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This study developed a probabilistic determination model with respect to cladding high-temperature burst conditions based on the Bayesian statistical method to reasonably evaluate fuel behaviors under loss-of-coolant accident conditions, including fuel fragmentation, relocation, and dispersal. The candidate models (correlations) were based on the widely accepted empirical model established based on nonirradiated fuel cladding data. Explanatory variables were added to improve the applicability of these models with respect to irradiated materials and generalization performance. The posterior predictive distribution of each candidate model was evaluated using Bayesian estimation comprising 238 sets of high-temperature burst test data. The generalization performance was evaluated using information criteria. The results of model evaluation showed improved predictive performance by considering the effect of hydrogen content. A comparison with burnup as an alternative explanatory variable confirmed that hydrogen content was the better parameter and other burnup-associated effects, such as irradiation hardening of the metal matrix and oxide growth (reduction of the metal matrix), were less dominant under burst conditions. [ABSTRACT FROM AUTHOR]

The seismic resistance of fuel cladding during the long-term core cooling after loss-of-coolant accidents (LOCAs) was investigated by performing cyclic four-point bending tests (4PBTs) of up to 1000 cycles with fresh fuel cladding samples that experienced integral thermal shock test, simulating LOCA conditions, including ballooning, rupture, oxidation, and quench. 4PBTs were performed on the samples that survived the quenching process. The results showed that up to 1000 cycles and 5.8 Nm of cyclic loading moment, there was no apparent effect on the bending fracture limit of the fuel cladding under the 4PBT. The scatter of the bending fracture limit for a given equivalent cladding reacted (ECR) evaluated by the Baker–Just oxidation rate equation (BJ-ECR) is attributed to two primary factors: first, the difference between the prescribed and the actual oxidation behavior, confirmed by comparing the BJ-ECR and the ECR evaluated based on metallographic observation (M-ECR), and second, the variated shape of the rupture-opening area after the integral thermal shock test. The strength of the α phase-dominant zone near the rupture opening seems to contribute to the bending fracture limit. [ABSTRACT FROM AUTHOR]

There is a recognized need to discuss the preservation of the coolable geometry of fuel rods during long-term cooling after a loss-of-coolant accident (LOCA). Mechanical load induced by an earthquake may be a primary cause of loss of the coolable geometry after a LOCA. In this research, aiming to confirm the preservation of coolable geometry of fuel rods during an earthquake after a LOCA, we investigated the stress applied to the cladding tube with a rupture opening under a vibration condition via finite element analyses for a fuel assembly that considered the mechanical interaction between a fuel rod and spacer grids. The analyses demonstrated that the stress on the cladding tube increased with the axial holding force of the spacer grid on the fuel rod. The stress concentrating at the periphery of the rupture opening was compared with the strength of the post-LOCA cladding tubes determined via bending tests. This research concludes that the fuel rods are likely to be prevented from fracture due to bending arising from an earthquake during post-LOCA cooling unless the oxidation of the cladding tubes exceeds 15% equivalent clad reacted. [ABSTRACT FROM AUTHOR]

The steel liner plays a key role in ensuring the leak-tightness of nuclear containment structures (NCSs). Once the liner is subjected to corrosion, its corrosion effects can severely impact the leak-tight function of the NCS, especially in severe accident scenarios. Therefore, evaluating the corrosion effects of the liner is essential to guaranteeing the safe operation of nuclear power plants. This paper presents a probabilistic safety analysis of the NCS under liner corrosion conditions. Firstly, the corrosion mechanism of the liner is elucidated, and a refined simulation method is developed to investigate the localized corrosion effects of the liner. Utilizing a probabilistic finite element method, the fragility of the NCS under loss-of-coolant accident (LOCA) conditions is evaluated. Finally, the leakage risk of the NCS at different degrees of liner corrosion is discussed. The results indicate that liner corrosion has a significant impact on the median pressure capacity and high-confidence–low-probability pressure capacity of the NCS. With the aggravation of liner corrosion, the safety margin of the NCS decreases, and the total probability of failure increases. Within the scope of this paper, the NCS can fulfill the probabilistic safety requirements. [ABSTRACT FROM AUTHOR]

Considering the development of deep learning and the emergence of intelligent control demands in nuclear reactors, along with the presence of plant‐level real‐time information monitoring systems in most nuclear power plants, there is a considerable accumulation of sensor measurements from long‐term operation. This makes it feasible to conduct medium to long‐term predictions for various real‐time conditions in nuclear power plants. Therefore, this paper proposes the utilization of a gate‐based recurrent neural network called GRU (Gated Recurrent Unit) and its variants for parameter prediction of LOCA (Loss of Coolant Accident) scenarios. The main content of this paper consists of two parts: (1) Experimental verification is conducted to demonstrate that GRU has excellent capability in capturing long‐term sequential information and generalization ability, making it suitable for predicting accident conditions in nuclear power plants. Two accident trend prediction methods based on the GRU network are proposed for scenarios with limited data. The results show that these methods can effectively provide short‐term development trends for accident conditions. Additionally, by considering the feature extraction capacity of CNN, the fusion of CNN and GRU models is employed for parameter prediction under different sizes of broken area. The results indicate an improvement in the model's generalization ability. (2) In scenarios with limited and incomplete data, a more robust variant of GRU called GRU‐D model is utilized for both univariate and multivariate synchronous prediction of accident conditions with different missing values. Experimental results demonstrate that even with a data missing rate of 90%, the GRU‐D network exhibits excellent predictive accuracy and generalization ability in parameter prediction for the given conditions. [ABSTRACT FROM AUTHOR]

A reactor core overheats due to decay heat generated in the fuel when an effective cooling medium is unavailable, such as in a loss-of-coolant accident combined with a loss of emergency core coolant. If the heat generated is not effectively dissipated, then at extreme temperatures, the structural strength of the bundle assembly may deteriorate, leading to slumping of fuel elements onto the inner wall of the pressure tube. It is essential to examine the temperature behavior of the channel containing fuel pins in a disassembled state in order to comprehend the impact of further thermally induced deformations in the channel during postulated accident conditions. Capturing the temperature of channel components at each circumferential position from experiments is extremely difficult; thus, a modeling tool is necessary to obtain a thorough circumferential temperature profile. This paper presents a numerical study that aims to study the temperature distributions in a 1-m-long pressurized heavy water reactor (PHWR) channel containing a disassembled fuel bundle. The channel geometry and the boundary conditions implemented were obtained from the experiment. A temperature profile for each channel element at every circumferential and axial location was obtained. A thorough comparison of the predicted and the reported experimental values was performed, and it was found that the predicted temperature behavior of the channel was consistent with the experimental data. Further simulations with different fuel element configurations and decay powers may be carried out; in addition, the results obtained may be used for coupled thermal-mechanical and thermal-mechanical-chemical simulations. [ABSTRACT FROM AUTHOR]

In this paper, we report an innovative approach of using spark plasma sintering (SPS) combining with finite element modeling (FEM) to investigate the transient behavior of nuclear fuels. The temperature ramping rates of the SPS with special tooling can be controlled from 5 to 10 °C/s up to 500 °C/s, mimicking prototypical thermal profiles of the loss-of-coolant-accident (LOCA) and reactivity-induced-accident (RIA) events, respectively. The fresh UO 2 pellets with micron grain size are synthesized by SPS. Their fracture and fragmentation behaviors are investigated under transient thermal testing, and temperature-gradient and thermal-stress are calculated by FEM. The micron-sized fresh UO 2 displays good thermal shock stability and anti-cracking performance under the LOCA thermal test. Under simulated RIA conditions, the UO 2 fuels crack due to the stress induced by the large thermal gradient. The unique capability of SPS with controlled temperature ramping enables a cost-effective approach for rapid screening and high throughput evaluation of nuclear fuels under power transients. [ABSTRACT FROM AUTHOR]

In this article we analyzed the translations into English of the categories of bottoms and tops present in two works by Reinaldo Arenas: Before Night Falls, translated by Dolores Koch, and The Color of Summer, by Andrew Hurley. To this end, we use the translation techniques proposed by Lucía Molina and Amparo Hurtado Albir (2002), and the translation methods proposed by Marti Ferriol (2006), following the new classification by Ericka Vargas Castro (2022). This is discussed in light of a theoretical framework that includes gender, with concepts as Judith Butler's (2009) performativity and precarity, or Pierre Bourdieu's (1998) paradox of the doxa, as well as the Cuban sociopolitical perspective. It is concluded that three techniques are the most used: the literal translation, the established equivalent, and the discursive creation. Finally, this discussion concerning gender, political regime, and translation might lead to the proposal of consistent translations of sex/gender designations in the work by Arenas. [ABSTRACT FROM AUTHOR]

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Nitrogen pressurization is widely used in nuclear reactor (such as passive nitrogen-type accumulators and nitrogen regulator) because of passive mechanism. However, nitrogen pressurization brings new challenges. Loss of coolant accident (LOCA) can lead to rapid reduction of pressure in primary circuit of pressurized water reactor (PWR), putting dissolved nitrogen in the water into a supersaturated state, which contributes to nitrogen release. In order to prevent the accumulation of released nitrogen, which may act as a threat to reactor safety, the migration and release characteristics of nitrogen were studied. In this paper, a nitrogen wall release model was established, the physical significance of the release source term was explained and the specific process of nitrogen release was revealed. In the case of LOCA in primary circuit of a PWR, the pressure in the primary circuit decreases and the susaturated dissolved nitrogen degas from the liquid and form nitrogen bubbles on nucleate sites. The susaturated dissolved nitrogen continuously degas at the gas-liquid interface of the bubble, leading to the sustained growth of nitrogen bubble. After reaching departure diameter, the buoyancy force on the nitrogen bubble is greater than surface tension and it escapes from the wall finally. In this paper, the mathematical modeling of nitrogen dissolution equilibrium, dissolved nitrogen convection and diffusion, the susaturated dissolved nitrogen wall release and nitrogen bubbles upward floating was carried out. Combined with the one-dimensional lumped parameter method, the mass conservation of the control volume was analyzed and the one-dimensional nitrogen mass conservation equation was deduced. To solve the one-dimensional nitrogen mass conservation equation, a one-dimensional lumped parameter nitrogen migration solution program was written based on MATLAB. The program was verified by the experimental data of small break loss of coolant accident (SBLOCA). The trends of dissolved nitrogen concentration and void fraction under pressure relief condition during SBLOCA were simulated. The results suggest that susaturated nitrogen degas on the wall in the form of bubbles. The bubble grows at a rate proportional to t1/2 and also proportional to (C-Ci)12. The influence of the initial nucleation radius on the growth period of the bubble is negligible and the bubble rises uniformly in the main stream after departure. Under pressure relief condition during LOCA, the nitrogen wall release effect is significant and the diffusion process of dissolved nitrogen is weak. Controlled by nitrogen wall release, the average concentration of dissolved nitrogen decreased by 19.5% and the void fraction became 4.33 times of the original, which indicates that wall release has significant effect on circuit, potentially affecting reactor safety. The simulation results are in good agreement with experimental data of SBLOCA, proving the reliability of wall release model and program.

This study explores the application of Liquid Optical Clear Adhesive (LOCA) in flexible display modules using inkjet printing technology. We started by selecting and testing the printing and mechanical characteristics of the LOCA material, ultimately determining the optimal viscosity window and suitable LOCA material for our needs. Next, we established the inkjet printing process for LOCA and validated the curing process based on peel strength results. Finally, we produced folding modules using LOCA and conducted folding tests, which demonstrated satisfaction of the 200,000‐cycles folding requirement. [ABSTRACT FROM AUTHOR]

Following a loss of coolant accident (LOCA), prestressing concrete containment vessels (PCCVs) may experience high thermal load as well as internal pressure. The high temperature stress would increase the risk of premature damage to the containment, which reduces the safety margin during the increasing internal pressure. However, current investigations cannot clearly address the issues of thermal-pressure coupling effect on damage propagation and thus safety of the containment. Thus, this paper offers three simple and powerful damage parameters to differentiate the severity of damage of the containment. Moreover, despite of the temperature action severely threatening the pressure performance of the containment, the research regarding the improvement of the resistant performance of the containment is quite scarce. Therefore, in this paper, a comprehensive comparison of damage propagation and mechanism between conventional and fiber-reinforced concrete (FRC) containments is performed. The effects of fiber characteristics parameters on damage propagation of structures following the LOCA are also specifically revealed. It is found that the proposed damage indices can properly indicate state of damage in the containment body and the addition of fiber can be used to obviously mitigate the damage propagation in PCCV considering the thermal-pressure coupling.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

A method was proposed for in-situ evaluating the thickness and resistivity of the oxide/hydroxide film formed on the surface of aluminum alloy exposed to sump water formed in the containment after a loss-of-coolant accident. The evaluation entailed fitting a model for the film impedance, which has film thickness and other variables describing the resistivity profile of the film along its thickness direction as fitting parameters, to the practically measured electrochemical impedance data. The obtained resistivity profiles implied that the films formed at pHs25°C 7, 8, 9, 10, and 11 all had a duplex structure; compared to the outer layer in contact with the solution, the inner layer of the film had a much higher resistivity and was inferred to be denser and provide most of the protectiveness of the film. Both the thickness and the total resistance of the film decreased with the increasing solution pH25°C, suggesting that the films formed in more alkaline solutions had less protectiveness against corrosion, consistent with the increasing aluminum alloy corrosion rates previously identified.

An industrial control process multiple-input multiple-output (MIMO) coupled system is analyzed in this study as an example of a Loss of Coolant Accident (LOCA) simulation system. Ordinary control algorithms can complete the steady state of the control system and even reduce the response time to some extent, but the entire system still consumes a large amount of energy after reaching the steady state. So a multivariable decoupled energy-saving control method is proposed, and a novel energy-saving function (economic function, Eco-Function) is specially designed based on the active disturbance rejection control algorithm. Simulations and LOCA simulation system tests show that the Eco-function algorithm can cope with the uncertainty of the multivariable system's internal parameters and external disturbances, and it can save up to 67% of energy consumption in maintaining the parameter steady state.

The steel liner plays a key role in ensuring the leak-tightness of nuclear containment structures (NCSs). Once the liner is subjected to corrosion, its corrosion effects can severely impact the leak-tight function of the NCS, especially in severe accident scenarios. Therefore, evaluating the corrosion effects of the liner is essential to guaranteeing the safe operation of nuclear power plants. This paper presents a probabilistic safety analysis of the NCS under liner corrosion conditions. Firstly, the corrosion mechanism of the liner is elucidated, and a refined simulation method is developed to investigate the localized corrosion effects of the liner. Utilizing a probabilistic finite element method, the fragility of the NCS under loss-of-coolant accident (LOCA) conditions is evaluated. Finally, the leakage risk of the NCS at different degrees of liner corrosion is discussed. The results indicate that liner corrosion has a significant impact on the median pressure capacity and high-confidence–low-probability pressure capacity of the NCS. With the aggravation of liner corrosion, the safety margin of the NCS decreases, and the total probability of failure increases. Within the scope of this paper, the NCS can fulfill the probabilistic safety requirements.

Accurately predicting severe accident data in nuclear power plants is of utmost importance for ensuring their safety and reliability. However, existing methods often lack interpretability, thereby limiting their utility in decision making. In this paper, we present an interpretable framework, called GRUS, for forecasting severe accident data in nuclear power plants. Our approach combines the GRU model with SHAP analysis, enabling accurate predictions and offering valuable insights into the underlying mechanisms. To begin, we preprocess the data and extract temporal features. Subsequently, we employ the GRU model to generate preliminary predictions. To enhance the interpretability of our framework, we leverage SHAP analysis to assess the contributions of different features and develop a deeper understanding of their impact on the predictions. Finally, we retrain the GRU model using the selected dataset. Through extensive experimentation utilizing breach data from MSLB accidents and LOCAs, we demonstrate the superior performance of our GRUS framework compared to the mainstream GRU, LSTM, and ARIMAX models. Our framework effectively forecasts trends in core parameters during severe accidents, thereby bolstering decision-making capabilities and enabling more effective emergency response strategies in nuclear power plants. [ABSTRACT FROM AUTHOR]

Copyright of Revista Digital de Derecho Administrativo is the property of Universidad Externado de Colombia and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

The specific configuration of the safety systems in VVER-1000/V320 reactors allows a comprehensive study of the Loss of Coolant Accident (LOCA). In the present paper, a verification of the success criteria of the event trees headers for the medium and large break LOCA sequences is conducted. A detailed TRACEV5P5 thermal-hydraulic model of the reactor has been developed, including all safety systems. When analyzing the results of all sequences, some conservatism is observed in certain specific configurations as the success criterion of some headers is not consistent with the classic PSA level 1. Therefore, new proposals for the LOCA event trees are performed based on a reconfiguration of LOCA break ranges and the use of the expanded event trees approach.

The present study focuses on the evaluation of the mechanical properties degradation of Cr-coated Zr-alloy fuel cladding. The main objective of the work is to find a suitable methodology to evaluate the mechanical properties degradation of coated cladding by performing several separate effects experiments. Apart from the many positive effects of protective coatings on the overall cladding properties, coatings’ general disadvantage is their reduced ability to tolerate plastic strain. Therefore, coating cracks might occur in the first stage of the hypothetical Loss of Coolant Accident (LOCA). The study is unique because of the consideration of coating cracks. Prior to the high-temperature (HT) oxidation, samples were subjected to either a scratch test or burst test, resulting in the creation of coating defects. The subsequent evaluation of the obtained data consisted of wavelength dispersion spectroscopy (WDS) and optical microscopy analysis and hydrogen content measurements.

Long-term aluminum (Al) corrosion tests were designed to investigate the condition that would generate severe Al corrosion and precipitation. Buffer agents of sodium tetraborate (NaTB), trisodium phosphate (TSP) and sodium hydroxide (NaOH) were adopted. The insulation materials, fiberglass and calcium silicate (Ca-sil), were examined to explore their effects on Al corrosion. The results show that significant precipitates were formed in both NaTB/TSP-buffered solutions at high pH. The precipitates formed in NaTB solution raise more concerns on chemical effects in GSI-191. A passivation layer formed on the surfaces of coupon in solution with the presence of insulations could effectively mitigate Al corrosion. The Fe-enriched intermetallic particles (IPs) embedded in coupon appeared to serve as seeds to readily induce precipitation via providing extra area for heterogeneous Al hydroxide precipitation. X-ray spectroscopy (EDS) and X-ray diffraction (XRD) analyses indicate that the precipitates are mainly boehmite (γ-AlOOH) and no direct evidence confirms the presence of sodium aluminum silicate or calcium phosphate.

The paper shows the occurrence of pressure waves and their propagation in the equipment of the first circuit of the WWER-1000 reactor plant in emergency situations associated with an instantaneous stop of the Main circulation pump or a two-way flow with an instantaneous two-way break in the cold thread of the main circulation pipeline. The influence of the time of initialization of the accident (pump stop, pipeline rupture) on the intensity of the process-amplitude, frequency of pressure changes is investigated. Pressure drops during an emergency on the main elements of the circuit are considered. It is shown that the maximum changes in the amplitude and frequency of both the pressure and the pressure drops on the circuit elements belong to the initial stage of the accident. The main attention is focused on the pressure drops on the equipment, because this parameter determines the dynamic loads on the equipment, which can lead to its failure.

The in-vessel Loss-of-Coolant Accident (LOCA) is one of the design basis accidents in the design of the EU DEMO tokamak fusion reactor. System-level codes are typically employed to analyse the evolution of these transients. However, being based on a lumped approach, they are unable to quantify localised quantities of interest, such as local pressure peaks on the vacuum vessel walls, to which the failure criteria are linked. To calculate local quantities, the 3D nature of the phenomenon needs to be considered. In this work, a 3D transient model of the in-vessel LOCA from a water-cooled blanket is developed. The model is implemented in the commercial CFD software STAR-CCM+. It simulates the propagation of the water jet in the vessel from the beginning of the accident, thus accounting for the phase change of the water, i.e., from the pressurised liquid phase to the vapour phase inside the vessel, being the latter at a much lower pressure than in the blanket coolant pipes. Due to the large pressure ratio (>1000), shocks are expected; therefore, an Adaptive Mesh Refinement (AMR) algorithm is employed. The physical models (in particular, the multiphase model) are benchmarked to a 2D reference problem before being applied to the 3D EU DEMO-relevant problem. The simulation results show that the pressure peaks in front of the vessel walls are not dangerous as they are below the design limit. The entire evolution of the water jet is followed up to the opening of the burst disks, in order to compare the average pressure evolution with that computed with system-level codes. A comparison with the in-vessel LOCA from a helium-cooled blanket is also carried out, showing that the accident evolution in the water case is less violent than in the helium case. [ABSTRACT FROM AUTHOR]

Reduction in ambient pressure within the containment of a light water reactor in the event of a hypothetical Loss of Coolant Accident (LOCA) is crucial for containment integrity. This article introduces the distinct concept of a Passive Containment Cooling System (PCCS). In the proposed PCCS concept, the containment is divided into two compartments with appropriate volumes. An external heat exchanger is connected between the two compartments of the containment for heat transfer and pressure reduction in the case of LOCA. This concept has unique characteristics and provides many advantages over previous designs. A RELAP5 model was developed to perform thermal-hydraulic analysis of the proposed PCCS concept. The performance of the PCCS after its deployment on a small reactor (998.6 MWth) has been assessed under various conditions following LOCA. The results indicate that partitioning of the containment creates a differential pressure that acts as a driving force for flow through the PCCS heat exchanger. Consequently, a decrease in long-term containment pressure is observed. In the case of equal volume fractions in the two containment compartments, the PCCS reduces long-term pressure in the containment by 16.5%. Sensitivity analysis was conducted in which the flow area of the PCCS inlet line, the heat transfer area, and the size of the heat exchanger tank were varied. The PCCS heat transfer capacity increases as these parameters are increased. However, the former is less sensitive than the latter two. The performance of the PCCS heat exchanger under steady-state conditions was validated using RELAP5 simulations, empirical correlations, and analytical models. The evaluation of heat exchanger parameters by various methods (presented in Appendix 1) was found in close agreement. [ABSTRACT FROM AUTHOR]

During a Loss Of Coolant Accident involving a primary circuit failure in a nuclear power plant, debris, generated by the impact of the steam jet, is transported to the hot liquid phase formed at the bottom of the nuclear containment. Water is recirculated after passing through filters. Chemicals are released and may precipitate on the filters, increasing the pressure drop. This paper investigates the characterization of precipitate formation in a soda/borate buffer, due to the presence of calcium, silicon, and zinc, released from mineral fibers, concrete particles, or the corrosion of metal surfaces. Experiments run at various temperatures and chemical compositions are presented, with the objective of identifying the precipitates and determining the critical thresholds of concentration and temperature beyond which the risk of chemical clogging disappears. The experimental results are compared to those calculated with the CHESS software, intended to calculate the speciation at thermodynamic equilibrium. The results show the formation of amorphous precipitates: silica, calcium borates, borosilicates and zinc silicate. Precipitation is temperature dependent, with the highest masses obtained between 50 and 60 °C. The comparison of experimental and calculated results indicates a good overall consistency. This work demonstrates the importance of considering chemical clogging in this context. [Display omitted] • Chemical effects on of sump filter clogging in LOCA situations are studied. • The precipitates are silica, calcium borate, borosilicate and zinc silicates. • Precipitation is influenced by temperature and is maximal around 60 °C. • Speciation calculations (CHESS) predict the experimental results fairly well. • Chemical effects contribute to the clogging of PWR sumps. [ABSTRACT FROM AUTHOR]

The 18th-century Enlightenment period manifests itself in all fields as an attempt to understand and redefine oneself and everything else. This intellectual discovery inevitably takes place in architecture as well. Through the realization of the historicity of architecture, actors who attempt to define architecture within a discipline regime strive to create a holistic language. While each encounter in this journey serves to construct the discipline of architecture, it also includes confrontations that can be destructive. Every devastating encounter is either evaluated in rationalized plane or ignored and expelled. These situations are signs that architecture can be conceived in many different ways. Food-beverage associations, coffee houses, saloons etc., which are different forms of creation of social environment, are significant paradigms of the Enlightenment period. Masonic lodges, built by/in architectural discourses, have an important place in the culture of socialization, shaped in non-institutional forms by their unique structuring, and spread to many parts of Europe in the 18th century. Therefore, this study is intended to be discussed not generation of Masonry but specifically how it became notable formation and how it transform into urban organization through architecture during the Enlightenment. In this context, the study also criticizes the totalitarian construction of architecture. [ABSTRACT FROM AUTHOR]

As part of the pre-conceptual design activities for the European DEMOnstration plant, a carefully selected set of safety analyses have been performed to assess plant integrated performance and the capability to achieve expected targets while keeping it in a safe operation domain. The DEMO divertor is the in-vessel component in charge of exhausting the major part of the plasma ions' thermal power in a region far from the plasma core to control plasma pollution. The divertor system accomplishes this goal by means of assemblies of cassette and target plasma facing units modules, respectively cooled with two independentheat-transfer systems. A deterministic assessment of a divertor in-vessel Loss-of-Coolant Accident is here considered. Both Design Basis Accident case simulating the rupture of an in-vessel pipe for the divertor cassette cooling loop, and a Design Extension Conditions accident case considering the additional rupture of an independent divertor target cooling loop are assessed. The plant response to such accidents is investigated, a comparison of the transient evolution in the two cases is provided, and design robustness with respect to safety objectives is discussed. [ABSTRACT FROM AUTHOR]

Compared with the traditional keyboard and mouse input method, touch screen technique has high convenience for users. With the development of technology, touch screen is widely application in various electronic products because of their high transmittance, clear images, and convenient operation. This review mainly introduces the optical clear adhesive (OCA), liquid optical clear adhesive (LOCA), and the most commonly used LOCA ultra violet (UV)‐curable adhesives. Preparation and performance of UV‐curable adhesives are deeply explored. The structure of touch screen, such as one glass solution (OGS), on‐cell and in‐cell is discussed. Moreover, lamination process includes full lamination and non‐full lamination is also discussed. This review introduces the alignment technique comprise mechanical alignment technique and non‐mechanical alignment technique (such as machine vision) and laminating equipment. The recent development and existing problems with the touch screen lamination process are discussed. Furthermore, this review describes the latest developments in the adhesive of touch screen lamination process and predicts possible future development trends, which could provide references for subsequent research work. [ABSTRACT FROM AUTHOR]

Accident-tolerant fuel concepts have been developed recently in diverse research programs. Recent research has shown clear advantages of Cr-coated Zr cladding over bare cladding tubes regarding oxidation behavior under the design basis loss-of-coolant accident condition. However, limited data are available about the hydriding behavior of the Cr coating. For that purpose, Cr-coated Zricaloy-4 tubes were tested to investigate the effects of hydriding, oxidation, and postquench ductility behavior on coated Zr cladding. A high-power impulse magnetron sputtering (HiPIMS) process was used to produce a high-density coating on the Zircaloy-4 tube surface. Coated and uncoated Zircaloy-4 tube specimens underwent one-sided hydriding in a tube furnace filled with pure hydrogen gas at 425 °C. The tubing specimen ends were sealed with Swagelok plugs before the hydriding runs. For uncoated specimens, H analysis of the hydrided specimens indicated that the H content increased as the test time and initial pressure increased. However, almost no change was observed for the coated specimens that were hydrided under the same test conditions. After one-sided hydriding, the hydrided coated and uncoated specimens were exposed to steam at high temperatures for two-sided oxidation studies to simulate accident conditions. The coated specimens showed a slower oxidation: oxygen pickup was 50% lower than the uncoated specimens tested under the same conditions. Ring compression testing was performed to evaluate the embrittlement behavior of the Cr-coated specimens after hydriding and oxidation. The results indicated that the HiPIMS coating provides excellent protection from hydriding and oxidation at high temperatures. [ABSTRACT FROM AUTHOR]

HPR1000 NPP is designed with Passive Containment Heat Removal System (PCS) following the technical route called active and passive combined concept. The installed PCS system will have effect on the thermal hydraulic (T-H) behavior in the containment, and the complex T-H environment will influence the heat removal and operating character of PCS system backwards. The main goal of this work is to evaluate the effectiveness of the PCS system of HPR1000 and analyse the complex T-H environment of HPR1000 when the PCS system is operating. The HPR1000 containment is modelled by the containment T-H code and the PCS system is modelled based on experiment results. Then, the severe accident sequence initiated by large break LOCA is calculated by an integrated computer code, which outputs mass and energy release sources from reactor coolant primary system as boundaries for the containment T-H code. The results show that HPR1000 PCS system has enough heat removal capability under selected severe accident, which could ensure the containment meet the design requirement. Also, the nonuniformity of temperature and gases in the containment will disappear when the mass and energy releases turn stable, and the operating of PCS system does not cause a huge effect on containment inhomogeneity.

• VIKTORIA: an integral-scale experiment of the sump filtration system. • The main objectives are to investigate the head loss of the filter (physical and chemical clogging) for prototypic upstream debris source term in relevant thermal hydraulic conditions (water temperature and flow velocity on the filter surface) in compliance with the temperature profile and the chemistry of the water in sumps during a LOCA transient. • The tests highlight the settling of the largest particles (concrete, painted chips) and part of the fibers; the transport of debris) leads to the physical clogging of the filter. • The debris carried to the filter generate at 80 °C (with chemistry) a very quick increase of the pressure drop across the filter that could be due to rapid chemical effects further to fibers corrosion. • The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior. • The observations made during these numerous campaigns of tests demonstrate the need to carry out tests at high temperature with the real "chemistry" in the water to highlight potential phenomenon and to provide relevant assessments. During a Loss of Coolant Accident (LOCA), in PWR's, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the ECCS and CSS systems. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop. The main objectives are to investigate the head loss of the filter (physical and chemical clogging) for prototypic upstream debris source term in relevant thermal hydraulic conditions (water temperature and flow velocity on the filter surface) in compliance with the temperature profile and the chemistry of the water in sumps during a LOCA transient. For that, the VIKTORIA loop was equipped successively with two types of 2 m2 filters used in 900MWe NPP's. The tests highlight the settling of the largest particles (concrete, painted chips) and part of the fibers; the transport of debris (roughly 55 to 65 % of the injected debris source term) leads to the physical clogging of the filter. The debris carried to the filter generate at 80 °C (with chemistry) a very quick increase of the pressure drop across the filter (≈7 kPa) that could be due to rapid chemical effects further to fibers corrosion. At the end of the 80 °C plateau, a pressure drop increase was observed due to the temperature decrease in agreement with the water viscosity evolution. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type; ≈1.5 kPa. Nevertheless, downstream debris source term appears to be more significant with the planar filter (compared to the filter with pockets) during the first hour of injection before the fibrous debris bed formation. We have also observed a more stabilized "cake", during the (7 days) medium term evolution compared to that for rectangular pockets filter due to motion of debris inside pockets. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior. [ABSTRACT FROM AUTHOR]

[Display omitted] • Burst testing of IPHWR zircaloy-4 claddings at 55–115 K/s and 15–45 bar. • Developed burst stress correlation as a function of oxygen pickup and burst temperature. • Developed burst criterion model and validated it with experimental data. • Studied the effect of various parameters on the burst behavior of zircaloy-4 cladding. This paper presents a detailed investigation of the creep and rupture behavior of Zircaloy-4 (Zry-4) fuel claddings used in Indian Pressurized Heavy Water Reactors (IPHWR) under simulated Loss-Of-Coolant Accident (LOCA) conditions. Fuel claddings are pre-oxidized in the steam environment at 500 °C to mimic the oxygen pickup during normal reactor operation. The burst tests are then performed on these preoxidized tubes at different heating rates (55–115 K/s) and internal overpressures (15–45 bar) in the steam environment, creating LOCA-like scenario in a high burnup condition, wherein the claddings further oxidize while undergoing deformation and rupture. The burst stress correlation is developed for IPHWR claddings from the obtained burst temperature and oxygen concentration data. A burst criterion model is developed by solving available creep rate, oxidation rate, and phase transformation equations simultaneously to study the effect of various parameters on burst characteristics of the fuel cladding. The proposed burst criterion model agrees well with the present and previous experimental burst data. Also, the ballooning progression predicted from the model is validated with the present and previous experimental data. In addition, the effect of available correlations for the creep rate, phase boundary temperature, and oxidation factor on the burst characteristics has been presented. [ABSTRACT FROM AUTHOR]

One of the coming Small Modular Reactor (SMR) designs is the integrated Pressurized Water Reactor (iPWR), which merges years of knowledge and experience in light water reactors with new demands from the market, such as flexibility, construction optimization, and reliability. Generally, SMRs are considered to be inherently safe and, in many cases, safer than generation III designs. Most of the SMR designs rely on passive safety systems at different levels of passive operation. One of the ways to examine new concepts is a numerical simulation using a reliable tool. However, when new features come up, there are also new challenges for existing tools, even though they are continuously updated following the technology's evolution. To investigate if implemented new features give expected results and to increase knowledge about SMRs modeling and simulations, accident transients of iPWR were calculated using MELCOR 2.2 code and analyzed. The work aims to understand better numerical simulations of the analyzed iPWR accident scenarios and code reliability. To do so, the paper is divided into two parts; in Part 1, the authors present MELCOR 2.2 iPWR input deck description, including nodalization and steady-state conditions as well as the analysis of an iPWR LOCA-type scenario in the design basis accident domain. In addition, design basis accident progression is compared to beyond design basis scenario in which safety systems are postulated partly failing. The analyses of beyond design basis scenarios in which core degradation is observed are published separately in second part of that paper. The obtained results indicate that MELCOR2.2 can simulate the thermal–hydraulic response of an iPWR in accident scenarios. Values received in this study in steady state and accident simulations are mostly expected and coherent with general knowledge of accident progression. [ABSTRACT FROM AUTHOR]

• The SCO 2 break blowdown model was established, and the LOCA progress was simulated. • After reactor outlet LOCA, the mass flow is low but the reactor temperature is raised. • The compressor outlet LOCA is the most severe, with the highest blowdown flow rate. • The pre-pressurized containment can effectively reduce the severity of LOCA. • If the break size is further increased, system supercritical failure is inevitable. The supercritical CO 2 (SCO 2) Brayton cycle has the advantages of high efficiency, flexibility, and safety. It is highly compatible with Generation-IV reactors, making it one of the most promising energy conversion systems. However, the special physical properties of the working medium will lead to complex blowdown behavior when loss of coolant accident (LOCA) occurs, and the system performance needs to be clarified. In this work, the SCO 2 blowdown model is developed. Combined with the system analysis program, the system transient behavior during the LOCA process is simulated, and the threat to the main equipment is evaluated. The results show that the accident process is closely related to the SCO 2 state at different break locations. When LOCA occurs at the reactor inlet and outlet, the break discharge rate is small, but the reactor fuel temperature will increase significantly at the accident late stage. The compressor outlet LOCA has the largest discharge rate and the shortest coolant phase transition time (720 s, break diameter is 4 mm). In addition, the pre-pressurized containment can effectively reduce the amount of coolant leakage and maintain SCO 2 in the supercritical state. However, for the compressor inlet LOCA, if the break area is further increased, even with the protection of the containment and inventory control system, the supercritical failure of SCO 2 is inevitable. This study can provide a reference for the parameter response and accident process of the SCO 2 Brayton cycle reactor system during LOCA. [ABSTRACT FROM AUTHOR]

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. A large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk. • Fuel performance depends on thermal hydraulic behavior during severe accidents. • The impact of using subchannel methods to predict loss-of-coolant behavior was investigated. • Using higher fidelity methods allows for a more accurate determination of clad surface temperature behavior. [ABSTRACT FROM AUTHOR]

• To validate the newly implemented FFR model in the SPACE code, the Halden IFA-650 test is calculated. • Overall prediction with the FFR model in SPACE is well matched with experimental results. • For some tests, the best estimated clad deformation model is necessary to validate the FFR model better. • Sensitivity analyses are conducted for fragmentation, packing fraction, and threshold gap to better understand the FFR model in SPACE. The new ECCS regulations were recently considered in South Korea. One of the major concerns in this revision is the consideration of fuel fragmentation, relocation, and dispersal (FFRD) phenomena during a loss of coolant accident (LOCA). In response, the FFR model was newly implemented in the SPACE code, which is a safety analysis code. The FFR model based on the QT model, which was developed by Quantum Technology Corporation, was developed with a threshold size of fuel fragments for fuel relocation, fuel power distribution, and so on. In this study, to validate the newly developed FFR model in SPACE code, calculations using Halden IFA-650 tests are conducted. Overall, the results show that the newly developed FFR model yields results that are consistent with experimental results in terms of fuel relocation and their effects on cladding temperature. However, when the cladding deformation is not well predicted, the fuel relocation prediction is poor. Thus, the cladding deformation is the most critical parameter for the FFR phenomena. In addition, for a fresh fuel condition, the threshold fragment size is additionally proposed in the fuel relocation model. For a better understanding of the FFR model behavior during a LOCA, some sensitivity tests for major FFR model parameters are conducted. This can give valuable information to develop FFR modeling in safety analysis methodology. [ABSTRACT FROM AUTHOR]

• New maximum curvature method to capture quench temperature for bottom reflood test. • Performance evaluation of Cr-coated ZIRLO cladding under simulated LOCA condition. • Modification of existing quench temperature correlation for thin wall cladding. • Multiple quench tests to investigate the mechanical integrity of Cr-coated cladding. Chromium (Cr) coated cladding is the most promising candidate for near-term accident-tolerant fuel (ATF) for current light water reactors (LWRs) because of its excellent oxidation resistance at high temperatures. For deployment of the Cr-coated cladding as ATF, the performance should be evaluated under various conditions, including the loss-of-coolant accident (LOCA). In this research, a single-rod reflood facility was constructed to investigate the quenching behavior and performance of Cr-coated cladding under bottom reflooding at high temperature conditions up to 1200 °C. Quench tests were performed using an uncoated, cold spray Cr-coated, and PVD Cr-coated Zr-alloy at various initial temperature ranges from 600 °C to 1200 °C with 20 K of water subcooling and 4.5 cm/s of reflood velocity. The quench temperatures were obtained at four different axial locations based on the maximum curvature method developed in this study. The bottom reflood quench test results indicate that the quench temperature increases with increasing initial cladding temperature and decreasing axial distance from the bottom. The quench temperature was not notably different for uncoated Zr-alloy and Cr-coated Zr-alloy, regardless of the coating deposition method. However, above 1000 °C, the Cr-coated Zr-alloy showed a consistent and predictable quench temperature in contrast with the uncoated Zr-alloy for which the quench temperature data were severely scattered. The underestimation by the existing quench temperature model was observed with the current test using typical cladding thickness (0.57 mm). The thin cladding thickness was suggested as the reason for the underestimation. The modified Kim and Lee's correlation is proposed for better prediction of the thin-walled Zr-alloy cladding. The Cr-coating appears to mitigate the degradation of the mechanical integrity of the underlying Zr-alloy tube under severe oxidation and repeated thermal cycling, by virtue of superior oxidation resistance. It indicates that the coated cladding provides better integrity and performance in power transient conditions such as load-following operation. Post-test characterization showed the Cr-coating oxide layer thickness to be an order of magnitude lower than that of the uncoated cladding. The change in wettability was minimal in all test cases despite the trend of increasing roughness with the increasing initial temperature. [ABSTRACT FROM AUTHOR]