Crosswind Sensitivity of Passenger Cars and the Influence of Chassis and Aerodynamic Properties on Driver Preferences

Results of vehicle crosswind research involving both full-scale driver-vehicle tests and associated analyses are presented. The paper focuses on experimental crosswind testing of several different vehicle configurations and a group of seven drivers. A test procedure, which utilized wind-generating fans arranged in alternating directions to provide a crosswind "gauntlet", is introduced and described. Driver preferences for certain basic chassis and aerodynamic properties are demonstrated and linked to elementary system responses measured during the crosswind gauntlet tests. Based on these experimental findings and confirming analytical results, a two-stage vehicle design process is then recommended for predicting and analyzing the crosswind sensitivity of a particular vehicle or new design.Peer Reviewedhttp://deepblue.lib.umich.edu/bitstream/2027.42/65022/1/MacAdam 1990 VSD Aerodynamic Crosswind paper.pd

Prioritization of VHTR system modeling needs based on phenomena identification, ranking and sensitivity studies.

Quantification of uncertainty is a key requirement for the design of a nuclear power plant and the assurance of its safety. Historically the procedure has been to perform the required uncertainty assessment through comparison of the analytical predictions with experimental simulations. The issue with this historical approach has always been that the simulations through experiments could not be at full scale for the practical reasons of cost and scheduling. Invariably, only parts of the system were tested separately or if integral testing was performed for the complete system, the size or scale of the experimental apparatus was significantly smaller than the actual plant configuration

Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ‘hot spots’ do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ‘first principles.’ Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate estimates of the flow and energy transport as applied to nuclear reactor safety. However, it is expected that these practices and procedures will require updating from time to time as research and development affect them or replace them with better procedures. The practices and procedures are categorized into five groups. These are: 1.Code Verification 2.Code and Calculation Documentation 3.Reduction of Numerical Error 4.Quantification of Numerical Uncertainty (Calculation Verification) 5.Calculation Validation. These five categories have been identified from procedures currently required of CFD simulations such as those required for publication of a paper in the ASME Journal of Fluids Engineering and from the literature such as Roache [1998]. Code verification refers to the demonstration that the equations of fluid and energy transport have been correctly coded in the CFD code. Code and calculation documentation simply means that the equations and their discretizations, etc., and boundary and initial conditions used to pose the fluid flow problem are fully described in available documentation. Reduction of numerical error refers to practices and procedures to lower numerical errors to negligible or very low levels as is reasonably possible (such as avoiding use of first-order discretizations). The quantification of numerical uncertainty is also known as calculation verification. This means that estimates are made of numerical error to allow the characterization of the numerica

Initial VHTR Accident Scenario Classification: Models and Data.

Nuclear systems codes are being prepared for use as computational tools for conducting performance/safety analyses of the Very High Temperature Reactor. The thermal-hydraulic codes are RELAP5/ATHENA for one-dimensional systems modeling and FLUENT and/or Star-CD for three-dimensional modeling. We describe a formal qualification framework, the development of Phenomena Identification and Ranking Tables (PIRTs), the initial filtering of the experiment databases, and a preliminary screening of these codes for use in the performance/safety analyses. In the second year of this project we focused on development of PIRTS. Two events that result in maximum fuel and vessel temperatures, the Pressurized Conduction Cooldown (PCC) event and the Depressurized Conduction Cooldown (DCC) event, were selected for PIRT generation. A third event that may result in significant thermal stresses, the Load Change event, is also selected for PIRT generation. Gas reactor design experience and engineering judgment were used to identify the important phenomena in the primary system for these events. Sensitivity calculations performed with the RELAP5 code were used as an aid to rank the phenomena in order of importance with respect to the approach of plant response to safety limits. The overall code qualification methodology was illustrated by focusing on the Reactor Cavity Cooling System (RCCS). The mixed convection mode of heat transfer and pressure drop is identified as an important phenomenon for Reactor Cavity Cooling System (RCCS) operation. Scaling studies showed that the mixed convection mode is likely to occur in the RCCS air duct during normal operation and during conduction cooldown events. The RELAP5/ATHENA code was found to not adequately treat the mixed convection regime. Readying the code will require adding models for the turbulent mixed convection regime while possibly performing new experiments for the laminar mixed convection regime. Candidate correlations for the turbulent mixed convection regime for circular channel geometry were identified in the literature. We describe the use of computational experiments to obtain correction factors for applying these circular channel results to the specialized channel geometry of the RCCS. The intent is to reduce the number of laboratory experiments required. The FLUENT and Star-CD codes contain models that in principle can handle mixed convection but no data were found to indicate that their empirical models for turbulence have been benchmarked for mixed convection conditions. Separate effects experiments were proposed for gathering the needed data. In future work we will use the PIRTs to guide review of other components and phenomena in a similar manner as was done for the mixed convection mode in the RCCS. This is consistent with the project objective of identifying weaknesses or gaps in the code models for representing thermal-hydraulic phenomena expected to occur in the VHTR both during normal operation and upsets, identifying the models that need to be developed, and identifying the experiments that must be performed to support model development

Next Generation Nuclear Plant Methods Technical Program Plan

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR

MaDDOSY (Mass Determination Diffusion Ordered Spectroscopy) using an 80 MHz bench top NMR for the rapid determination of polymer and macromolecular molecular weight

YesMeasurement of molecular weight is an integral part of macromolecular and polymer characterization which usually has limitations. Herein, we present the use of a bench-top 80 MHz NMR spectrometer for diffusion-ordered spectroscopy as a practical and rapid approach for the determination of molecular weight/size using a novel solvent and polymer-independent universal calibration.Royal Society. Grant Number: URF∖R1∖180274. Engineering and Physical Sciences Research Council. Grant Numbers: EP/V037943/1, EP/V007688/1, EP/V036211/

Alpha kinase 3 signaling at the M-band maintains sarcomere integrity and proteostasis in striated muscle

Muscle contraction is driven by the molecular machinery of the sarcomere. As phosphorylation is a critical regulator of muscle function, the identification of regulatory kinases is important for understanding sarcomere biology. Pathogenic variants in alpha kinase 3 (ALPK3) cause cardiomyopathy and musculoskeletal disease, but little is known about this atypical kinase. Here we show that ALPK3 is an essential component of the M-band of the sarcomere and define the ALPK3-dependent phosphoproteome. ALPK3 deficiency impaired contractility both in human cardiac organoids and in the hearts of mice harboring a pathogenic truncating Alpk3 variant. ALPK3-dependent phosphopeptides were enriched for sarcomeric components of the M-band and the ubiquitin-binding protein sequestosome-1 (SQSTM1) (also known as p62). Analysis of the ALPK3 interactome confirmed binding to M-band proteins including SQSTM1. In human pluripotent stem cell-derived cardiomyocytes modeling cardiomyopathic ALPK3 mutations, sarcomeric organization and M-band localization of SQSTM1 were abnormal suggesting that this mechanism may underly disease pathogenesis

Primeiro registro do gastrópode africano invasor Melanoides tuberculatus (Gastropoda: Prosobranchia: Thiaridae) na Bacia do Rio Paranã, GO, Brasil

The Thiarid snail Melanoides tuberculatus (Müller, 1774), native to Asia and East Africa was recorded for the first time in the Paranã River basin, Goiás State. There is no evidence concerning introduction vectors but aquarium releases is the most probable vector. Specimens were collected at three different water bodies after twenty-seven rivers were investigated. The possible spread of this species to other habitats and potential effects on native thermal water communities are discussed.O gastrópode Thiaridae Melanoides tuberculatus (Müller, 1774), nativo da Ásia e do Leste Africano, é registrado pela primeira vez na Bacia Hidrográfica do Rio Paranã (Estado de Goiás). Não se conhecem os vetores de introdução da espécie, mas o setor de aquariofilia foi apontado por alguns moradores locais como a mais provável causa. Os espécimes foram coletados em três corpos d´água depois de serem investigados vinte e sete rios e lagoas. A possibilidade de dispersão dessa espécie para outros habitats e os efeitos potenciais dessa introdução sobre a comunidade nativa são discutidos no artigo