq, Anton Evgenivich Kruglikov

Abstract: The focus of this work is to analyse the suitability of two-group diffusion coefficients and macro constants generated from SERPENT using out-scattering approximation (OSA), transport correction (TRC) and cumulative migration methods (CMM) for fuel and non-fuel materials. For this purpose, various assembly and core models of ACP-100 SMR were designed. Assessment of these constants was conducted using COMSOL Multiphysics. For six distinct fuels, the best models were proposed with the least error margin in keff. Fuel material affects the group constants of non-fuel components except for radial reflectors. The sufficiency of two-group calculation was justified through spectrum analysis. Additional analysis revealed that MOX-RG has the hardest spectrum among all the fuels. Moreover, the effectiveness of boric acid to control excess reactivity was observed. Subcriticality was achieved for all fuel types except MOX-RG at a boric acid concentration of 4500 ppm. The influence of variation of boric acid concentrations on group constants was investigated using TRC and OSA. The reactivity difference between SERPENT and COMSOL was determined. It was found that OSA generates the most accurate results for MOX-RG with maximum 863 pcm error, while TRC produces higher accuracy with maximum error of approximately 250 pcm for other fuels.

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DOI: 10.3897/nucet.10.144559

Authors: Gleb W. Karpovich, Yuri A. Kazansky, Nikita D. Vasechkin, Kirill A. Bakhantsov

Abstract: There is currently a consensus among the scientific and engineering community regarding the solution to the problem of minor actinides (MAs) formed in the process of nuclear power operation: MAs need to be converted to fission products during burnup in power reactors. Fast neutron reactors (BN, BREST) and molten salt reactors (MSR) are considered largely to this end. Despite the advantages of using fast reactors, there are no currently power unit designs with BN or BREST reactors available for commercial operation. The possibility of using VVER reactors for this purpose is rarely covered in scientific literature, despite the fact that the technology of light water reactors has long been mastered, and preparing a VVER-based MA burner reactor is technically simpler than on the basis of a pilot commercial BN technology or BREST or MSR reactors at different R&D stages. In the previously published first part of the study by Kazansky and Karpovich 2022, the fuel cycle for VVER reactors was investigated: minor actinides obtained during reactor operation were extracted from spent nuclear fuel and added to fresh fuel for the same reactor. It has been found that implementing this cycle and bringing the concentration of MAs up to 4 wt. % makes it possible to reduce the amount of minor actinides produced in the VVER reactor by a factor of 8, and without a loss in the NPP unit power generation. This paper investigates, based on an idea of closing the fuel cycle in terms of minor actinides, the relationship between the MA burnup depth in a VVER-1200 reactor and the enrichment of fresh fuel, the dynamics of unloaded heavy nuclei, and the amount of MAs in additionally loaded fuel. Under investigation are two types of additionally loaded fuel with oxides of minor actinides added to it: based on enriched uranium dioxide or a mixture of (1 – x) 232ThO2+x (96% UO2) (Kazansky et al. 2023). At the same time, the number of heavy nuclei in fuel does not change when there is a change in the number of loaded MA nuclei, which is measured in the number of the VVER-1200 reactors “served” (the mass of accumulated MA nuclei is about 1% and is removed annually from spent fuel). The number of minor actinides entering fuel is kept at a level that allows having a reactivity margin for the reactor operation at a power of over 2% before the annual refueling. Calculations were undertaken using a fuel assembly model with fuel elements broken down into a number of groups with different fuel burnup depths to simulate the actual loading of the reactor core. This model makes it possible to avoid the power peaking effects and give an answer about the neutronic characteristics of the fuel cycle involving MAs. The calculations made it possible to make a number of important conclusions: more refueling cycles lead to a dynamic equilibrium taking place between the minor actinide amounts loaded and burnt; the number of the reactors served is directly proportional to the enrichment of the make-up fuel and a 20% UO2 enrichment makes it possible to serve up to 23 VVER-1200 reactors, while using 80% 232ThO2 + 20% (96% UO2) fuel allows 17 VVER-1200 reactors to be served; using MAs extracted from SNF with a burnup of 20 to 50 MW*day/kg leads to the MA composition effect on the reactivity balance being within 4 to 5 βeff.

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DOI: 10.3897/nucet.10.144292

Authors: Andrey A. Andrianov, Olga N. Andrianova, Ilya S. Kuptsov, Anastasia A. Uvarova

Abstract: The paper analyses the neutron data and covariance matrixes, which are crucial for the computational prediction of neutronic characteristics of fast neutron reactors with uranium-plutonium fuel and currently available in the state-of-the-art versions of evaluated nuclear data libraries: ENDF/B-VIII, JENDL-5, TENDL 2021, JEFF 4T1. The newly obtained covariance data are compared to the data that was presented in the Russian evaluated nuclear data library BROND 3.1. A simplified model of a fast neutron reactor with mixed dense nitride uranium-plutonium fuel and heavy liquid metal coolant was used to calculate the spread in neutronic performance values and their uncertainties due to neutron data using different evaluated nuclear data libraries for the following characteristics: effective multiplication factor, effective delayed neutron fraction, Doppler reactivity coefficient, breeding ratio and gain, reactivity margin for fuel burnup and other fuel burnup characteristics. It has been observed that, on the whole, the uncertainties of the reactor functionals have decreased when estimated using state-of-the-art versions of evaluated neutron data libraries, compared to previous releases of those libraries. The paper also examines changes in target accuracies for predicting the main neutronic characteristics of fast neutron reactors over the past decades, as well as evaluates the requirements for neutron data uncertainties to achieve the recently declared target accuracies. The main findings of this study are presented in two aspects: first, in terms of the effects evaluated neutron data from different libraries have on the accuracy of calculations for the key neutronic characteristics of fast reactors; and second, in terms of the potential for improving the accuracy of predicted neutronic characteristics of fast reactors by considering the results of reactor physics measurements.

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DOI: 10.3897/nucet.10.142930

Authors: Anatoly N. Shmelev, Vladimir A. Apse, Vasily B. Glebov, Gennady G. Kulikov, Evgeny G. Kulikov, Anton E. Kruglikov

Abstract: The paper analyzes the possibility for large-scale production of highly pure plutonium suitable for use in radioisotope thermoelectrical generators in a VVER-SKD light-water power reactor with supercritical coolant parameters. Neptunium dioxide (NpO2) was used as the starting material. Neptunium can be extracted from minor actinides, i.e. from the main components of radioactive transuranium waste. Large-scale production of useful high-purity plutonium via neutron irradiation of useless radioactive waste can be estimated as one of the possible ways to convert harmful waste to a harmless form. The assembly with the starting material was placed in the central part of the reactor core with the maximum neutron flux to increase the plutonium accumulation. Numerical studies were based on a multi-physical approach with coupled analyses of neutronic and thermal-hydraulic processes in the downward and upward reactor zones. The results obtained in numerical studies have demonstrated that large quantities of high-purity plutonium with a low content of 236Pu and a high content of 238Pu could be accumulated in the central NpO2 assembly if the following conditions are satisfied. Firstly, it is desirable that the lattice of NpO2-containing rods would be sufficiently wide with a high volume fraction of light-water coolant. Secondly, the central NpO2 assembly needs to be surrounded by protective assemblies to form a barrier against high-energy fission neutrons which are capable to intensify the threshold 237Np(n,2n)236Pu reaction. The protective function can be performed by the assemblies containing rods with natural lead or radiogenic lead. The paper estimates the production scales of high-purity plutonium in the central assembly with a wide lattice of NpO2-containing rods and surrounded by a layer of protective Pb-containing assemblies.

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DOI: 10.3897/nucet.10.135254

Authors: Guadie Degu Belete, Yihunie Hibstie Asres, Senamaw Mequanent Zegeye, Yesigat Enawugaw Alemu

Abstract: In this study deuteron induced nuclear reaction on 20Ne target was studied to get nuclear information about 18F and 21Na radioisotopes which are widely used in medical and nuclear reactor technologies respectively. This nuclear reaction is important to get nuclear data since the produced radioisotope is widely used in nuclear medicine due to its appropriate short half-life and positron emitter. The main objective of the study was to analyse and interpret the behaviour of the resulting reaction cross section. It also aimed to compare the computed theoretical result with the experimental data retrieved from IAEA EXFOR for its validity. From such nuclear reaction, the reaction cross sections of 20Ne(d, α)18F, 20Ne(d, x)18F, and 20Ne(d, n)21Na reaction channels were computed using a nuclear computational code called COMPLET. The computed reaction cross sections for each channel were found in good agreement with the experimental data within the specified energy range and show a strong correlation as assessed by Pearson’s correlation coefficient. The analysis of the result shows that compound nucleus reaction is found the dominant reaction mechanism at the lower energies of the projectile.

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DOI: 10.3897/nucet.10.139455

Authors: Sergey L. Soloviev, Denis G. Zaryugin, Sergey G. Kalyakin, Sergey T. Leskin, Dmitry S. Soloviev

Abstract: The paper describes the new principle of passive NPP safety systems designing, in which the natural circulation of working fluid or gas is replaced by forced circulation in order to increase it’s capacity. The energy required by the system is converted from the emergency process itself, which this system resists when performing a given safety function. The goal is to expand the variety and capacity of passive NPP safety systems in order to optimize and reduce costs of NPP power units with various types of reactors while increasing the safety level. The proposed technical solutions are based on the use of direct non-mechanical conversion of the energy, comprised in emergency process, to electricity, and then to mechanical energy of hydraulic machine for working medium moving in a forced circulation mode. To demonstrate the proposed principles, the following technical solutions have been considered: • new passive electrochemical hydrogen recombiner for NPPs with light water reactors. Application of new techniques allowed increasing the performance by weight of the recommended hydrogen simultaneously with increasing the hydrogen torch threshold, as well as other important characteristics; • passive emergency core cooling system for heavy liquid metal cooled reactors. Application of new techniques allowed increasing the core remove heat capacity in order to scale down the main equipment; • passive system for in-vessel core melt retention of Pressurized Water Reactors of high thermal capacity (more than 3000 MWe). Application of new techniques allowed increasing the system thermal capacity in order to be able to apply to high thermal power reactors like VVER-1000 and more.

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DOI: 10.3897/nucet.10.139463

Authors: Yulia A. Kuzina, Alexander P. Sorokin

Abstract: The results of studies on hydrodynamics and heat transfer processes in fast neutron reactors are presented. Data on turbulent momentum transfer in rod bundles are analyzed. It is shown that the intensification of turbulent momentum transfer in the rod bundle channels is due to large-scale turbulent momentum transfer (secondary currents). The intensification of interchannel turbulent exchange in close-packed lattices of rods is explained. A dependence is obtained for the dissimilarity coefficients of interchannel convective exchange forced by wire wrapping in bundles of rods. The methods and results of numerical modeling of thermal hydraulics using the Monte Carlo method, thermomechanical analysis of the temperature field in fuel rod assemblies in the lifetime process are presented. The results of modeling based on a water model of temperature fields and the structure of coolant movement in the primary circuit of the reactor in various regimes are presented. A stable temperature stratification of the coolant was revealed in the peripheral zone of the upper chamber of the reactor above the side screens. It is shown that the process of boiling liquid metals in fuel assemblies has a complex structure, characterized by stable and pulsating regimes and a heat transfer crisis. The agreement between the results of experimental and numerical modeling is shown. A cartogram of the flow regimes of a two-phase flow of liquid metals in fuel rod assemblies has been plotted. The influence of the surface roughness of fuel elements on the boiling process and heat transfer during boiling of liquid metals is analyzed. Long-term cooling of a fuel assembly with a “sodium cavity” above the reactor core in accident regimes with boiling of liquid metals is shown. The objectives of further research are formulated.

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DOI: 10.3897/nucet.10.139467

Authors: Vladimir M. Troyanov, Andrey V. Gulevich, Olga S. Gurskaya, Viktor M. Dekusar, Vladimir A. Eliseev, Andrey L. Moseev

Abstract: The paper deals with the influence of the strategy for using separated plutonium from spent fuel of thermal and fast reactors, as well as homogeneous burning of americium in the BN reactor core on the balance of americium in the Russian nuclear fuel cycle throughout the 21st century. The assessment is carried out with the use of mathematical modeling of nuclear materials movement including nuclide transformations throughout the nuclear energy system based on the CYCLE code. Scenario modeling of the americium and Pu-241 accumulation was carried out in Russia’s two-component nuclear energy system model with thermal (VVER) and fast (BN) reactors. In so doing, spent nuclear fuel (SNF) reprocessing was simulated in 2 options: as priority reprocessing of SNF of VVER reactors (1) or SNF of BN reactors (2). In addition to americium accumulation in the system without burning, the accumulation of this actinide was studied taking into account its homogeneous burning in the MOX-fuel of fast reactors at the level of its equilibrium content of ~1%. It has been shown that the priority reprocessing of VVER spent fuel makes it possible to reduce the americium accumulation by the end of the century by ~8 tons, and the effect is achieved by using freshly separated plutonium with short-term cooling, thus, by priority the americium source is eliminated, without directly handling it. Homogeneous addition of americium to the fuel of fast reactors of the BN-1200 type at a level of ~1% makes it possible to stop accumulation of americium in a two-component energy system by 2070, stabilizing it at a level of ~40 tons in the scenario with priority reprocessing of VVER SNF and ~50 tons in the scenario with priority reprocessing of BN SNF.

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DOI: 10.3897/nucet.10.133677

Authors: Achraf Radi, Ouadie Kabach, El Mahjoub Chakir

Abstract: The VVER-1200/AES-2006 is recognized as a leading Gen III+ nuclear reactor design, meeting stringent international safety standards. This study evaluates the use of a novel PuO2-ThO2 duplex fuel, incorporating weapon-grade plutonium (WgPu) and thorium, for a hypothetical VVER-1200 assembly. The research also explores incorporating minor actinides (MAs) for transmutation, comparing two methods: MAs coated on WgPuO2 and MAs mixed with WgPuO2 as integral fuel burnable absorber rods. Neutronic properties of these fuels are compared to those of LEU-fueled assemblies. The results show a 135% higher burnup for the duplex fuel compared to LEU, with extended criticality, reduced reactivity swings, and lower Pu-239 concentrations upon discharge. While Np-237 and Am-241 concentrations decrease, Am-243, Cm-244, and Cm-245 increase, but overall radiotoxic waste is reduced. Enhanced safety coefficients are also observed, within acceptable LWR ranges.

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DOI: 10.3897/nucet.10.137527

Authors: Elizaveta P. Averchenkova, Yana V. Dyachenko, Svetlana V. Zabrodskaya, Gennady N. Manturov, Vyacheslav A. Mishin, Daria V. Panova, Anton A. Peregudov, Mikhail Yu. Semenov, Ivan V. Tormyshev, Evgeny P. Lyapin

Abstract: The subject of constant support development is currently becoming increasingly important in connection with the national strategy of transition to a closed nuclear fuel cycle. At the same time, the importance of the tasks of refining calculation models and minimizing methodological, statistical, and constant errors is increasing. In this connection, the idea of creating a universal system of constants describing with equal accuracy both uranium loading and loading with mixed oxide uranium-plutonium fuel and the possibility to perform on its basis both precision Monte Carlo calculations (data format – ACE) and multigroup calculations (formats – ABBN and ANISN) was laid in the basis of this work. In this paper we analyze the freely available libraries of evaluated nuclear data in order to justify the choice of source files for the formation of a new system of group constants for neutron-physical calculations of fast reactor cores, describe the process of formation of a new library of reactor constants, and verify the obtained system on computational test models of BN-800 reactor and critical systems. The process of formation of the new system of group constants included selection of initial neutron data files, updating of data tables of basic neutron cross-sections, self-shielding factors and Doppler coefficients, as well as data on fission spectra for the main fuel nuclides. The methodological component of the error for test models of the BN-800 core was evaluated. Using the ABBN-RF22 system of group constants it was possible to estimate the correction in the 299 groups calculation, which amounted to 0.3%. Previously, when using the ABBN-93 library, there was no such possibility due to the lack of continuity between the files of evaluated neutron data and the group constants used.

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DOI: 10.3897/nucet.10.137522

Authors: Artem Z. Gaiazov, Anton Iu. Leshchenko, Valery P. Smirnov, Elena A. Zvir, Pavel A. Ilyin, Vadim G. Teplov

Abstract: The paper presents an experimental evaluation of drying the damaged VVER SNF after the wet storage. The aim of the experiments was to justify the fire safety of air-tight canisters with the dried SNF during transportation, storage, and reprocessing. Keeping the SNF in boric acid during the experiments determined the concentration of uranium, plutonium, and specific activity of fission products in the solution inside the model canisters with the SNF. The canisters with the SNF underwent the thermal vacuum drying to determine the process parameters (temperature, residual pressure, drying rate), a hydrogen release, 85Kr activity during the drying, and an airborne radionuclide release onto the filters of the SNF thermal vacuum drying system. The results of weighing the baskets of the canisters before soaking in the boric acid and after drying at 160 °С to the residual pressure below 6 mbar showed that the difference in weight is almost the same as the calculated difference in weight of H3BO3 loaded into the canister and the weight after the complete decomposition into B2O3. Thus, we have discovered that the boric acid undergoes complete decomposition into boric anhydride, and the SNF corrosion products generate hydrogen that accumulates in the air-tight canister. We also considered the hydrogen that accumulated in the air-tight model canister with the dried SNF as a result of the radiolysis of hydrated corrosion products of the spent oxide nuclear fuel, as well as a release of gaseous fission products (GFP) inside the canister. We calculated the quantity of the hydrated corrosion products of uranium oxide after the wet storage and determined their chemical composition after the thermal-vacuum drying. The equation with an exponential decay function describes the hydrogen accumulation. The calculated maximum volume of hydrogen and uranium in the SNF corrosion products suggests that the SNF corrosion product is schoepite with the number of gyrated water molecules corresponding to the formulas from UO3·0.95H2O to UO3·1.5H2O. The received data allows for the fire safety justification of drying the wet damaged SNF and handling the dried SNF during transportation and storage.

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DOI: 10.3897/nucet.10.137455

Authors: Vladimir I. Folomeev, Svetlana M. Ganina, Natalya E. Astakhova

Abstract: There is a substantial amount of experimental data that confirm the peculiarity of the oxide fuel behavior during the first hours after a reactor reaches its power. With an increase in temperature during the said period, oxide fuel pellets crack because of a significant temperature gradient. Further developments occur due to the accumulation and redistribution of fission products, and manifest themselves as changes in the fuel matrix porosity and the formation (or diameter increase) of a central hole. The zones that differ in their microstructure, density and thermal conductivity are formed along the fuel pellet radius. As the oxide fuel composition and restructuring result in its thermal conductivity change, it is important to pay attention to the formulas used in the stress-strain state calculations of a fuel element. The methodology for calculating changes in the oxide fuel porosity and the pellet’s internal diameter is proposed and based on the published research papers dedicated to the study of oxide fuel properties and behavior during the first hours after a reactor reaches its power. Тhe methodology was tested using real experimental data on the porosity redistribution along the fuel pellet radius. The presence and the size of the pellet’s inner hole, as well as changes in the fuel matrix porosity, have a noticeable effect on the maximum temperature value. Taking into account the pellet structure evolution while performing computational simulation of the fuel rod operation under irradiation allows assessing the fuel element’s operability more accurately. The proposed methodology can be used in computer codes designed to calculate the stress-strain state of cylindrical fuel elements of fast reactors.

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DOI: 10.3897/nucet.10.129596

Authors: Nikita Popov, Alexey Suchkov, Mikhail Zharkov, Veronika Kirillova, Alexander Bazhenov, Ivan Fedotov, Irina Bajenova, Alexandra Khvan, Ilya Kozlov, Alexey Vertkov, Oleg Sevryukov

Abstract: Development of new plasma facing components (PFC) draw attention of numerous scientific groups in fusion energy field. Tungsten as a main PFC material main have poor machinability, thus various designs were proposed to overcome this problem. However, such technologies as functional graded layers, sintering and additive technologies are limited in production size and have low cost-efficiency. This research considers the alternative approach of steel melt injection or melt infiltration of tungsten mesh. The results obtained establish the mechanism of phase evolution during interaction of solid W with liquid steel and iron. Energy dispersive spectroscopy, X-ray diffraction and Thermocalc calculations used to analyze the phase composition of tungsten wetted with various steels and pure iron. Results show that interaction rate significantly depends on melt temperature and overheating above its liquidus. Then the overheating exceeds 150 °C erosion of tungsten occur.

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DOI: 10.3897/nucet.10.133710

Authors: Varvara Lagaki, Georgios Siltzovalis, Ioannis Madesis, Polytimos Vasileiou, Theo J. Mertzimekis

Abstract: Radioactivity in the marine environment, although present since the Earth’s formation, is comparatively understudied in contrast to aerial and terrestrial environments. A thorough examination of the radioactivity levels in aquatic environments can establish a robust foundation for comprehending various geochemical processes and phenomena within the water column and near the seabed, and as a result estimate the impact of radioactivity on local ecosystems. To achieve this objective, in-situ, long-term, and continuous monitoring is required. The present part I of the white paper highlights the fundamentals of marine radioactivity, describes the main objectives of an ambitious EU-funded project (RAMONES: Radioactivity Monitoring in Ocean Ecosystems) and introduces the innovative aspects of the technology developed as novel solutions to open problems.

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DOI: 10.3897/nucet.10.125376

Authors: Fadi El Banni, Bogbe L. H. Gogon, Ouadie Kabach, El Mahjoub Chakir

Abstract: This article investigates alternate fuel options for Pressurized Water Reactors (PWRs), focusing on thorium use to address safety, efficiency, and waste issues associated with standard UO2 fuel. Challenges in thorium utilization, such as the lack of a fissile isotope, are handled using approaches such as homogeneous mixtures and heterogeneous arrangements, promoting the exploration of (Th-233U-235U)O2 fuel in various assembly configurations. According to recent research, the annular dual-cooled assembly design has promising results in terms of fuel efficiency and safety while lowering the requirement for higher fissile enrichment levels. Studies additionally demonstrate that annular dual-cooled duplex fuel configurations can produce higher discharge burnup and lower power peaking factors than traditional UO2 fuel. The purpose of this work is to analyze and compare the performance of (Th-233U-235U)O2 fuel in various configurations against conventional UO2 fuel, focusing on key characteristics such as reactivity change, criticality, discharge burnups, and reactivity feedback coefficients.

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DOI: 10.3897/nucet.10.128016

Authors: Ivan M. Russkikh, Evgeniy N. Seleznev, Alexandra A. Zyryanova, Nikita A. Kotelnikov, Oleg L. Tashlykov

Abstract: When solving several research problems, the need arises for reactor testing of non-standard objects. The IVV-2M reactor is a pool, heterogeneous, light-water reactor. The 500 mm-high reactor core comprises six sections of six fuel assemblies (FA) with central water traps and a water trap in the center of the core. The personnel of INM JSC were tasked with testing an object whose length exceeded the height of the IVV-2M reactor core. A non-standard core layout was developed with six sections of seven fuel assemblies to accomplish this task.. The test object is located in the center of the core, surrounded by special aluminum blocks. Several fuel assemblies are installed on special supports, which allow the fuel assemblies to be installed 200 mm higher. Shim rods have been moved to the center of the sections. The safety of the developed arrangement was confirmed by neutron-physical calculations using the certified MCU-PTR software tool. Calculations have shown that the maximum maximum power in the fuel assembly is 236 kW, which is 50% less than the permissible value. To confirm the safety of the layout and ensure the required test conditions, tests of a mock-up of the object were carried out. The efficiency of the control members was experimentally determined. The overall efficiency of the control members was 4.85% ∆k/k, the physical weight of the object was 1.46% ∆k/k. Based on the requirements of the Nuclear Safety Rules for Research Reactors, the test duration is no more than three days. The distribution of neutron flux density over the height of the core was determined using indium neutron activation detectors. The developed design solutions made it possible to form a unique IVV-2M reactor core layout for testing objects exceeding its height.. Neutron-physical calculations and tests of the irradiated device mock-up confirmed the safety and performance of this core arrangement.

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DOI: 10.3897/nucet.10.125815

Authors: O. Ashraf, Alaa Hamdy El-Kholy, Tarek Elzayat, Ayman M. Abdalla, A. H. Ashry

Abstract: The performance of the spectral shift control (SSC) method is evaluated and compared to the conventional poison method in the VVER-1000 fuel assembly design. The SSC method can be implemented by gradually adjusting the ratio of heavy water to light water moderator (D2O/H2O) during the fuel cycle. In this study, the efficiency of using the SSC design with or without a thermal absorber (gadolinium) is investigated. We apply the SSC with both 12 burnable absorber rods containing 4.0 wt.% Gd2O3 (Case 1) and without Gd2O3 (Case 2). The neutronic calculations indicate that the discharge burnup is enhanced by 60% and the conversion ratio (CR) is increased by 64.4% at the beginning of the cycle (BOC) compared to the benchmark data. The breeding of Pu239 and Pu241 is extended to 33.7% and 29.5%, respectively, for the SSC design case (2), and better utilization of U-235 and U-238 has been achieved compared to BM.

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DOI: 10.3897/nucet.10.122847

Authors: Dmitriy G. Veretennikov, Ahmed H. Ali, Sergey V. Bedenko

Abstract: For long-term and safe operation of the reactor, nuclear fuel is modified by doping with various homogeneous compounds and heterogeneous inclusions. Such modified fuel has demonstrated satisfactory performance when irradiated at elevated temperatures and high burnup. However, the issues of radiation safety when handling modified fuel remain less studied. Elements of low and medium atomic mass are often targets for (α, n) reactions, so their use as alloying additives, as well as the use of α-emitting additives, can complicate the radiation situation at the stages of the nuclear fuel cycle. In this work, the neutron components of the radiation characteristics of UO2 with Gd2O3 and AmO2 additives were analyzed. Americium has been investigated as a possible alternative to gadolinium. In fuel containing AmO2, the neutron yield is higher compared to Am-free fuel and is formed mainly by (α, n) reactions in AmO2 in fresh fuel and spontaneous fission of 244Cm nuclides in spent fuel. The research was carried out with the aim of developing procedures and regulations for handling new fuel during its manufacture and after irradiation in the reactor. This work contributes to the study of the neutronic and radiation characteristics of Am-containing fuel, which has the potential for use in modern reactors. Calculations were performed using verified computational codes SOURCES-4C and WIMS-D5B.

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DOI: 10.3897/nucet.10.130134

Authors: Marine T. Hakobyan, Nikolay R. Avakyan, Alexander I. Ksenofontov

Abstract: This paper discusses a detailed description of the management of radioactive waste (RW) generated during operation at power unit No. 2 of the Armenian NPP during the design and additional (extended) life of the Armenian NPP power unit. The resulting RW of different types (solid, liquid, gaseous) and of different classes, from very low to high activity, which are regulated according to the rules and norms of radiation safety, are described. Options for the preparation of RW for long-term storage with solid medium-active and low-active RW at the industrial site of the Armenian NPP are considered. The principles of continuous step-by-step improvement of safety through modernization are proposed, which implies performing analyses of compliance of the power unit with the requirements of modern safety regulations, including international ones, based on modern technologies, IAEA recommendations, probabilistic safety assessments and analysis of local and international operating experience, lessons learned from accidents and incidents at nuclear power plants, elimination of deviations from existing standards and rules. It is proposed to introduce new technologies and facilities for processing intermediate storage of RW before disposal, improve and modernize existing storage facilities, and create new storage volumes that ensure the safe operation of the Armenian NPP power unit, including extending the service life and decommissioning of power units. The omissions in the management system that create difficulties in handling RW in the Republic of Armenia are indicated. The models of increasing the safety level of RW storage at nuclear power plants by introducing a unified RW management system are formulated, which will reduce the formation of RW of various types and activities, improve and expand the system of safe management of RW and SNF in the Republic of Armenia.

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DOI: 10.3897/nucet.10.130131

Authors: Vladimir A. Solntsev, Dmitry M. Soldatkin, Vladimir N. Nuzhin

Abstract: This paper describes the development of a dispersion-type uranium-zirconium fuel rod. Uranium is distributed in the zirconium matrix material in the form of axis-oriented fibers. The fuel rod is designed for the conversion of the IVG.1M research reactor (Republic of Kazakhstan) from highly enriched uranium (HEU) to low enriched uranium (LEU). The need for the HEU-LEU conversion arose in connection with Kazakhstan joining the program to convert research and test reactors to fuel with reduced enrichment (RERTR 2023). The study solves the problem of deformation of a low-tech U-Zr alloy (located in a zone of low plasticity) by replacing them with a heterogeneous compound. The manufacture of fuel rod is based on metal forming processes. Initially, a fuel rod wire with a core of fiber structure is formed by triple co-extrusion of cylindrical uranium and coaxial zirconium billets. At the next stage, the wire is processed to the required diameter by drawing, then the operation of flattening, twisting and cutting into specified lengths is carried out. Upon reaching a high total degree of deformation obtained during cold work, relaxation annealing is carried out at temperatures of 550 to 600 °C, which leads to the formation of a transboundary layer of the UZr2 intermetallic compound with a thickness of 1 to 2 μm. The intermetallic layer, without having a significant effect on the strength and thermal conductivity of the compound, ensures high quality diffusion bonding of all fuel rod components. The final operations are melting of the ends of the fuel rods and sealing by electroplating with nickel. As a result, blade-profile fuel elements are obtained with a thickness of 1.5 mm, a diameter of the circumcircle of 2.8 mm and an average effective diameter of uranium fiber of 40 μm. A set of 14040 fuel rods was manufactured and loaded into an operating IVG.1M reactor. The power start-up took place in 2023. Due to the unification and wide variability in the loading of the fuel component, in size and shape of the cross section, in structure and materials of the matrix compound, the fiber fuel element design can be used in the development of fuel rods for advanced reactors for various applications.

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DOI: 10.3897/nucet.10.130487

Authors: Alexander V. Salyaev, Svetlana G. Usynina, Valeriya P. Kalinina

Abstract: A comparative study has been conducted to find out how design approximations and simulation methods of a prismatic-type fuel block of a high-temperature gas-cooled reactor (HTGR) may affect the calculation accuracy of neutronic characteristics of fuel assemblies. To study the impact, a detailed three-dimensional computational model of a typical fuel block including fuel compacts, burnable absorber compacts, and coolant passages was developed. Changes in neutronic characteristics in the process of fuel assembly irradiation were calculated. The burnup was analyzed based on the SCALE 6.2.4 software package using a calculation module implementing the Monte Carlo method with a multigroup library of cross-sections on the basis of ENDF/B-VII.1 files of assessed nuclear data and the ORIGEN burnup analysis module included in this package. Different ways of modeling fuel compacts and burnable absorber compacts have been considered: using a built-in tool (DOUBLEHET cell type), by specifying fuel particles in the graphite matrix, and their combination. The calculations were made using the 252-group library of constants except for the option in which fuel compacts and burnable absorber compacts were simulated explicitly by particles in the graphite matrix. In the latter case, a library with a pointwise (CE) representation of cross-sections was used. A series of calculations were also made to assess the way computational statistic parameters affect the results. The results confirm correct operation of the SCALE complex built-in tool, i.e. cells with the DOUBLEHET-type double heterogeneity, and its prospective use to calculate neutronic characteristics of HTGR fuel. The calculations have also shown that it is acceptable to model burnable absorber compacts both by setting a DOUBLEHET-type cell and explicitly by particles in the graphite matrix. In general, the calculation results for these options agree quite well, within 1-2%, with the direct calculation using the pointwise library of cross-sections. Based on the computational statistic parameters, it may be recommended to set at least 200,000 histories and the number of particles in a generation or the number of generations should be at least 250.

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DOI: 10.3897/nucet.10.130485

Authors: Ulugbek T. Mirkhusanov, Dariya Yu. Semenova, Vladimir V. Kharitonov

Abstract: The paper presents a new analytical methodology for predicting the volume and cost of natural uranium production depending on uranium resources and scenarios for the development of traditional nuclear energy. The proposed methodology is based on a mathematical model of the dynamic process of exhaustion of fossil resources previously developed at MEPhI and uses the parameters of the explored mass of the fossil resource, the annual volume of production and the set rate of production growth. The three parameters known at the beginning of forecasting make it possible to describe the change in the resource base over time and solve a number of economic problems related to the extraction of valuable minerals. The paper provides forecasts of the NPP’s demand for natural uranium, depending on the scenarios for the development of global nuclear energy developed by the World Nuclear Association (WNA), uranium resources and volumes of its production with different costs for 2021–2022. It is shown that at a low rate (1.8%/year) of the installed capacity of nuclear power plants, uranium resources will last for more than a century, but the contribution of nuclear energy to electricity production will continuously decrease from the current 10%. Under high scenarios for the development of nuclear energy (>5%/year), which is possible under the conditions of intensification of the “green energy transition”, the peak of uranium production may occur before the middle of this century. Based on these data, forecasts of the dynamics of uranium exhaustion and annual production volumes at fields with different costs are presented until the second half of this century, depending on the initial production growth rates determined by the scenario of nuclear energy development.

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DOI: 10.3897/nucet.10.130483

Authors: Andrey Yu. Podchufarov, Anastasia N. Galkina, Svetlana S. Vanina

Abstract: The future of nuclear power has been widely discussed by the scientific community and the public, which is explained by the high potential of peaceful nuclear technologies in providing mankind with affordable energy, preserving the climate, developing industry, medicine, science, and education. The materials of the study are devoted to assessing the impact of nuclear power on modern challenges and achieving indicators that are reflected in the Sustainable Development Goals. The paper uses the provisions of the intersystem interaction theory, system models for managing basic target indicators, methods of correlation, regression, autoregression analysis. The study confirmed the positive impact of nuclear power on the integral indicators of sustainable development. Considering the resource constraints, the preferred scenario for nuclear power development was substantiated, and its impact on achieving the integral indicators of the Sustainable Development Goals was assessed.

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DOI: 10.3897/nucet.10.129573

Authors: Vladimir M. Troyanov, Andrey V. Gulevich, Olga S. Gurskaya, Viktor M. Dekusar, Vladimir A. Eliseev, Valeriy V. Korobeynikov, Andrey L. Moseev

Abstract: The stepwise conversion of Russian nuclear power industry to a two-component nuclear energy system (NES) with thermal and fast reactors in a single closed NFC allows a number of deferred system problems of modern power industry to be solved, such as accumulation of spent fuel from thermal reactors and spent fuel repatriated from foreign Russian-designed NPPs; inefficient use of highly limited raw uranium inventories; recycling of minor actinides and other long-lived high-level waste accumulated in the course of thermal reactors’ operation. The paper examines the fundamental capabilities of two-component NES, which appear due to the availability of commercial fast neutron sodium reactors. Due to the synergistic development of thermal and fast reactors, it is expected to achieve an economic effect associated with introduction of new fuel services in the nuclear energy system, such as production of plutonium and improvement of its isotopic composition for thermal reactors, transmutation of “external” minor actinides, production of isotopes for various purposes, production of hydrogen, etc. The expansion of Russia’s export opportunities is also being considered, it can be done through fuel supply for both power units built in compliance with Russian technologies and foreign power units with MOX fuel. In this approach, it is important to analyze the economy of the whole system, where fast reactors, due to their inherent surplus of neutrons, perform the functions required to the entire NES, rather than to analyze the technical and economic indicators of individual reactors. The article makes a preliminary technical and economic assessment of such system functions provided by fast sodium reactors. It is shown that fast and thermal reactors operating together in a two-component nuclear energy system, with the same electrical power and similar technical and economic indicators, can significantly reduce the necessary infrastructure costs and/or generate additional income, thereby reducing the specific levelized cost of electricity production.

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DOI: 10.3897/nucet.10.129000

Authors: Artur V. Lyubarskiy, Gennady V. Tokmachev, Aleksandr V. Mikhalev, Irina B. Kuzmina

Abstract: The probability of a large early release is one of the safety indicators that reflects the risk of fatalities during an accident at a nuclear power plant. Despite the fact that this indicator is very important, it is not defined in the federal rules and regulations of the Russian Federation. The article analyzes the worldwide experience related to determining criteria for large early release taking into account the statistical data of the specific country under consideration. Based on the analysis, the definition of a large early accidental release and the criteria for its achievement are formulated taking into account measures to protect the population living within the protective action planning zone (PAPZ), and an approach for their definition is proposed. A conclusion has been made about the sustainable nature of the goals identified using the NRC approach. A critical analysis of the safety indicators established worldwide allowed for the conclusion that it is necessary to improve the methodology for assessing the safety goals taking into account the specifics of NPP design, regional characteristics, phases of accident progression and other factors. The paper outlines the proposed approaches to determining high-level safety goals based on probabilistic targets – quantitative values for probabilistic health objectives (PHO) designated as PHO1 and PHO2 for the Russian Federation. A definition of a “large early release” is formulated and criteria for its achievement are proposed taking into account measures to protect the population living within the PAPZ. Thus, when assessing the degree, to which these goals are achieved in practice, it is proposed to take into account not only the likelihood of a large release during an accident at a nuclear power plant, but also measures aimed at protecting the population, in particular, evacuation. In this case, early deaths occur only when the population is not evacuated before the critical health dose loads from the release are reached.

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DOI: 10.3897/nucet.10.121449

Authors: Andrey V. Kaplienko, Boris A. Gabaraev, Yuriy S. Cherepnin

Abstract: Thermal Power Plants (ThPP), along with transport facilities, are the major sources for industrial emissions of carbon dioxide (CO2) believed to be responsible for the greenhouse effect leading to overheating of the lower atmosphere. In the opinion of many scientists, there is a threshold value of the average atmospheric temperature exceeding, which entails the potential for the development of irreversible processes threatening the existence of humankind. To avoid this danger, governments in nearly 200 countries have chosen voluntarily to reach net-zero CO2 emissions by 2050. Renewable Energy Sources (RES), including wind and solar power plants, have been selected as substitutes for ThPPs. However, energy systems based on RES only need to be multiply redundant in terms of installed capacity due to their efficiency being heavily dependent on daily, seasonal and weather factors, leave alone the scale of the required material resources (metals, polymers, concrete, glass, etc.). The major drawback of such energy systems is, however, the RES common-cause failure, e.g., in the event of a global volcanic eruption, when no energy-security requirement can be met to provide energy for satisfying the most vital needs. A need is fully evident for furnishing such energy system with another power source to be not dependent on the event that has caused the mass RES failure. With the net-zero-carbon requirement taken into account, Nuclear Power (NP) appears to be the best option in this respect. Modern NP does not however fully suits this role due to its inherent drawbacks (limited fuel resources, pending Spent Nuclear Fuel (SNF) and RadioActive Wastes (RAW) handling and nuclear-material nonproliferation issues). A potential solution to these drawbacks is a two-component NP technology in a closed nuclear-fuel cycle currently in the process of development. In Russia, where the greatest progress has been achieved in this field of development, under construction is a pilot and demonstration energy complex with the BREST-OD-300 nuclear unit expected to be started up in 2026–2027. Another promising designs to be developed are Small Modular Reactors (SMRs) / Small Nuclear Power Plants (SNPPs).

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DOI: 10.3897/nucet.10.119642

Authors: Rafael Sh. Isayev, Pavel S. Dzhumaev, Irina A. Naumenko, Maria V. Leontieva-Smirnova

Abstract: The processes of corrosion damage of the inner surface of the cladding are determined by corrosive reagents aggressive with respect to the cladding and the type of fuel used. Reactor irradiation of cladding made of EP823-Sh steel with mixed nitride fuel planned for use in the BREST-OD-300 reactor revealed non-uniform corrosion of the inner surface of the cladding. In this paper, the use of the chromium coating is proposed to prevent the corrosion of the inner surface of the steel fuel cladding. The results of corrosion tests of chromium coating applied to the inner surface of cladding made of EP823-Sh steel by electrolytic deposition are presented. Electron-microscopic studies of the chromium coating on EP823-Sh steel showed no significant signs of corrosion damage when tested in the environment of simulant fission products (CsI+Te) and in liquid lead at 650 °C.

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DOI: 10.3897/nucet.10.123398

Authors: Gennady V. Arkadov, Vladimir I. Pavelko, Mikhail T. Slepov

Abstract: Studies of maneuverable modes of VVER to confirm the possibility of participation of nuclear power plants in the mode of daily carrying capacity have been conducted for quite a long time. Tests at various nuclear power plants with VVER-1000 (Zaporizhzhia NPP in 1998, Khmelnitsky NPP in 2005, Tianwan NPP in 2007) have shown the practical possibility of NPP participation in the daily schedule of carrying capacity, however, the commissioning of nuclear power plants with VVER-1200 requires similar work on all new units with VVER-1200: NVAES-2, LNPP-2 Belarusian NPP. The article presents some aspects of the use of noise control methods for analyzing the condition of equipment and the core. Since the emergence of the technology of noise analysis of signals from VVER reactor installations, researchers have formulated several criteria for obtaining results of appropriate quality. The fundamental requirement for conducting noise experiments was the registration of data in stationary modes of operation of power units, since any non-stationarity made significant changes in spectral estimates, which ultimately complicated the work and “distorted” the results obtained. This requirement was included in the operating instructions of various diagnostic systems using noise signal analysis methods (the SUS system, manufactured by Siemens). For a long period of time, the current situation suited both developers of various diagnostic systems and NPP personnel operating them at power units. On the one hand, this was due to the imperfection of the technical means used (low speed of analog-to-digital converters, limited storage capacity, bulky equipment, etc.), on the other hand, the use of domestic NPP power units only in the base load mode without tracking daily power fluctuations in the power system. The standard archives of the upper block level system, the in-reactor control system and additionally produced multi-channel “fast” measurements with a frequency of 1 kHz for the analysis of maneuverable mode 95-55-95% of the VVER-1200 reactor plant were analyzed. Global disturbances of the core have been detected after one step of the regulatory body of the control and protection system, which attenuates within one second if the next step of the control and protection system has not occurred during this time. Such fast neutron processes can be controlled only by neutron-noise measurements with an upper frequency of at least 20 Hz.

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DOI: 10.3897/nucet.10.123054

Authors: Anton N. Yusupov, Pavel A. Mikhailov, Viktor D. Kizin, Mikhail O. Gromov, Aleksei V. Kusovnikov, Vasilii V. Avdonin

Abstract: Described shortly here is a procedure of demounting, removal, transport and long-term storage of the SM-3 core, based on the previous experience of reactor refurbishment undertaken in 1991. Prior to performing refurbishment, computations and calculated data analysis were performed to prove radiation safety of this work, which included estimation of the activity level for activation products in the structural materials of the nuclear research reactor core and the radiation conditions at different stages of its handling. As evidenced by the calculated data, the activity of the main dose-forming radionuclide 60Co attains equilibrium in about 12 years of radiation exposure. Taking into account the fact that the time period between two refurbishments was longer than 12 years, the calculated values of the equivalent dose rate were normalized to the radiation monitoring data obtained during the previous refurbishment, taking into account the calculated activity of 60Co radionuclide. The normalization made it possible to confirm reliability of estimates. The obtained activity data of activation products and taking into account the time spent during the SM-3 refurbishment in 1991, the radiation impact on personnel was estimated. Calculated values of the anticipated effective radiation exposure doses to the personnel engaged in the refurbishment revealed that the main limits of the personnel radiation exposure established in accordance with NRB-99/2009 were not exceeded. Comparison of the results of calculating the equivalent dose rate with the results of radiation monitoring at various points allowed us to establish that during the calculation and analytical justification of the radiation safety of work, the assessment of reflected radiation was significantly underestimated. But the radiation monitoring data, personal radiation monitoring, as well as recorded data of automatic radiation monitoring system show that all work was performed in compliance with the requirements of regulatory documents in the field of radiation safety.

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DOI: 10.3897/nucet.10.123052

Authors: Inga R. Makeyeva, Vasiliy Yu. Pugachev, Olga V. Shmidt, Anastasiya A. Rykunova, Andrey Yu. Shadrin

Abstract: There exist different variants of organizing the closure of nuclear fuel cycle (CNFC) depending on fast reactor type, fuel types, station or centralized allocation of closed nuclear fuel cycle stages. One of the ways to verify and estimate engineering solution is mathematical modeling of radiochemical technology which in the end will allow to optimize composite technological process in order to increase effectiveness and reduce cost. In order to calculate the balance of material flows of process circuits and individual production sections in the stationary and dynamic modes, with taking into account the isotopic composition evolution, a software package VIZART (Virtual Plant of Radiochemical Technologies) was developed, allowing the user to assemble the required sequence of operations for any part of the process scheme and perform the calculation of material balance for all flows of the circuit, as well as to optimize the equipment operating modes and provide the necessary data to justify the safety of certain limits and the entire process circuit. The following capabilities of code VIZART for computational substantiation of CNFC technology design and characteristics are considered: material balance calculation, cyclogram creation, determination of the most loaded parts of processing lines, estimation of fissile materials accumulating in devices and intermediate vessels, optimization of productivity of nodes and devices.

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DOI: 10.3897/nucet.10.122526

Authors: Aleksei V. Panov, Ludmila N. Komarova, Elena R. Lyapunova, Anzhelika A. Mel’nikova

Abstract: The paper presents an analysis of the current radiation situation in the settlements of the Russian Federation located on the most radioactively contaminated territory after the Chernobyl accident. The assessment of 137Cs content in local agricultural (milk, beef, pork, potatoes), natural (mushrooms) foodstuffs, doses of external, and internal irradiation of residents was carried out. 37 years after the Chernobyl NPP accident, 72 settlements with a total number of residents of 63869 remain in the zone of radioactive contamination whose average effective radiation doses exceed 1 mSv/year. All these settlements are located in five southwestern districts of the Bryansk region and are rural except for Novozybkov. In the private sector of settlements, potatoes and pork fully comply with radiological standards for the content of 137Cs. In milk, the content of the radionuclide can exceed the standards up to 3, in beef up to 5, in mushrooms up to 8–17 times. Over the past 30 years, the number of settlements exceeding the legally established standard for the total exposure dose to the population (1 mSv/year) has decreased by 7.5 times. Due to the decrease in the levels of 137Cs contamination of local foodstuffs and the volume of their consumption, the role of internal exposure dose has been decreasing in recent years. It was noted that in addition to a comprehensive assessment of the radiation situation in settlements, agricultural and natural ecosystems, when planning the rehabilitation of settlements with average annual radiation doses of residents above 1 mSv, it is important to take into account the current diets of residents and the real demographic situation in the territories affected by the Chernobyl accident.

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DOI: 10.3897/nucet.10.118046

Authors: Andrey A. Kashirsky, Vladimir M. Solomatin, Sergey A. Panov, Elena A. Rodina, Alexander V. Egorov, Sergey A. Dmitriev, Alexey Y. Shulga

Abstract: Within the framework of “Proryv” project a radiation-equivalent approach to radioactive waste management is being envisioned with U and Pu recycling and MA transmutation. Successful industry-wide implementation of the design approaches should be planned in order to avoid considerable financial and radiological encumbrances caused by the NFC final stage for two-component nuclear power system (NPS) under formation on the basis of thermal and fast reactors. In order to ensure a successful industry-wide implementation of the approaches being developed, the back-end of the NFC should not constitute considerable a financial and radiological burden for the emerging two-component nuclear power system (NPS). This article addresses the problems concerning justification of radiological and technical-and-economic feasibility of MA partitioning and subsequent transmutation in FNR. The extent of MA accumulation as a result of TNR SNF reprocessing confirms the need for the introduction of MA partitioning technologies not only at all reprocessing plants planned for commissioning, but also at the plants now in operation. Based on available data, the study has shown that the implementation of the closed NFC with FNR contributes to significant reduction in the cost of disposal of radwaste compared to the scenario based exclusively on the development of VVER and open fuel cycle technologies. Recycling plutonium in fast reactors should be implemented in conjunction with MA to address environmental, non-proliferation and economic concerns of the back-end of advanced NFC. Within the scale of the future nuclear power system in Russia, an option such as this can only be realized on the basis of developing a FNR fleet.

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DOI: 10.3897/nucet.10.122353

Authors: Alexander V. Avdeenkov, Oleg I. Achakovskii, Vladimir V. Ketlerov, Sergei L. Soloviev, Quang Huong Duong

Abstract: For a consistent CFD substantiation of the recombiner performance, a detailed mechanism of hydrogen and oxygen recombination is used. The detailed mechanism of chemical kinetics (multi-step recombination reaction) makes it possible to claim universality, both in the numerical justification of the recombiner performance and in the justification of the flameless recombination threshold and makes it possible to justify the method for optimizing the recombiner to improve its characteristics. The models developed based on this approach were applied to both flat and cylindrical catalytic elements, which are used in FR and RVK recombiners, respectively. As part of the numerical studies, the detailed recombination mechanism was verified, namely the temperature distribution along the catalytic elements was compared and the performance of catalytic elements was compared as well. Good agreement was obtained between the calculated and experimental data. The approach considers not only the mechanism of surface recombination of hydrogen and oxygen on platinum, but also the mechanism of recombination in the gas phase. This makes it possible to calculate the onset of intense combustion outside the catalytic plates, which is a sign of volumetric ignition of the hydrogen-air environment. The concentrations at which such ignition is possible were obtained at different contents of water vapor in the medium. Thus, the proposed approach and the created models make it possible to fully describe the performance of recombiners of distinct designs without the use of additional experimental data, which is extremely necessary when justifying the hydrogen explosion safety of nuclear power plants.

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DOI: 10.3897/nucet.10.122352

Authors: Timur R. Smetanin, Vasilii V. Pakholkov, Sergey A. Rogozhkin, Sergey F. Shepelev

Abstract: The paper presents the results of a numerical simulation for sodium fluid dynamics and mixing in the tubing system of an air-cooled heat exchanger (AHX), which is a part of the emergency cooldown system (ECS) of sodium fast reactors (SFRs). Non-uniform sodium flows in the AHX tubing system may lead to the mixing of different-temperature sodium flows, temperature fluctuations and tube breaks. It was found in the course of investigating accidents involving breaks in the PFR and Phénix reactor AHX tubing systems that the failure was caused by the metal temperature fluctuations (Cruickshank and Judd 2005). The numerical simulation used three- and one-dimensional computer codes. It has been found that the calculations of the AHX sodium flow rate distribution with a practically acceptable accuracy can be performed using a one-dimensional code. The factors that influence the non-uniform distribution of sodium flows in the AHX tubing system have been analyzed. Calculations have been performed for the AHX sodium flow distributions and for the mixing of different-temperature sodium flows in the AHX outlet header. The results are presented from calculating the amplitude of sodium fluctuations near the AHX header walls. The effect from shutting down several modules on the non-uniform flow distribution and temperature fluctuations in the AHX has been investigated. Approximations of numerical solutions have been obtained for the sodium flow distribution as a function of the number of the modules shut down.

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DOI: 10.3897/nucet.10.122292

Authors: Sergey L. Soloviev, Andrey V. Shishov, Vladimir P. Povarov, Sergey V. Yaurov

Abstract: The paper presents information on the key approaches to the design of the containment ventilation system for units 1 and 2 of Novovoronezh NPP II (NPP-2006 project). The authors have developed a CFD model for the containment of Novovoronezh NPP II’s units 1 and 2, which includes the key structural components and the basic equipment installed within the containment. A series of the containment air temperature measurements was undertaken during power operation of the units. Based on the measured temperature values, a series of calculations was undertaken to determine the air temperature field inside the containment. It is revealed that when ensuring the design characteristics of the cooling capacity of the ventilation system stages, the design parameters of the containment air, wall and equipment temperature are achieved. In addition, with proper mixing of the containment air, it is possible to significantly reduce the average air temperature in the most “hot” rooms. Based on the calculation results, causes have been identified for the low efficiency of the ventilation system, and specific measures have been proposed for increasing significantly the system capacity. The proposed approach to determining the characteristics of ventilation systems using modern methods of three-dimensional computational hydro-gas dynamics makes it possible to optimize and modernize existing ventilation systems, as well as to assess the efficiency of ventilation at the design stage of nuclear power plants. The developed and proposed CFD model makes it possible to do this at the modern level without resorting to bench/experimental modeling issues.

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DOI: 10.3897/nucet.10.122284

Authors: Oleg Yu. Kochnov, Valery I. Stepanov, Denis A. Pakholik, Valery V. Kolesov, Evgeny V. Nikulin

Abstract: The key industrial method for producing 99Mo is production of the radionuclide as one of the 235U fission fragments. 235U is irradiated with neutrons in a nuclear reactor (both heterogeneous and homogeneous nuclear reactors can be used) and then processed in radiochemical laboratories, where 99Mo is chemically extracted from fission products. Both highly enriched uranium (HEU) and low enriched uranium (LEU) can be used to produce 99Mo by the fragmentation method. To date, almost all world producers, with the exception of Russia, are either in the final stages of transferring production from highly enriched uranium to low enriched uranium, or are already producing 99Mo using LEU. This is due to the problems of non-proliferation of nuclear materials and the prevention of the likelihood of terrorist threats. A number of experimental studies have been carried out on the basis of the VVR-ts research reactor. Experimental studies included the study of the effect of LEU targets on the reactivity reserve of the VVR-ts reactor, irradiation of these targets in experimental channels and separation of 99Mo from them. The paper presents the results of producing and separating 99Mo from targets with LEU material. It is shown that it is necessary to improve the processing technology to increase the production of fragmented 99Mo from LEU.

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DOI: 10.3897/nucet.10.114088

Authors: Yaping Guo, Peng Nie, Ruizhi Li, Lijun Zhang, Xingwang Zhang, Ren Ren, Zelong Zhao

Abstract: The calculation and visualization of the dose field in the decommissioning of nuclear facilities is one of the important functions of the decommissioning virtual simulation system. The dose field simulation tools can provide radiation field distribution and play an important role in determining the decommissioning plan and protecting personnel during the engineering implementation process. This article investigates the development of dose field calculation and visualization in the reactor decommissioning virtual simulation systems. A preliminary technology plan suitable for the development of the decommissioning dose field calculation and visualization display programs of the first Heavy Water Research Reactor (HWRR) in China is proposed. The applicability of the selected scheme is analyzed. The functional requirement and development direction of the HWRR reactor decommissioning dose field tool are preliminarily determined. Furthermore, the reactor vessel of HWRR reactor is modeled, the dose field distribution is calculated and visualized based on the preliminary decommissioning code. This research can provide technical support for the development of the decommissioning simulation system for the first HWRR reactor in China.

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DOI: 10.3897/nucet.9.117199

Authors: Anatoly N. Shmelev, Nikolay I. Geraskin, Vladimir A. Apse, Gennady G. Kulikov, Evgeny G. Kulikov, Vasily B. Glebov

Abstract: The paper presents the estimates for the possibility for large-scale production of 238Pu in the core of a VVER-1000 power reactor. The Np-fraction of minor actinides extracted from transuranic radioactive waste is proposed to be used as the starting material. The irradiation device with NpO2 fuel elements is installed at the reactor core center. The NpO2 fuel lattice pitch is varied and the irradiation device is surrounded by a heavy moderator layer to create the best possible spectral conditions for large-scale production (~ 3 kg/year) of conditioned plutonium with the required isotopic composition (not less than 85% of 238Pu and not more than 2 ppm of 236Pu). Plutonium with such isotopic composition can be used as the thermal source in thermoelectric radioisotope generators and in cardiac pacemakers. It has been demonstrated that the estimated scale of the 238Pu production in a VVER-type power reactor exceeds considerably the existing scale of its production in research reactors.

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DOI: 10.3897/nucet.9.117198

Authors: Andrey A. Andrianov, Olga N. Andrianova, Yury A. Korovin, Iliya S. Kuptsov, Anastasiya A. Spiridonova

Abstract: The paper describes in brief the functional capabilities of a computer code for optimizing the neutronics model parameters (neutron data, technological parameters, and their covariance matrices) based on results of reactor physics experiments using conditional nonlinear multi-parameter optimization algorithms. The code’s application scope includes adjustment of neutron constants, technological parameters and their covariance matrices based on integral measurement results, formulation of requiremen117198ts with respect to the neutron data uncertainties for achieving the target accuracies in calculation of the reactor functionals, and estimation of the reactor performance prediction accuracy, as well as the informativity and similarity metrics of reactor physics experiments with respect to each other and in relation to the target reactor system. The paper also considers some examples of using the code to refine the neutronics models of nuclear reactor and fuel cycle systems based on results of reactor physics experiments.

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DOI: 10.3897/nucet.9.116662

Authors: Ruben A. Shaginyan, Valery V. Korobeinikov, Viktor Yu. Stogov

Abstract: At present, the existing approaches to production of artificial isotopes are mostly based on the development experience from previous years. This work aims to develop an algorithm for selecting the most effective irradiation modes for target materials. The study is based on sequential modeling of irradiation of target isotopes by neutrons of different ‘single-group’ fluxes at the same neutron flux density within each energy group (BNAB-93). In this study, a flux density equal to 2×1015 n/(cm2×s) was used for each energy group. This approach will help ‘designing’ and selecting the actual neutron spectrum that has the highest efficiency compared to alternatives. The study modelled Co-60 and Lu-177g production for each energy group. The kinetics was analyzed in the most efficient groups in terms of specific activity. The maximum specific activity for Co-60 is reached in group 17 and is equal to 1 kCi/g. For the scheme of Lu-177g production through Lu-176 the maximum specific activity is reached in group 26 and is equal to 58.5 kCi/g. For the scheme of Lu-177g production through Yb-176, the maximum specific activity is reached in group 17 and is equal to 260 Ci/g, advantageous for production are groups 15–17 and 26.

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DOI: 10.3897/nucet.9.116661

Authors: Nikolay M. Barbin, Stanislav A. Titov, Dmitry I. Terentiev, Anton M. Kobelev

Abstract: The paper presents the results from a thermodynamic analysis of the behavior of Sr and Ca radionuclides in the process of heating radioactive graphite in an air atmosphere. The TERRA software package was used for the thermodynamic analysis in a temperature range of 300 to 3600 K to determine the possible composition of the ionized, gaseous and condensed phases. It has been found that strontium is in the form of condensed SrCl2(c) and gaseous SrCl2 in a temperature range of 300 to 1600 K, and in the form of gaseous SrCl2, SrO, SrCl and Sr and ionized SrCl+, Sr+ and SrO+ when the temperature is increased from 1600 to 3600 K. Calcium is in the form of condensed CaCl2(c), CaUO4(c), CaO(c) and gaseous CaCl2 in the temperature interval between 300 and 2100 K, and in the form of gaseous Ca, CaCl and CaO and ionized Ca+, CaO+ and CaCl+ when the temperature is increased from 2100 to 3600 K. The paper determines the key reactions within individual phases and among condensed, gaseous and ionized phases. The equilibrium constants of their reactions have been calculated. Based on the results obtained, dependence plots are presented for the Sr and Ca radionuclide distribution by phases.

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DOI: 10.3897/nucet.9.116659

Authors: Artur V. Lyubarskiy, Gennady V. Tokmachev, Irina B. Kuzmina

Abstract: The paper considers the implementation of the “practical elimination” principle in the design of the Akkuyu NPP with VVER1200 reactors being under construction in Turkey. The “practical elimination” principle is defined as follows: for accident sequences or phenomena that contribute or lead to unacceptable radiological consequences for the public or the environment, it shall be shown with a high level of confidence that their occurrence is highly unlikely. “Practical elimination” is proved predominantly by results of a Level 2 probabilistic safety assessment. The implementation of the “practical elimination” principle was considered at the level of accident sequences leading to a large radioactive release. It has been shown that each accident sequence leading to an unacceptable release has the probability of occurrence below 4.45Е–8 per reactor per year, while their total probability not exceeding the value of 6.17Е–7 per reactor per year. For the phenomena inside the containment area during severe accidents, including hydrogen detonation, a large thermal explosion, direct containment heating, overpressure in the containment volume, and the containment damage at later stages due to the basement melting through, their “practical elimination” has been demonstrated. The paper also considers specific hard-to-assess scenarios of beyond design basis accidents, for which the applicability of the “practical elimination” principle is assessed as well: a major positive reactivity insertion, a rupture of the reactor pressure vessel and other largescale components, damage of fuel elements in the spent fuel pool, severe accidents with the containment bypass or containment failure, and severe accidents accompanied the means for mitigating with their consequences being unavailable. Criteria have been developed and used for the “practical elimination” assessment. A sensitivity analysis was undertaken as part of the Level 2 probabilistic safety assessment which has shown that estimated values slightly depend on the analytical assumptions, as well as on the random change in the parameters that the affect the progression of severe accidents.

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