Tube Rupture Research Articles

Review of Design Extension Conditions Experiments and Analyses for Non-degraded Core

The second generation nuclear power plants were designed and built to withstand without loss to the systems, structures, and components necessary to ensure public health and safety during design basis accidents (DBAs). In the transient and accident analysis the effects of single active failures and operator errors were considered. There are also accident sequences that are possible but were judged to be too unlikely and therefore were not fully considered in the design process of second generation reactors. In that sense, they were considered beyond the scope of design-basis accidents that a nuclear facility must be designed and built to withstand. Such accident sequences have been analysed in the past to fully understand the capability of a design. The requirements to analyse such sequences for existing reactors have been introduced after Fukushima Dai-ichi accident. In 2012 the design extension conditions (DECs) were introduced in the International Atomic Energy Agency (IAEA) requirements for the design of nuclear power plants (NPPs). Western European Nuclear Regulators Association (WENRA) requirements of existing reactors for DEC were introduced in 2014. The purpose of considering DEC is to further improve safety by enhancing the plant’s capability to withstand the conditions generated by accidents that are more severe than DBAs. This concept by IAEA and WENRA (WENRA definition of DEC is consistent with IAEA definition from 2012, in which DEC with prevention of core melt is called DEC A) is not completely new, since some multiple failures have already been considered in the design of existing reactors, for example anticipated transients without scram and station blackout. The research for beyond design basis accidents with non-degraded core (i.e. DEC A) for existing reactors has been already done in 80’s and 90’ of the previous century. The purpose of this paper is to review that research. The tests performed include total loss of feedwater, station blackout, small break without high pressure safety injection, steam generator tube rupture with no high pressure safety injection etc. Besides review of experiments performed on integral test facilities, examples of DEC A tests, which have been analysed at Jožef Stefan Institute using RELAP5 or TRACE computer code in the last three decades, will be presented too.

A receptor-channel trio conducts Ca2+ signalling for pollen tube reception.

Precise signalling between pollen tubes and synergid cells in the ovule initiates fertilization in flowering plants1. Contact of the pollen tube with the ovule triggers calcium spiking in the synergids2,3 that induces pollen tube rupture and sperm release. This process, termed pollen tube reception, entails the action of three synergid-expressed proteins in Arabidopsis: FERONIA (FER), a receptor-like kinase; LORELEI (LRE), a glycosylphosphatidylinositol-anchored protein; and NORTIA (NTA), a transmembrane protein of unknown function4-6. Genetic analyses have placed these three proteins in the same pathway; however, it remains unknown how they work together to enable synergid-pollen tube communication. Here we identify two pollen-tube-derived small peptides7 that belong to the rapid alkalinization factor (RALF) family8 as ligands for the FER-LRE co-receptor, which in turn recruits NTA to the plasma membrane. NTA functions as a calmodulin-gated calcium channel required for calcium spiking in the synergid. We also reconstitute the biochemical pathway in which FER-LRE perceives pollen-tube-derived peptides to activate the NTA calcium channel and initiate calcium spiking, a second messenger for pollen tube reception. The FER-LRE-NTA trio therefore forms a previously unanticipated receptor-channel complex in the female cell to recognize male signals and trigger the fertilization process.

Foaming biocompatible and biodegradable PBAT/PLGA as fallopian tube stent using supercritical carbon dioxide

• A macroporous PBAT/PLGA foam was prepared using supercritical CO 2 . • The foam displayed suitable mechanical property for the fallopian tube stent. • A rapid biodegradation rate was observed in simulated internal environment. • Implantation in the fallopian tube verified good biocompatibility of the foam. Tubal pregnancy is a common abnormal pregnancy manifestation, and the ordinary conservative treatment of tubal adhesion usually leads to the rupture of fallopian tube, which increases the risk of a second ectopic pregnancy. To avoid this symptom, it is suitable to implant a stent to separate the adhesion. Here we prepared the PBAT/PLGA foam as the stent material using supercritical CO 2 foaming technology. With uniform macroporous structure and thin-wall feature, the foam possessed low compressive modulus in prevention of the possible second injury to the fallopian tube. The introduction of PLGA 50/50 improved the biodegradable capability of the foam, with a mass loss about 20% after a 12-week hydrolysis. After implanted into the ruptured fallopian tube of the rabbit model, the foam displayed excellent biocompatibility, and provided a good support to prevent tubal adhesion. As such, this work provides the foam material as a promising candidate for fallopian tube stent to remedy the tubal adhesion.

Transient Simulation of an Industrial Steam Boiler

The safety analysis of an industrial installation is extensively based on modeling and simulation. However, the analysis of simulation using realistic computer codes like RELAP5/Mod3.2 will help understand thermal-hydraulic behavior of an installation during normal and accidental conditions. In a steam boiler operation, it is very important to evaluate numerous accident scenarios under real plant conditions. One of the main accidental transients is tube rupture of steam boiler. The main objective of this study consists toinvestigate the behavior of an industrial steam boiler installed in the complex of natural gas liquefaction (LNG) at Skikda in Algeria during feed water line break accident. A RELAP5 model was set up to simulate the entire system, the model represents all steam boiler components to be suitable for the analysis of the several accidents. The control and regulation systems are also considered. The model was qualified against the steam boiler data at steady-state conditions. As seen from the results, a good agreement was obtained. The transient simulation results show that the thermal-hydraulic code correctly predicts the behavior of the main steam boiler parameters and how the control system, when required, can successfully mitigate the accident.

Updating risk model for SGTR accident based on success criteria analysis

Abstract Success criteria analysis plays a key role in the development of realistic probabilistic safety/risk assessment (PSA/PRA) model because it provides supporting information regarding the response of complex nuclear power plant systems to different accident conditions. The current paper performs plant specific success criteria analysis for steam generator tube rupture (SGTR) accident in a typical pressurized water reactor (PWR) and demonstrates implementation of the obtained best estimate results on a risk model which has been initially developed based on expert judgment and general plant design data. The modifications on the risk model include configuration of the safety systems as well as the event tree structure. The updated PSA model shows 50% reduction in the plant core damage frequency (CDF) in comparison to the base case risk model. This highlights the importance of success criteria analysis for the development of a realistic PSA model in risk informed applications.

Simulation of the processes occurring during steam generator tube rupture in the lead cooled reactor unit using HYDRA-IBRAE/LM code

The article is devoted to the new and improved models, implemented in system thermohydraulic HYDRA-IBRAE/LM code, being developed in IBRAE RAN, which make it possible to carry out more realistic simulation of thermohydraulic parameters in case of a steam generator tube rupture in reactor plant with lead coolant: a model to estimate initial bubble diameter, a model of steam-water mixture behavior in liquid lead and interfacial area transport equation to define the evolution of the dispersed phase size.The paper also contains the models validation results on experiments conducted in Kutateladze Institute of Thermophysics Siberian Branch of the Russian Academy of Sciences (IT SB RAN). The experimental researches were devoted to the study of the initial bubble size dependence on the gas flow rate and the heat exchange between steam-water mixture and liquid lead. The comparison of calculation results and experimental data demonstrates not only qualitative but also good quantitative agreement.

Evaluation of optimized machine learning models for nuclear reactor accident prediction

Several studies have proposed machine learning models to diagnose and predict accidents in nuclear power reactors. However, the training data in these studies are deterministic, and they don't consider the parameter uncertainty caused by sensor failure during an accident. The performance of the ML must be weighted with the uncertainties and possible malfunctions of the detectors to be considered feasible. Consequently, this work presents a novel training approach for machine learning models using an augmented dataset that reflects the sensor status. In this study, three machine learning models, Support Vector Machines (SVM), Decision Trees (DT) and Multilayer Perceptron (MLP), are developed, trained and compared. By simulating the loss of coolant accident (LOCA) and Steam Generator Tube Rupture (SGTR) accident, several augmented reactor coolant parameters were used to train and test the machine learning models to predict reactor accidents. The model performances were evaluated using the F1 score, precision, accuracy, learning curves, and the Confusion Matrix to determine the suitable algorithm for accurate diagnosis of the two reactor accidents. From this study, we conclude that the SVM and DT models performed better than MLP for these accident scenarios.

Study of transient events during the full-system chemical decontamination operation of the Kori-1 nuclear power plant

The Kori-1 nuclear power plant (NPP), a Westinghouse 2-loop pressurized water reactor in Korea, permanently ceased operations on June 18, 2017. Before the main decommissioning activities, full chemical decontamination of primary systems will be performed to secure the safety of workers and minimize the quantity of radioactive waste. In this study, the steady and transient flow characteristics for the chemical decontamination operations were analyzed via a simulation using the Multidimensional Analysis of Reactor Safety, KINS Standard version. Then, based on the accumulated flow mass released from the reactant cooling systems during transient events, the radioactivity levels were estimated using the oxide layer removal ratio obtained from the chemical decontamination experiment. The loss of residual heat removal, steam generator tube rupture (SGTR), power-operated relief valve open (PORV Open), and combination of SGTR and PORV Open simulations were investigated in terms of the transient events. Finally, a strategy for managing transient events during the full-system chemical decontamination operation of the Kori-1 NPP is proposed.

Heat exchanger network synthesis with process safety compliance under tube rupture scenarios

While there exists considerable research on HEN Synthesis (HENS), little attention has been paid to incorporating process safety of HENS at the conceptual design stage. In this work, we extend the optimization-based HENS to incorporate safety using a Safety Rating (SR) to safely navigate dynamic tube rupture scenarios for all plausible heat exchanger matches. This metric allows one to assess the risk of overpressure resulting in failure of the equipment. To avoid such complexity, we develop a piecewise linear approximation of SR using data obtained from a dynamic tube rupture model. Our approximation is such that it always provides guaranteed and tight underestimation of the SR values, thereby ensuring process safety. Applying the methodology, we obtain HEN configurations with lowest costs that can safely handle overpressure events for the entire network. By rethinking how to mitigate process safety challenges using systems approaches, plants can incorporate safer designs for minimal cost.

Dynamic response of square recycled aggregate concrete-filled steel tube columns subjected to close-range blast loads

To study the dynamic response of square recycled aggregate concrete-filled steel tube (RACFST) columns under explosion loads, explosion examinations and numerical analyses were performed. Based on the measured longitudinal strain and residual displacement of the columns, the effects of recycled aggregate replacement percentage, steel tube thickness, explosive charge weight, blast center distance, and explosion height on the dynamic response of square RACFST columns are analyzed, and the failure modes are investigated. Then, the established numerical model was verified through test results, and further explore the dynamic response of square RACFST columns. It indicates that square RACFST columns suffer localized damage after detonation, and no obvious global deformation occurs. The localized failures, including blast pit, steel tube bulge or rupture, and weld fracture, could be observed on these columns. The square RACFST columns strain and displacement values are positively correlated with TNT equivalent and RA content, while negatively correlated with steel tube thickness and blast center distance. Furthermore, the concrete absorbs most of the total energy, implying the important role of core concrete in the explosion resistance of RACFST columns.

Flexural performance of cast-in-place concrete-filled steel tube piles under varying axial load

Concrete-filled steel tube (CFST) piles with thin-walled tubes are widely used as foundations for buildings and bridges in high seismic regions. To design these piles for large earthquakes, it is necessary to test specimens of realistic size under axial loads representative of actual design conditions. However, previous experimental studies of CFSTs were carried out mostly on small-diameter specimens with small or no axial load. To address this issue, cyclic flexural tests were conducted on five large-scale CFST pile specimens under varying axial load, with the range of axial load ratio (−0.21 to 0.38) and internal reinforcement as test variables. The specimens had a diameter of 1200 mm and diameter-to-thickness ratio of 125. This study discusses issues relevant to the design and performance assessment of piles such as moment capacity, hysteretic response, damage progression, strain distribution, and strain limits. It is observed that the ultimate condition of CFST piles is governed by the rupture of the tube, and these piles retain axial load-carrying capacity even after the tube ruptures. Internal reinforcement is effective to improve the deformation capacity and energy dissipation of piles. These findings contribute towards the rational seismic design of CFST pile foundations for large earthquakes.

증기발생기 세관파단사고 모의 시험설비 내 에어로졸 손실량 예측을 위한 PLC 모델링

The amount of aerosol retention in the pipe of the test facility was analytically evaluated in the steam generator tube rupture experimental facility installed in the Korea Atomic Energy Research Institute (KAERI). The amount of aerosol removal in the pipe can be varied greatly depending on the thermo-hydraulic conditions (gas velocity, pressure, temperature, etc.) of the gas containing the aerosol. It is also known that the geometry of the pipes, such as length, curvature, and diameter can be important factors in determining the amount of aerosol retention inside the pipe. Thus, sensitivity calculations were performed on various uncertain variables to identify a decontamination factor of the test facility. In this study, the PLC (Particle Loss Calculator) program was used to evaluate the decontamination factor of aerosol inside the pipes. The calculation results were also compared to the experimental data obtained from the KAERI. It was found that the analysis results underestimate the decontamination factor due to the limitation of the models considered in the PLC program.

Prediction of thermohydraulics in shell side of CAP1400 steam generator by two-fluid porous media model

Three-dimensional thermohydraulics in the shell side of the u-tube steam generator are essential for the thermal design and tube flow-induced vibration analysis, which are crucial for the equipment economy and integrity. In this work, the Eulerian two-fluid framework along with porous media model was employed to predict the thermal-hydraulic parameters in the shell side of the CAP1400 steam generator. The two-phase flow in the shell side was treated as flow in porous media with additional momentum and energy source terms for each phase to consider the effects of structure resistance and primary-to-shell heat transfer. The primary side was simplified by the multi-channel model with 16 equivalent tubes. The distributions of thermal-hydraulic parameters for each phase and the mixture phase were obtained. High non-uniformity can be observed for the flow and heat transfer parameters. The ratio between heat released in the hot side and cold side was as large as 2.7. The crossflow energy related to the flow-induced vibration was also derived to find the possible location of tube damage. The most probable location for a potential tube rupture locates at approximate 0.4 rad in the hot side.

An HFM‐CREAM model for the assessment of human reliability and quantification

AbstractHuman reliability analysis (HRA) is a proactive approach to model and evaluate systematic human errors and has been extensively implemented in various complicated systems. The assessment of human errors relies heavily on the knowledge and experience of experts in real‐world cases. Moreover, there are usually specific sorts of uncertainty while experts use linguistic labels to evaluate human failure events. In this context, this paper seeks to establish a new model based on the hesitant fuzzy matrix (HFM) and the cognitive reliability and error analysis method (CREAM) to conduct a quantitative analysis of human errors. This model handles the multiple crisp scores of the common performance conditions (CPCs) given by experts according to the context description in terms of CPCs, determines the weights of CPCs by the HFM, and elicits the human error probability (HEP) point estimation formula considering consequences based on the CREAM. Finally, the effectiveness and practicality of the presented HFM‐CREAM model are demonstrated through the emergency response analysis of the steam generator tube rupture (SGTR) in nuclear power plant.

RALF peptide signaling controls the polytubey block in Arabidopsis.

Fertilization of an egg by multiple sperm (polyspermy) leads to lethal genome imbalance and chromosome segregation defects. In Arabidopsis thaliana, the block to polyspermy is facilitated by a mechanism that prevents polytubey (the arrival of multiple pollen tubes to one ovule). We show here that FERONIA, ANJEA, and HERCULES RECEPTOR KINASE 1 receptor-like kinases located at the septum interact with pollen tube-specific RALF6, 7, 16, 36, and 37 peptide ligands to establish this polytubey block. The same combination of RALF (rapid alkalinization factor) peptides and receptor complexes controls pollen tube reception and rupture inside the targeted ovule. Pollen tube rupture releases the polytubey block at the septum, which allows the emergence of secondary pollen tubes upon fertilization failure. Thus, orchestrated steps in the fertilization process in Arabidopsis are coordinated by the same signaling components to guarantee and optimize reproductive success.

Safety Analysis in VVER-1000 Due to Large-Break Loss-of-Coolant Accident and Station Blackout Transient Using RELAP5/SCDAPSIM/MOD3.5

It is crucial to do safety evaluation of different postulated transient scenarios in actual nuclear power plants (NPPs). Some of the common analyzed scenarios are primary coolant tube rupture and station blackout (SBO). In this research, it was supposed that after establishment of a steady-state condition, an instantaneous guillotine large-break loss-of-coolant accident (LB-LOCA) of 850-mm inside diameter in one of the reactor vessel cold legs occurred, accompanied with SBO. The event progression and the variation of different reactor parameters like loop pressures, mass flow rates, fuel and clad temperature, injection rate of accumulators (ACCs), decay, and reactor power were investigated using the RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program. The reactor consequences due to availability and unavailability of passive ACCs were compared. These kinds of analyses assist in estimating the time available to perform operator safety actions. This in turn aids in emergency planning and severe accident management. The results reveal that fuel damage decreased after the introduction of ACCs. Actuation of ACCs at their actuation setpoints provided core cooling by injecting water into the reactor core. However, ACCs alone are inadequate to contain long-term core cooling during a persistent LB-LOCA. The results obtained in the research were compared with MELCOR 2.1 and ASTEC V1.3, and a cohesive agreement was obtained. Therefore, RELAP5/SCDAPSIM/MOD3.5 is capable of modeling a LB-LOCA and SBO in VVER-1000, and it provides a significant analytical capability of safety systems for specialists in the field in NPP safety.

Separate and integral effect tests of aerosol retention in steam generator during tube rupture accident

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release.AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences.The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number.The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

Investigation on GFRP Pultruded tube confined double skin tubular columns under axial loading

Composite construction with concrete filled steel tubes (CFT) as structural columns gained prominence in the recent construction scenario to make use of the benefits of both steel and concrete. Double skinned tubular (CFDST) filled with concrete between tubes is subjected to passive confinement pressure increasing the load carrying capacity of columns. This paper delineates the effect of confinement by outer steel and GFRP tube on the axial load capacity on CFDST short as well as slender columns made of steel inner tube. The experimental study was carried out on twenty composite columns consisting of ten numbers of double skin tubular short columns and ten numbers of double skin slender columns (with slenderness ratio 12.5). Ten short columns consist two pairs of columns with outer tube material as steel tube, two pairs of columns with outer tube materials as GFRP tube and one pair of reinforced concrete column. Diameter of the outer tube to thickness ratio (Do/t) of 21 and 30 is considered for all the columns irrespective of the length. The finite element analysis was undergone by modelling the columns in the Abaqus software, simulated the experimental work in the analytical study and the results are compared in terms of load deflection curves and failure modes of the columns. The SCFDST short columns failed by local buckling of outer tube whereas the failure of FCFDST short columns failure is characterized by rupture of the outer tube. The column with Do/t ratio 21 confines the concrete significantly than the tubes with Do/t ratio 30 irrespective of the outer tube material. The slender columns failed by global buckling and the ultimate load taken by the slender columns is reduced by 23% compared to short columns in SCFDST and 28% in FCFDST columns.

Death due to rupture of tubal pregnancy

Ectopic pregnancy (EP) is defined as a pregnancy in which the implantation of the embryo occurs outside the uterine cavity, most frequently in one of the two fallopian tubes or, more rarely, in the abdominal cavity. The result of an undiagnosed ectopic pregnancy is rupture of the fallopian tube with intra-abdominal bleeding, leading to hemorrhagic shock and death. Here, we present a case of 26-yearold recently married female who was brought for postmortem examination with history of sudden death following pain abdomen. Postmortem examination revealed findings suggestive of ectopic pregnancy. Histopathology revealed multiple vascularized branching villi lined trophoblasts and syncytiotrophoblasts along the lumen. The significance of ectopic pregnancy lies in the fact that it often goes undiagnosed with massive intraperitoneal bleeding. This case emphasizes the need for mandatory pregnancy screening in female patients of reproductive age groups who present with abdominal pain.

Simulation Research of Once-through Steam Generation Tube Rupture in an Improvement Model

Simulation Research of Once-through Steam Generation Tube Rupture in an Improvement Model