Hasil untuk "Nuclear engineering. Atomic power"

Ye Sung Kim, Myung Jin Jeong, San Lee et al.

Heat Pipe-cooled Microreactors (HPMRs) are emerging as a sustainable solution for off-grid power. The multiphysics simulations are required for the safety analysis of HPMRs, where the modeling of transient phenomena within alkali metal heat pipes is important for cooling capacity. This study introduces SNUHTP, a transient heat pipe analysis code developed based on a thermal resistance network model. The melting/solidification effect of the working fluid, which influences the startup and shutdown behavior, was modeled through an effective heat capacity method. The code underwent transient verification against analytic solutions of the 1D Stefan problem and a lumped heat pipe model. Steady-state validation used sodium heat pipe experimental data, showing good agreement. Transient validations on a water heat pipe and the SAFE-30 sodium heat pipe module test were also conducted. Both cases matched the experiments well. Especially in the sodium heat pipe, the latent heat effect resulted in a delay in temperature rise. Additionally, an OpenFOAM-SNUHTP coupled calculation was performed on the sodium heat pipe for the realistic modeling of adjacent structures. The validations of SNUHTP demonstrated its capability to predict transient behavior in heat pipes, including frozen startup.

WU Weiming, LI Xin, LUO Hongyi

The accumulation of excessive internal pressure within the hermetic containment of α-decay radioisotope heater units (RHU) may cause cladding rupture and risk of radioactive material release. To address this safety issue, a fully-coupled model was developed which integrating radioactive decay, temperature attenuation, helium states, and cladding mechanical response. This model overcomes the limitation of conventional methods, which only consider the state change of helium. Taking light weight radioisotope heater unit (LWRHU) as an example, numerical simulations were performed to analyze cladding temperature and deformation under two typical scenarios: normal storage and accidental re-entry. The results show excellent agreement with RMSE and R2 values of (0.06, 1.001 9) and (0.09, 0.986 8) for the two scenarios, confirming the model’s accuracy in predicting the RHU temperature decay. As decay progresses and helium accumulates, the slope of p-V curve increases significantly, which imparts a strong volume-sensitive characteristic to the internal pressure. Based on this, a p-V-t evolution equation for helium and a pressure-bearing p-V characteristic curve of the cladding were established. Two methods for determining the internal pressure are proposed: the graphical calibration method, which estimates internal pressure by matching helium and cladding curves in p-V space, though with some subjectivity in selecting the intersection point; and the self-consistent iterative method, which achieves a self-consistent solution by establishing a coupled isothermal p-V state between the helium and the cladding. This method is particularly suitable for high-temperature conditions where the cladding undergoes significant nonlinear deformation, and provides high accuracy in approximating the actual internal pressure. The proposed model and methods provide a theoretical basis and practical tools for the design optimization, safety assessment, and storage management of α-decay RHU systems. This method also facilitates the development of computational software for estimating the internal pressure in α-decay RHUs, thereby demonstrating substantial potential for practical engineering applications.

Hirofumi Tosaki

This study examines the Kishida administration’s nuclear arms control and disarmament policies through the lens of Japan’s dual responsibilities: the normative responsibility to pursue a world without nuclear weapons and the strategic responsibility to ensure national security. While these responsibilities are not inherently contradictory, they have often posed dilemmas for Japan’s nuclear policies, particularly amid worsening global security conditions. The study categorizes the administration’s nuclear arms control and disarmament policies into four types: (1) those emphasizing strategic responsibility while also aligning with normative responsibility, (2) those prioritizing normative responsibility while converging with strategic responsibility, (3) efforts attempting to balance the two, and (4) policies where conflict between them necessitated prioritization of one over the other. The analysis draws on official statements, policy documents, and supplementary expert and media sources. The paper evaluates the administration’s policies based on their success in bridging the dual responsibilities, enhancing Japan’s security, and advancing global disarmament norms. It concludes that while the Kishida administration sought convergence where possible, when conflicts arose, it prioritized national security, leading to criticism from nuclear abolition proponents. The study highlights key challenges and implications for Japan’s future nuclear policy.

Jeseok Song, Chang-Lak Kim

This study compares how the change in decontamination scope affects the total decontamination factor (DF) in system decontamination of Kori Unit 1 (Kori-1), by development of decontamination factor evaluation program (DFEP). The DFEP calculates the total DF from initial activity inventory and its measurement date of each part in whole scope of decontamination, decontamination date and number of decontamination cycles, and DF of each cycle in each part. For every cycle, DF of each part is driven by DF of each cycle in whole system, and DF of each part in whole cycles, both from overseas full system decontamination (FSD) cases.By the analysis of DFEP, total DF cannot reach the Kori-1 target DF, 30, in case of excluding both SGs. In the case of excluding only 1 SG, total DF can be over 30 if additional 2 more cycles of decontamination with excluding 1 SG are done after 2-cycle FSD. Comparing which of the SG is excluded, the total DF is higher in the case of excluding the SG connected to pressurizer side.

Weiguang Zhao, Pei Yu, Xiaobo Zeng et al.

The Component Cooling System (CCS) and the Sea Water System (SWS) of HPR1000 are heat removal systems related to the safety of nuclear power plants. In both CCS and SWS, the configuration of pumps is closely associated with system design and energy consumption. This paper addresses the issue of high energy consumption and poor economic performance of the current configuration scheme that relies on a single large-capacity constant frequency pump. Based on a mathematical model for CCS and SWS that can be utilized for optimization calculations, we propose and validate an Improved Non-dominated Sorting Genetic Algorithm II (INSGA-II) with good algorithm performance, optimizing for system design costs and total operational costs as objectives, and provide a sensitivity analysis of relevant variables. The results demonstrate that compared to the prototype values of HPR1000, the investment costs of CCS and SWS can be reduced by up to 4.65 %, and the total operational costs can be decreased by as much as 63.6 %, with the optimization effect being most significant when variable frequency pumps are used in CCS and SWS.

Róbert Farkas, Zoltán Hózer, Zoltán Kis et al.

A loss-of-coolant accident in a nuclear power plant was simulated in the CODEX-SLIM experiment. The electrically heated fuel rod bundle contained cladding material used at a VVER-440 power plant. The maximum temperature reached 900 °C and the experiment was terminated by water quenching. The condition of the bundle was investigated using non-destructive and destructive techniques. According to the experimental data, the fuel rods would maintain their integrity during a design basis loss-of-coolant accident.

ZHANG Ye, LI Xiang, WANG Wen-tao

Plutonium(Pu) is a very critical radioactive element, and its separation and purification are the core technology in the reprocessing of spent fuel. At the same time, research on the separation of Pu is also of great significance to environmental protection and radioactive pollution remediation. Currently, precipitation, extraction, and adsorption are the primary techniques for separating and purifying plutonium. Of these, adsorption is a crucial method for the process. The technique offers benefits such as simplicity of operation and high-speed adsorption. Additionally, adsorbents have advantages in the form of low cost and high adsorption capacity. In the review, according to the properties of adsorption materials, we introduced the research progress on the adsorption and separation of Pu from five aspects: natural mineral, polymer resins, inorganic oxides, carbon materials and nanomaterials. Furthermore, the adsorption ability and mechanisms of these different adsorption materials as well as the influencing factors during the adsorption process were also summarized and discussed. Based on these discussions, the advantages and disadvantages of these different materials were also evaluated. Finally, the problems and future development trend of the plutonium adsorption and separation were analyzed and prospected, which should be provided guidance for the research and application in the adsorption and separation of Pu and other radionuclides.

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

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

LIU Huanling, WANG Shengli, SU Haijun et al.

BackgroundThe 1.2 MeV/10 mA electron accelerator, as one of electron irradiation sources for comprehensive irradiation test chamber in Space Environment Simulation and Research Infrastructure (SESRI), can provide electron beams on the millimeter scale. However, the electron beam, as the electron irradiation source in space environment ground simulation experiment for aerospace, must be uniformly irradiated on large objects. Therefore, a well-suited irradiated technique is significant.PurposeThis study aims to obtain beam scanning with less than 10% of inhomogeneity for 0.6 MeV, 1.0 MeV, 1.2 MeV electron beam for SESRI.MethodsBased on the overall irradiation requirements, a specific beam-scanning system, including the scanning magnet with customized design, digital power supply and a dedicated apparatus for beam uniformity measurement, was developed. Particularly, in order to eliminate the influence of 45° incidence of electron beam upon scanning uniformity, an asymmetric and non-standard triangular waveform for the magnetic field excitation current was employed and implemented. The measurements for the beam non-uniformity were carried out on field experiments.ResultsExperimental results show that the scanning area of this electron accelerator reaches 1 000 mm×1 000 mm, and the scanning nonuniformity is less than 10% for variable beam energy from 0.6 MeV to 1.2 MeV, achieving the design goal and satisfying the irradiation requirements of SESRI.ConclusionsA specific beam-scanning system developed for SESRI is verified in this study, offering a good reference for any similar beam-scanning scenarios.

Yong Han, Ling Zhou, Ling Bai et al.

The ball valve is an important device in the pipeline transportation system of nuclear power plants. Its operational stability and safety directly affect the normal working of nuclear power plants. In this study, the transient numerical simulation of the opening and closing process of a ball valve was conducted on the basis of the flow interruption capability experiment of the ball valve by using the moving mesh method and inlet and outlet variable boundary conditions. The flow rate and pressure difference with time of the opening and closing process of the ball valve were studied. The internal flow characteristics of the ball valve under different relative openings were analyzed in conjunction with the typical back-step flow structure. Results show that the transient numerical results agree well with the experimental results. The internal flow characteristics of the ball valve are similar at the same opening during opening and closing process. At small opening, the spool and outlet channels easily form a back-step flow structure. The disappearance and generation of backflow vortices during opening and closing occur at 85% opening and 75% opening, respectively. With the decrease in opening degree, the difference in vortex core area in the flow channel of the ball valve spool in the opening and closing process gradually appears. The research results provide some reference value for the design and optimization of ball valves.

H.Q. Wang, X. Ma, R. Maurizio et al.

Dedicated experiments in DIII-D find that magnetic shaping and divertor target geometry significantly affect the divertor plasma conditions and divertor detachment process in the small-angle-slot (SAS) divertor. The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core–edge integration solutions. A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher electron temperatures near the divertor target plate and requires higher upstream densities to achieve the same level of divertor detachment. In addition, with the ion B × ∇B drift away from the SAS divertor and the outer strike point (OSP) near the outer corner, the target temperature is lower for a particular upstream density than with the OSP on a slanted or flat surface, leading to lower heat flux even when the particle flux remains similar. In contrast, with the ion B × ∇B drift into the SAS divertor, a strike point at the inner slanted surface exhibits a lower upstream density to achieve divertor detachment than a strike point either at the outer corner or the outer slanted target. Experimental results and SOLPS-ITER simulations with full drifts suggest the strong interplay between drift flows and the neutral distribution resulted from target shaping. Furthermore, in-slot gas puffing has been shown to achieve global divertor detachment with an onset density about 10 % lower than that using main-chamber gas puffing when the outer strike point is placed at the inner slanted surface. Corresponding modelling reveals that the local gas puffing enhances the neutral ionization which potentially facilitates the achievement of divertor dissipation. However, such improvement diminishes when the strike point is at the outer corner, which also indicates the geometric dependence on divertor performance in the SAS divertor. Even with different strike point locations, complete divertor detachment with very low particle and heat fluxes at the divertor targets and a high confinement core with normalized energy confinement factor H98 > 1.0 can be simultanesouly achieved with the SAS divertor with ion B × ∇B drift into SAS divertor, demonstrating the benefit of a closed divertor for exploration of core–edge integration.

O. A. Gubeladze, A. R. Gubeladze

The article deals with one of the nuclear terrorism problem aspects, namely the attempts to implement “nuclear” blackmail by individual states. The possible destructive aircrafts impact to objects using atomic energy is considered. An express assessment of the guided air bomb strike result on the reinforced concrete object structure was carried out.

Sreenivasulu Sunkara, M. Sreenivasa Reddy

Gold is one of the valuable metals and an important asset class for investors. People in India are emotionally attached to gold. Thousands of tonnes of idle gold are lying with Indian temples, trusts, and individuals. Investors consider capital appreciation; interest income and safety are major factors that influence buying of gold [1][4][7]. India is one of the biggest importers of gold every year. The government of India has introduced a few gold-related schemes to reduce gold imports. Sovereign gold bond [2] scheme and Re-vamped gold deposit scheme are introduced in the year 2015 under the Swarna Bharath initiative [9]. The present study is an attempt to find the association between awareness of investors and demographic factors towards gold ETFs, gold bonds, and gold deposits. The results reveal that there is a significant association at a 5% significance level between awareness of investors and all demographic factors used in the study except the gender of the respondent.

Vicente Navarro, Laura Asensio, Heidar Gharbieh et al.

This work describes the development of a numerical module with a multiphysics structure to simulate the thermo-hydro-chemo-mechanical behavior of compacted bentonites. First, the conceptual model, based on the state-of-the-art formulation for clay-based engineered barriers in deep geological repositories, is described. Second, the advantages of multiphysics-based modules are highlighted. Then, the guidelines to develop a code using such tools are outlined, presenting an example of implementation. Finally, the simulation of three tests that illustrate the behavior of compacted bentonites assesses the scope of the developed code. The satisfactory results obtained, and the relative simplicity of implementation, show the opportunity of the modeling strategy proposed. Keywords: Deep geological repository, Clay-based barrier, Compacted bentonite, Multiphysics modeling, Thermo-hydro-chemo-mechanical behavior

Mikhail A. Belonosov, Vladimir L. Kishkin, Sergey A. Korolev

The article describes an automated verification method used for application software of control safety systems based on the TPTS-SB equipment. Verification is performed by comparing two mathematical models (oriented graphs): one obtained by processing the original design data, i.e., graphical functional diagrams, and the other formed by reversing the program code loaded from the controller. The vertices in both graphs are functional blocks of mathematical and logical operations; the edges are connections between them. The constructed mathematical models undergo a comparison, covering the vertices and edges of the graphs as well as the memory cells and values of constants. The equivalence of mathematical models proves the correspondence between the program code and the initial set of design functional diagrams. The proposed automated verification method makes it possible to prove that no distortion is introduced into the program during the process of converting graphical functional diagrams into the program code with its subsequent translation and loading into the controller. It is postulated that any distortions will be detected during the verification procedure, which is performed every time after loading the code into the controller. The solution provides an acceptable speed when large volumes of vector graphics stored in a relational database are processed, and makes it possible to visualize the verification results. The proposed method is implemented in the GET-R1 instrumentation tools for TPTS-SB and is used in designing and verifying the application software of the safety systems at the Belarusian NPP.

N. Yamashita, K. Omori, Y. Kimura et al.

Silicon carbide (SiC) and its composites are candidate materials for the blanket components and for the first wall in a fusion reactor. If the SiC is used without any armor materials for the first wall, it is exposed by helium (He) plasma as well as hydrogen plasma. Characteristic surface morphology changes are reported for various materials by He plasma exposure. Thus, we exposed SiC specimens to He or simultaneous deuterium (D) and He (D + He) plasma by various conditions and then observed surface morphology changes by SEM. As a result, needle-like structures and whiskers-like structures at the tip were formed in He plasma and D + He irradiation, while only needle-like structures were formed in D plasma. Therefore, it indicated that the effects of He were attributed to form whiskers-like structures. Although the structures are different among He plasma, simultaneous D + He plasma and D plasma irradiations, sputtering is considered to be a dominant process for the formation of the structure formation. However, the effects of He atoms in the structure could also be attributed to form whiskers-like structures. Keywords: Helium induced nanostructure, Morphology change, Silicon carbide

E. Wolfrum, M. Beurskens, M.G. Dunne et al.

Plasmas in machines with all metal plasma facing components have a lower Zeff, less radiation cooling in the scrape-off layer and divertor regions and are prone to impurity accumulation in the core. Higher gas puff and the seeding of low-Z impurities are applied to prevent impurity accumulation, to increase the frequency of edge localised modes and to cool the divertor. A lower power threshold for the transition from low-confinement mode to high confinement mode has been found in all metal wall machines when compared to carbon wall machines. The application of lithium before or during discharges can lead to ELM free H-modes. The seeding of high-Z impurities increases core radiation, reduces the power flux across the separatrix and, if applied in the right amount, does not lead to deterioration of the confinement. All these effects have in common that they can often be explained by the shape or position of the density profile. Not only the peakedness of the density profile in the core but also the position of the edge pressure gradient influences global confinement. It is shown how (i) ionisation in the pedestal region due to higher reflection of deuterium from high-Z walls, (ii) reduced recycling in consequence of lithium wall conditioning, (iii) the fostering of edge modes with lithium dropping, (iv) increased gas puff and (v) the cooling of the scrape-off layer by medium-Z impurities such as nitrogen affect the edge density profile. The consequence is a shift in the pressure profile relative to the separatrix, leading to improved pedestal stability of H-mode plasmas when the direction is inwards. Keywords: Plasma, Pedestal, Confinement, Wall material, Reflection, Lithium, Edge localised modes, Impurities, Stability

Masrukan Masrukan, Sungkono Sungkono, Yanlinastuti Yanlinastuti et al.

ABSTRAK PEMBUATAN SERBUK PADUAN U-6Zr DENGAN PENGKAYAAN URANIUM 19,75 % UNTUK BAHAN BAKAR REAKTOR RISET. Telah dilakukan pembuatan serbuk paduan U-6Zr dengan pengkayaan 19,75 % untuk bahan bakar reaktor riset. Pembuatan bahan bakar U-6Zr ini dalam rangka mencari bahan bakar baru yang mempunyai densitas tinggi untuk mengganti bahan bakar yang sudah ada U3Si2-Al. Tujuan dari percobaan ini untuk mengetahui sifat-sifat serbuk paduan U- 6Zr yang diperoleh dari proses hydriding-dehydriding sebagai kandidat bahan bakar reaktor riset. Serbuk yang diperoleh dari proses hydriding-dehydriding dikenai pengujian, diantaranya pungujian komposisi kimia, densitas, kandungan hidrogen, fasa dan sifat termal. Hasil pengujian komposisi kimia menunjukkan beberapa unsur seperti Al, Ca, Cu, dan Ni melebihi batas yang diijinkan dimana masing-masing unsur terdapat sebesar 202,21 ppm; 214,05 ppm; 61,25 ppm dan 134,53 ppm. Pada pengujian diperolah densitas serbuk U-6Zr sebesar 13,58 g/cm3 dan pada pengujian kandungan hidrogen sisa diperoleh kandungan hidrogen sebesar 0,16 %. Untuk pengujian fasa, diperoleh fasa αU dan δU, sedangkan pada pengujian sifat termal yakni transformasi temperatur terdapat dua puncak yakni puncak pertama terjadi pada temperatur 274 hingga 311 oC dan puncak kedua terjadi pada temperatur 493 hingga 527oC. Puncak pertama terjadi reaksi endotermik dengan menyerap panas sebesar ∆H = 6,23 cal/g tetapi tidak terbentuk fasa baru, sedangkan puncak kedua terjadi reaksi eksotermik dengan mengeluarkan panas sebesar ∆H = -9.34 cal/g dan terbentuk fasa αZr. Sementara itu, dari pengujian kapasitas panas pada temperatur 34 hingga 75 oC, terjadinya penurunan nilai kapasitas panas yang disertai dengan penyerapan panas. Pada temperatur yang lebih tinggi hingga temperatur 437oC nilai kapasitas panas menjadi lebih kecil disertai pengeluaran panas. Reaksi termokimia antara Zr dengan hidrogen sisa menunjukkan terbentuknya fasa αZr yang diindikasikan oleh reaksi eksotermik dengan mengeluarkan panas sebesar ∆H = 9,3449 cal/g. Dari hasil analisis dapat diketahui bahwa paduan U-6Zr tersebut dapat digunakan sebagai bahan bakar pengganti untuk reaktor riset. Kata kunci: Serbuk U-6Zr, pengkayaan U 19,75 %, bahan bakar, reaktor riset. ABSTRACT MAKING OF U-6Zr ALLOY POWDER WITH URANIUM ENRICHMENT OF 19.75 % FOR RESEARCH REACTOR FUEL. Making U-6Zr alloy powder with enrichment of 19.75 % for a research reactor fuel has been done. Making of U-6Zr fuel in order to find new fuels that have a high density to replace the existing fuel U3Si2-Al. The purpose of this experiment was to determine the properties of U-6Zr alloy powder obtained from the hydriding-dehydriding process as a candidate research reactor fuel. Initially was made U-6Zr ingot by melting U and Zr metals using electric arc melting furnace. U-6Zr ingots were found then cut into pieces and put into hydriding-dehydriding equipment that operates at pressures 14,46054 psi and hydriding temperatures of 350 oC. Ingots that out of the hydriding-dehydriding equipment becomes brittle then was crushed so becomes powder. Powder that obtained subjected to the test, including chemical composition testing, density, hydrogen content, phase and thermal properties. The results testing show that chemical composition of some elements such as Al, Ca, Cu, and Ni exceeded the permitted limit, each of which contained elements of 202.21 ppm, 214.05 ppm, 61.25 ppm, and 134.53 ppm. On testing the density obtained that U-6Zr powder density of 13.57 g/cc and the testing of hydrogen residual content obtained hydrogen residual content of 0.16 %. For the testing phase were obtained the αU and δU phases, while in testing of the transformation temperature, there are two peaks, the first peak occurs at a temperature of 274 to 311 °C and a second peak occurs at a temperature of 493 to 527 oC. The first peak occurs endothermic reaction by absorbing heat of 6.23 cal/g but not formed a new phase, while the second peak occurs exotermic reaction with brought out heat of -9.34 cal/g and formed αZr phase. Meanwhile, the heat capacity of the test at temperature of 34 to 75 °C, the decrease in the value of the heat capacity accompanied heat absorption. At higher temperatures up to 437 °C, heat capacity value becomes negative and accompanied by heat expenditure. It can be concluded that the alloy the U-6Zr when viewed from the chemical composition can still be used for fuel research reactor. Residual hydrogen still found in small quantities in the U-6Zr powder that can be eliminated by heating for longer. Reaction between Zr with residual hydrogen to form the αZ phase is accompanied exothermic reaction with brought out heat of 93 449 cal/g. Keywords: U-6Zr powder, U enrichment of 19.75 %, fuel, research reactor.

YANG GYUN OH, JIN KWON JEONG, CHANG JAE LEE et al.

For the improvement of APR1400 Diverse Protection System (DPS) design, the Advanced DPS (ADPS) has recently been developed to enhance the fault tolerance capability of the system. Major fault masking features of the ADPS compared with the APR1400 DPS are the changes to the channel configuration and reactor trip actuation equipment. To minimize the fault occurrences within the ADPS, and to mitigate the consequences of common-cause failures (CCF) within the safety I&C systems, several fault avoidance design features have been applied in the ADPS. The fault avoidance design features include the changes to the system software classification, communication methods, equipment platform, MMI equipment, etc. In addition, the fault detection, location, containment, and recovery processes have been incorporated in the ADPS design. Therefore, it is expected that the ADPS can provide an enhanced fault tolerance capability against the possible faults within the system and its input/output equipment, and the CCF of safety systems.

Néstor Cornejo Díaz