Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Iterative”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Fusion materials science and technology research opportunities now and during the ITER era

Several high-priority near-term potential research activities to address fusion nuclear science challenges are summarized. General recommendations include: (1) Research should be preferentially focused on the most technologically advanced options (i.e., options that have been developed at least through the singleeffects concept exploration stage, technology readiness levels >3), (2) Significant near-term progress can be achieved by modifying existing facilities and/or moderate investment in new medium-scale facilities, and (3) Computational modeling for fusion nuclear sciences is generally not yet sufficiently robust to enable truly predictive results to be obtained, but large reductions in risk, cost and schedule can be achieved by careful integration of experiment and modeling.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

A fresh look at electron cyclotron current drive power requirements for stabilization of tearing modes in ITER

ITER is an international project to design and build an experimental fusion reactor based on the “tokamak” concept. ITER relies upon localized electron cyclotron current drive (ECCD) at the rational safety factor q=2 to suppress or stabilize the expected poloidal mode m=2, toroidal mode n=1 neoclassical tearing mode (NTM) islands. Such islands if unmitigated degrade energy confinement, lock to the resistive wall (stop rotating), cause loss of “H-mode” and induce disruption. The International Tokamak Physics Activity (ITPA) on MHD, Disruptions and Magnetic Control joint experiment group MDC-8 on Current Drive Prevention/Stabilization of Neoclassical Tearing Modes started in 2005, after which assessments were made for the requirements for ECCD needed in ITER, particularly that of rf power and alignment on q=2 [1]. Narrow well-aligned rf current parallel to and of order of one percent of the total plasma current is needed to replace the “missing” current in the island O-points and heal or preempt (avoid destabilization by applying ECCD on q=2 in absence of the mode) the island [2-4]. This paper updates the advances in ECCD stabilization on NTMs learned in DIII-D experiments and modeling during the last 5 to 10 years as applies to stabilization by localized ECCD of tearing modes in ITER. This includes the ECCD (inside the q=1 radius) stabilization of the NTM “seeding” instability known as sawteeth (m/n=1/1) [5]. Recent measurements in DIII-D show that the ITER-similar current profile is classically unstable, curvature stabilization must not be neglected, and the small island width stabilization effect from helical ion polarization currents is stronger than was previously thought [6]. The consequences of updated assumptions in ITER modeling of the minimum well-aligned ECCD power needed are all-in-all favorable (and well-within the ITER 24 gyrotron capability) when all effects are included. However, a “wild card” may be broadening of the localized ECCD by the presence of the island; various theories predict broadening could occur and there is experimental evidence for broadening in DIII-D. Wider than now expected ECCD in ITER would make alignment easier to do but weaken the stabilization and thus require more rf power. In addition to updated modeling for ITER, advances in the ITER-relevant DIII-D ECCD gyrotron launch mirror control system hardware and real-time plasma control system have been made [7] and there are plans for application in DIII-D ITER demonstration discharges.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2022

Design and Testing of a Prototype Eddy Current Actuated Valve for the ITER Shattered Pellet Injection System

Reliably mitigating disruptions is essential for ITER to meet its long-term operational research plan without damage to the in-vessel components. Currently, the shattered pellet injection (SPI) technique is the most effective radiator of thermal energy and has been chosen for the baseline disruption mitigation system (DMS) for ITER. The SPI process uses cryogenic temperatures to desublimate material into the barrel of a pipe gun forming a solid cylindrical pellet. Pellets for ITER will initially be hydrogen and hydrogen-neon mixtures. Once formed, pellets are dislodged and accelerated using high-pressure gas (40-60 bar) delivered by a fast-opening valve. The solenoid valves currently used for SPI experiments will not operate in an ITER environment due to the large background magnetic field. An ITER prototype fast-opening valve, called a flyer plate valve (FPV), has been designed and has undergone a wide range of testing. The FPV operates by pulsing current through a pancake coil that is closely coupled with a ``flyer plate.'' The flyer plate is an aluminum plate in which eddy currents are generated creating a repulsive force from the pancake coil. The force generated in the flyer plate rapidly lifts the valve tip off the seat and delivers a pulse of gas to the rear of the pellet, breaking it free from the barrel and accelerating the pellet downstream to its intended target. The design of the valve has been iterated on over the lifetime of this project, as the DMS for ITER shifted from massive gas injection (MGI) to SPI. The most recent design has been tested, and operational ranges have been mapped. The valve must survive 3000+ cycles in an ITER-like magnetic field. The principal functional requirement of this valve is to reliably dislodge and accelerate hydrogen (or H-Ne mixture) pellets into ITER. The valve was mated with an ITER SPI test stand and has been shown to be capable of launching pellets reliably. The valve and power supply design will be discussed in this article, along with the various testing setups used to determine the feasibility of this valve for use on ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Volumetric quantification of lung nodules in CT with iterative reconstruction (ASiR and MBIR)

Purpose: Volume quantifications of lung nodules with multidetector computed tomography (CT) images provide useful information for monitoring nodule developments. The accuracy and precision of the volume quantification, however, can be impacted by imaging and reconstruction parameters. This study aimed to investigate the impact of iterative reconstruction algorithms on the accuracy and precision of volume quantification with dose and slice thickness as additional variables.Methods: Repeated CT images were acquired from an anthropomorphic chest phantom with synthetic nodules (9.5 and 4.8 mm) at six dose levels, and reconstructed with three reconstruction algorithms [filtered backprojection (FBP), adaptive statistical iterative reconstruction (ASiR), and model based iterative reconstruction (MBIR)] into three slice thicknesses. The nodule volumes were measured with two clinical software (A: Lung VCAR, B: iNtuition), and analyzed for accuracy and precision.Results: Precision was found to be generally comparable between FBP and iterative reconstruction with no statistically significant difference noted for different dose levels, slice thickness, and segmentation software. Accuracy was found to be more variable. For large nodules, the accuracy was significantly different between ASiR and FBP for all slice thicknesses with both software, and significantly different between MBIR and FBP for 0.625 mm slice thickness with Software A and for all slice thicknesses with Software B. For small nodules, the accuracy was more similar between FBP and iterative reconstruction, with the exception of ASIR vs FBP at 1.25 mm with Software A and MBIR vs FBP at 0.625 mm with Software A.Conclusions: The systematic difference between the accuracy of FBP and iterative reconstructions highlights the importance of extending current segmentation software to accommodate the image characteristics of iterative reconstructions. In addition, a calibration process may help reduce the dependency of accuracy on reconstruction algorithms, such that volumes quantified from scans of different reconstruction algorithms can be compared. The little difference found between the precision of FBP and iterative reconstructions could be a result of both iterative reconstruction's diminished noise reduction at the edge of the nodules as well as the loss of resolution at high noise levels with iterative reconstruction. The findings do not rule out potential advantage of IR that might be evident in a study that uses a larger number of nodules or repeated scans.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Neutronics Analysis for ITER Diagnostic Generic Upper Port Plug

This paper presents new results of neutronics analysis performed for the updated design of the ITER Diagnostic Generic Upper Port Plug (DGUPP). Such analysis is performed regularly during the development of the ITER Ports design. Neutronics analysis is an important step in support of the development of ITER components irradiated by 14 MeV neutrons generated in D-T fusion thermonuclear reactions. The neutrons are born in the ITER tokamak plasma chamber and then impinge on the surrounding blanket and port plugs, after that they diffuse through shielding material of vacuum vessel components or could stream through the diagnostic channels and assembly gaps far away from plasma up to the bio-shield and in a small amount beyond it. While ITER machine is in operation it is inaccessible, after the shut-down it is planned to provide personnel access to the Upper Port Inter-Space Structure (UP ISS) behind the Port Plug via the corridors. Neutrons activate the ITER materials and those radioactive materials emit decay gammas during the shut-down causing a Shut-Down Dose Rate (SDDR) field distributed all around the activated materials. Decay gammas could also stream through the void channels. That emphasizes the need for calculation of decay gamma transport in evaluations of SDDR, which is reflected in the use of transport - activation interface code described in Section of Computational Modelling. The assessment of radiation environment (fluxes and SDDR) at DGUPP was focused on the port back-side area of UP ISS, where personnel access is planned along the ISS maintenance corridors. Minimization of SDDR at the front part of the ISS maintenance corridors was the target value for optimization of the DGUPP shielding and analysis of related neutronics effects. Following the dose ALARA (As Low As Reasonably Achievable) principle, every possibility to reduce SDDR should be investigated, emphasizing the importance of this analysis. The challenges of such analysis are inextricably connected with the engineering complexity of the ITER machine, with integration of many diagnostics systems having channels and gaps inside the ITER ports. Those gaps result in radiation streaming and negatively affect the shielding performance of the port plugs. The computational complication is caused by the need to model precisely the ITER geometry and then to run deep penetrating radiation transport in the substantially heterogenic toroidal geometry of ITER large-size tokamak, with 30 m height and diameter of its bio-shield. The whole mass of Upper Port Plug (UPP) could be up to 25 tons and it must be very precisely positioned inside the port opening in blanket and Vacuum Vessel. Facing the engineering challenges for the assembly of DGUPP, there were proposed two Project Change Requests (PCRs) No. 439 and 709 concerning the design updates of UPP/blanket interface and its remote handling.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Methodological Aspects of Shielding and Activation Analyses for ITER Diagnostics Systems - Paper 125

It is expected that the diagnostics systems of the ITER tokamak will provide valuable measurement data about the first in its kind D-T fusion plasma with emission of 14 MeV neutrons. These systems must withstand the neutron and secondary photon radiation along the ITER operation. Under the neutron irradiation the ITER materials are getting radioactive and therefore an adequate shielding must be provided along all the time of ITER operation and shut-down for the reliable machine operation and for the safe personnel access to its components, such as the diagnostics ports. These are the prerequisites for the shielding and activation analysis of the ITER diagnostics systems discussed in this paper. This paper provides insight into the current tasks of neutronics support for the development of ITER diagnostics systems. The tasks themselves range from relatively straightforward ones such as Monte Carlo radiation transport calculations for shielding analyses of the diagnostics systems, to more advanced tasks of Shut-Down Dose Rate (SDDR) analyses for which the dedicated Rigorous 2-Step mesh-based (R2Smesh) and Direct 1-Step (D1S) methodologies were applied. The accuracy and precision of the ITER neutronics analysis is dependent on the methodological approaches and the implying modeling assumptions. In addition to the SDDR challenges, another distinguishing feature of fusion neutronics is a wide use of CAD-based neutronics modeling to design components of fusion devices such as ITER and DEMO. Basically, there are two possibilities of CAD geometry settings in Monte Carlo (MC) radiation transport models, either by the geometry format conversion of CAD models into the format of MC models by means of the interface programs such as McCad and MCAM or by the direct use of CAD geometry engine in MC calculations such as it was realized in the DAGMC code approach. Several conversion interfaces and the DAGMC software library have been tested on the 3D ITER model developed for the CAD-MCNP benchmark. It was concluded that the considered approaches have reached the maturity level to allow their application to real ITER design analyses.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

ITER Central Solenoid Module Fabrication

The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first ITER module is in progress. The seven modules will be individually shipped to Cadarache, France upon their completion. This paper describes the processes and status of the fabrication of the CS Modules for ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Final report on the MAGARC software consulting support contract (WFO No. 15904) for ITER magnet safety assessment

he ITER (not an acronym but Latin for "the Way”) International Organization (IO) is building a large-scale scientific experiment that is now under construction in Cadarache, France. The ITER Tokamak fusion reactor aims to demonstrate that it is possible to produce energy from fusion. The ITER’s power output goal is to deliver ten times the power it consumes. During its operational lifetime, ITER will test key technologies necessary for the next step: the demonstration fusion power plant that will prove that it is possible to capture fusion energy for commercial use. General information on the ITER project can be found at http://www.iter.org. The ITER reactor’s superconducting magnet systems create ~41GJ of stored magnetic field energy when fully energized. The consequences of such energy dissipating via electrical arcs under faulted conditions gives rise to safety concerns even though these fault events are extremely unlikely. As part of the ITER licensing procedure, and as a result of such safety concerns, this contract was established to benchmark a developing magnet safety analysis tool at Culham Centre for Fusion Energy against the MAGARC computer code. MAGARC was developed by the U.S. Idaho National Laboratory (INL) under the ITER Engineering Design Activity (EDA) Project and was applied to produce previous analyses, in order to support magnet fault and safety assessment for the ITER design description document. This contract report documents the completion of requested modifications to the MAGARC code by the ITER IO, a time-step size sensitivity analysis with MAGARC during an unmitigated quench event, a comparison of the modeling approaches taken in developing MAGARC and the CCFE analysis package, and the data required for the benchmarking activity for these codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Reducing the effects of acoustic heterogeneity with an iterative reconstruction method from experimental data in microwave induced thermoacoustic tomography

Purpose: An iterative reconstruction method has been previously reported by the authors of this paper. However, the iterative reconstruction method was demonstrated by solely using the numerical simulations. It is essential to apply the iterative reconstruction method to practice conditions. The objective of this work is to validate the capability of the iterative reconstruction method for reducing the effects of acoustic heterogeneity with the experimental data in microwave induced thermoacoustic tomography. Methods: Most existing reconstruction methods need to combine the ultrasonic measurement technology to quantitatively measure the velocity distribution of heterogeneity, which increases the system complexity. Different to existing reconstruction methods, the iterative reconstruction method combines time reversal mirror technique, fast marching method, and simultaneous algebraic reconstruction technique to iteratively estimate the velocity distribution of heterogeneous tissue by solely using the measured data. Then, the estimated velocity distribution is used subsequently to reconstruct the highly accurate image of microwave absorption distribution. Experiments that a target placed in an acoustic heterogeneous environment are performed to validate the iterative reconstruction method. Results: By using the estimated velocity distribution, the target in an acoustic heterogeneous environment can be reconstructed with better shape and higher image contrast than targets that are reconstructed with a homogeneous velocity distribution. Conclusions: The distortions caused by the acoustic heterogeneity can be efficiently corrected by utilizing the velocity distribution estimated by the iterative reconstruction method. The advantage of the iterative reconstruction method over the existing correction methods is that it is successful in improving the quality of the image of microwave absorption distribution without increasing the system complexity.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Iterative metal artifact reduction for x-ray computed tomography using unmatched projector/backprojector pairs

Purpose: Metal artifact reduction (MAR) is a major problem and a challenging issue in x-ray computed tomography (CT) examinations. Iterative reconstruction from sinograms unaffected by metals shows promising potential in detail recovery. This reconstruction has been the subject of much research in recent years. However, conventional iterative reconstruction methods easily introduce new artifacts around metal implants because of incomplete data reconstruction and inconsistencies in practical data acquisition. Hence, this work aims at developing a method to suppress newly introduced artifacts and improve the image quality around metal implants for the iterative MAR scheme. Methods: The proposed method consists of two steps based on the general iterative MAR framework. An uncorrected image is initially reconstructed, and the corresponding metal trace is obtained. The iterative reconstruction method is then used to reconstruct images from the unaffected sinogram. In the reconstruction step of this work, an iterative strategy utilizing unmatched projector/backprojector pairs is used. A ramp filter is introduced into the back-projection procedure to restrain the inconsistency components in low frequencies and generate more reliable images of the regions around metals. Furthermore, a constrained total variation (TV) minimization model is also incorporated to enhance efficiency. The proposed strategy is implemented based on an iterative FBP and an alternating direction minimization (ADM) scheme, respectively. The developed algorithms are referred to as “iFBP-TV” and “TV-FADM,” respectively. Two projection-completion-based MAR methods and three iterative MAR methods are performed simultaneously for comparison. Results: The proposed method performs reasonably on both simulation and real CT-scanned datasets. This approach could reduce streak metal artifacts effectively and avoid the mentioned effects in the vicinity of the metals. The improvements are evaluated by inspecting regions of interest and by comparing the root-mean-square errors, normalized mean absolute distance, and universal quality index metrics of the images. Both iFBP-TV and TV-FADM methods outperform other counterparts in all cases. Unlike the conventional iterative methods, the proposed strategy utilizing unmatched projector/backprojector pairs shows excellent performance in detail preservation and prevention of the introduction of new artifacts. Conclusions: Qualitative and quantitative evaluations of experimental results indicate that the developed method outperforms classical MAR algorithms in suppressing streak artifacts and preserving the edge structural information of the object. In particular, structures lying close to metals can be gradually recovered because of the reduction of artifacts caused by inconsistency effects.

60 APPLIED LIFE SCIENCES↗