A novel ground motion simulation framework for long-period multi-damping spectral and PSD matching in nuclear engineering

The development of advanced integrated small reactors is a significant trend in modern nuclear engineering. Designed by the Institute of Nuclear and New Energy Technology at Tsinghua University, NHR200-II is a nuclear heating reactor that incorporates a range of advanced passive and inherent safety features. Different from general pressurized water reactors, the safety injection system is excluded from the NHR200-II design to enhance system safety and simplicity. Therefore, the potential occurrence of a loss-of-coolant accident (LOCA) will be a key concern in the NHR200-II reactor design. The main objective of this study is to develop a Reactor Excursion and Leak Analysis Program 5 (RELAP5) model suitable for small-break loss-of-coolant accident (SB-LOCA) analysis and to explore the dynamic characteristics of system thermal-hydraulic parameters during the accident, as well as to assess the safety of NHR200-II. In this paper, a detailed thermal-hydraulic analysis of two typical SB-LOCA scenarios is presented: (1) pipe rupture of the hydraulic control rod driving system with isolation failure and (2) rupture of a small-diameter pipe on the head of the reactor pressure vessel (RPV). Several key safety characteristics, including total coolant loss from the RPV, liquid level remaining in the RPV, and maximum temperature of fuel rods, are examined to assess the safety performance of NHR200-II. The results indicate that in both accident scenarios, the remaining coolant in the primary loop is adequate to cover the reactor core and effectively remove decay heat via the passive residual heat removal system, which prevents the reactor core from overheating damage and ensures the reactor remains in a safely controllable state. Thus, the safety characteristics of the NHR200-II reactor in LOCAs are demonstrated.

Lagrangian particle-tracking models are increasingly used to simulate radionuclide transport in marine environments, especially for assessing the consequences of accidental releases. However, existing models generally neglect radioactive decay chains, limiting their ability to reproduce the complete behavior of radionuclides and their progeny. To the authors’ knowledge, this work presents the first implementation of radioactive decay chains within a fully three-dimensional Lagrangian marine radionuclide transport model, explicitly coupling stochastic particle tracking with decay kinetics and dynamic sediment–water interactions, enabling a realistic simulation of parent–daughter transformations in the ocean. The approach is tested for the chain in the Western Mediterranean Sea, following a hypothetical nuclear accident. Results confirm that the stochastic treatment accurately reproduces analytical decay solutions and can be seamlessly incorporated into operational-scale transport simulations. The framework can be extended to other radionuclide series and marine domains, providing a versatile and computationally efficient tool for emergency response, environmental impact assessment, and safety analysis in nuclear engineering applications.

Nuclear engineering uncertainty quantification for modeling and simulation has historically focused on first-order perturbation theory–based approaches to rapidly generate sensitivity and similarity coefficients for the effective multiplication factor to assess the accuracy of nuclear data, guide criticality experiment design, and propagate uncertainty effects. This work extends these efforts from the steady-state to the transient regime with the Coupled Adjoint Perturbation Theory for dynAmIcs and heat traNsfer (CAPTAIN) framework. Building on previous verification work of the CAPTAIN methodology using single physics, this work couples the reactor dynamics and heat conduction to verify the accuracy of the CAPTAIN framework on a tightly coupled transient with temperature feedback for the reactivity insertion, effective delayed neutron fraction, temperature reactivity coefficient, decay constant, change in temperature, density, specific heat capacity, and power density. Tightly coupled power, delayed neutron precursor, and temperature sensitivity coefficients are generated for all of these inputs with respect to time. The sensitivity and similarity coefficients computed using CAPTAIN can be leveraged to rapidly quantify uncertainty for coupled multiphysics nuclear transients, guide experiment design to target largest uncertainties, calibrate uncertain nuclear data and physics parameters, and increase confidence in nuclear data for performing transient simulations.

Abstract The experiments of W. Bothe and P. Jensen from the year 1940 to determine of the thermal neutron absorption cross sections of graphite and carbon were recalculated with current neutron transport codes and latest nuclear data in order to find out the reasons for the negative results, which excluded the use of both graphite and pure carbon as moderators in a natural uranium reactor. One result of the recalculation is that the diffusion theory underlying the measurement of the graphite cross section was appropriate to the problem, that the boron content of the graphite investigated was at 6 appm, and that the thermal absorption cross section of 8 mb (or 7.58 mb in a re-evaluation of the experiment) as given by W. Bothe and P. Jensen was correct. On the other hand, the measured absorption cross section of the graphite impurities of 1.61 mb was too low by a factor of 2.64, resulting in a too high carbon absorption cross section of 6.4 mb. According to the current calculation, the absorption cross section of the carbon should only have been 3.38 mb. The recalculation of the ash disk experiment, on the basis of which W. Bothe and P. Jensen had determined the neutron absorption of the impurities, shows that there are four reasons for their underestimation, two of which can be quantified. The measurements of W. Bothe and P. Jensen take center stage of a more comprehensive investigation of the graphite work of the German Uranium Club (Uranverein) in the years 1939–1942. This shows, among others, that the negative result of the Bothe-Jensen measurement was not of decisive importance in the decision to reject graphite as a moderator in a nuclear reactor, but rather that it was erroneously assumed that it was practically impossible to remove the boron impurity from the graphite. The successful approach of L. Szilard and E. Fermi to solve the graphite problem during the early phase of the Manhattan Project is also escribed. Thereby, it reveals, that the measuring method attributed to W. Heisenberg, which was used in Germany by W. Bothe and P. Jensen to determine the thermal neutron absorption cross section of graphite, most probably originates from L. Szilard. Overall, it turns out that the members of the German Uranium Club, compared to the American scientists, made only minor or no efforts in order to develop procedures for the accurate characterization, the lossless ashing, and the sufficient purification of graphite, which are necessary for a successful application of this material in nuclear engineering.

With the ever increasing beam power at particle accelerator-based facilities for nuclear and particle physics, radioactive isotope production, and nuclear engineering, it becomes increasingly important to have targets that can withstand this power, and shielding to block the secondary particles produced. Here we present Monte Carlo (MC) calculations using the well-established Geant4 software to predict the antineutrino yield of a 8Li Decay-At-Rest (DAR) source. The source relies on 600 kW of beam power from a continuous wave proton beam impinging on a beryllium target, where spallation neutrons are captured by 7Li to produce the 8Li. We further present an in-depth treatment of the neutron shielding surrounding this target. We show that we can produce the high antineutrino flux needed for the discovery-level experiment IsoDAR, searching for “sterile” neutrinos (predicted new fundamental particles) and other beyond standard model physics, while maintaining a neutron flux in the detector that is below natural backgrounds. The methods presented in this paper are easily transferable to other high-power targets and their associated shielding.

Electrical energy plays a very important role in human life, as almost all aspects of daily activities depend on it. Therefore, the provision of sufficient and sustainable electrical energy is a key factor in improving the quality of human life. Most of the electricity supply in Indonesia still relies on fossil fuels, which are one of the causes of the increase in the Earth’s atmospheric temperature; therefore, the use of carbon-based energy must be reduced. Energy from sunlight can be utilized by developing a Solar Power Plant (PLTS) to reduce the consumption of carbon-based energy. The building selected for the installation of the Solar Power Plant is the Department of Nuclear Engineering and Engineering Physics (DTNTF) building. The DTNTF building is one of the facilities located within the Faculty of Engineering, Universitas Gadjah Mada. On average, almost all buildings at Universitas Gadjah Mada (UGM) are equipped with adequate air conditioning (AC) systems. More than 50% of the electrical energy consumption in UGM buildings is used for cooling lecture rooms and office spaces. Similarly, the DTNTF building is equipped with more than 50 air conditioning units, where each room contains one to two AC units depending on the room size. The remaining electricity consumption comes from office equipment and lighting systems used to illuminate each room in the DTNTF building. Based on these data, the average monthly electricity demand of the DTNTF building ranges from approximately 1,000 kWh to 1,500 kWh. To meet this electricity demand, approximately 30 units of 350 Wp solar panels are required to be installed on the roof of the DTNTF building. Considering the average daily solar irradiation duration in Indonesia of about 5 hours per day, under ideal conditions a 350 Wp solar panel receiving 5 hours of sunlight per day would produce 350 W × 5 hours = 1,750 Wh (1.75 kWh). Thus, with 30 solar panels, the solar power plant would generate 52.5 kWh per day.

A high-temperature piezoelectric accelerometer, employing a pair of radially polarized BiScO3-PbTiO3 (BS-PT) ceramic rings, operating in symmetric shear-bending coupling vibration mode is proposed. This coupling mode enhances effective electromechanical coupling through a charge amplification factor, yielding a high sensitivity (22.8 pC/g), approximately three times that of conventional shear-type accelerometers at room temperature. The device exhibits stable performance, maintaining a sensitivity of 14 pC/g after repeated thermal cycling between room temperature and 260 °C and during forced vibration tests, while exhibiting a linear response across a wide frequency range. Operation at even higher temperatures is feasible by employing piezoelectric materials with higher Curie temperatures. The demonstrated combination of high sensitivity, thermal reliability, and structural robustness underscores the potential of this accelerometer for applications in aerospace, energy, and nuclear engineering.

The thermal conductivity of epoxy resin decreases sharply to approximately 0.01 W·m−1·K−1 at 4 K, significantly limiting its application in nuclear fusion engineering. To address this limitation, this study prepared epoxy composite reinforced with silver, diamond, or wollastonite at 50 vol % and 70 vol %, and systematically investigated their 4-K thermal conductivity. The results demonstrated that at 70 vol %, the silver epoxy composite achieved a thermal conductivity of 5 W·m−1·K−1, which is more than two orders of magnitude higher than that of the diamond-epoxy composite (0.05 W·m−1·K−1). At 50 vol %, the silver epoxy composite (0.12 W·m−1 ·K−1) exhibited similar thermal conductivity to the wollastonite epoxy composite (0.1 W·m−1·K−1). Increasing the silver filler from 50 to 70 vol % enhanced the thermal conductivity from 0.12 to 5 W·m−1·K−1, representing a 42-fold improvement. Through combined theoretical analysis, microstructural characterization, and finite element analysis, the following conclusions were drawn. The electron conduction–dominated fillers were the preferred choice for the cryogenic thermal conductivity epoxy composite. In the discrete composite, even the electron conduction–dominated filler could not significantly improve the cryogenic thermal conductivity of the epoxy composite; even 50 vol % silver only reached approximately 0.12 W·m−1 ·K−1. The synergistic effect between the filler type and the distribution (electron conduction–dominated filler combined with continuous filler distribution) can substantially enhance the cryogenic thermal conductivity of the epoxy composite. This work provides technical support for the design of high-performance thermal management materials for superconducting magnet cooling systems in fusion reactors.

Abstract Accurate fission cross section data are essential for nuclear science and nuclear engineering. The traditional fission cross section measurement using the fission ionization chamber can hardly meet the accuracy requirements. The time projection chamber (TPC) is supposed to be one of the potential detectors for high accuracy fission cross section measurement based on its track reconstruction and particle identification capacities. Using the Multi-purpose Time Projection Chamber (MTPC) at the China Spallation Neutron Source (CSNS), we have already measured the cross sections of 232Th(n, f) reaction based on the mono-energetic neutron source at Peking University (PKU), showing the potential of high accuracy fission cross section measurement. In the present work, cross sections of the 235U(n, f) reaction were measured both at 43 energies (10 energy bins per magnitude in equal logarithm intervals) and at 215 energies (50 energy bins per magnitude in equal logarithm intervals) in the neutron energy range from 0.5 eV to 10 keV using the MTPC based on the CSNS Back-n white neutron source. The results are consistent with the data in the evaluation libraries, showing the reliability of the fission cross section measurement method using the MTPC. The measurement of the fission cross sections using the MTPC has been expended from the mono-energetic neutron source to the white neutron source. Our results are the first cross section results measured by the MTPC based on the CSNS Back-n white neutron source. With longer beam time in the future measurement, the uncertainties of the fission cross sections are expected to be greatly reduced.

Corrigendum to "Verification of PWR-Core power distribution based on precisely calculated SPND response currents" <[ Nuclear Engineering and Design 448 (2026) 114726]>

Innovative projects and technologies of nuclear power engineering. Review of Proceedings of the VI International Scientific and Technical Conference (ISTC NIKIET-2023) (14–17 November 2023, JSC «NIKIET», Moscow, Russia)

ABSTRACT The Riga plate produces a Lorentz force to control boundary layers (BL) and improve cooling purposes for effective electromagnetic flow control in nuclear and aeronautical engineering systems. Furthermore, the synergistic interactions of different nanoparticles optimize heat transfer. A Riga surface is a specialized electromagnetically active device that controls the flow regime by modifying the Hartmann number. Riga surfaces have been used in industrial setups for chemical engineering, biomedicine procedures, and environmental engineering. In addition, the boundary layer controlled by the electromagnetic field provided by the rigid surface is used extensively in the production process of extruding polymers, spinning of metals, and fiberglass production. Advancement in thermal capability of the Riga surface by inducing ternary nanoparticles is necessary to fulfill the demand. So, the primary focus is to investigate the upsurge in thermal capability of the Riga surface by adding ternary nanoparticles, considering activation energy and convective heating. Water‐based tri‐HNF is used because of its improved thermal characteristics, which are anticipated using the Gharesim model viscosity and Hamilton‐Crosser thermal conductivity models. The assumptions of the normal heat and mass fluxes are considered for practical purposes. Computational simulations are executed to handle the developed non‐linear mathematical model by using the Runge–Kutta procedure in combination with the shooting approach. The wall friction factor, heat, and mass fluxes are optimized by applying the statistical Response Surface Methodology (RSM) technique. The ideal conditions to improve heat and mass transmission are found using sensitivity analysis, and they are subsequently validated by analysis of variance testing. Sensitivity analysis showed that in a thinner boundary layer, the skin friction increases with an augmentation in nanoparticle concentration. A higher Nusselt number indicates improved heat transfer with increased nanoparticle load. The activation energy uplifts the mass transfer rates, but decreases with nanoparticle concentration.

Radionuclides 137Cs, 238, 239+240Pu, 40K and 210Po in water areas on the river-sea border and assessment of their action levels to hydrobionts.

The paper is devoted to the description of the method of joint analysis of deformation and fracture of structural elements that are in conditions of high-temperature creep, the accumulation of hidden damage associated with it, and in which some surfaces are exposed to aggressive environments. The boundary value problem is solved using the finite element method, for the initial one the finite difference method of integration over time was used. For the analysis of corrosion cracking of the inner surface of the tube, an approach was used that consists in excluding from the calculation model the “destroyed” finite elements, i.e. those in which the damage parameter has reached its critical value, and reformulating the boundary value problem for the model with a new geometry and preserved initial conditions for the components of the stress-strain state and the damage parameter in the remained finite elements. To simulate the accumulation of hidden damage due to creep and corrosion, an approach was used that takes into account the contributions of the increments of the corresponding processes at each time step. Corrosion damage is modeled by an auto-model evolution equation taking into account a specially defined equivalent stress on the surface of the body. As an example of using the numerical modeling method, the creep process with material damage accumulation and subsequent corrosion cracking in a thick tube is considered. The process of damage growth in its material is analyzed, the ratio between the rates of damage accumulation due to creep and corrosion cracking on the surface is estimated. The developed approach and finite element calculation method are proposed to be further applied to the analysis of deformation and corrosion cracking of structural elements of complex geometry used in the power and nuclear engineering.

The third millennium is associated with many achievements in science and technology, one of which is nanomaterials, i.e. discrete particles of material, as well as materials with an internal or surface structure, one of the characteristic dimensions of which usually lies in the range from 1 nm to 100 nm. Due to their unique properties, primarily thermophysical and mechanical, nanomaterials are used in heat transfer processes, which are common in thermal power engineering, nuclear power engineering, chemical and food technology, metallurgy, electronics, mechanical, and instrument engineering. Nanomaterials increase the efficiency of thermal conductivity and convection and are used in all heat transfer processes, namely heating, cooling, boiling, and condensation. Almost all classes and types of nanomaterials are used, including nano-objects such as nanoparticles, nanofibers, and nanoplates, as well as nanostructured materials such as nanostructured powders, nanocomposites, nanoporous materials, and fluid nanodisperse systems. Nanomaterials are most widely used in coolants in the form of nanosuspensions and nanoemulsions, as well as in the design of heat exchange equipment in the form of coatings for heat exchange elements and structural materials for the manufacture of these elements. Currently, the main trends in the application of nanomaterials in heat exchange processes and equipment are the development of effective compositions of fluid nanodispersions and nanocoatings of heat exchange surfaces, which can be implemented on existing heat exchange equipment directly or with minor modernization. Less attention is paid to the development of structural nanomaterials for the manufacture of heat exchange elements, since they involve a more profound change in existing heat exchange equipment or the creation of fundamentally new heat exchanger designs. In any case, one should not forget about the possible negative impact of nanomaterials when handling them, which they can have on the environment and humans, and, if possible, take measures to eliminate or minimize this negative impact. Bibl. 103, Fig. 9.

The Riga plate is arrangement of electrodes and permanent magnets allows for efficient regulation of fluid flow. The Riga surface leverages Lorentz forces to control boundary layers (BL) and improve cooling purposes for effective electromagnetic flow control in nuclear and aeronautical engineering systems. Furthermore, by utilizing synergistic interactions of different nanoparticles, heat transfer rat can be optimized in industrial setup. The primary focus of this work is to investigate the unsteady BL flow of water-based tri-hybrid nanolfuid (tri-HNF) flow over a Riga plate senor under the influence of activation energy, cross-diffusion, and convective heating. Three different nanoparticles A{l}_{2}{O}_{3}, CuO and Ti{O}_{2} are dispersed in a pure water. The model equations are constructed using BL theory and transformed into ordinary differential equations using an appropriate similarity rule. The Runge–Kutta fourth-order (RK-4) method, along with shooting approach, is used to address the problem numerically. The skin friction and Nusselt and Sherwood numbers are assessed using optimized statistical Response Surface Methodology (RSM) technique. The Gharesim model viscosity and Hamilton-Crosser thermal conductivity models are deployed in the governing model. A mathematical model is designed and developed using RSM to obtain an optimal skin friction, heat and mass transfer rate. Sensitivity analysis (SA) is performed to investigate the response of input on these coefficients. SA shows that in narrow BL, the skin friction rises with nanoparticle concentration. Velocity of tri-HNF boost with the Hartman number and the electrode-magnet distance parameter. The Soret number, and activation energy increases the concentration profile. Higher Nusselt number indicates improved heat transfer with increased nanoparticle load. Activation energy uplift the mass transfer rates, but dwindle with nanoparticle concentration.

In the context of the development of nuclear power engineering, there is an increasing demand for next-generation building materials capable of simultaneously providing thermal and radiation protection while maintaining reduced weight and improved performance characteristics. This study is devoted to the development of a synthesis methodology for heat-resistant foam concrete that combines radiation-shielding and thermophysical properties, intended for use in enclosure and technological structures of nuclear power facilities. It is demonstrated that the combination of functionalized foam with heavy fillers and water-containing components can ensure attenuation of γ- and neutron radiation while simultaneously reducing the thermal conductivity and density of the material.

Fluid-conveying pipes represent a fundamental dynamic problem within the realm of fluid-structure interaction. They find extensive applications in various industries, including petroleum, nuclear engineering, aviation, aerospace, and nanostructures. This paper applies the Green’s function method to solve the stability problem of a fluid-conveying pipe, hinged at both ends and supported by intermediate linear-elastic supports. The objective is to examine the influence of the number and rigidity of these supports on the critical fluid velocity, which is the velocity at which the pipe loses stability. A numerical solution was performed for a straight pipe conveying fluid with specified geometric and physical characteristics, where the number and rigidity of the elastic supports were considered as parameters. The numerical analysis presented herein includes graphs illustrating the dependence of the critical fluid velocity on the number of elastic supports for varying support rigidities. These results reveal that the elastic supports affect both the vibrational characteristics and the critical velocity of the conveyed fluid. The solution results are compared with those obtained using one of the most widely employed methods for analyzing the dynamic stability of pipe systems (Transfer Matrix Method – TMM). A good agreement between the results is observed. The paper aims at presenting a method for obtaining the exact solution to the differential equation governing the lateral displacements of a pipe system. This paper discusses the authors' perceived pros and cons of the Green's function method in comparison to the most popular methods for the dynamic investigation of fluid-conveying pipes.

Zirconium and its alloys are widely used in nuclear power engineering due to their favorable physical and mechanical properties and their low thermal-neutron absorption cross-section. Their high corrosion resistance in aqueous and steam environments at elevated temperatures is essential for the reliable operation of fuel assemblies and is associated with the formation of a stable, compact ZrO2 oxide layer. However, under reactor conditions, the presence of hydrogen, iodine and other fission products can reduce corrosion resistance, making detailed corrosion assessment necessary. Manufacturing technology, alongside alloy composition, also plays a decisive role in determining corrosion behavior. This study presents corrosion test results for a Zr-1%Nb alloy processed under thermomechanical conditions corresponding to rolling in a special type of three-roll skew rolling–Radial-Shear Rolling (RSR). The applied rolling technology ensured the formation of a pronounced ultrafine-grained (UFG) structure in the near-surface layers, with an average grain size below 0.6 µm. EBSD and TEM observations revealed a largely equiaxed microstructure with refined grains and increased grain boundary density. The corrosion testing was performed in high-temperature steam vessels at 400 °C and 10.3 MPa for 72, 336, 720 and 1440 h. The results demonstrate that RSR processing is an efficient alternative to conventional multi-pass normal bar rolling with vacuum heat treatments, allowing a significant reduction in processing steps and eliminating the need for expensive tooling and intermediate thermal or chemical treatments. Bars manufactured using this method meet the ASTM B351 requirements. The specific weight gain did not exceed 22 mg/dm2 after 72 h and 34.5 mg/dm2 after 336 h. After 1440 h, the samples exhibited a continuous, uniform dark-grey oxide layer with an average thickness below 5.3 µm.