Spark plasma sintering of tungsten-based WTaVCr refractory high entropy alloys for nuclear fusion applications

W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a powder metallurgy process combining mechanical alloying and spark plasma sintering (SPS). The SPSed samples contained two phases, in which the matrix is RHEA with a body-centered cubic structure, while the oxide phase was most likely Ta2VO6through a combined analysis of X-ray diffraction (XRD),energy-dispersive spectroscopy (EDS), and selected area electron diffraction (SAED). The higher oxygen affinity of Ta and V may explain the preferential formation of their oxide phases based on thermodynamic calculations. Electron backscatter diffraction (EBSD) revealed an average grain size of 6.2μm. WTaVCr RHEA showed a peak compressive strength of 2997 MPa at room temperature and much higher micro-and nano-hardness than W and other W-based RHEAs in the literature. Their high Rockwell hardness can be retained to at least 1000°C.

Shielding and corrosion properties of the Alloy 709 as canister material for spent nuclear fuel dry casks

The shielding and corrosion properties of the Alloy 709 advanced austenitic stainless steel have been investigated as a candidate canister material in spent fuel dry casks.The results revealed that the experimental and computational data of the linear and mass attenuation coefficients of the alloy are in good agreement,in which the attenuation coefficient values decreased with increasing photon energy.Alloy 709 was shown to exhibit the highest linear attenuation coefficient against gamma rays when compared to 304 and 316 stainless steels.On the other hand,Alloy 709 exhibited no considerable weight change over a 69-day period in circulating salt brines corrosion testing,while it showed an exponential increase of corrosion current density with temperature in acidic and basic corrosive solutions during electrochemical polarization corrosion testing.Furthermore,Alloy 709 was the least corroded steel compared to other austenitic stainless steels in both acidic and basic solutions.The optimistic results of the shielding and corrosion properties of Alloy 709 due to its chemical composition,suggest utilizing it as a canister material in spent nuclear fuel dry casks.

Control system design for a pressure-tube-type supercritical water-cooled nuclear reactor via a higher order sliding mode method

Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor(SCWR)is a fourth-generation conceptual reactor.In an SCWR,the non-linear dynamics of the reactor require a controller capable of control-ling the nonlinearities.In this study,a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode,and the reactor outgoing steam temperature and pressure were controlled simultaneously.In an SCWR,the temperature,pressure,and power must be maintained at a setpoint(desired value)during power maneuvering.Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation.Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers speci-fied in previous studies.The controlled parameters were reactor power,steam temperature,and pressure.Notably,for these parameters,the PI controller had certain instabilities in the presence of disturbances.The classic sliding mode controller had a higher accuracy and stability;however its main drawback was the chattering phenomenon.HOSMC was highly accurate and stable and had a small computational cost.In reality,it followed the desired values without oscillations and chattering.

Experimental investigation on effective aerosol scavenging using different spray configurations with pre-injection of water mist for Fukushima Daiichi decommissioning

During the decommissioning of the Fukushima Daiichi nuclear power plant,it is important to consider the retrieval of resolidified debris both in air and underwater configurations.For the subsequent retrieval of debris from the reactor building,the resolidified debris must be cut into smaller pieces using various cutting methods.During the cutting process,aerosol particles are expected to be generated at the submicron scale.It has been noted that such aerosols sizing within the Greenfield gap(0.1-1μm)are difficult to remove effectively using traditional spraying methods.Therefore,to improve the aerosol removal efficiency of the spray system,a new aerosol agglomeration method was recently proposed,which involves injecting water mist to enlarge the sizes of the aerosol particles before removing them using water sprays.In this study,a series of experiments were performed to clarify the proper spray configurations for effective aerosol scavenging and to improve the performance of the water mist.The experimental results showed that the spray flow rate and droplet characteristics are important factors for the aerosol-scavenging efficiency and performance of the water mist.The results obtained from this study will be helpful for the optimization of the spray system design for effective aerosol scavenging during the decommissioning of the Fukushima Daiichi plant.

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.

Investigating the Effects of Injection Pipe Orientation on Mixing and Heat Transfer for Fluid Flow Downstream a T-Junction

At T-junctions, where hot and cold streams flowing in pipes join and mix, significant temperature fluctuations can be created in very close neighborhood of the pipe walls. The wall temperature fluctuations cause cyclical thermal stresses which may induce fatigue cracking. Temperature fluctuation is of crucial importance in many engineering applications and especially in nuclear power plants. This is because the phenomenon leads to thermal fatigue and might subsequently result in failure of structural material. Therefore, the effects of temperature fluctuation in piping structure at mixing junctions in nuclear power systems cannot be neglected. In nuclear power plant, piping structure is exposed to unavoidable temperature differences in a bid to maintain plant operational capacity. Tightly coupled to temperature fluctuation is flow turbulence, which has attracted extensive attention and has been investigated worldwide since several decades. The focus of this study is to investigate the effects of injection pipe orientation on flow mixing and temperature fluctuation for fluid flow downstream a T-junction. Computational fluid dynamics (CFD) approach was applied using STAR CCM+ code. Four inclination angles including 0 (90), 15, 30 and 45 degrees were studied and the mixing intensity and effective mixing zone were investigated. K-omega SST turbulence model was adopted for the simulations. Results of the analysis suggest that, effective mixing of cold and hot fluid which leads to reduced and uniform temperature field at the pipe wall boundary, is achieved at 0 (90) degree inclination of the branch pipe and hence may lower thermal stress levels in the structural material of the pipe. Turbulence mixing, pressure drop and velocity distribution were also found to be more appreciable at 0 (90) degree inclination angle of the branch pipe relative to the other orientations studied.

Loss of offsite power (LOOP) accident analysis by integration of deterministic and probabilistic approaches in Bushehr-1 VVER-1000/V446 nuclear power plant

The results of an accident analysis for the loss of offsite power(LOOP)scenario in a reference Bushehr-1 VVER-1000/V446 nuclear power plant(NPP)are presented in this paper.This study attempted to provide a better analysis of LOOP accident management by integrating deterministic and probabilistic approaches.The RELAP5 code was used to investigate the occurrence of specific thermal–hydraulic phenomena.The probabilistic safety assessment of the LOOP accident is presented using the SAPHIRE software.LOOP accident data were extracted from the Bushehr NPP final safety analysis reports and probabilistic safety analysis reports.A deterministic approach was used to reduce the core damage frequency in the probabilistic analysis of LOOP accidents.The probabilistic approach was used to better observe the philosophy of defense in depth and safety margins in the deterministic analysis of the LOOP accident.The results show that the integration of the two approaches in LOOP accident investigations improved accident control.

Numerical analysis on element creation by nuclear transmutation of fission products

A burnup calculation was performed to analyze the Apr`es ORIENT process, which aims to create highlyvaluable elements from fission products separated from spent nuclear fuels. The basic idea is to use nuclear transmutation induced by a neutron capture reaction followed by a β-decay, thus changing the atomic number Z of a target element in fission products by 1 unit. LWR(PWR) and FBR(MONJU) were considered as the transmutation devices. High rates of creation were obtained in some cases of platinum group metals(44Ru by FBR,46 Pd by LWR) and rare earth(64Gd by LWR,66 Dy by FBR). Therefore, systems based on LWR and FBR have their own advantages depending on target elements. Furthermore, it was found that creation rates of even Z(= Z + 1) elements from odd Z ones were higher than the opposite cases. This creation rate of an element was interpreted in terms of "average 1-group neutron capture cross section of the corresponding target element σc Z defined in this work. General trends of the creation rate of an even(odd) Z element from the corresponding odd(even) Z one were found to be proportional to the 0.78th(0.63th) power of σc Z, however with noticeable dispersion. The difference in the powers in the above analysis was explained by the difference in the number of stable isotopes caused by the even-odd effect of Z.

Molecular dynamics study of interactions between edge dislocation and irradiation-induced defects in Fe–10Ni–20Cr alloy

Irradiation-induced defects frequently impede the slip of dislocations, resulting in a sharp decline in the performance of nuclear reactor structural materials, particularly core structural materials. In the present work, molecular dynamics method is used to investigate the interactions between edge dislocations and three typical irradiation-induced defects(void,Frank loop, and stacking fault tetrahedron) with the sizes of 3 nm, 5 nm, and 7 nm at different temperatures in Fe–10Ni–20Cr alloy. The critical resolved shear stresses(CRSSs) are compared among different defect types after interacting with edge dislocations. The results show that the CRSS decreases with temperature increasing and defect size decreasing for each defect type during the interaction with edge dislocations, except for the case of 3-nm Frank loops at 900 K. According to a comparison, the CRSS in Frank loop is significantly higher than that of others of the same size, which is due to the occurrence of unfaulting and formation of superjog or stacking-fault complex during the interaction. The atomic evolution of irradiation-induced defects after interacting with dislocations can provide a novel insight into the design of new structural materials.

Pitting corrosion behavior and corrosion protection performance of cold sprayed double layered noble barrier coating on magnesium-based alloy in chloride containing solutions

Nitrogen processed, cold sprayed commercially pure(CP)-Al coatings on Mg-based alloys mostly lack acceptable hardness, wear resistance and most importantly are highly susceptible to localized corrosion in chloride containing solutions. In this research, commercially pure α-Ti top coating having good pitting potential(~1293 mV_(SCE)), high microhardness(HV_(0.025): 263.03) and low wear rate was applied on a CP-Al coated Mg-based alloy using high pressure cold spray technology. Potentiodynamic polarization(PDP) curves indicated that the probability of transition from metastable pits to the stable pits for cold spayed(CS) Al coating is considerably higher compared to that with the CS Ti top coating(for Ti/Al/Mg system). In addition, CS Ti top coating was in the passivation region in most pH ranges even after 48 h immersion in 3.5 wt% NaCl solution. The stored energy in the CS Ti top coating(as a passive metal) was presumed to be responsible for the easy passivation. Immersion tests indicated no obvious pits formation on the intact CS Ti top coating surface and revealed effective corrosion protection performance of the CS double layered noble barrier coatings on Mg alloys in 3.5 wt% NaCl solution even after 264 h.

Comparison of Small Modular Reactor and Large Nuclear Reactor Fuel Cost

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. They are factory built, requiring smaller initial capital investment and facilitating shorter construction times. SMRs also promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are public parameters for many near-term SMR projects. The fuel cost of electricity generation for selected SMRs and large reactors is simulated, including calculation of optimal tails assay in the uranium enrichment process. The results are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 70% higher fuel costs than large light water reactors using 2014 nuclear fuels market data. Fuel cost sensitivities to reactor design parameters are presented.

Assessment of Sustainable Energy Strategy with Long-Term Global Energy Model Incorporating Nuclear Fuel Cycle

This paper investigates long-term energy strategy compatible with significant reduction of world carbon dioxide （CO2） emissions, employing a long-term global energy model, Dynamic New Earth 21 （called DNE21）. The model seeks the optimal energy mix from 2000 to 2100 that minimizes the world total energy system cost under various kinds of energy and technological constraints, such as energy resource constraints, energy supply and demand balance constraints, and CO2 emissions constraints. This paper discusses the results of primary energy supply, power generation mix, CO2 emission, CCS （carbon capture and storage） and total system costs for six regions including world as a whole. To evaluate viable pathways forward for implementation of sustainable energy strategies, nuclear power generation is a viable source of clean and green energy to mitigate the CO2 emissions. Present research shows simulation results in two cases consisting of no CO2 regulation case （base case） and CO2 REG case （regulation case） which halves the world CO2 emissions by the year 2050. Main findings of this research describe that renewable and nuclear power generation will contribute significantly to mitigate the CO2 emission worldwide.

Influence of the electrolyte conductivity on the critical current density and the breakdown voltage

The work investigates influence of the electrolyte conductivity on the onset of partial contact glow discharge electrolysis(CGDE)in a water electrolysis.Critical current density(CCD)and breakdown voltage were measured together with in situ observation of hydrogen bubble behavior,whose influence has not been focused on.For a fixed current during normal electrolysis,hydrogen coalescence adjacent to cathode surface was invigorated at a lower conductivity.Photographic analyses elucidated the hydrogen coalescence characteristics by quantifying size and population of detached hydrogen bubbles.The CCD increased about 104% within given range of conductivity(11.50-127.48 mS·cm^(-1))due to impaired bubble coalescence,which delays hydrogen film formation on the cathode.Meanwhile,decreasing trend of breakdown voltage was measured with increased conductivity showing maximum drop of 74%.It is concluded that onset of partial CGDE is directly affected by hydrodynamic bubble behaviors,whereas the electrolyte conductivity affects the bubble formation characteristics adjacent to cathode electrode.

Dynamic scaling characteristics of single-phase natural circulation based on different strain transformations

To understand the dynamical system scaling(DSS)analysis theory,the applicability of DSSβ-andω-strain transformation methods for the scaling analysis of complex loops was explored.A simplified model consisting of two loops was established based on the primary and secondary sides of a nuclear reactor,andβ-andω-strain transformation methods were used to ana-lyze the single-phase natural circulation in the primary circuit.For comparison with the traditional method,simplified DSSβ-andω-strain methods were developed based on the standard scaling criterion.The strain parameters in these four methods were modified to form multiple groups of scaled-down cases.The transient process of the natural circulation was simulated using the Relap5 code,and the variation in the dynamic flow characteristics with the strain numbers was obtained using different scaling methods.The results show that both the simplified and standard DSS methods can simulate the dynamic characteristics of natural circulation in the primary circuit.The scaled-down cases in the simplified method exhibit the same geometric scaling and correspond to small core power ratios.By contrast,different scaled-down cases in the standard DSS method correspond to different geometric scaling criteria and require more power.The dynamic process of natural circula-tion can be simulated more accurately using the standard DSS method.

Numerical Analysis of Heating Technique in Corium Melt Pool Convection Flow Field &Thermal Interaction in a Volumetrically Heated Molten Pool

In-Vessel Retention (IVR) is one of the existing strategies of severe accident management of LWR, which intends to stabilize and isolate corium & fission products inside the reactor pressure vessel (RPV) and primary containment structure. Since it has become an important safety objective for nuclear reactors, it is therefore needed to model and evaluate relevant phenomena of IVR strategy in assessing safety of nuclear power reactors. One of the relevant phenomena during accident progression in the oxidic pool is non-uniform high heat generation occurring at large scale. Consequently, direct experimental studies at these scales are not possible. The role computer codes and models are therefore important in order to transpose experimental results to reactor safety applications. In this paper, the state-of-the-art ANSYS FLUENT CFD code is used to simulate Non-uniform heat generation in the lower plenum by the application of Cartridge heating under severe accident conditions to derive the basic accident scenario. However, very few studies have been performed to simulate non-uniform decay heat generation by Cartridge heaters in a pool corresponding lower plenum of power reactor. The current investigation focuses on non-uniform heating in the fluid domain by Cartridge heaters, which has been done using ANSYS FLUENT CFD code by K-epsilon model. The computed results are based on qualitative assessment in the form of temperature and velocity contour and quantitative assessment in terms of temperature and heat flux distribution to assess the impact of heating method on natural convective fluid flow and heat transfer.

Assessment of Radiation Hazard from External and Internal Exposures at Adham and Surroundings in KSA

Twenty-eight environmental samples (eight well water, sixteen granitic rocks and four soils) were collected from different parts of Adham governorate (Adham, Haqal and Al-Jaizah), to assess the radiological hazard and cancer risk from different perspectives. Adham is situated in a valley between two granitic mountain chains, where much of water supply for drinking, house use and irrigation comes from wells collecting water rains. The activity concentrations of naturally occurring 40K, 226Ra and 232Th and radionuclides were measured by gamma-ray spectrometry for all samples using RGK-1, RGU-1 and RGTh-1, IAEA reference standards issued by the International Atomic Energy Agency, for detector efficiency calibration. The measured values were utilized to evaluate the internal and external exposures both outdoors and indoors. Different standard room models were adopted for this respect to evaluate the indoor gamma-rays exposure from construction materials as well as internal exposure to radon gas emanating from them. Radon concentration indoors, exceeded the upper reference level in dwellings set at 300 Bq/m3 by the world health organization, in many scenarios. The mean value of the total excess lifetime cancer risk (due to external exposure from gamma-rays) was 2.29 × 10-3, above the world average value of 1.45 × 10-3. Furthermore, the measured radon concentrations in all water samples exceeded the EPA (Environmental Protection Agency) 11.1 Bq·L-1 standard for drinking water, ranging from 12 to 38 Bq·L-1 with a mean value of 27 Bq·L-1. The total annual effective dose (due to inhalation and ingestion) from radon in water, ranged from 58 to 192 μSv/y (for adults) exceeding the international permissible limit of 100 μSv/y, in seven out of eight samples. According to obtained results, the internal exposure from radon in directly used water from wells, might be the major reason of any suspected radiological health hazard especially in Haqal. The second reason might be the internal exposure from indoor radon gas inhalation in poorly ventilated dwellings.

System Variables Design of Safety Analysis for Fast Reactors

This research aims to examine the risk in the technology design of fast breeder reactors while the development depends on safety considerations. The project explored the variables, which could affect positively the expected average fuel burn-up, breeding ratio, and decay heat removal. That is accomplished using features such as guard vessels and elevated pipe routing to prevent the cracked state of both core components and fuel cladding interface conditions. So, the cracked region of fuel was detected by thermal-hydraulic analysis. We used ZrFeCr alloys to estimating of the rise in fuel cladding and coolant that can be incorporated in the design ZrFeCr alloys to uniform corrosion in temperature and 10.3 Mpa pressure. Fast creep of the reactor vessel during the coolant heat-up transient is another issue to be considered corrosion resistance of structural material can be achieved by controlling oxygen content in steel alloy. In this trend, S4337 S5140 steels are wide and can be used in future fossil power plants because of their excellent high-temperature strength.

Decomposition mechanisms of nuclear-grade cationic exchange resin by advanced oxidation processes:Statistical molecular fragmentation model and DFT calculations

The treatment and disposal of radioactive waste are presently facing great challenges.Spent ion exchange resins have become a focus of attention due to their high production and serious environmental risks.In this paper,a simplified model of cationic exchange resin is proposed,and the degradation processes of cationic resin monomer initiated by hydroxyl radicals(·OH)are clarified by combining statistical molecular fragmentation(SMF)model and density functional theory(DFT)calculations.The prediction of active sites indicates that the S-O bonds and the C-S bond of the sulfonic group are more likely to react during the degradation.The meta-position of the sulfonic group on the benzene ring is the most active site,and the benzene ring without the sulfonic group has a certain reactivity.The C11-C14 and C17-C20 bonds,on the carbon skeleton,are the most easily broken.It is also found that dihydroxy addition and elimination reactions play a major role in the process of desulfonation,carbon skeleton cleavage and benzene ring separation.The decomposition mechanisms found through the combination of physical models and chemical calculations,provide theoretical guidance for the treatment of complex polycyclic aromatic hydrocarbons.

Gradient Optimizer Algorithm with Hybrid Deep Learning Based Failure Detection and Classification in the Industrial Environment

Failure detection is an essential task in industrial systems for preventing costly downtime and ensuring the seamlessoperation of the system. Current industrial processes are getting smarter with the emergence of Industry 4.0.Specifically, various modernized industrial processes have been equipped with quite a few sensors to collectprocess-based data to find faults arising or prevailing in processes along with monitoring the status of processes.Fault diagnosis of rotating machines serves a main role in the engineering field and industrial production. Dueto the disadvantages of existing fault, diagnosis approaches, which greatly depend on professional experienceand human knowledge, intellectual fault diagnosis based on deep learning (DL) has attracted the researcher’sinterest. DL reaches the desired fault classification and automatic feature learning. Therefore, this article designs a Gradient Optimizer Algorithm with Hybrid Deep Learning-based Failure Detection and Classification (GOAHDLFDC)in the industrial environment. The presented GOAHDL-FDC technique initially applies continuous wavelettransform (CWT) for preprocessing the actual vibrational signals of the rotating machinery. Next, the residualnetwork (ResNet18) model was exploited for the extraction of features from the vibration signals which are thenfed into theHDLmodel for automated fault detection. Finally, theGOA-based hyperparameter tuning is performedtoadjust the parameter valuesof theHDLmodel accurately.The experimental result analysis of the GOAHDL-FD Calgorithm takes place using a series of simulations and the experimentation outcomes highlight the better resultsof the GOAHDL-FDC technique under different aspects.

Numerical Simulation of Direct-contact Condensation from a Supersonic Steam Jet in Subcooled Water

The phenomenon of direct-contact condensation,used in steam driven jet injectors,nuclear reactor emergency core cooling systems and direct-contact heat exchangers,was investigated computationally by introducing a thermal equilibrium model for direct-contact condensation of steam in subcooled water.The condensation model presented was a two resistance model which takes care of the heat transfer process on both sides of the interface and uses a variable steam bubble diameter.The injection of supersonic steam jet in subcooled water tank was simulated using the Euler-Euler multiphase flow model of Fluent 6.3 code with the condensation model incorporated. The findings of the computational fluid dynamics（CFD） simulations were compared with the published experimental data and fairly good agreement was observed between the two,thus validating the condensation model.The results of CFD simulations for dimensionless penetration length of steam plume varies from 2.73-7.33,while the condensation heat transfer coefficient varies from 0.75-0.917 MW·（m ^2 ·K）^ -1 for water temperature in the range of 293-343 K.