Passive Safety Features Research Articles

Safety Trends in the Automobile Industry in India

The automobile industry in India has experienced significant advancements in recent years, driven by rapid technological innovations and evolving consumer expectations. Safety has emerged as a critical focus area, with stakeholders ranging from government bodies to manufacturers and consumers increasingly emphasizing the need for stringent safety measures. This paper explores the key safety trends shaping the Indian automotive sector, including the adoption of advanced driver-assistance systems (ADAS), the implementation of new crash test standards, and the increased integration of passive and active safety features in vehicles. Additionally, it highlights the regulatory changes brought forth by the Indian government, such as the Bharat New Vehicle Safety Assessment Program (BNVSAP) and mandatory safety norms like airbags and anti-lock braking systems (ABS). The study also examines the role of technological innovations, including connected vehicle technologies and the potential of autonomous driving in enhancing road safety. By analyzing these trends, the paper aims to provide insights into how safety measures in the Indian automobile industry are evolving to meet global standards and reduce the rising number of road accidents in the country.

Transient thermal hydraulic analysis of a small passive lead–bismuth eutectic cooled fast reactor

The lead–bismuth eutectic cooled fast reactor (LFR) emerges as a cutting-edge technology in nuclear reactor design. This study introduces a compact, passive, and long-life LFR reactor LFR-180. The LFR-180 incorporates modular helical-coiled once-through steam generators (H-OTSG) and direct reactor auxiliary cooling systems. This integration enables the generation of superheated steam and passive safety features of the LFR-180, eliminating the necessity for a water-steam separation device and contributing to enhanced power generation efficiency. A detailed transient thermal hydraulic analysis is conducted with tailored models for LBE solidification and H-OTSG thermal hydraulics. Results demonstrate that LFR-180 is a fully passive reactor, requiring no intervention up to 149.1 h after shutdown. The H-OTSG can serve as the heat exchangers of the active safety systems that can effectively reduce the peak core outlet temperature from 1003.6 K under the SBO accident to 724.2 K after shutdown.

Investigation of Natural Helium Circulation Inside Dual Channels Prismatic Modular Reactor

Natural helium circulation inside a dual-channel prismatic modular reactor was numerically investigated to comprehend the passive safety features using the new RHYS/ASYST VER4 package. The natural circulation was studied at Depressurized Cooldown (DC) at heat fluxes of 860–2293 W/m2 with a pressure of 413.7 kPa and Pressurized Cooldown (PC) at pressures of 413.47–689.12 kPa with a flux of 1720 W/m2. A reversal of the heat transfer near the riser exit was seen, combined with a decrease in helium and riser wall temperatures due to the end effect. The pressure has an inverse effect on the heat transfer, where the temperature decreases by 19 % as the pressure increases from 413.47 to 689.12 kPa. The heat transfer was studied under the DC and PC. Both entrance and end effects are introduced with the heat transfer coefficient. Also, the nondimensional groups (Nu, Gr, Pr, and Ra) were discussed.

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO2) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. The reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

Transient Analysis of LBLOCA with Subsequent Loss of Offsite Power and Loss of All AC Power in VVER-1200 Reactor

The generation III+ VVER-1200 (AES-2006) reactor is equipped with improved safety features with component redundancy compared to the previous models. This paper is concerned with the performance of active and passive safety features during the transient of a major beyond design basis accident (BDBA) scenario of the reactor. The study is done considering large break loss of coolant accident (LB LOCA) with loss of offsite power (LOOP) and LB LOCA with loss of all AC power (LOACP) sources namely station blackout (SBO) using VVER-1200 NPP simulator. It was found that the reactor is capable of mitigating such BDBA scenarios by the effectiveness of the active and passive safety systems. Moreover, LB LOCA with LOACP is observed to be more severe compared to LBLOCA with LOOP. This comparative analysis is compatible with IAEA safety guidelines.

Multiphysics Analysis of PLOFC and DLOFC Accidents in the HTGR-TR

High-temperature, gas-cooled reactors (HTGRs) are promising advanced reactor designs. Leveraging passive safety features, higher core outlet temperatures, previous commercial operational experience, and potential coupling with nonelectrical systems, HTGRs present many advantages over traditional light water reactors and other advanced reactor designs. The High-Temperature, Gas-Cooled Test Reactor (HTGR-TR) is a point design developed by Idaho National Laboratory as a response to the need for the expansion of U.S. test reactor capabilities. The HTGR-TR was designed to serve as a test reactor for Generation-IV small modular prismatic block HTGRs and to provide irradiation data, two crucial contributions to the development of advanced reactors. As in any reactor design, it is necessary to understand the behavior of the reactor during accident scenarios to determine the overall safety of the design. Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two accidents that can occur in HTGRs and that present challenges to decay heat removal. In order to understand the behavior of the fuel during these accidents, the resulting mechanical behavior and fission product migration in the fuel are evaluated. Using the HTGR-TR design, which utilizes tristructural isotropic (TRISO) fuel, a multiphysics analysis during the PLOFC and DLOFC transients was performed. This study utilizes neutronics, thermal hydraulics, and fuel performance modeling to analyze the behavior of TRISO particles during steady-state operation and the described transients. This analysis aims to characterize the hoop stress and strain during the transients, as well as predict fission product migration and radiological release. Due to the startup conditions of the HTGR-TR, the silicon carbide (SiC) layer for both of the transients and steady-state conditions are initially in compression for both the hoop stress and strain. The DLOFC transient at the beginning of cycle was the most challenging overall to the fuel, having the closest to tensile behavior and the closest to tensile strain. Because the integrity of the SiC layer was not challenged during any of the transients, the release of 90Sr and 137Cs was found to be negligible. The PLOFC transient at the end of cycle resulted in the largest 110mAg release and subsequent radioactivity increase. However, the increase in radioactivity was minor when compared to typical operating conditions due to the relatively large diffusivity of 110mAg though SiC at normal operating temperatures. Highlights include Neutronics and thermal-hydraulic modeling generating steady-state and transient conditions for use in fuel performance analysis.2. Detailed BISON analysis demonstrating the excellent performance of TRISO fuel during PLOFC and DLOFC transients.3. Fuel behavior showing negligible fission product release.

Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor

Abstract 200 MWth lead-cooled fast reactor (LOTUS) reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at Vietnam National University (VNU) University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refueling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the startup core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers are withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refueling. Good neutronic behavior of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add to the burnable poison rods.

Comparative Analysis of Controller Effect on Anti-lock Braking System Performance using MATLAB/ Simulink

Active safety features aim to avoid or reduce road accidents, in contrast to passive safety features, which are meant to protect occupants after a collision. An Anti-lock Braking System (ABS) is a category of active safety breaking that stops the wheels from locking during a braking action, keeping them in contact with the ground. Different researchers have proposed various control methods for developing the ABS controller. A controller is effective if it reduces stopping distance and time and maintains the desired slip ratio. In this research paper, a quarter-vehicle model is utilized to evaluate the effect of the controller on the Anti-lock Braking System (ABS) and study longitudinal braking dynamics. The Bang-bang controller, Proportional Integral Derivative (PID) controller and the Fuzzy Logic (FL) controller for the ABS were modelled in MATLAB/ Simulink software. A feedback controller that abruptly flips between the two states is called a bang-bang controller, also called an on-off, two-step, or hysteresis controller. PID (Proportional - Integral - Derivative) parameters were obtained using the Ziegler Nichols tuning Method. A comparative analysis of the controllers concerning braking distance, braking time, and normalized relative slip was performed. ABS with the Fuzzy Logic (FL) controller outperforms both the Proportional - Integral - Derivative (PID) and bang-bang controllers.

Thermal-Hydraulic comparative analysis of G-III + BWR and PWR passive safety features under Loss of coolant accident

As one of symbolic design features of Generation-III and III+ (G-III and G-III + ) nuclear reactors, the passive safety system has been implemented on majority of G-III and III + PWRs and BWRs. The passive safety systems in those reactors are designed differently yet adhere to identical fundamental principles. It would be interesting to examine the similarities and differences between the passive safety systems of G-III + PWR and BWR. System code simulations for Small Break and Large Break Loss of Coolant Accident for the Economic Simplified Boiling Water Reactor (ESBWR) and China’s advanced PWR had been conducted, respectively. The results showed that the passive safety systems of ESBWR and China’s Advanced PWR were comparable in design, and when dealing with the similar type of SBLOCA and LBLOCA, the passive safety components with the same functionality could all contribute to the core's cooling with stability and reliability, and the injection flow rate values were nearly equivalent. Separating the injection lines of ESBWR's safety systems guarantees that the action sequence could well remain stable. This design could prevent the injection flow rate and action sequence from being impacted by the interaction of safety injection systems. This work could serve as a reference for the design of subsequent new reactors, thereby contributing to the design of reactors with enhanced economics and safety.

Comparison on Safety Features among HTGR’s Reactor Cavity Cooling Systems (RCCSs)

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

Safe and effective use of a passive safety needle by healthcare professionals in a simulated environment, including perceptions and preferences

ABSTRACT Background Needlestick injuries (NSIs) may potentially expose healthcare professionals (HCPs) to bloodborne pathogens. Safety needles are designed to protect against NSIs. We evaluated whether a new fully passive safety needle could be used safely by HCPs. Research design and methods The passive safety needle was tested by physicians, nurses, and pharmacists in subcutaneous or intramuscular injection scenarios in simulation studies (1–3). Data collected included successes, close calls, difficulties, use errors, and failures. In study 4, HCPs rated the device safety (21-item questionnaire). Results Overall, 104 participants completed 4772 simulated tasks, including 932 injections. 915 injections (98.18%) were performed successfully and no NSIs (0%) were observed in any of the studies. Studies 1 & 2: 84.15% tasks and 96.06% injections were completed successfully, but use errors occurred, mostly arising from the participants’ mental model. There were no failures in Study 3. In Study 4, >98% of participants responded positively to every question, while all felt that the passive safety feature could eliminate NSIs and would better protect against bloodborne pathogens than other existing devices with active or semi-passive safety mechanisms. Conclusions The passive safety needle was used successfully by HCPs, did not lead to any NSIs, and was rated as the safest compared to similar devices.

Core design and neutronic analysis of a long-life LBE-cooled fast reactor NCLFR-Oil

The exploitation of heavy oil reservoirs calls for the implementation of thermal recovery techniques. To provide the heat source for oil recovery energy system, NCLFR-Oil, a Natural Circulation Lead-bismuth Fast Reactor was designed for heavy oil extraction, with a designed power of 40 MW and a lifetime of at least 15 years. In order to achieve the designed object for a reactor core that is durable, compact, transportable, and exhibits strong passive and active safety features, the selection of core materials such as fuel and absorber, evaluation of reactivity control system and corresponding neutronic performance have been optimized by iterative design in this paper with the reference of many small modular reactors. The depletion calculation confirms that the designed reactor can operate for 19 years without refueling. Neutronic performance were analyzed for multiplication factor within lifetime, isotopes conversion, neutron energy spectrum, radial and axial power distributions, nuclear fission reaction rates, reactivity feedback coefficients, and control rod worth with shutdown margins. The core analysis shows that the reactor has good power and safety characteristics, which can fully meet the energy demand of heavy oil exploitation.

Thermal-hydraulic behavior simulations of the reactor cavity cooling system (RCCS) experimental facility using Flownex

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

Traffic compliance effect of more women behind the wheel: Pride or prejudice?

Driving offences can cause road accidents with serious outcomes and are a topic where differences in attitudes and behaviour can be found between sexes. While men are more prone to taking risks and, statistically, are involved in a greater number of traffic accidents, even today, sex stereotypes persist with women regarded as worse drivers. To explore the truth behind the myth, this study analyses how the increase in the number of females in the driver census in Spain, could impact the effectiveness of traffic law enforcement in terms of traffic offences recorded by the traffic police (including the main violations such as passive safety feature use, speed limits and drunk driving) while considering a set of control variables related to the quality of infrastructure, legal reforms and other socio-economic attributes. The results point to Spanish women paying closer attention to traffic regulations, behaving more carefully, which should be considered in designing specific preventive policies and road safety awareness strategies.

Direct Numerical Simulation of High Prandtl Number Fluid Flow in the Downcomer of an Advanced Reactor

The design of advanced nuclear reactors [Generation IV (Gen IV)] involves an array of challenging fluid-flow issues that affect its safety and performance. Given that Gen IV designs have improved passive safety features, the downcomer plays a crucial role in loss-of-power scenarios. Fluid-flow behavior in the downcomer can involve forced to mixed to natural convection, and characterizing the heat transfer for these changing regimes is a daunting challenge. The creation of a high-resolution heat transfer numerical database can potentially support the development of precise and affordable reduced-resolution heat transfer models. These models can be designed based on a multiscale hierarchy developed as part of the recently U.S. Department of Energy–funded Center of Excellence for Thermal Fluids Applications in Nuclear Energy, which can help address industrial-driven issues associated with the heat transfer behavior of advanced reactors. In this paper, the downcomer is simplified to heated parallel plates, and high Prandtl number fluid (FLiBe) is considered for all simulations. The calculations are performed for a wide range of Richardson numbers from 0 to 400 at two different FLiBe Prandtl numbers (12 and 24), which result in 40 simulated cases in total. Time-averaged and time series statistics, as well as Nusselt number correlations, are investigated to illuminate mixed convection behavior. The calculated database will be instrumental in understanding flow behavior in the downcomer. Ultimately, we aim to evaluate existing heat transfer correlations, and some modifications are proposed.

The development of micro and small modular reactor in the future energy market

Micro and Small Modular Reactor (MSMR) is an emerging energy technology that meets the requirements of market demand, safety, efficiency, and sustainability. This paper summarizes the advantages, application scenarios, and advanced technologies to support MSMR. Now that the energy market is more flexible and the requirements are more complex, while MSMR can meet the market demand and has a lower cost compared with other clean energies such as wind and solar photovoltaic. The United States is vigorously developing MSMRs into residential energy markets. The MSMR developed around the world has more than three generations of safety characteristics that have adopted passive safety features. MSMR can be manufactured in the factory which reduces construction schedule, cost, and waste. The nuclear fuel supply chain for MSMR is complete and perfect, including the front end and back end. An increasing number of advanced technologies support the development of MSMR, including advanced materials (TRISO fuel and accident-tolerance fuel), advanced control knowledges (DI&C, cybersecurity, and AI), and an advanced computational platform (MOOSE framework).

Exploring the benefits of molten salt reactors: An analysis of flexibility and safety features using dynamic simulation

There has been a growing interest in Molten Salt Reactors (MSRs) in recent years due to the significant potential for increasing flexibility, security, and reliability of the grid, as well as the inherent passive safety features when compared to traditional pressurized water reactors (PWRs). MSRs can help meet many future nuclear energy goals, such as improved sustainability, high security, high efficiency, and high safety passive features, and help reduce nuclear waste. In this study, to investigate MSRs’ passive safety features, a dynamic model of 9 graphite nodes and 18 fuel salt nodes are simulated in 7 safety scenarios. These simulation results are compared with a traditional PWR dynamic simulation. The simulation shows the stability of MSR operations during these 7 safety scenarios, showing that the coolant and graphite temperature within the system stay within the safety limits of operation. The negative feedback coefficient of the fuel salt within MSR cores plays a significant role in stabilizing the power response inside the core, keeping the power from significant excursions. A one-year simulation is also conducted to test the load-following capabilities of MSRs in comparison with traditional PWRs. It is found that MSRs increase the flexibility, reliability, and security of the grid by operating in load-following mode without the need to change the position of the control rods. MSR's increased efficiency also leads to a reduction in backup fossil-fuel based electricity generation by 82% when compared to traditional PWRs operating in load-following mode.

Numerical analysis of dynamic load following response in a natural circulation molten salt power reactor system

The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. An investigation into the power transient behavior of an autonomous load-following, closed-loop, natural circulation MSR system is important to quantifying operational and safety performance under dynamic conditions. This paper presents the results of a STAR-CCM+ and a comparatively simple asymmetric, one-dimensional, numerical model to solve the compound dynamic MSR power behavior subject to flow and temperature reactivity feedback only. Results show that reactor power is affected by fuel salt flow velocity (global) and temperatures (local bulk volume averages) in a coupled, time-delayed manner that results in a unique compound dynamic, closed-loop power feedback mechanism. The 1-D simulation approach opens the possibility of performing inexpensive computations to evaluate time-dependent reactor performance relative to thermo-physical fuel salt limitations. Results also potentially motivate a convincing conclusion that natural circulation MSRs provide a leap in safety and reliability.

Extraordinary Passive Safety in Cars Using a Sensor Network Model

Context: The automobile industry has included active and passive safety. Active safety incorporates elements to avoid crashes and collisions. Some elements are ABS brakes and stabilization bars, among others. On the other hand, passive safety avoids or minimizes damage to the occupants in the event of an accident. Some passive safety features include seat belts and front and curtain airbags for the driver and other occupants. Method: In this research work, we propose a new category called Extraordinary Passive Safety (XPS). A model of a sensor network was designed to inspect the conditions inside the car to detect fire, smoke, gases, and extreme temperatures. The sensors send data to a device (DXPS) capable of receiving and storing the data. Results: Each sensor collects data and sends it to the DXPS every period. The sensor sends 0s while there is no risk, and 1s when it detects a risk. When the DXPS receives a 1, the pattern is evaluated, and the risk is identified. Since there are several sensors, the reading pattern is a set of 0s (000000). When a pattern with one or more 1s (000100, 010101) is received, the DXPS can send an alert or activate a device. Conclusions: The proposed solution could save the lives of children left in the car or people trapped when the car catches fire. As future work, it is intended to define the devices to avoid or minimize damage to the occupants such as oxygen supply, gas extraction, regulating the temperature, among others.

Pre-test analysis for definition of steam injection tests through multi-hole sparger in PANDA facility

Pressure Suppression Pool (PSP) is a passive safety feature in Boiling Water Reactors (BWR) and Advanced Pressurized (AP) reactors. Steam released from the primary coolant system is condensed in a large water pool to prevent containment overpressure. Injected steam induces sources of heat (buoyancy force) and momentum (inertia). The competition between the sources might result in the development of thermal stratification or mixing of the pool. Increased temperature of the top pool layer leads to higher partial pressure of steam in the containment and thus reduces pressure suppression capacity of the pool. Models with predictive capabilities are needed for the analysis of the reactor pool transients. Development and validation of the models require adequate experimental data. In this work we discuss results of the pre-test analysis that was carried out to select conditions for the tests with steam injection through sparger head and Load Reduction Ring (LRR) in a large scale PANDA facility. The aim of the tests was to obtain data on pool thermal stratification and mixing under different regimes of steam injection. Effective Heat Source (EHS) and Effective Momentum Source (EMS) models were implemented in a computational fluid dynamics (CFD) code in order to carry out the analysis. Evolution of the pool temperature and velocity characteristics were analyzed in the scoping analysis to provide suggestions for selection of (i)pool depth, (ii)elevations of the sparger head and LRR, (iii)number of open LRR holes, (iv)layout of instrumentation, and (v)steam injection procedure for each test.