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AbstractIn the long-term operation of nuclear power plants, the aging of systems, structures, and components can lead to maintenance issues that must be dealt with to maintain cost-effective plant operations. One common issue affecting the currently operated boiling water reactors is the onset of unexpected level oscillations in feedwater heaters. This phenomenon can cause excessive cycling of drain valves and lead to premature failures. In this work, we develop a dynamic model of a set of feedwater heaters to determine the root cause of oscillations observed in an operating plant. Simulation results of various transient scenarios were used to investigate the effects of the controller parameters, boundary conditions, and possible valve and instrument issues. The analysis led to the conclusion that the most likely causes of the observed self-sustained oscillations in the system are the nonlinear behaviors of the drain valve and the level transmitter induced by degraded equipment condition. A partial plug of the pressure line used for level sensing in the system can account for a significant deadtime in the level transmitter, a nonlinear effect shown to induce self-sustained oscillatory behaviors. [ABSTRACT FROM AUTHOR]

The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data. [ABSTRACT FROM AUTHOR]

To evaluate the effectiveness of the wet cavity strategy, the authors developed a stochastic evaluation method that considers the uncertainties of the molten material conditions ejected from reactor pressure vessels. This study analyzed the probability of ex‐vessel debris coolability under the wet cavity strategy. The first step was uncertainty analysis using the severe accident analysis code MELCOR to obtain the melt condition. Five uncertainty parameters related to the core degradation and transfer process were chosen. With the assumed probabilistic distributions, the input parameter sets were generated using the Latin hypercube sampling (LHS) method. Analyses were conducted, and the conditions of the melt were obtained. The second step was to analyze the melt behavior in the water and the spreading radius using the JASMINE code and to calculate the height of the debris on the floor. The probabilistic distribution of parameters for the JASMINE analyses was determined from the MELCOR analysis results. LHS generated 200 parameter sets. The depths of the water pool in the analysis were 0.5, 1.0, and 2.0 m. The debris height was compared with the criterion to judge its coolability. Consequently, the probability of successful debris cooling was obtained through the sequence of calculations. The feasibility and technical difficulties in the MELCOR‐JASMINE combined analysis were also discussed. [ABSTRACT FROM AUTHOR]

Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors. [ABSTRACT FROM AUTHOR]

The application of a two-phase ejector allows for the mixing of liquid and gas and provides effective heat transfer between phases. The aim of the study is a numerical investigation of the performance of a water-driven, condensing two-phase ejector. The research was performed using CFD methods, which can provide an opportunity to analyze this complex phenomenon in 2D or 3D. The 2D axisymmetric model was developed using CFD software Siemens StarCCM+ 2022.1.1. The Reynolds-Averaged Navier–Stokes (RANS) approach with the Realisable k-ε turbulence model was applied. The multiphase flow was calculated using the mixture model. The boiling/condensation model, where the condensation rate is limited by thermal diffusion, was applied to take into account direct contact condensation. Based on the mass balance calculations and developed pressure and steam volume fraction distributions, the ejector performance was analyzed for various boundary conditions. The influence of the suction pressure (range between 0.812 and 0.90) and the steam mass flow rate (range between 10 g/s and 25 g/s) is presented to investigate the steam condensation phenomenon inside the ejector condenser. The provided mixture of inert gas (CO2) with steam (H2O) in the ejector condenser was investigated also. The weakening of the steam condensation process by adding CO2 gas was observed, but it is still possible to achieve effective condensation despite the presence of inert gas. [ABSTRACT FROM AUTHOR]

After the Fukushima nuclear disaster, the nuclear materials community has been vastly investing in accident tolerant fuel (ATF) concepts to modify/replace Zircaloy cladding material. Iron–chromium–aluminum (FeCrAl) alloys are one of the leading contenders in this race. In this study, we investigated FA-SMT (or APMT-2), PM-C26M, and Fe17Cr5.5Al over a time period of 6 months in simulated BWR environments and compared their performance with standard Zirc-2 and SS316 materials. Our results implied that water chemistry along with alloy chemistry has a profound effect on the corrosion rate of FeCrAl alloys. Apart from SS316 and Zirc-2 tube specimens, all FeCrAl alloys showed a mass loss in hydrogen water chemistry (HWC). FA-SMT displayed minimal mass loss compared to PM-C26M and Fe17Cr5.5Al because of its higher Cr content. The mass gain of FeCrAl alloys in normal water chemistry (NWC) is significantly less when compared to Zirc-2. [ABSTRACT FROM AUTHOR]

Neutron flux modal decomposition is a key tool for analytical and reduced order modeling of boiling water reactor (BWR) stability and oscillations. As a minimum, the fundamental flux mode is used for representing global oscillations while the addition of at least one azimuthal harmonic is needed for simulating the regional out-of-phase mode. Unlike the fundamental and first azimuthal modes, the excitation of an axial flux mode alters the axial power shape but not the total power in the channel and therefore cannot be self-sustained when coupled to density wave–generated reactivity, presumably explaining why it has not been explicitly included in previously published models. Although not self-sustained, the axial mode excitation driven by density wave propagation and interactions with other spatial modes play important roles in interpreting observed BWR stability and oscillations particularly in the nonlinear regime when the oscillation magnitude is large. In this paper, the characteristics of the steady-state axial modes are presented, and their impact on oscillation dynamics for small and large amplitudes of both the global and the regional oscillations is studied using reduced order analytical tools. Aside from the oscillating component, our research results identify an average nonzero axial mode component to develop during limit cycle oscillations that causes the average axial power profile to shift toward the bottom of the core and thus contributes a negative reactivity component. The emergence of this nonzero average axial mode component and the associated negative reactivity were found to diminish the power increase due to global mode power oscillations and contribute to nonlinear stabilization of regional oscillations. The physical interpretation of nonlinear power oscillations with the inclusion of the axial mode component resolves previously unexplained results obtained from high-fidelity numerical models. [ABSTRACT FROM AUTHOR]

The intricate internal structure of fuel rods results in a non-uniform mass distribution, making it imperative to employ analytical methods for accurate assessment. The study utilizes Euler beam theory to derive the transverse vibration equation for beams with varying mass distribution. The approach involves transforming the non-uniform mass beam into a multi-segment beam with concentrated mass points. Modal function relationships between adjacent uniform segments are established based on continuous conditions at connection points. This transformation leads to the conversion of the variable coefficient differential equation into a nonlinear matrix equation. The Newton- Raphson method is then applied to calculate the characteristic equation and mode shapes, essential for determining natural frequencies. To validate precision, the results obtained are compared with those derived from the finite element method. Furthermore, the developed method is employed to assess the impact of gas plenum location and length on the natural frequency of fuel rods. The proposed methodology serves as a rapid design tool, particularly beneficial during the design phase of fuel rods with non-uniform mass distribution, aiding in configuring structural aspects effectively. [ABSTRACT FROM AUTHOR]

The search for novel materials to protect stainless steel from corrosion at different working conditions has attracted researchers' attention worldwide, where ZrO2 has been of great interest. In this work, the effect of a coating of ZrO2 nanoparticles smaller than 5 nm on the electrochemical corrosion potential—standard hydrogen electrode—(ECPSHE) of 304 stainless steel (304SS), at boiling water reactor conditions, is presented. A ZrO2 coating was applied over the surface of oxidized specimens of 304SS, by deposition of ZrO2 nanoparticles with mean size of 3.75 nm. The ECPSHE of this system was determined under BWR operation conditions, through polarizing tests. It was possible to abate ECPSHE to less than − 230 mV, which is enough to protect 304SS against stress corrosion cracking. The effect of ZrO2 nanoparticles concentration on the coating characteristics and on the electrochemical performance was evaluated. Coatings were analyzed by transmission electron microscopy, scanning electron microscopy, energy dispersive x-ray spectroscopy (EDS) and by x-ray diffraction. The optimal concentration determined and the size of synthesized ZrO2 nanoparticles, allowed to obtain a coating without cracks, forming a ZrO2 physical barrier for oxidant species to avoid water and oxidants reach the 304SS surface. [ABSTRACT FROM AUTHOR]

This study introduces a novel method for enhancing Boiling Water Reactor (BWR) operation simulations by integrating machine learning (ML) models with conventional simulation techniques. The ML model is trained to identify and correct errors in low-fidelity simulation outputs, traditionally derived from core physics computations. These corrections aim to align the low-fidelity results closely with high-fidelity data. Precise predictions of nuclear reactor parameters like core eigenvalue and power distribution are crucial for efficient fuel management and adherence to technical specifications. Current high-fidelity transport calculations, while accurate, are impractical for real-time predictions due to extensive computational demands. Our approach, therefore, utilizes the standard two-step simulation process-assembly-level lattice physics calculations followed by whole-core nodal diffusion computations-to generate initial results, which are then refined using the ML-based error correction model. The methodology focuses on improving simulation accuracy in regular BWR operations rather than developing a universal ML predictor for reactor physics. By training an advanced neural network model on the difference in high-fidelity and low-fidelity simulations, the model can reduce the nodal power error from low-fidelity simulations to around 1% on average and the core eigenvalue down to under 100 pcm. This result is under the condition of the normal variations of control rod pattern and core flow rate changes in standard BWR operations used in the training and evaluation of the machine learning model. This work suggests a promising approach for achieving more accurate, computationally feasible simulation solutions in nuclear reactor operation and management. [ABSTRACT FROM AUTHOR]

The object of this study is the process of thickening organic polycomponent compositions based on Jerusalem artichoke, apples, pumpkin, cranberry, and dogwood to produce high-level semi-finished products employing the proposed equipment and technological solutions implemented in a unified reactor. Conventional thickening devices have a steam heat supply system, which leads to overheating of certain layers of raw materials and prevents the mobile properties of the apparatus under the conditions of significant execution of organic properties of raw materials. The introduction of apratur-technological solutions could provide the competitive advantage of the reactors by heating the film-like emitting electric heater using a mixing device with a heating surface. Increasing the total area of heat exchange to 1.24 m2 (prototype – 0.98 m2). There is a decrease in specific heat consumption by 20 %, specific metal capacity by 47 %, and the duration of the process by 30 %. We have determined rheological properties in the thickening of the organic polycomponent composition of the paste at a temperature of 55 °C and a shift rate within 3.0…4.5 s-1, effective viscosity is in the range of 22...3.5 Pa•s. In this case, the kinetics of boiling in an improved reactor are by 27 % smaller than that in the basic structure R-P-150 m. According to organoleptic indicators, the composition "B" has a pleasant harmonious taste of all components. The results of the studies are relevant for practical implementation in the temperature range from 35 °C to 55 °C. The hardware and technological solutions form the competitiveness of mobile reactors for the thickening of plant semi-finished products under conditions of elimination of the steam component and an increase in the surface of the heat exchange of the mixing device. Providing preference in operating duration with the simultaneous use of Peltier elements to convert secondary heat to the autonomous work of exhaust fans in the motor compartment. [ABSTRACT FROM AUTHOR]

This paper presents an investigation of the oxidation of Alloy X-750 containing 5 wt% iron in a simulated boiling water reactor (BWR) environment. The specimens were exposed by a water jet (10 m/s) at 286 °C for durations ranging from 2 to 840 h, and the development of the oxide microstructure was mainly studied using electron microscopy. The results showed that the oxide scale consists of blocky crystals of trevorite on top of a porous inner layer rich in Ni and Cr. After the longest exposure time, the trevorite crystals completely covered the specimen surface. The study further revealed that the rate at which the oxide grew and the metal dissolved both decreased with time, and the metal thinning process appeared to be sub-parabolic. Given the significant variation in iron content in the X-750 specification, the influence of this element on the material's corrosion performance in BWR was examined by comparing the results from this investigation with those from previous work on material containing 8 wt% Fe. The study demonstrates that the oxide growth, metal dissolution and metal thinning were slower in the material with a higher iron content, indicating the importance of this element in limiting the degradation of Alloy X-750 in BWR environments. [ABSTRACT FROM AUTHOR]

An approach for computing the thermal balance and energy consumption of pits during boiling of frozed logs intended for veneer production has been presented. With the help of our own non-stationary model, the boiling times of beech logs with a diameter of 0.4 m, initial temperature of −10°C, and moisture content of 0.6 kg·kg−1 were determined at water temperatures in the pit equal to 70°C, 80°C and 90°C. Using the determined logs' boiling durations and the mentioned approach, the change in energy required for the entire boiling process and that for each of the components of the thermal balance was calculated. Computer simulations were performed for concrete pit with a volume of 20 m3 and a degree of filling with logs 45%, 60%, and 75%. It was found that the total energy consumption of the pit increases from 169 to 205 kWh·m−3 when the temperature of the boiling water increases from 70°C to 90°C at a pit filling level of 75%. The approach can be applied to compute thermal balances of concrete pits during boiling of frozed and non-frozed logs to any desired final average mass temperature required for the mechanical processing of the plasticized logs. [ABSTRACT FROM AUTHOR]

Flash evaporation processes are widely adopted in the desalination, food processing, waste heat recovery and other industries for heat extraction or product separation. In this paper, a pressure-driven phase transition model is developed by improving the Lee model and combined with the VOF (Volume of Fluid) method to numerically simulate the flash evaporation process. In this modified Lee phase transition model, the driving force for the rates of the local phase transition is calculated using the local temperature and static pressure magnitude. Numerical simulations are carried out in a water-circulating flash chamber and compared with the experimental results to obtain the values of the time relaxation parameters. And the non-equilibrium fraction of the outlet water can be effectively obtained under different conditions of flow rate, inlet temperature and initial liquid level height. The time relaxation factor takes values from 0.195 to 0.43 (Pout,v = 19.9 kPa) and from 0.31 to 0.92 (Pout,v = 31.2 kPa) with increasing superheat. In addition, the model can effectively represent the evolution of the unstable flow flash evaporation from the initial rapid boiling state to dynamic equilibrium. [ABSTRACT FROM AUTHOR]

This study aims to improve the measurement accuracy of liquid film thickness using a liquid film sensor with an optical waveguide film (OWF). The measurement principle of employing the OWF is based on the detection of light reflection at the liquid film surface with high spatial resolution. Because the curved surface of the liquid film reflects light and increases measurement error, we propose a signal processing method to remove the error factor in the calculation of the time-averaged thickness. This method requires prediction of the surface curvature, and we numerically investigated the characteristics of the output signal related to the reflected light intensity. The analysis results showed that the effect of the curved surface up to the surface curvature of 5.0 mm−1 was negligible because the liquid film thickness showed good agreement with that of the flat liquid film surface within 7% accuracy. Furthermore, we consider the applicable range of liquid film thicknesses under the operating conditions of boiling water reactors (BWRs). We estimated the surface curvature of the liquid film based on the calculation of the critical Weber number and confirmed that the curvature caused under the BWR operating conditions was covered by the analysis conditions of this study. Therefore, our proposed method for signal processing via the OWF enabled us to improve the measurement accuracy of the time-averaged thickness with respect to the base film thickness by extracting accurate surface curvature data. [ABSTRACT FROM AUTHOR]

The main objective of the EU-funded project mitigating environmentally-assisted cracking through optimisation of surface condition (MEACTOS) was to gain knowledge on the ability of different surface machining procedures to mitigate environmentally-assisted cracking (EAC) in typical light water reactor structural materials and environments. Surfaces of cold-worked (CW) type 316L austenitic stainless steel and nickel-based weld metal Alloy 182 flat tapered tensile specimens were machined using different processes. EAC initiation susceptibility of these specimens was evaluated using constant extension rate tensile (CERT) tests under simulated boiling water reactor (BWR) and pressurized water reactor (PWR) conditions and assessed using constant load experiments. More than a hundred tests were performed covering about 10 years of autoclave testing time. Only minor or no measurable improvements in EAC initiation susceptibility as a function of surface treatments (grinding or advanced machining) compared to the standard industrial face milling were demonstrated. In most cases, the stress thresholds for EAC initiation determined in constant load tests confirmed the trend obtained from CERT tests. This paper summarises the most important results and conclusions concerning the EAC initiation behaviour for the CW 316L and Alloy 182 under reducing PWR and oxidizing BWR conditions. [ABSTRACT FROM AUTHOR]

Current plans and efforts of reactor operators and vendors to include extended-enrichment (EE) fuel and accident-tolerant fuel (ATF) in current reactor fleets motivate the study of these changes in reactor physics analysis. This work uses the U.S. Nuclear Regulatory Commission's core simulator PARCS to do the core calculation and the SCALE Polaris lattice physics code to generate the homogenized, few-group constants. In this work, both pressurized water reactor and boiling water reactor (BWR) colorset models are used to verify the proposed approach. The accuracy presented in the colorset models verified the capability of the PARCS/Polaris procedures for the transition core analysis in light water reactors. For the whole-core calculation, the ATF and EE-ATF transition core models were incorporated, in addition to the nominal core model. The BWR model was chosen to represent the entire core calculation due to its challenging design. The core parameters studied are the core power distribution, power peaking factor, Doppler temperature coefficients, and control rod worth at cold zero power and hot full power. When the core parameters of the transition cores are compared with those of the nominal core in PARCS, the results suggest that there is no drastic change in the core parameters for the implementation of ATF and EE fuels. [ABSTRACT FROM AUTHOR]

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This paper investigates the possibility to extend the cycle length of boiling water reactor bundles to 15 years of operation with three different burnable poisons; gadolinium, erbium, and boron carbide. This can be carried out by mixing highly enriched UO2 fuel (15–19.9% U-235) with high concentrations of Gadolinium oxide (3–14% Gd2O3) or Erbium oxide (2–4% Er2O3).The boron carbide B4C was modeled as (Al2O3-B4C) rods in the bundle guide tubes. MCNPX code 2.7 was used to evaluate infinite multiplication factor (K-inf), power distribution, peaking factor, void reactivity coefficient, fuel cycle length, depletion of U-235, and fissile inventory ratio for the three designs at 40% void. The MCNPX simulation showed that introducing gadolinium rods at the bundle periphery has the advantage of lowering reactivity swing throughout the exposure range. The uniform distribution of erbium in all fuel rods contributed to the flattening of peaking factor at all the burnup stages. For the B4C design, the author found that the assembly with B4C–Al performs best in terms of reactivity flattening when five of the B4C–AL2O3 rods are positioned in the central region of the assembly. Furthermore, the fuel temperature coefficient is more negative for gadolinium design at all burnup stages. On the other hand, the boron model delivers the lowest control rod worth. Finally, the moderator temperature coefficient is more negative for erbium and WABA designs due to the enhanced thermal neutrons capture by the effect of the strategic arrangement of WABA rods and the uniform distribution of erbium. [ABSTRACT FROM AUTHOR]

Recent performance breakthroughs in artificial intelligence (AI) and machine learning (ML), especially advances in deep learning, the availability of powerful and easy-to-use ML libraries (e.g., scikit-learn, TensorFlow, PyTorch), and increasing computational power, have led to unprecedented interest in AI/ML among nuclear engineers. For physics-based computational models, verification, validation, and uncertainty quantification (VVUQ) processes have been very widely investigated, and many methodologies have been developed. However, VVUQ of ML models has been relatively less studied, especially in nuclear engineering. This work focuses on uncertainty quantification (UQ) of ML models as a preliminary step of ML VVUQ, more specifically Deep Neural Networks (DNNs) because they are the most widely used supervised ML algorithm for both regression and classification tasks. This work aims at quantifying the prediction or approximation uncertainties of DNNs when they are used as surrogate models for expensive physical models. Three techniques for UQ of DNNs are compared, namely, Monte Carlo Dropout (MCD), Deep Ensembles (DE), and Bayesian Neural Networks (BNNs). Two nuclear engineering examples are used to benchmark these methods: (1) time-dependent fission gas release data using the Bison code and (2) void fraction simulation based on the Boiling Water Reactor Full-size Fine-Mesh Bundle Tests (BFBT) benchmark using the TRACE code. It is found that the three methods typically require different DNN architectures and hyperparameters to optimize their performance. The UQ results also depend on the amount of training data available and the nature of the data. Overall, all three methods can provide reasonable estimations of the approximation uncertainties. The uncertainties are generally smaller when the mean predictions are close to the test data while the BNN methods usually produce larger uncertainties than MCD and DE. [ABSTRACT FROM AUTHOR]

Hitachi is advancing their designs for a conceptual reactor called the resource-renewable boiling water reactor (RBWR), a concept reactor similar to the advanced boiling water reactor with a harder neutron spectrum. This design aims to minimise construction costs and waste production as well as to utilise separated plutonium and minor actinide fuel. However, the axial heterogeneity of the design poses calculation difficulties. The aim of this work is to use a known method, reactivity-equivalent physical transformation (RPT), for calculating fuel with double heterogeneity and apply it to a BWR-type fuel pin. This could reduce the calculation time needed for optimisation of the design of the RBWR. The objective of the study is to use SCALE 6.2 to produce an equivalent axial pin model by comparison with the burnup and neutron spectra of a radial model of the fuel. This model can then be used for 2D burnup calculations, and in future work will be used for the generation of two-group and multigroup cross-sections for further deterministic calculations as part of a two-step approach for analysis of the RBWR. The RPT method has been extensively tested on spherical fuel, and SCALE is a standard industry code. The initial radial model is a hexagonal assembly with 20% enriched UO2 fuel in a zircaloy cladding, surrounded by light water moderator. The derived axial model has a water density distribution taken from Hitachi's RBWR designs. Criticality over 70 GWd/tU burnup is estimated using the model. The application of the RPT to the BWR pin was shown to be possible, but to have limitations with the introduction of additional radial complexity. For a single pin, excellent agreement between the radial and axial models could be found across a range of water densities, but in the case of an assembly level calculation distinct equivalence models were required for each water density. In addition, the produced RPT model is validated using SCALE's 3D Monte Carlo module, KENO. [ABSTRACT FROM AUTHOR]

As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment, and fission product analysis in the vessel and the containment, respectively. The severe accident progression in TMI unit 2 has been analyzed as a part of a validation of the CINEMA computer code. This analysis has been performed to validate CINEMA models on the core melt progression, in particular, RCS thermal hydraulic behavior during core melt progression, fuel cladding oxidation with hydrogen generation, and fuel melting with relocation to the lower part of the core. The CINEMA results on main parameters, such as RCS pressure and an integrated hydrogen generation mass are compared with the TMI-2 data. The CINEMA results have shown that the RCS pressure is very similar to the TMI-2 data. The CINEMA results and measured total hydrogen production are very similar, which were approximately 465 kg and 460 kg, respectively. [ABSTRACT FROM AUTHOR]

In recent years, there has been an increased motivation to reduce meat consumption globally due to environmental and health concerns, which has driven the development of meat substitutes. Filamentous fungal biomass, commonly known as mycoprotein, is a potential meat substitute since it is nutritious and has filaments to mimic meat fibrils. The current study aimed to investigate the potential use of a cheap substrate derived from the food industry, i.e., residual water in a tempeh factory, for mycoprotein production. The type of residual water, nutrient supplementation, optimum conditions for biomass production, and characteristics of the mycoprotein were determined. The results showed that the residual water from the first boiling with yeast extract addition gave the highest mycoprotein content. The optimum growth condition was a pH of 4.5 and agitation of 125 rpm, and it resulted in 7.76 g/L biomass. The mycoprotein contains 19.44% (w/w) protein with a high crude fiber content of 8.51% (w/w) and a low fat content of 1.56% (w/w). In addition, the amino acid and fatty acid contents are dominated by glutamic acid and polyunsaturated fatty acids, which are associated with an umami taste and are considered healthier foods. The current work reveals that the residual boiling water from the tempeh factory can be used to produce high-quality mycoprotein. [ABSTRACT FROM AUTHOR]

Activated corrosion products deposited on the surfaces of fuel rods and pipelines contribute the majority of the radiation level in the primary system piping of a light water reactor and would have a significant impact on the safety of maintenance personnel or those involved in future decommissioning work. A computer model for site-specific applications, by the name of ACP_BWR, was developed to predict the distribution of activated corrosion products in the primary coolant circuit of a boiling water reactor (BWR). The prediction results were in reasonably good agreement with the data taken by periodic and in situ measurements at three locations after permanent shutdown of the BWR. Our analyses indicated that the 60Co, 54Mn, 58Co, and 59Fe activities in the core bypass, upper plenum, and lower downcomer regions were higher than those at other regions of the Chinshan Unit 1 reactor. Accordingly, the dose rates resulting from the activated corrosion products deposited at regions close to either side of the core shroud were comparatively high, surpassing those induced by neutron activation at these regions. [ABSTRACT FROM AUTHOR]

During the 2011 nuclear catastrophe at Fukushima Daiichi, Unit 3 had a sharper increase in containment pressure than Unit 2, with thermal stratification of the suppression pool cited as one of the contributing factors. In the present work, the buoyancy-induced circulation consequent to steam condensation in a large, toroidal pool of water is studied using computational fluid dynamics (CFD) simulations with a view to understanding the role of important design parameters of the suppression pool system. The tunnelling phenomenon observed in the development of the thermal stratification process is delineated in terms of the establishment of a thermocline. The effects of the number of steam injection points and the cross-section of the pool on thermal stratification characteristics have been investigated through a number of case studies. In all the cases, the surface temperature, which is responsible for over-pressurization of the containment, is found to be significantly higher than the bulk pool temperature. Multiple injection points with the same overall steam flow rate are found to lead to higher surface temperatures due to a shortened circulation path. For the same volume of pool water, the simulations show that a deeper and narrower pool gives rise to significantly higher temperatures than a wider and shallower pool. This is attributed to the relatively deeper penetration of the buoyancy-induced circulation into the pool. [ABSTRACT FROM AUTHOR]

This paper will demonstrate the potential of modern design for additive manufacturing by using computational fluid dynamics with design optimization. The Transformational Challenge Reactor Instrumentation and Control (I&C) Team has specifically requested that an instrumentation plane be designed for monitoring of the core-coolant-flow average temperature within ±5°C of the core outlet average temperature. However, because of systemwide constraints, this design space is allotted a pressure drop of only 0.5 psi. A successful design optimization study is discussed along with the thought process leading to the successful conclusion of an I&C plane with an average temperature of 497.3°C and a standard deviation of 1.03°C, all while maintaining a 3.38-kPa (0.49-psi) pressure drop across the outlet plenum. [ABSTRACT FROM AUTHOR]

The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro-posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. 1%w-channel models were tested (the average and the hot channel). Ther-mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective. [ABSTRACT FROM AUTHOR]

The aim of this work is to introduce the novel concept of an m-polar fuzzy soft set, including various types of algorithms and their fundamental operations. We created mathematical modeling to analyze operational rules and discuss the advantages, disadvantages, and natural aspects of algorithms for six types of nuclear power plants. It has been determined that emerging trends and the benefits of algorithms are increasing step by step. The suggested modeling with an m-polar fuzzy soft set is integrated into the fuzzy mean environment to analyze the effect of the correlation between decision factors and decision results without an excessive duty cycle, thus minimizing energy use and other adverse effects. Based on a new group decision-making technique considering an asymmetric weight vector, we proved that Gas Cooled, Graphite-Moderated, and Pressurized Water Reactors are the optimal choices for nuclear power plants. In the end, a numerical illustration is provided for selecting the best photo to demonstrate the use of the generated technique and to exhibit its adequacy. [ABSTRACT FROM AUTHOR]

The development of coatings for accident-tolerant fuels (ATFs) for light water reactor (LWR) applications promises improved corrosion resistance under accident conditions and better performances during operation. CrN and TiN coatings are characterized by high wear resistance coupled with good corrosion resistance properties. They are generally used to protect materials in applications where extreme conditions are involved and represent promising candidates for ATF. Zr cladding tubes coated with 5 µm-thick CrN or TiN, exposed in an autoclave to simulated PWR chemistry and BWR chemistry, were characterized with SEM, EDS, and STEM. The investigation focused on the performance and oxidation mechanisms of the coated claddings under simulated reactor chemistry. Both coatings provided improved oxidation resistance in a simulated PWR environment, where passivating films of Cr2O3 and TiO2, less than 1 µm-thick, formed on the CrN and TiN outer surfaces, respectively. Under the more challenging BWR conditions, any formed Cr2O3 dissolved into the oxidizing water, resulting in the complete dissolution of the CrN coating. For the TiN coating, the formation of a stable TiO2 film was observed under BWR conditions, but the developed oxide film was unable to stop the flux of oxygen to the substrate, causing the oxidation of the substrate. [ABSTRACT FROM AUTHOR]

Addition of Zn2+ and Mg2+ ions to the coolant water is followed in boiling and pressurized water reactors to control the radiation field. These cation additions modify the composition of double-layered spinel nickel chromium ferrite oxide films formed on stainless steel. The primary objective of this paper is to offer a systematic perspective for comprehending the dissolution behaviour of non-stoichiometric oxides upon Zn2+ and Mg2+ addition. In this regard single-phased nickel chromium ferrites were synthesized and characterized by various techniques. The crystal morphology and band gap changed with an increase in lattice dimensions upon incremental substitution of Ni. The oxidative dissolution rates increased with increasing Mg2+ or Zn2+ in the Ni–Cr–Fe–O lattice with a maximum at x = 0.6. Magnesium-substituted nickel ferrites showed a higher dissolution kinetics compared to their zinc-substituted counterparts. The dissolution behavior and kinetics are explained on the basis of cation redistribution in the octahedral or tetrahedral sites of the spinel lattice. The Laser Raman Spectroscopic (LRS) analysis corroborated the X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) observations. These results will provide an insight into the mechanism of radioactivity removal kinetics for developing effective dose rate reduction practices for nuclear plants. • Effect of Zn/Mg substitution in chromium containing nickel ferrites was probed. • M x Ni 1- x CrFeO 4 was synthesized by combustion method using 1:1 fuel to oxidiser ratio. • The band gap and lattice parameters increased with Mg/Zn substitution in NiCrFeO 4. • Cation redistribution in octahedral and tetrahedral sites was confirmed from XPS. • Redistribution of cations affected morphology, composition and dissolution rates. [ABSTRACT FROM AUTHOR]

• Decay heat distribution in the spent RBMK-1500 FA was calculated using SCALE code. • Influence of the application of the axial power profiles was estimated. • The decay heat in outer pins is higher by 10 ÷ 14 % compared to the inner pins. • Evaluation of spatial heterogeneities results up to 3% difference in decay heat. The RBMK-1500 reactor is a boiling water reactor. Thus, the coolant density and the neutron moderation efficiency vary axially over the fuel assembly height influencing the distribution of decay heat in the discharged fuel. The underestimation of this effect on the decay heat may lead to biases in safety assessments of the spent fuel management. Axial coolant density and power profiles were used in this work to determine axial and radial decay heat distributions in the RBMK-1500 fuel assembly over the first 100 years of cooling using the SCALE code system. It was found that the application of axial profiles influences the increase of estimated total decay heat by 3 % just 2 years after the irradiation is finished. However, the estimation of total decay heat is lower by 2 % after 100 years of cooling compared to the case with no applied axial profiles. The maximum of the normalized decay heat profile also varies between 1.31 and 1.40 over the cooling time. The estimated decay heat in outer pins over the cooling time is higher by 10 ÷ 14 % compared to the decay heat produced in inner pins. The variation of results over the cooling phase and reasons behind the difference in estimated decay heat was explained by analysing nuclides which mainly contribute to the total decay heat. [ABSTRACT FROM AUTHOR]

• Identified the contact angle on ZrO 2 using Molecular Dynamics for both main crystallographic planes at Boiling Water Reactor in-situ operating conditions. • Connected the nanoscale Molecular Dynamics simulations to mesoscale Lattice Boltzmann simulations through the contact angle. • Modelled the nucleation of bubbles on a ZrO 2 surface at Boiling Water Reactor in-situ operating conditions taking into account the micro-roughness. • Obtained a correlation between the roughness and the shape of the surface and the nucleation site density. The mechanism of bubble nucleation in boiling water at solid surface remains poorly understood. In this study, the water boiling and bubble nucleation density under nuclear reactor operation conditions were investigated by multiphase Lattice Boltzmann and Molecular Dynamics simulations. The developed multiscale model takes into account the surface energy of ZrO 2 cladding in different crystallographic planes obtained by molecular dynamics simulations and uses it to inform the corresponding contact angle of two-phase fluid in the thermal multiphase Lattice Boltzmann model. The model describes the bubble formation on a rough surface due to boiling and predicts the number of active nucleation sites on surfaces of different roughness. Further, the effect of the different contact angles on bubble dynamics while in close contact with the surface is investigated. The obtained results are in good agreement with experimental observations and provide functional relationships between interface properties (roughness, surface energy) and the nucleation site density, necessary for macroscopic simulations of the boiling phenomena. [ABSTRACT FROM AUTHOR]

This work demonstrates the use of 14.1-MeV fast-neutron interrogation in associated particle (AP) mode for the nondestructive assay of SNF canisters. We performed Monte Carlo simulation-based studies to inspect BWR GE (Boiling Water Reactor General Electrics)-14 fuel assemblies in an MPC-68 canister and basket and reconstructed a cross-sectional view of a quarter canister. Detection of the time-correlated characteristic gamma rays emitted by neutron inelastic scattering allowed us to perform elemental imaging of selected materials, such as iron, which would reveal damages to the canister structure and other steel components of the canister. Using this method, structural damages located in the peripheral or central area of an empty canister can be imaged in a 7.7-days measurement, and in a 52-days measurement in a fully loaded canister, when using a commercial AP neutron generator with a source strength of 1 0 8 n/s. Additionally, single-pin resolution can be obtained using this method when detecting the scattered gamma rays in the 50 keV–15 MeV energy range. As AP generator technology advances and higher source strengths will become available, shorter measurement times will be possible. This unique approach to inspecting the integrity on the fuel and ancillary canister structures is highly penetrating, non-destructive, material-specific, and does not require the detector to be directly in contact with the outer canister surface. [ABSTRACT FROM AUTHOR]

During a severe power reactor accident, the plant core can melt. The resulting mixture of molten nuclear fuel and other in-core materials is known as corium. For a Canada Deuterium Uranium (CANDU) reactor, the corium is expected to settle at the bottom of the calandria vessel, but there is a potential for some melt to flow through connecting piping and other penetrations. The flow of corium through these structures can be contained if melt solidification and thus plugging occur. A numerical model was created to simulate the flow of molten metal through an empty vertical pipe. This model was benchmarked to a previous analytical model and validated against experimental results with gallium metal (which is a metal with low melting temperature) as an alternative for corium. The numerical model predicted the penetration length of gallium with an average percent error of 10.3% when compared to the experimental penetration length results of gallium. The model was also updated to predict the corium penetration length in cooling pipes of the CANDU reactor during a severe accident. [ABSTRACT FROM AUTHOR]

A demonstration plant designed for methanol synthesis with steel mill gases was operated continuously for the first time over a period of 5 weeks. The reactor represents a single tube with typical dimensions of an industrial tube bundle reactor with a boiling water cooling jacket and is incorporated into a full loop of unreacted gases. Temperature profile along reactor axis and gas composition in gas recycle stream are measured online. In several experiments, the operating range of the plant was determined. Additionally, the reactor and the loop were simulated and the results agree well with experimental values. [ABSTRACT FROM AUTHOR]

10.3390/en13061431 14 Fang R., Cacuci D.G. Comprehensive second-order adjoint sensitivity analysis methodology (2nd-ASAM) applied to a subcritical experimental reactor physics benchmark: V. Computation of 2nd-order sensitivities involving isotopic number densities. Higher-order sensitivities have also been incorporated into a predictive modeling formalism [[17]], which combines experimental and computational information in the joint phase-space of responses and model parameters, including not only the 1st-order response sensitivities, but also the complete hessian matrix of 2nd-order second-sensitivities and also the 3rd-order sensitivities. 2006 17 Cacuci D.G. Towards overcoming the curse of dimensionality: The third-order adjoint method for sensitivity analysis of response-coupled linear forward/adjoint systems, with applications to uncertainty quantification and predictive modeling. Each successively higher-order sensitivities are computed using the same principles as used for computing the 2nd-order sensitivities, i.e., the computation of the 3rd-order sensitivities can be prioritized by using the magnitudes/importances of the 1st-order and 2nd-order sensitivities as guiding indicators, and so on. [Extracted from the article]

Underfloor air distribution (UFAD) systems are increasingly used for their advantages in improving energy savings, indoor air quality, and thermal comfort. In UFAD systems, an underfloor plenum delivers conditioned air to the air supply diffusers. The distribution of internal air velocity and static pressure in plenums determines the uniformity of the airflow to the occupied zones. As a result, the plenum has a detrimental effect on the characteristics of the supply air and, thus, the resulting indoor air quality and thermal comfort. Nevertheless, most existing studies on underfloor plenums focused on small-scale plenums with a single internal air duct. Large plenums and multiple air ducts in UFAD equipped in large premises are underexplored. In this study, a circular underfloor plenum with a large scale (radius of 15 m, height difference of 0.9−2.9 m) and 503 under-seat diffusers in a conference room was studied using computational fluid dynamics (CFD) simulation (ANSYS Fluent (16.0)). The distributions of airflow velocity and static pressure inside the plenum were analyzed and compared to one concentrated air supply mode and three uniform air supply modes. Based on the air velocity at the center of under-seat diffusers, the outgoing airflow uniformity from the diffusers under four cases was evaluated by the index of air velocity uniformity. The results showed that the multiple supply ducts with bottom air outlets yielded the best uniformity of supply air. The findings of this paper are expected to provide a technical basis for realizing the optimal design of the UFAD system in terms of uniformity of supply air. [ABSTRACT FROM AUTHOR]

This article presents (nonlinear) control system synthesis for a boiling water reactor (BWR) by using artificial intelligence (AI)-based reinforcement learning (RL), where the pertinent algorithm is deep deterministic policy gradient (DDPG). The BWR model, used in this article, exhibits limit cycling and/or chaotic behavior in different regions of operation. The performance of the RL control system is compared with that of a control system synthesized by the standard $\mathcal {H}_\infty $ theory. The results of comparison show that the RL control system outperforms the $\mathcal {H}_\infty $ control system for disturbance rejection, stability under perturbation, and set-point tracking in a majority of the test cases. [ABSTRACT FROM AUTHOR]

Surface treatments of primary circuit components in light water reactors are regarded as possible ways to mitigate environmentally-assisted cracking (EAC). To date, it is not fully conclusive which surface condition is suitable to reduce the EAC initiation susceptibility. Constant extension rate tensile (CERT) tests were performed by several labs using flat tapered tensile specimens with different surface conditions (ground, industrial face milled, advanced face milled and shot peened), exposed to a boiling water reactor normal water chemistry environment at 288°C. Despite some scatter in the results, the CERT tests revealed that the EAC initiation susceptibility seems lowest for the advanced face milled surface and highest for the shot peened surface. However, it must be emphasised that the differences were moderate and that the surprising behaviour of the shot peened surface can be explained. The mechanical grinding of the surface did not significantly retard EAC initiation compared to industrial face milling. [ABSTRACT FROM AUTHOR]

The marine environment may change the force on the fluid and inevitably influence bubble behavior and the two-phase flow in the reactor core, which are vital to the safety margin of a nuclear reactor. To explore the effect of the marine motion on the flow and heat transfer features of subcooled flow boiling in the reactor core, the volume of fluid (VOF) method is employed to reveal the interaction between the interface structure and two-phase flow in the subchannel under rolling motion. The variations of several physical parameters are obtained, including the transverse flow, the vapor volume fraction, the vapor adhesion ratio, and the phase distribution of boiling two-phase flow with time. Sensitivity analyses of the amplitude and the period of the rolling motion were performed to demonstrate the mechanisms of the influence of the rolling motion. We found that the transverse flow in the subchannel was mainly affected by the Euler force under the rolling motion. In contrast to the two-phase flow in the static state, the vapor volume fraction and vapor adhesion ratio show different characteristics under rolling motion. Additionally, the onset of significant void (OSV) point changes periodically under rolling motion. [ABSTRACT FROM AUTHOR]

Techno-economic prognostications for the manufacture of fuel assemblies with mixed uranium-plutonium fuel for the VVER-SKD power reactor, which is now under development, show that at the current level of technological development of pressurized water-cooled reactors in our country and the world the share of the fuel component in the price of electricity generated by NPP with VVER-SKD will not exceed the level realized by modern VVER, PWR, and BWR reactors. Moreover, considering the higher efficiency of VVER-SKD compared to the customary VVER/PWR reactors, maintaining the share of the fuel component in the cost of electricity at the current level will give this innovative reactor technology significant competitive advantages in regional electricity markets. [ABSTRACT FROM AUTHOR]

Thermal-induced two-phase flow instabilities could badly slow the runtime performance of two-phase systems, such as boiling water reactor core, and even endanger the safe operation. Hence, many researchers have carried out numerous researches on flow instability characteristics, but few publications cover the effects of local throttling on two-phase flow instabilities in open-channel systems such as coolant rod bundle subchannel with spacer grids in the nuclear reactor core. This paper provides a numerical study on the effects of local throttling on two-phase flow instabilities in a simplified typical coolant open-channel system in a nuclear reactor core by using the NUSOL-SYS code. The effects of local throttling ratio, throttling position, and other throttling parameters on the stability of the boiling channel system were carried out. The results show that usually in uniformly distributed throttling conditions, the stability of the system and high throttling ratio are positively correlated. In fixed throttle ratio conditions, the stability of the system is positively correlated with the distance from the throttling region to the entrance and local void fraction. If throttling regions are uniformly arranged along the heating channel with a certain value, the number of throttle regions will enhance the instability of the system. Besides, this paper preliminarily proposed a criterion function f(n, RT, α) to study the complicated throttling effects on flow instability which could provide technical reference for the safe design and operation of two-phase systems, especially, the reactor core. [ABSTRACT FROM AUTHOR]