Modern Reactors Research Articles

The Role of Nuclear Energy in Achieving Sustainable Energy Solutions

Purpose: The aim is to assess the role of nuclear energy as a stable and environmentally friendly energy source, as well as its advantages over fossil fuels and renewable energy. The research identifies the benefits, challenges, and technological advances in nuclear energy, as well as the potential for its expanded use in global efforts to reduce carbon emissions and transition to more environmentally friendly energy sources. Study Design/Methodology/Approach: The study used qualitative methodology. Data was collected from secondary sources; scientific papers, government reports, and industry assessments. Findings: Nuclear energy is a stable, low-carbon energy source that is independent of weather conditions, supporting grid stability, especially during times of peak demand or low renewable energy availability. But there are significant challenges. Modern reactor designs and new cooling systems can overcome these challenges and improve nuclear energy efficiency. International co-operation and oversight by the IAEA are essential to ensure global nuclear security. Originality/Value: Comprehensively analyses the role of nuclear energy in meeting global energy needs, combating climate change, and driving technological innovation. Evaluates the benefits and challenges of nuclear energy, with an emphasis on safety, waste management, and technological advancement. Policy recommendations, investment guidelines, and strategies to extend the lifetime of nuclear power plants.

Importance of Nuclear Energy to Accomplish Net Zero Carbon

The world is entangled with the urgent need for existing energy modes transition to clean energy for mitigating global warming and climate change. It ultimately demands to set stringent goals for achieving net zero carbon for each country. In this direction different approaches are being stipulated and one of the important options emerges as nuclear power, a compelling option for sustainable development. Unlike renewable energy sources such as solar, wind and nuclear energy which provides a continuous and reliable supply of electricity with minimal greenhouse gas emissions. On the other hand nuclear power plants operate with high capacity factor compared to variable renewable sources that depend on weather conditions and its reliability makes nuclear energy a strong candidate for baseload power. Moreover, nuclear energy's low carbon footprint and high energy density output makes it an attractive alternative to fossil fuels. Although, it is crucial to acknowledge the challenges associated with nuclear energy, including the safe management of nuclear waste, the risk of nuclear accidents, and the high upfront costs of building nuclear power plants. Hence, a balanced perspective on the future of nuclear energy and its potential to contribute to a sustainable future is to be adopted carefully. However, a periodic advancements in technology and stringent safety measures, modern Nuclear reactors are designed to minimize environmental impact and enhance safety.

Risk-Informing Nuclear Reactor Safety: The Prediction of the Probability of Core Damage Due to Loss of Power and Cooling

Abstract The modern idea of risk-informed decision-making (RIDM) is here critically examined for all existing, advanced and, generation-IV nuclear reactor systems. Motivated by the predictive difficulties of probabilistic risk assessment (PRA) in regard to occurred accidents, it is evident that the real (not hypothetical) consequences of nuclear core damage accidents that must be considered and quantified are the financial losses, infrastructure damages, societal disruptions, and adverse political policies, and not solely the traditional exceedance of regulated radiation release or public exposure limits. With this perspective, a new dynamic analysis is presented for estimating the probability of core damage due to extended loss of power and cooling in a modern nuclear reactor, giving results different from current standardized PRA/PSA analyses. Using existing data, we approach the multitude of different events in a new way: whatever the initial event in the finite event set {external flood, fire, hurricane, ice storm, typhoon, earthquake, cyber attack…}, the fundamental concerns are the consequent loss and nonrestoration of power, and the inadequate cooling of the core. The present proposed quantitative evaluation uses applicable and fully “exchangeable” severe event data for nuclear and nonnuclear systems, including active and passive emergency back-up systems for a wide range of power losses that lead to widespread damage and societal disruption. While not design-specific, this new independent “order-of-magnitude” estimate for the probability of core damage is some two to two hundred times larger than that shown or reported in recent modern and formal PSA/PRA for licensing submissions using generic failure rate data rather than actual severe event data. This new finding suggests greater uncertainties exist than presently assumed for risk-informed decision-making (RIDM), and points to the need for a major reconsideration and updating of risk assessment and regulatory risk-informed approaches for nuclear plant safety evaluation.

New perspective crosslinking electrochemistry and other research fields: beyond electrochemical reactors.

Over the years, electrochemical reactors have evolved significantly, with modern reactors now able to achieve a high current density and power output in compact sizes. This leap in performance has not only greatly accelerated the rate of electrochemical reactions but also had a broader impact on the environment. Traditional research perspectives, focused primarily on the internal working systems of reactors, possibly overlook the potential of electrochemical systems in regulating their surrounding environment. A novel research perspective considering the interaction between electrochemical processes and their environmental context as a unified subject of study has gradually emerged alongside the dramatic development of electrochemical techniques. This viewpoint introduces a paradigm shift: electrochemical reactors are not isolated entities but rather are integral parts that interact with their surroundings. Correspondingly, this calls for an innovative research methodology that goes beyond studying the electrochemical processes in isolation. Rather, it integrates the design of the electrochemical system with its specific application environment, ensuring seamless integration for optimal performance under various practical conditions. Therefore, performance metrics should include not only the basic parameters of the electrochemical reactions but also the adaptability of the electrochemical system in real-world scenarios beyond the laboratory. By focusing on environmental integration and application-driven design, the applications of electrochemical technology can be more effectively leveraged. This perspective is exemplified by an electrochemical system based on coupled cathodic oxygen reduction and anodic oxygen evolution reactions. By adopting this new research paradigm, the applications of this electrochemical system can be extended to fields like medical treatment, food science, and microbial fermentation, with an emphasis on tailored designs for these specific application fields. This comprehensive and systematic new research approach aims to fully explore the potential applications of electrochemical technology and foster interdisciplinary collaboration in the electrochemical field.

Optimization analysis of nuclear thermal coupling for a small nuclear thermal propulsion (SNTP) reactor

As a main structure, tie-tube elements are used both in the traditional and modern LEU nuclear thermal propulsion reactors. This complex structure and its corresponding complex convective heat transfer process have been studied, including CFD calculations. But these studies did not effectively evaluate the applicability of the models and typical problems. In this study, a general small nuclear thermal propulsion (SNTP) reactor model is established by using three-dimensional Monte Carlo code and commercial CFD program Fluent. The applicabilities of material properties, turbulence models and grids are clarified by analyzing various influencing factors. The calculation results show that due to the small enthalpy difference and heat transfer, the convection effect of inner flow based on hydrogen can be ignored. There is no factor that can significantly reduce the maximum fuel temperature, but the porosity of the insulator can significantly improve the heat transfer ratio. In terms of thermal safety and overall performance, 2:1 support ratio is still the best choice. Based on the simulation results, the phenomenon of steep temperature rise of outer tie-tubes is found, and the SST k-omega model with Y + <3 is recommended for simulation of similar problems.

Scale analysis using FSA methodology for single-phase natural circulation in a simplified RCCS

Improving the safety and reliability of innovative Generation IV nuclear reactors can be achieved by adopting passive safety systems for accident prevention. These systems rely only on physical phenomena such as natural circulation and eliminate the need for an external response in the event of an accident. While passive natural circulation systems are widely used in modern nuclear reactors, their complexity and high thermal power output make them difficult to simulate at full scale. One solution is to study reduced-scale systems that are more feasible for experimental investigation. The correct scale procedure is essential to ensure that the results of these systems represent those expected in a real facility. This article describes a small-scale, single-phase water natural circulation circuit representing a simplified reactor cavity cooling system. The standard scale method is combined with an innovative fractional scaling analysis to determine the conditions and similarity parameters for a test facility. This analysis outlines the necessary simplifications required to achieve a certain similarity between natural circulation systems in the transient regime. The methodology is applied to two studies from the literature, demonstrating the feasibility of comparing different passive heat removal systems.

Enhancing sensitivity to leptonic CP violation using complementarity among DUNE, T2HK, and T2HKK

After the landmark discovery of non-zero theta _{13} by the modern reactor experiments, unprecedented precision on neutrino mass-mixing parameters has been achieved over the past decade. This has set the stage for the discovery of leptonic CP violation (LCPV) at high confidence level in the next-generation long-baseline neutrino oscillation experiments. In this work, we explore in detail the possible complementarity among the on-axis DUNE and off-axis T2HK experiments to enhance the sensitivity to LCPV suppressing the theta _{23}-delta _{textrm{CP}} degeneracy. We find that none of these experiments individually can achieve the milestone of 3sigma LCPV for at least 75% choices of delta _{textrm{CP}} in its entire range of [-180^{circ }, 180^{circ }], with their nominal exposures and systematic uncertainties. However, their combination can attain the same for all values of theta _{23} with only half of their nominal exposures. We observe that the proposed T2HKK setup in combination with DUNE can further increase the CP coverage to more than 80% with only half of their nominal exposures. We study in detail how the coverage in delta _{textrm{CP}} for ge 3sigma LCPV depends on the choice of theta _{23}, exposure, optimal runtime in neutrino and antineutrino modes, and systematic uncertainties in these experiments in isolation and combination. We find that with an improved systematic uncertainty of 2.7% in appearance mode, the standalone T2HK setup can provide a CP coverage of around 75% for all values of theta _{23}. We also discuss the pivotal role of intrinsic, extrinsic, and total CP asymmetries in the appearance channel and extrinsic CP asymmetries in the disappearance channel while analyzing our results.

Microstructural characterization of the CGB graphite grade from the molten salt reactor experiment

In the 1960s, the Molten Salt Reactor Experiment demonstrated the feasibility of molten salt reactors for civil applications. The Molten Salt Reactor Experiment used a nuclear graphite grade known as CGB in the core as the fast neutron moderator. For application in this purpose, CGB graphite underwent additional impregnation steps that are not typically used in conventional nuclear graphites. These additional impregnations significantly lowered the permeability and reduced the size of pore openings in CGB graphite, for the purpose of inhibiting molten salt ingression into the graphite. Although several publications report the mechanical properties and irradiation response of CGB graphite, little information has been published about the microstructure or sealant used for this graphite grade. Here the results of multiple advanced microscopy techniques are presented with the objective to investigate the unique microstructure and sealing technology of legacy material from the Molten Salt Reactor Experiment so it can be potentially readopted for modern reactors. The microstructural characterization shows that some pores were fully sealed by the impregnation whereas other regions contained foam-like materials that partially filled the void content. This manuscript also investigates and documents the nature and possible source of the sealant material. The pore sealing methods used for CGB are discussed and compared with other common technologies used to reduce the impregnation of molten salts into graphite components. Knowledge of CGB graphite's microstructure can enable development of new pore infiltration technologies to improve in operando behavior advancing the next generation of molten salt reactors.

Modern technologies in processing and utilization of hydrocarbon waste by thermal methods

As a result of the extraction, processing, transportation and storage of oil, oil waste is formed, which are carriers of valuable products. The processing of these wastes makes it possible to obtain alternative energy sources in the form of fuel components, thereby making the oil and gas industry more sustainable, stabilizing the load on it. This article is devoted to the analysis of modern technologies for the thermal processing of hydrocarbon waste. An overview of existing technologies for handling organic waste is presented. The main advantages of thermal methods of processing and utilization are highlighted. Their advantages and disadvantages are given in general terms. It was revealed that the most promising method is thermal destruction due to the advantages of the purity of the feedstock. Modern technical solutions of each of the technologies are analyzed. A complex installation for thermal destruction with intensifying devices is presented. It has been established that the use of various intensifying devices in modern reactor equipment can increase the efficiency of thermal destruction. Keywords: hydrocarbon waste; oily sludge; modern technologies; thermal methods; alternative sources; thermal destruction; reactor.

Interaction with the Public as a Condition for the Development of Nuclear Energy

The subject of the study is the practice of interaction of nuclear energy specialists with the public, whose representatives have different opinions on the need for further operation of nuclear power plants, the construction of new ones in modern conditions. On the one hand, the creators of nuclear power plants prove that modern nuclear reactors make it possible to minimize the threat of emergency situations. On the other hand, in the public opinion of different countries, the idea of the danger posed by these industrial facilities is steadily preserved. The purpose of the work is to disclose practical actions taken by the international atomic energy organization IAEA, Rosatom to ensure dialogue between specialists in this area and interested social groups. The object of study is publications that have appeared in recent years in foreign and domestic literature on this issue. It has been shown that there is a search for such interaction of atomic physicists with the public in the media, the Internet, which will overcome existing biases about the state and prospects of nuclear energy. The absence of an open discussion of current problems of NPP development as a way to solve not only problems related to the economic development of countries, but also the protection of nature from the harmful effects of enterprises using gas, coal, oil for energy production, is one of the main reasons for the public's wary attitude towards nuclear power plants. Data on new technologies for creating objective ideas about the safety of modern nuclear energy in different public groups used in Russia and supported by specialists from other countries are presented. Conclusions were drawn about the need to create a system for the formation of knowledge about the state of nuclear power plants among the population, modern means of preventing emergency situations. These issues should be considered in special educational programs for different groups of the population, in the media with the involvement of qualified specialists.

(Digital Presentation) Pulsed-Waveform Electrodeposition of Fe-W Brazing Interlayers for Fusion Applications

Numerous government and commercial entities are currently engaged in research toward net-energy-positive nuclear fusion electricity generation technology, which if successfully developed would revolutionize the energy economy and have a significant beneficial impact on carbon emissions and anthropogenic climate change. One fusion subsystem with significant remaining engineering challenges comprises the multitude of divertor modules installed on the plasma-facing surface of modern nuclear fusion reactors. These divertors must absorb/conduct steady-state heat fluxes on the order of 10–20 MW/m2 in an extremely harsh environment, including high-energy neutron bombardment. One common material choice for the plasma-facing portion of these components is tungsten, which provides a high melting point, high sputtering resistance, low tritium retention, and other benefits. As the thermal conductivity of tungsten is somewhat low, however, these tungsten components are often paired with a heat sink made from a more thermally-conductive material, such as copper-chromium-zirconium alloy, CuCrZr. While these materials can be joined by a number of methods, such as diffusion bonding and brazing, the dramatic mismatch in coefficient of thermal expansion (CTE; ~4 μm/m/K and ~17 μm/m/K for tungsten and CuCrZr, respectively) has a significant detrimental effect on the joint strength under these high-temperature, high-heat flux conditions. Development of a means for mitigating this CTE mismatch is imperative for the use of these materials to be feasible in a practical fusion device.This talk will present results to date on development of an electrodeposited functionally-graded iron-tungsten (Fe-W) interlayer system for mitigation of the CTE mismatch between tungsten and CuCrZr in brazed joints. In theory, an interlayer with a CTE functionally graded between “tungsten-like” and “copper-like” values should significantly mitigate the thermally-induced stresses at the tungsten/CuCrZr joint. A key objective of the work is to demonstrate sufficient control over the local interlayer composition, residual stress, and other properties by adjusting primarily the pulse/pulse-reverse electrical waveform parameters, so as to allow interlayer preparation from a single electrodeposition bath. Results from electrodeposition trials and mechanical testing of brazed joints under ambient conditions as an initial screen for interlayer performance will be discussed, in anticipation of high-heat flux testing of joint performance under fusion-relevant conditions, where the ultimate goal is demonstration of increased tungsten/CuCrZr joint lifetime under thermal load via inclusion of graded interlayers, as compared to similar joints without interlayers. Figure Caption (Left) Tungsten substrate with electrodeposited Fe/W interlayer. (Right) Brazed tungsten/CuCrZr sample with Fe/W interlayer. Figure 1

Possible Implementation of Ex-Core Measurement in TEPLATOR Graphite Reflector

Abstract An ex-core neutron flux measurement is a crucial system for all common power reactors. It is necessary to monitor the neutron flux and control the chain reaction, therefore, the ex-core neutron flux measurement is one of the main safety and control systems. The main advantage of this arrangement of detectors is a fast response to neutron flux change, which determines the reactor power change. Regarding to the new reactor concepts, it is important to deal with improved detection systems suitable for these reactors. Many of the modern reactor concepts are based on a graphite moderator or reflector, which is also the case of the TEPLATOR. The TEPLATOR is a solution of a district heating system based on heavy water as a moderator and graphite as a reflector. The TEPLATOR is designed to use irradiated fuel from the commercial pressurized water reactor or boiling water reactor, which has low to intermediate burnup. This work is focused on the verification of the possible use of the special neutron measuring system placed in the graphite reflector. The Monte Carlo code serpent was used for the calculations performed in this work.

Актуальні проблеми фізики динамічних пошкоджень в елементах першого контуру реакторів типу ВВЕР

In the context of the actual problems of the physics of operational damage of modern reactor steels produced in the leading countries of the world (USA, Russia, Western Europe) and used for the manufacture of nuclear reactor vessels and other equipment of the first circuit of nuclear power plants, the characteristic features of possible dynamic damage in the responsible elements of this are considered. The mentioned problems are systematized from the standpoint of analyzing the effects of radiation embrittlement, as well as physical and chemical processes that, under certain conditions, are capable of developing in the operating equipment of Ukrainian NPPs, which are already working out their design operational resource. The characteristic features of possible dynamic damage in the operating reactor equipment of Ukrainian and foreign nuclear power plants are considered. The problem is systematized, first of all, from the standpoint of analyzing the operational stability of domestic and foreign reactor steels in relation to their radiation embrittlement. The peculiarities of the course of this physical process have been analyzed, which should be taken into account when determining the maximum possible terms of extension of safe operation of nuclear power units with reactors of the VVER type at the NPP of Ukraine. The main metal-physical properties of reactor steels of various types and possible problems caused by neutron irradiation, physical and chemical processes, vibrational and thermomechanical fatigue, which threaten the unexpected sudden destruction of reactor vessels, are considered. Special attention is paid to mechanical damage and processes accompanying the operation of reactor housings under conditions of cyclic and dynamic loads. A warning has been given regarding the unjustified extension of the period of reactors safe operation. The significant technological lag of the former Soviet, and now Russian, metallurgy from the level of metallurgy of the leading Western countries was noted. Data are provided on the high operational properties of the latest American steels, from which modern reactors of the AR1000 type are manufactured in the USA, and the safety, technical, economic and environmental advantages of using these reactors in Ukraine in comparison with new models of reactors of the VVER-1000 and VVER-1200.

Щодо будівництва нових ядерних енергоблоків в Україні

The article provides information on the pace of construction and commissioning of nuclear power units at nuclear power plants in the world over the past 60 years. Nuclear energy is the most important factor in ensuring the energy independence of Ukraine. As of 2022, 80% of the entire fleet of Ukrainian nuclear power units (12 out of 15) have already been in operation for more than 30 years. In the world, this figure is 68%. Two power units with VVER-440 reactors have been in operation for more than 40 years, and by 2028 the number of power units at Ukrainian NPPs, that have been in operation for more than 40 years, will increase to 10. Therefore, the need and importance of introducing a program for the construction of new nuclear power units to ensure energy independence of Ukraine are obvious. The article discusses the most important characteristics of the modern AP1000 reactor unit, which is licensed in the USA, Canada and some other countries. Information is presented on the advantages of the AP1000 project over other modern reactor plants EPR-1750, APR-1400, VVER-1200, as well as issues that need to be paid special attention, when performing the relevant stages of work on the justification and implementation of the AP1000 technology in the nuclear power industry of Ukraine. For example, the installed capacity utilization factor of operating power units with EPR-1750, APR-1400, VVER-1200 is lower than for АР1000. When justifying the decision to choose AP1000, it is necessary to pay attention to the already known problematic issues of implementing Westinghouse technology. It is recommended to involve the institutes of the National Academy of Sciences of Ukraine in the work, related to the scientific and technical substantiation of the choice of promising nuclear installations for Ukrainian NPPs.

Hydraulic resistance of supercritical pressure water flowing in channels – A survey of literature

The survey is devoted to the problem of the hydraulic resistance of channels with water at supercritical pressures. The pressure drop in fuel assemblies is one of the main characteristics of water-cooled nuclear reactors. It is expected that pressure drops in SCWRs, which are considered now as a perspective of the IV Generation PWR, may be essentially less than in modern reactors. However, the flow resistance of the reactor equipment, as previously, is an important element among the various parameters of the future plants with SCWR. From 1968 up today, in spite of the obvious peculiarities of hydrodynamics and heat transfer processes in fluids at supercritical pressures, comparatively few experimental investigations of this problem have been carried out. Up to 2000, the experiments at supercritical pressure of water were conducted mainly with smooth round tubes, some data on pressure drops in annular channels were obtained, and it was known only one work in which the hydraulic resistance of a tight-lattice 7-rod bundle with helical fins was studied. Later, the similar investigations began intensively to develop in the People's Republic of China. On the whole, the experimental results, which have been published, are contradictory, and the correlations, based on these data, may be used in limited ranges of geometric and regime parameters. The conclusion was made that there is the urgent necessity to continue the investigations in the here examined field. The purpose of future researches must be the removal of existent contradiction in experimental results, receiving the new data on hydraulic resistance of channels simulating of the real fuel assemblies of SCWRs and the development of general correlations suitable for engineering calculations of hydrodynamic characteristics of the SCWR core in the range of operating parameters.

Pressurizer water level control with estimation of primary circuit coolant mass in nuclear power plants via robust observer based dynamic sliding mode control

Pressurizer is one of the most important components in a pressurized water nuclear power plant. Since it is responsible for coolant mass balance and non-boiling heat transfer in the primary circuit. The pressurizer water level control system in a nuclear power plant with the Pressurized Water Reactor (PWR) is responsible for the coolant mass balance. The main control goal is to stabilize the water level at a reference value and to suppress the effect of time-varying disturbances such as the coolant leakage in the primary circuit pipeline system. In the process of PWR power plant operation, the incorrect pressurizer water level may disturb pressure control or may cause damage to electric heaters which could threaten plant security and stability. In modern reactors, standard Proportional–Integral–Derivative (PID) controllers are used to control the water level in a pressurizer. In this paper, for the first time, in order to improve the performance and transient response of the pressurizer control system, an observer-based dynamic sliding mode control using Lyapunov-approach is designed and applied for the level control problem of the pressurizer in a pressurized water nuclear power plant based on the validated nonlinear model. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. The comparison between the Dynamic Sliding Mode Control (DSMC) and the PID controller shows a significant improvement in water set-point tracking and increased ability in disturbance rejection for the proposed observer-based DSMC. Simulation results confirm the improvement in the transient response obtained by using the proposed controller.

A new generation of uranium coincidence fast neutron collars for assay of LWR fresh fuel assemblies

The active uranium neutron coincidence collar provides a means of non-destructively assaying the fissile linear density of Light Water Reactor fresh fuel assemblies containing low enriched uranium. These neutron collars can operate in two modes: a thermal and a fast mode. In fast mode, a neutron collar has an added cadmium (Cd) liner in the sample cavity of the detector to reduce the impact of the burnable poison (thermal neutron absorber) on the detector signal (doubles). The main advantage for operating in fast mode is a detected signal that is less dependent of the burnable neutron poison content and thus less dependent on facility operator declarations. The drawback is that operating in fast mode requires a longer measurement time (∼hour vs tens of minutes for thermal mode) to achieve the statistically needed precision in the measurements. The trend in the modern reactor fuel assemblies is moving to higher burnup by using higher initial enrichment and, consequently, a higher number of burnable poison rods to compensate the initial neutron reactivity. The increase of the burnable poison loading has motivated the development of a new generation of high efficiency fast neutron collars to allow practical measurements in-field by nuclear inspectors. This paper describes the development and performance evaluation of a new generation of neutron collars, for both boiling water reactor (BWR) and pressurized water reactor (PWR) fuels, jointly developed between the US Department of Energy, through Los Alamos National Laboratory, and the Euratom Safeguards Directorate of the European Commission. We present here calibrations with reference fuel assemblies at Los Alamos National Laboratory as well as the results of in-field measurement campaigns in fuel fabrication plants with modern commercial fuel assemblies. The experimental results show that a typical PWR verification can be made in a total time of 30 min with an uncertainty in the measured mass of 2% at one standard deviation (1σ). A BWR verification can be made in 47 min with an uncertainty in the measured mass of 1.9% at 1σ, or a total time of 20 min with 1σ uncertainty in the measured mass of 2.5%.

Updating of dry shielding of nuclear power plant reactor vessel

Introduction. Dry shielding is a cylindrical structure made of serpentinite concrete in a metal casing with an inner diameter of 5.6 m, an outer diameter of 6.7 m, and a height of 5.3 m, which surrounds the VVER reactor vessel in the vicinity of the core. The purpose of serpentinite concrete, containing an increased amount of chemically bound water, is to soften the spectrum of the neutron flux outside the reactor, increasing the fraction of thermal neutrons in the spectrum, which is necessary for the operation of ionization chambers (IR) of the reactor control and protection system. Dry shielding also performs the functions of radiation and thermal protection, reducing the flux of radiation on ordinary concrete of biological protection. Before the installation of the dry shielding in the reactor shaft, heat treatment (drying) of concrete is carried out at temperatures up to 250 °С to remove unbound water in order to avoid radiolysis. Quality control of concreting and then heat treatment is carried out using a radioisotope device — a neutron moisture meter. These works are very lengthy and costly. Materials and methods. The design of the dry shielding casing was considered in order to perform additional perforation in order to avoid the formation of air pockets during concreting. The possibility of using modern plasticizing additives was considered in order to minimize the consumption of mixing water and, as a result, free water in the body of serpentinite concrete. Results. The possibility of exclusion of the stages of quality control of concreting and heat treatment in their traditional form is shown. Additional perforation of the metal casing, its internal diaphragms in problem areas, the use of a mixture of 20 cm slump or more allows you to completely eliminate the formation of internal voids. According to preliminary estimates, given the intensity of radiation in the NW for a modern reactor with a capacity of 1200 MW, the intensity of the release of hydrogen outside the shell due to radiolysis does not pose any danger. The concentration of hydrogen in the air surrounding the dry shielding is many orders less of magnitude than the dangerous 4 %. Conclusions. The cost of work on the construction of the SZ power unit of a nuclear power plant with a capacity of 1000–1200 MW can be reduced by 70–100 million rubles, the duration of work by 5 months.

DESIGN OF NEUTRONIC PARAMETERS OF MTR REACTOR USING WIMSD-5B/BATAN-FUEL CODES

BATAN has three aging research reactors, so it is necessary to design a new, more modern MTR type reactor using high-density, low enrichment uranium molybdenum fuel. The thermal neutron flux at the irradiation position is an important concern in the design of research reactors. This analysis is performed using standard computer codes WIMSD-5B and Batan-FUEL. The purpose of this study is to analyze the effect of the core configuration with safety control rods and neutronic parameters using the diffusion method calculation. The reactor core consists of 16 fuel elements and four control rods placed in the 5 x 5 position of the grid plate and is loaded the reflector elements outside the core. The cycle length is also a concern, not less than 20 days, and the reactor can be operated safely with a power of 50 MW. The calculation results show that for the highest fuel loading, which is 450 grams of U7Mo/Al fuel with D2O as a reflector, it will provide the lowest thermal neutron flux at the center of the core irradiation position, namely 1.0 x1015 n/cm2s. The core fuel cycle length will be up to 39 days, meeting the expected acceptance and safety criteria.

Phonon Scattering and Thermal Conductivity of Actinide Oxides with Defects

In the present study, we examine the effect of point defects and fission gases on thermal transport in representative actinide oxides used in modern reactors. In particular, oxygen interstitials and Kr/Xe fission gas bubbles are of primary focus. Reverse non-equilibrium molecular dynamics is employed to investigate thermal transport in UO2 and PuO2 with oxygen interstitials at the defect concentrations of 0.1%, 1%, and 5%. The results show that any alteration to the lattice structures of these fuels reduce their thermal conductivities significantly. For the largest UO2 structure simulated in the present study, for example, 0.1% oxygen interstitials decreased the thermal conductivity by 18.6%. For the case of the effect of fission gas bubbles, serious modification to phonon dispersion in oxide fuels is caused by the presence of a single fission gas bubble, resulting in a large temperature drop in their temperature profiles. The average interfacial thermal resistance across a fission gas bubble (comprised of 30 Kr and/or Xe atoms) is estimated to be 2.1 × 10−9 Km2/W.