Y1000 gene family
All relative analysis data and codes used in this study are available on the Figshare data repository. You can also find the codes on the Github repository at https://github.com/vnuii/Y1000_gene_family.
budding yeast↗
SEARCH · Engineering Papers
Results for “analysis codes”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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Microbes’ effect on switchgrass germination and growth under drought depends on life stage
Repository for code used for analysis of effects of the belowground microbiome on switchgrass responses to drought.
drought↗
SPARClite v. 8/17/16
SPARClite is a compressible CFD code useful for analysis of problems involving high speed flows of fluids around solid structures. It is being built for efficient use on next-generation high performance computers, and is intended to be developed at a high level of software quality to meet the rigors of Sandia’s production software needs.
Howard, Micah↗
SPARC v. 8/17/2016
SPARC is a compressible CFD code useful for analysis of problems involving high speed flows of fluids around solid structures. It is being built for efficient use on next-generation high performance computers, and is intended to be developed at a high level of software quality to meet the rigors of Sandia’s production software needs.
Howard, Micah↗
Extreme Load Computational Fluid Dynamics Analysis and Verification for a Multibody Wave Energy Converter: Preprint
A wave energy converter (WEC) must be designed to survive the extreme sea states that it will be subject to throughout its lifetime. Although there are many analysis methods and codes available to accomplish this, there are currently several engineering challenges to WEC survival design. Foremost, the computational design approach will typically involve a trade-off between accuracy and computational efficiency. Additionally, most computational fluid dynamics (CFD) codes are not ideally suited to modeling extreme events for WECs with multibody dynamics, power-take-off systems, and mooring systems. Finally, although WEC design standards and CFD guidelines are emerging, with the current immaturity of the WEC industry, they are not yet well established. In this study, loads on a 1:35-scale, moored, multibody WEC are evaluated with CFD. The CFD results are compared with results obtained from a computationally efficient, midfidelity model based on linearized potential flow hydrodynamics. For these model verification comparisons, both operational and survival configurations are considered. The extreme load results obtained, using both codes, indicate that the survival configuration successfully sheds loads during extreme sea states. It is also found that WEC-Sim, when appropriately applied, can provide reasonable load results, at a fraction of the computational expense of CFD. However, for the more extreme sea states, and for higher-order effects not included in the WEC-Sim model, the linear-based results have significant errors in comparison to the CFD-based results, and should be used judiciously.
16 TIDAL AND WAVE POWER↗
Comparison between PARFUME and Bison Using the AGR-2 Irradiation Experiment
This report documents comparisons between Fuel Model (PARFUME) model predictions versus Bison for selected compacts from the second irradiation test of the Advanced Gas Reactor (AGR) program that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). PARFUME is a fuel performance analysis and modeling code, used for evaluating gas-reactor tristructural isotropic (TRISO) coated particle fuel for prismatic, pebble bed, plate, and cylindrical type fuel geometries. PARFUME is an integrated mechanistic computer code that evaluates the thermal, mechanical, and physico-chemical behavior of TRISO coated-fuel particles and the probability for fuel failure given the particle-to-particle statistical variations in physical dimensions and material properties that arise during the fuel fabrication process. Bison is a nuclear fuel performance application built using the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element library (Permann 2020). Bison is capable of modeling multiple fuel forms in a wide variety of dimensions and geometries. It solves coupled nonlinear partial differential equations, including heat conduction, mechanics, fission product species transport etc., in a fully implicit manner. Comparisons between PARFUME and Bison were performed using four compacts from the AGR-2 experiment. Selected outputs were chosen based on their impact on the probability of the SiC layer failing. In general there was good agreement between PARFUME and Bison with the exception of predicting the gap formed between the buffer and IPyC. Further comparisons between PARFUME and Bison are planned to further develop Bison’s capabilities.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
IEA Wind Energy Task 37 System Engineering Aerodynamic Optimization Case Study: Preprint
This paper presents the results from the first IEA Wind Task 37 aerodynamic optimization case study. Eight participants applied their optimization tools to a purely aerodynamic problem and the results were compared. Overall, the different tools produced widely different designs, while there was better agreement in the improvement achieved. This highlights the fact that further investigation is needed to try to understand these differences and develop best practices. There were too many differences between the different analysis and optimization codes to determine the sources of these differences. However, several potential sources of discrepancy were identified for further investigation. One hypothesis is that one potential source of discrepancy is that the design problem itself is relatively flat in design directions of constant loading. This flatness would impact the convergence in the optimization, and at the same time mean that discrepancies in the design itself would be less severe.
49 EE - Wind and Water Power Program - Wind (EE-4W↗
A high-fidelity Monte Carlo evaluation of CANDU-6 safety parameters
Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANDU-6 (CANada Deuterium Uranium) reactor have been evaluated by using a modified MCNPX code. For accurate analysis of the parameters, the DBRC (Doppler Broadening Rejection Correction) scheme was implemented in MCNPX in order to account for the thermal motion of the heavy uranium nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted by using the MCNPX and the FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated by using several cross section libraries such as ENDF/B-VI, ENDF/B-VII, JEFF, JENDLE. The PCR value is also evaluated at mid-burnup conditions to characterize safety features of equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, huge number of neutron histories are considered in this work and the standard deviation of the k-inf values is only 0.5{approx}1 pcm. It has been found that the FTC is significantly enhanced by accounting for the Doppler broadening of scattering resonance and the PCR are clearly improved. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Effects of cooling time on a closed LWR fuel cycle
In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Use of Monte Carlo simulation to quantify CIPS risk
Crud Induced Power Shift (CIPS), caused by the accumulation of boron compounds in corrosion product deposits on high power fuel assemblies, is an important operational issue for Pressurized Water Reactors (PWRs). The Boron-induced Offset Analysis (BOA) computer code has been developed by EPRI in collaboration with Westinghouse to help utility engineers avoid onset of CIPS. The code, used with an established procedure, predicts the maximum boron mass that will be accumulated in the upcoming cycle for a given core design and plant chemistry and operating program. The calculated boron mass above a given threshold predicts the risk of CIPS to the upcoming cycle. BOA has been successful in correctly predicting the risk for onset or lack of CIPS in many cycles. However, like most codes modeling complex physical phenomena, it is subject to uncertainties in several of the inputs used to describe these phenomena. The current BOA procedure results in a single 'best estimate' prediction, which has been judged to provide engineers with a practical way to assess CIPS risk. As such, however, there is a probability that the actual plant CIPS performance may be underestimated. This paper describes a method to assess the probability (with an associated confidence level) of no CIPS, mild to moderate CIPS, or severe CIPS using BOA for a given core design. This is accomplished through the use of a Monte Carlo simulation. The simulation is applied to several plant examples, and the results are evaluated in terms of the likelihood of mild or severe CIPS. A suggested set of criteria for decision making regarding CIPS risk is presented in the conclusion. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Cyber security using multi-threaded architecture data diode at the NBSR - 114
Ongoing upgrades for the control room of the National Bureau of Standards Reactor (NBSR) located at the National Institute of Standards and Technology (NIST) require the ability to collect sensor data from the secured reactor network and transfer to the outside network for trend analysis, while maintaining the security of the reactor network. To accomplish this, a data diode was created to provide a physical barrier that prevents any outside data from reaching the secured reactor network. The hardware of the data diode consists of two embedded computers, running Unix based operating systems, connected by three DB-9 serial cables. The serial cables have been modified so that only two wires, a ground and a transmission line, provide connection between the computers. The data diode relies on a multi-threaded architecture to collect and simultaneously transfer data over three serial connections. A software program, implemented in Java, has been written to request data from the digital recorders installed to monitor the NBSR process instrumentation and transfer that data across the serial connections. The data diode can collect over 1200 data points from multiple digital recorders each second. The collected data can then be stored in a relational database for later analysis. The source codes for the software are available open source on GitHub. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Estimation of LOCA break size using cascaded support vector regression - 334
A failure of emergency core cooling system (ECCS) during a certain small break loss of coolant accident (SBLOCA) can cause a severe core uncovery and fuel failure. The safety injection systems (SISs) might not function properly in case of a SBLOCA, due to the slight change of pressure in the pipes. Early trend of the major parameters has to be observed and the accurate information has to be given to the nuclear power plant (NPP) operators by precisely identifying the break size to effectively manage accidents. In this study, the objective is to provide to operators the information on the break size in a short time by considering the accident situations such as hot-leg LOCA, cold-leg LOCA, steam generator tube rupture (SGTR). A cascaded support vector regression (CSVR) is used in order to estimate the break size. The simulation data set was obtained from the optimized power plant reactor 1000 (OPR1000) using modular accident analysis program (MAAP) code. And genetic algorithm (GA) was used to select the input variables of the CSVR model and optimize its parameters. As a result of this study, the CSVR model estimate very well the break size of LOCAs. If the operators can predict the break size in the LOCA, they can response quickly and properly to LOCA circumstances to prevent the core uncovery and fuel failure. Also, it will be possible to more efficiently manage beyond design basis accidents. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Identification of NPP transients using artificial intelligence - 335
If the accidents happen in nuclear power plants (NPPs), the NPP operators identify the transients through observing the behaviors of the major parameters. However, the integrity of the instrumentation signals is not ensured under the accident circumstances and the major parameters essential in identifying the transients can rapidly change although the instruments generate accurate signals. Thus, to successfully control the accidents by taking necessary measures relatively faster, it is very important to provide the operators with the information that can identify the transients and accidents accurately. In this study, the transients of severe accidents in NPPs are identified and classified using artificial intelligence methodologies. Several transients are identified using support vector classification (SVC) and probabilistic neural network (PNN) as artificial intelligence methodologies. The used data for the transients are the simulation data using the modular accident analysis program (MAAP) code for the optimized power reactors (OPR1000s). The time-integrated values of chosen simulated sensor signals are used as the input variables for identifying the transients. As a result of this study, the proposed models identified the transients quite precisely. Therefore, the quite accurate identification of the transients indicates the excellence of these artificial intelligence methods and they will produce useful supporting information for the operators even under extreme transient circumstances in NPPs. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
TRACG benchmark study: 2015 Fermi thermal-hydraulic instability event
On March 19, 2015, the Boiling Water Reactor (BWR) Fermi 2 (referred to hereafter as Fermi), located in Michigan, U.S., experienced a scram initiated by the Oscillation Power Range Monitor (OPRM) system. The plant had been operating in Single Loop Operation (SLO) for approximately 10 minutes before the scram occurred. During this time, the plant operating conditions reached a region within the power/flow (P/F) map susceptible to coupled neutronic/thermal-hydraulic (T-H) instabilities. Limited code benchmarking is available in literature for T-H Instability (THI) transients occurring in plants operating at SLO conditions due to scarce plant data for this type of operational condition. This study aims to model and benchmark the THI event at SLO conditions using TRACG, the GE Hitachi (GEH) proprietary best-estimate transient system code. The TRACG model was benchmarked by comparing the results with measured plant data, such as the total reactor power and core flow, from the event. Additionally, to evaluate the inherent stability of the plant at different P/F points, Decay Ratio (DR) calculations were performed. The capability of TRACG for calculating DRs is well documented, and it has also been approved by the U.S. Nuclear Regulatory Commission (NRC) as part of ESBWR TRACG stability licensing topical reports. The results of TRACG DRs were compared to the GEH best-estimate frequency domain code ODYSY, which is the current Fermi licensing basis code for reload analysis DR calculations.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
A new model for the dynamic of criticality excursions in fissile solutions - 14446
A recent international inter-code comparison exercise for criticality excursion analysis demonstrated that the transient criticality codes participating the exercise are showed to reproduce well the initial peak power compared with measurements performed with TRACY and SILENE reactors. However the oscillatory behavior after peak power is less well reproduced. The integrated behavior in term of energy release is quite well reproduced. The recent experiments performed with TRACY JAEA's reactor and the analytic interpretation of void clustering during prompt critical have suggested a simplified vibration model of a damped oscillator. Finally a recent ICNC paper by present authors offered a perfect correlation between peak pressure and peak power. Owing to these three facts the present paper aims at proposing a new system of equations for the dynamic of a critical excursion taking into account gas bubble generation after exceeding the solubility threshold of dissolved radiolytic gas. The model explains why the simple vibration model of the JAEA school fairly represent pressure dynamic records at moderate peak pressure as measured with SILENE reactor. Also, an important parameter of the proposed model, namely the rate of transfer from dissolved gas to gas bubbles, is fixed by the correlation between peak pressure and peak power. Thus the model has one less tunable parameter and can predict a large range of critical low pressure excursions. After presentation of the model and discussion of some of its important features, some numerical simulations are made based upon SILENE and TRACY transient experiments ranging from tenths of kPa to several bars and comparisons are presented between calculated results and experimentally recorded one's. (authors)
42 ENGINEERING↗
Dose Consequence and Probabilistic Risk Assessment Integration into Digital Documented Safety Analysis
This is an intern poster presentation. The current reactor authorization process is complex and error prone. The development of a digital Documented Safety Analysis has been proposed to provide an automated and integrated solution to enhance the design and authorization process. The digital DSA will consist of interlinked models, analyses, and reports, all of which will be updated using automated workflows when design changes are made. This poster examines the integration of the probabilistic risk assessment (PRA) with the transient and dose consequence analyses. A PRA for a generic high temperature gas-cooled reactor (HTGR) has been constructed which will drive the input parameters for a transient analysis model currently being constructed. Dose consequence will be determined using the results of the transient analysis and the Radiation Safety Analysis Computer (RSAC) code. Dose consequence data will then be input back into the PRA to drive design parameters.
42 ENGINEERING↗
Creation of the VADER Code in SCALE [Slides]
This presentation details the creation of Validation Analysis Data Evaluation Resource (VADER) code in SCALE. This lecture highlights the USLSTATS program written in JAVA and its history. Additionally, it covers the integration of VADER with SCALE along with features and background information.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Visualization and Analysis Requirements for In Situ Processing for a Large-Scale Fusion Simulation Code
In situ techniques have become a very active research area since they have been shown to be an effective way to combat the issues associated with the ever growing gap between computation and I/O bandwidth. In order to take full advantage of in situ techniques with a large-scale simulation code, it is critical to understand the breadth and depth of its analysis requirements. In this paper, we present the results of a survey done with members of the XGC1 fusion simulation code team in order to gather their requirements for analysis and visualization. We look at these requirements from the perspective of in situ processing and present a list of XGC1 analysis tasks performed by its physicists, engineers, and visualization specialists. This analysis of the specific needs and use cases of a single code is important in understanding the nature of the needs that simulations have in terms of data movement and usage for visualization and analysis, now and in the future.
Kress, James M.↗