Next Generation Nuclear Research Articles

CFD modelling and simulation of anaerobic digestion reactors for energy generation from organic wastes: A comprehensive review.

Anaerobic digestion (AD) has been recognized as one of the most viable options for the treatment of a wide range of waste materials. Complex structure of wastes is safely broken down to destroy pollutants and pathogens. Biogas is produced as a by-product of this process which is considered as a clean energy resource. However, provision of controlled environment for microbial activities is critical to ensure the required process efficiency. This can only be achieved with the efficient design of controlled vessels used for anaerobic digestion, termed as AD reactors. AD functions such as mixing, hydrodynamics, multiphase interaction, heat transfer, temperature distribution and bio kinetics are significantly affected by the reactor shape, design and configurations, hence making it essential to optimize the reactor design before installation at large scale. Mostly, such optimization is carried out with the help of lab scale experimentations and testing protocols which result in high costs for repeating several design experiments. Computational fluid dynamics (CFD) is an applied mathematical tool which helps to understand and predict the fluid dynamics, heat flow as well as species transport in different domains. This approach contributes to minimize the experimental costs while optimizing the reactor configurations in less time. The current review is presented to summarize and discuss the core characteristics of AD process followed by concerned CFD attributes. Research gaps and critical challenges are identified in different aspects such as reactor design, and configuration, mixing, multiphase flow, heat transfer, biokinetics as well as machine learning approaches.

Influence of Bed Medium on Bubbling Fluidized Bed Reactors for Biomass Gasification: Experimental Study

Abstract This research provides a contribution to biomass gasification technology development and to its establishment into the market as mature and functional technology, competitive for large and medium-scale applications. In this article, an experimental analysis of Bubbling Fluidized Bed Reactors (BFBR) for power generation with different sand mixtures is presented. This experimental study is based on a 50 kW and a 10 kW BFBR, respectively, with inner diameters of 10 cm and 22.2 cm. In the experiments, three different bed sands (molochite, silica and alumina) have been tested attending to their properties, such as operating temperatures, bulk density for fluidisation, homogeneous grain size and commercial availability. According to the implemented tests, the obtained results demonstrate that molochite (calcined kaolin) has best effectiveness as bed additive for optimizing the reactor pressure drop, operating temperature, mean particle size and fluidization conditions. On the other side, the most appropriate range for the relation between the sand bed height and the reactor inner diameter to reduce the energy consumption was determined. Finally, the distributor pressure drop and the range of pressured drop were experimentally evaluated.

Investigation of Lane-consistent dispersive optical-model potential for <sup>208</sup>Pb

Lead is an important alloy material nuclide. And lead eutectic is also an important coolant, which is applied in the construction of Lead-cooled Fast Reactor such as The European Lead-cooled System (ELSY) and the China Lead-based Research reactor (CLEAR-I), as well as in research related to Generation-IV reactor. The study and calculation of lead nuclear data have important theoretical value and application prospects. <sup>208</sup>Pb is the most stable and abundant isotope in lead nuclei, and high quality description of <sup>208</sup>Pb nuclear scattering data is the key to achieving theoretical calculations of nuclear reaction data for lead nuclei. Based on the dispersive optical model, this work describes nucleon scattering on <sup>208</sup>Pb by using the dispersive optical potential. The dispersive optical model potential is defined by energy-dependent real potentials, imaginary potentials, the corresponding dispersive contributions to the real potential which are calculated analytically from the corresponding imaginary potentials by using a dispersion relation, and isospin dependence is reasonably considered by introducing isovector component (i.e. Lane term) in the potential depth constants of the real Hartree-Fock potential <i>V</i><sub>HF</sub> and the surface imaginary potential <i>W</i><sub>s</sub>. Unlike K-D potential, which requires two different sets of parameters to describe neutron and proton induced scattering data, this optical potential uses the same set of parameters to simultaneously describe nucleon-nucleus scattering data. The derived potential in this work shows a very good description of nucleon-nucleus scattering data on <sup>208</sup>Pb up to 200 MeV. Calculated neutron total cross sections, neutron and proton elastic scattering angular distributions, as well as neutron and proton elastic analyzing powers are shown to be in good agreement with experimental data. Additionally, the difference in potential between the neutron and protons induced is described by the isovector term, a reasonable good prediction of quasielastic (p, n) scattering data is achieved.

Coal-to-nuclear modernization of an existing power plant with a IV-th generation nuclear reactor

The paper presents an analysis of a transition from a coal-fired power plant to a nuclear unit. The main focus is set on the extensive usage of the existing parts of the already operating system. The key problem is the correct matching of a nuclear reactor and the steam island. It is assumed here that the reactor module operates under nominal conditions and the steam turbine is adapted to fit the reactor. The paper describes the numerical model of the steam turbine cycle for the off-design simulations. The developed model allows us to determine the changes in the steam cycle in order to match the required water and steam temperature values at the inlets and the outlets of the steam generator. The paper presents the suggested modifications and the evaluation of the operation after the transition.

Impact of Hall current on radiative Ree-Eyring nanofluid flow across rectangular frame with Joule and viscous dissipation effects

Fluid flow and heat transfer inside a rectangular framework have numerous engineering applications, including energy generation, power systems, cooling technologies, and nuclear reactors. Researchers are currently focusing on improving heat transfer rates through the use of nanofluids. With this motive, the current study investigates the constant, incompressible flow of nanofluids containing copper nanoparticles suspended in ethylene glycol and water, within a rectangular framework. The study investigates the non-Newtonian Ree-Eyring nanofluid flow with impact of Hall Effects. Furthermore, the impacts of Joule dissipation, thermal radiation, and viscous dissipation are examined. Additionally, the physical quantities skin friction coefficient and Nusselt number are investigated. The mathematical modeling is examined, followed by the application of similarity transformation to the nonlinear governing partial differential equations (PDEs) with particular boundary conditions. This procedure simplifies the derivation of dimensionless ordinary differential equations. The dimensionless scheme of equations is then solved numerically using the bvp4c numerical technique, which is integrated into the MATLAB software. Graphical and tubular representations are used to study the logical impact of fascinating aspects on subjective dimensions. According to the findings, greater magnetic parameters reduce velocity components, whereas larger Ree-Eyring fluid values enhance them. The temperature profile improves when magnetic parameters increase, but declines as Ree-Eyring fluid parameters rise. As suction parameters increase, the skin friction coefficient drops. The Nusselt number drops with a rising Biot number. This study have examined the first time to the best of authors knowledge.

Optimization of the route of the refueling machine to reduce the refueling time: Case study of the BN-800 reactor

Sodium-cooled fast reactors (SFR) are included in the list of fourth-generation nuclear energy systems, the so-called generation IV (GEN-IV). It is the only technology of GEN-IV that possesses the significant practical experience of design, construction, and operation of high-power reactors. Due to the high temperature of the coolant at the BN-600 and BN-800 reactor core outlets, the steam generator produces steam with higher enthalpy, and this significantly increases the efficiency of cogeneration at SFR power plants in comparison with pressured water reactor (PWR) or boiling water reactor (BWR) units. The primary nuclear fuel effectiveness rises due to cogeneration with higher efficiency of a sodium-cooled fast reactor nuclear power plant (NPP) in comparison with thermal reactors. In addition, it is necessary to reduce the reactor refueling outages to improve the nuclear power plant utilization factor.It is possible to reduce the duration of the reactor refueling by the route optimization of the refueling machine movement. The prevailing in the world thermal neutron reactors with water cooling have a refueling machine that points the certain fuel assembly using two coordinates. The fast neutron reactors use sodium as a coolant; thus, there is a problem with its violent reaction with water and air oxygen. Therefore, it is necessary to exclude the contact of sodium and surrounding air during the refueling. To achieve this, the system of refueling machine pointing is used, consisting of two or three eccentrically located rotating plugs with hydraulic locks. The article presents the results of the creation of mathematical models of refueling machine movement and fuel assembly gripping. The time-optimal algorithms for the operation of refueling machines with three rotating plugs are proposed. The use of the new algorithm allows to reduce the time of movement of the grip of the refueling machine by 30–37%.

Lessons from national approaches: a long uphill struggle in search of sites for repositories for nuclear waste locations

Abstract. A quarter of a century ago, a long-term expert in the nuclear waste scene stated that “the management of spent nuclear fuel and high-level nuclear waste has the deserved reputation as one of the most intractable policy issues facing the United States and other nations using nuclear reactors for electric power generation” (North, 1999, p. 751). Apart of exceptions, this statement is still true. At some points, however, there is light at the end of the tunnel if we can read the signs of the times. It will be a long hike, in steep terrain, poor visibility and with an approximate destination. We need a safe and acceptable site, tolerated by the affected parties, where a repository can be built, operated and, finally, closed down in reasonable course by a generation to come and with a clear conscience. This contribution does not present the silver bullet (which does not exist) but suggests some criteria and characteristics which have not been respected in the history of final disposal – but they should be. It needs adequate resources: stable structures, competent institutions, learning personnel (in institutions and civil society), mature and open discourse as well as sufficient time. Based on https://doi.org/10.5194/egusphere-egu24-6514 (Flüeler, 2024a).

Evaluating Nuclear Forensic Signatures for Advanced Reactor Deployment: A Research Priority Assessment

The development and deployment of a new generation of nuclear reactors necessitates a thorough evaluation of techniques used to characterize nuclear materials for nuclear forensic applications. Advanced fuels proposed for use in these reactors present both challenges and opportunities for the nuclear forensic field. Many efforts in pre-detonation nuclear forensics are currently focused on the analysis of uranium oxides, uranium ore concentrates, and fuel pellets since these materials have historically been found outside of regulatory control. The increasing use of TRISO particles, metal fuels, molten fuel salts, and novel ceramic fuels will require an expansion of the current nuclear forensic suite of signatures to accommodate the different physical dimensions, chemical compositions, and material properties of these advanced fuel forms. In this work, a semi-quantitative priority scoring system is introduced to identify the order in which the nuclear forensics community should pursue research and development on material signatures for advanced reactor designs. This scoring system was applied to propose the following priority ranking of six major advanced reactor categories: (1) molten salt reactor (MSR), (2) liquid metal-cooled reactor (LMR), (3) very-high-temperature reactor (VHTR), (4) fluoride-salt-cooled high-temperature reactor (FHR), (5) gas-cooled fast reactor (GFR), and (6) supercritical water-cooled reactor (SWCR).

Pure metals for creation the nuclear structural materials

The paper presents the main results of research on the development of structural materials for equipment elements of existing and prospective nuclear power plants. The further development of nuclear energy largely depends on the development of advanced structural materials for new generation reactors and the improvement of materials for operating nuclear power plants. The high requirements for structural materials regarding the content of impurity elements can be met only when high-purity metals are used as initial components and new technologies for their production are used. The prospects of using new methods and approaches for evaluating and improving the physico-mechanical and physico-chemical properties (microhardness, corrosion resistance, radiation resistance, etc.) of structural materials for nuclear technologies were considered.

Development of a liquid metal subchannel code applied to ocean conditions and study on the effect of core geometry parameters and ocean motions on flow and heat transfer characteristics

The Generation IV International Forum (GIF) proposes a type of liquid metal reactor, due to its thermal properties and efficient heat transfer in a relatively small system. This special heat transfer characteristic allows the liquid metal reactor to be modular in design and flexible in installation on a nuclear-powered ship. Accurate prediction of the coolant temperature and flow distribution between subchannels is important to ensure nuclear safety in the complex ocean environment. In this study, an ocean version of subchannel code is developed by adding the motion terms in the momentum equations for the liquid metal reactor core based on the previously land-based subchannel code. Since the research on the characteristics of flow resistance and heat transfer mechanism under ocean conditions is not mature, the original empirical model is still used in the ocean version code. The constitutive models of the liquid metal are analyzed and incorporated into the ocean code. The code has been verified and validated by comparing with Computational Fluid Dynamics (CFD) simulation results, and experimental data. The effect of geometric parameters such as rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) is studied in the liquid metal lead 7 rod bundles. In addition, the effect of heaving start-up and shutdown motions between liquid metal lead and water is discussed in the 5 × 5 bare rod bundles. In general, the modified subchannel analysis code can well complete the calculation of liquid metal reactor under ocean conditions, as well as provide support for the design and safety analysis of a liquid metal cooled fast reactor.

Data and modeling sensitivity analysis for molten salt fast reactor benchmark – Static calculations

The Molten Salt Reactor (MSR) idea is increasingly being recognized in the nuclear field due to its potential safety, sustainability, and economic efficiency advantages. The Molten Salt Fast Reactor (MSFR) benchmark, introduced in 2019, highlighted variations in results tied to different neutron cross-section libraries. This study investigates the impact of utilizing the ENDF/B-VIII.0 and JEFF-3.3 cross-section libraries for MSFR benchmark assessment compared to the ENDF/B-VII.1 database. Monte Carlo based open source code OpenMC is used for the analyses. Rigorous sensitivity analyses assess the influence of individual components, including the cross-section database, resonance elastic scattering, and Thermal Scattering Law (TSL). Beyond the criticality assessments, parameters such as delayed neutron fraction, temperature coefficient of reactivity, and neutron spectrum are compared for different cross-section libraries. Our analyses reveal that incorporating new evaluations for 233U (n,γ) and fission cross-sections in ENDF/B-VIII.0 significantly alters criticality results, i.e., more than 1700 pcm difference is seen between libraries. Similarly, critical concentration using ENDF/B-VII.1 and JEFF-3.3 is over-predicted by approximately 3%. The variations in Thermal Scattering Law (TSL) files do not yield substantial differences in outcomes due to the fast spectrum of the reactor. In some cases, the treatment of resonance elastic scattering leads to reactivity differences greater than 50 pcm. The benchmark compares 233U-started and Minor Actinide (MA)-started core. From the reactor physics point of view, the MA-started core leads to a 29% higher (n, γ) reaction rate than the 233U-started core. A 3–4% smaller value of thermal reactivity coefficient is obtained using the ENDF/B-VIII.0 library compared to the ENDF/B-VII.1 value. Using the ENDF/B-VIII.0 for the MSFR benchmark signifies using newer and better data for the GEN-IV reactors neutron physics calculations.

Thermo-neutronic integrated coupling effects on nuclear reactor core calculations

Small Modular Reactors (SMRs) have gained significant attention as the next generation of nuclear reactors due to their unique features, such as improved economic viability and safety. This study focuses on the SMART reactor, the first certified design of an SMR, and investigates various Thermo-Neutronics aspects using an integrated coupling scheme of nuclear codes. The Neutronics cell and core calculations are performed using the DRAGON/DONJON codes, while the COBRA code is employed for thermal-hydraulic calculations. The analysis considers parameters including axial and radial power peaking factors (PPFs), critical heat flux (CHF), minimum departure from nucleate boiling ratio (MDNBR), axial average coolant temperature in hot channel and core, and maximum fuel temperature. These parameters are evaluated using the developed coupling system. Code-to-code verification is conducted to ensure the accuracy of the results, demonstrating good agreement. The coupling approach reveals significant effects compared to stand-alone code modeling, emphasizing the importance of considering the coupling between Neutronics and thermal-hydraulics, especially during the cycle length. This study contributes to the understanding of SMRs, particularly the SMART reactor, by providing insights into key parameters and their interdependencies. The integrated coupling scheme enhances the accuracy of the analysis and highlights the significance of considering coupling effects in the modeling process during the steady state and the cycle length.

Structural Changes in High-Entropy Alloys CoCrFeNi and CoCrFeMnNi, Irradiated by He Ions at a Temperature of 700 °C.

High-entropy alloys (HEA) are promising structural materials that will successfully resist high-temperature irradiation with helium ions and radiation-induced swelling in new generations of nuclear reactors. In this paper, changes in the elemental and phase composition, surface morphology, and structure of CoCrFeNi and CoCrFeMnNi HEAs irradiated with He2+ ions at a temperature of 700 °C were studied. Structural studies were mainly conducted using the X-ray diffraction method. The formation of a porous surface structure with many microchannels (open blisters) was observed. The average diameter of the blisters in CoCrFeMnNi is around 1.3 times smaller than in CoCrFeNi. It was shown that HEAs' elemental and phase compositions are stable under high-temperature irradiation. It was revealed that, in the region of the peak of implanted helium, high-temperature irradiation leads to the growth of tensile macrostresses in CoCrFeNi by 3.6 times and the formation of compressive macrostresses (-143 MPa) in CoCrFeMnNi; microstresses in the HEAs increase by 2.4 times; and the dislocation density value increases by 4.3 and 7.5 times for CoCrFeNi and CoCrFeMnNi, respectively. The formation of compressive macrostresses and a higher value of dislocation density indicate that the CoCrFeMnNi HEA tends to have greater radiation resistance compared to CoCrFeNi.

Study of radiation-induced structural changes in the near-surface layer of ZrO2 ceramics caused by He2+ irradiation

Abstract The aim of this study is to determine changes in the near-surface layer of ZrO2 ceramics caused by irradiation with low-energy He2+ ions, as well as the associated formation of defects and structural deformations. During the experimental work conducted, it was established that the accumulation of deformation-structural distortions is of two stage nature, having a direct dependence on irradiation fluence and, therefore, on atomic displacement value. It was determined that at small atomic displacement values (less than 10 dpa), the main mechanisms of structural distortions are caused by tensile residual stresses, the value of which is less than 0.1 GPa. At the same time, the deformation distortion of chemical bonds has a pronounced anisotropy associated with a more pronounced distortion of the Zr–OI chemical bonds, the distortion of which results in the formation of vacancy defects in the form of VO. During determination of alterations in optical characteristics depending on the atomic displacement value, it was found that the dominant role at small values of dpa (less than 10 dpa) is played by point defects, which influence the formation of obstacles in the form of absorbing centers. In this case, an increase in the irradiation fluence above 1017 ion/cm2 results in a growth in the linear refractive index, the change of which has a direct correlation with the value of residual stresses in the damaged layer. Certain dependencies of changes in structural features and their relationship with deformation distortions, as well as the accumulation of vacancy defects, can subsequently be used to predict the potential of using these ceramics as materials for new generation nuclear reactors.

Properties of W–Ta Materials of the Neutron-Producing Target of the Subcritical Assembly at the National Scientific Centre ‘Kharkiv Institute of Physics and Technology’ of the National Academy of Sciences of Ukraine

The works in the field of radiation materials science of target materials of neutron sources based on the subcritical assemblies controlled by linear accelerators of electrons or protons, so-called accelerator driven systems (ADS), are reviewed. Now, electronuclear ADS systems are the prototype of safe nuclear reactors of the 5th generation. In connection with the physical start-up of the neutron source facility of the National Scientific Centre ‘Kharkiv Institute of Physics and Technology’ of the N.A.S. of Ukraine (NSC ‘KhIPT’ NASU), the target of which is fabricated from powdered tungsten covered with tantalum, the issues of preparation technology, construction, and physical and mechanical properties of W–Ta materials of targets in non-irradiated and irradiated states are considered. The nuclear-physical processes of radiation damage to the target during co-irradiation of it with both neutrons of a subcritical assembly and high-energy e- and γ-beams are analyzed. The target resources of ADS systems are estimated. As noted, the difficulty of predicting the ‘survivability’ of the W–Ta target also relates to the fact that, except for the works carried out at the NSC ‘KhIPT’ NASU in 70–90th of the last century, there are no experimental works in the world on the radiation damage of reactor materials by high-energy electrons with an energy of 100 MeV and above.

Influence of injected ions on α’ formation under ion irradiation in Oxide Dispersion Strengthened Steels

Oxide Dispersion Strengthened (ODS) steels hold great promise for applications in next generation reactors. Under irradiation, a phase separation α/ α’ can occur within the Fe-Cr matrix of ODS steels that can alter their mechanical properties. This work presents, for the first time, the characteristics of α’ precipitates enhanced by ion irradiation at 400 °C and examines the influence of the implanted ions. Far from the implanted region, α’ is reported in significant density while at the implanted peak, the α’ density is considerably reduced. This suggests that ion implantation either reduces the fraction of α’ phase formed after irradiation or delays considerably its formation. Through atom probe tomography analysis and comparison with existing literature, the low impact of the damage rate and fluence on the α’ formation in ODS steels is highlighted. Interestingly, the efficiency of ballistic mixing of α’ appears to be less pronounced in ODS steels than in Fe-Cr systems.

Comparison of Monte Carlo Tools for Displacement Calculations for Protons and Heavy Ions Irradiation on Iron and EUROFER97

Abstract Ion irradiation of structural materials for the current and future generation nuclear reactors is a widespread technique used to simulate neutron irradiation in the context of material science research. During the designing phase of irradiation experiments Monte Carlo transport codes are often employed to quantify the radiation exposure, by estimating the displacements per atom (dpa) quantity. From the available displacement calculation codes, four popular choices were selected: cern-fluka, mcnp, phits, and srim. These codes will be used to simulate the irradiation effects of proton and Fe ions on pure iron (G379) and EUROFER97 alloy (a common material in fusion applications) targets. Results coming from the different software are presented in the form of damage profiles and of number of displacements as a function of the primary beam energy. This study identifies discrepancies between the codes' outputs and provides optimized simulation setup parameters.

Evidence that Femto-H2 Exists Leads to the Correct Mechanism of Cold Fusion (Cold Fusion is Caused by Femto-D2)

Femto-H2 is generated by compression of H- H bond of H2 confined at expandable T site, and authors’ mechanism of Cold fusion is that femto-D2 created at expandable T site cause fusion of D+D due to the shielding of coulomb repulsive force because femto- D2 has so dense electron between d-d that coulomb repulsive force is shielded to cause Fusion. One of the controversial heat generation reactors is QHe (Quantum Hydrogen Energy) developed by clean-planet, who explains that QHe is a heat-generating technology with hydrogen quantum diffusion. The diffusion is induced by heating a small amount of hydrogen saturated in nano-sized nickel-based composite material. Although they did not describe it cold fusion, it is commonly understood to be cold fusion. Therefore, this kind of Cold fusion debate is so confusing that no common standard theory has been established. Therefore, I would like to explain the mechanism of QHe based on author’s Mechanism of Cold Fusion. with H2 loading into multi-stacked Ni/Cu which has large number of reaction site of grain-boundary, where the hydrogen segregates to enhance the reaction of femto- H2 generation. And femto-H2 can fuse to Cu or Ni to be mainly Ze to generate 15~17MeV, which is comparable of energy by D+D=4He+24MeV.Because femto-H2 is so small to fuse to the metal nucleus at room temperature, and at high temperature, the nucleus vibrates at high speed and femto-H2 can fuse with the metal nucleus and once the heat generation start transmutation continues. Author also discovered phenomena that seems to be related to femto-H2, which are transmutation reactor by Dr. Ohmasa and car on H2O (Water Fuel CELL) by Stanly A Meyer. Ohmasa’s transmutation reactor uses Pd coating plate and loaded hydrogen by electrolysis generate femto-H2 which enter into H2O to transmute H2O to be helium-3 and oxygen-18 clusters. Water Fuel Cell can generate huge heat by the transmutation of metal with femto-H2. Therefore, author would like to request clean-planet to study the transmutation with femto-H2 in QHe reactor to prove author’s mechanism hypo, and Cold fusion community to start the discussion on the standard theory of Cold Fusion because author’s mechanism is inconsistent with nucleus model.

Long term reactivity operation for a high temperature testing reactor (HTTR) loaded with uranium nitride fuel

A High Temperature Testing Reactor (HTTR) is one of the fourth generation nuclear reactors, the reactor uses UO2 as typical fuel with different axial and radial enrichments. In the present research, uranium nitride (UN) is proposed as a fuel for the reactor for the same core enrichment distribution. The core multiplication factor, cycle length, discharge burnup are calculated and the results of UO2 fuel are compared with UN fuel results. Neutron spectrum, fissile isotopes behavior and Breeding capability are compared for the two fuel types. Kinetic parameters, Control rods and Burnable poisons worth are also evaluated. The results indicated that uranium nitride (UN) has a longer fuel cycle, higher breeding capability and harder thermal flux compared to UO2 fuel.

Site effect on seismic response of buried nuclear power plant structure

The development of multifunctional, multipurpose small modular next generation nuclear reactors has become a trend in designing small-scale nuclear power stations. In this study, taking a new next-generation buried nuclear power structure as the target structure, a 3D finite element model of the site-nuclear power plant structure system was established, and its response at different sites was studied. Using the internal substructure method, this study initially explored the impact of earthquake input and artificial boundary locations on the structural response. The results indicate that the internal substructure method enhances computational efficiency without compromising accuracy. For this model, it is recommended to input earthquake motion close to the structure, set the lateral boundary location to three times the width of the structure, and set the bottom boundary location to two times the depth of the structure. Subsequently, an investigation into the response of the structure at varying site wave velocities is conducted. The results show that as the wave velocity increases, the acceleration response intensifies, while the displacement response diminishes. However, the relationship is nonlinear; the response exhibits less variation with increasing site wave velocity. The seismic response of a structure is profoundly influenced by site conditions, emphasizing the imperative need for careful consideration of specific site characteristics in seismic analyses. In addition, the response pattern of the overall structure is significantly different from that of conventional large surface-sited nuclear power plants, and further research should be conducted on the seismic response characteristics of buried nuclear power plants.