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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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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Reduction of asymmetric wall force in ITER disruptions with fast current quench

One of the problems caused by disruptions in tokamaks is the asymmetric electromechanical force produced in conducting structures surrounding the plasma. The asymmetric wall force in ITER asymmetric vertical displacement event (AVDE) disruptions is calculated in nonlinear 3D MHD simulations. It is found that the wall force can vary by almost an order of magnitude, depending on the ratio of the current quench time to the resistive wall magnetic penetration time. In ITER, this ratio is relatively low, resulting in a low asymmetric wall force. In JET, this ratio is relatively high, resulting in a high asymmetric wall force. Previous extrapolations based on JET measurements have greatly overestimated the ITER wall force. Here, it is shown that there are two limiting regimes of AVDEs, and it is explained why the asymmetric wall force is different in the two limits.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Reduction of asymmetric wall force in JET and ITER disruptions including runaway electrons

It was shown that asymmetric vertical displacement event (AVDE) disruptions in JET and ITER do not produce a large asymmetric force on conducting structures surrounding the plasma, provided that the current quench (CQ) time is less than the resistive wall penetration time. This is verified with JET data. In ITER the CQ is expected to be comparable to the resistive wall penetration time, which gives a small wall force. It is shown that runaway electrons (REs) in JET do not produce a large wall force, and this is verified with JET data. This occurs if the RE current is about half of the initial plasma current. In ITER, the RE current might be comparable to the initial current, and REs might lengthen the CQ time sufficiently to produce a large wall force.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Simulation of plasma and neutral transport in PISCES-RF using SOLPS-ITER

In this research, the fluid plasma transport code SOLPS-ITER is applied and validated against experimental data from the plasma interaction surface component experimental station (PISCES)-RF linear plasma device to establish a physics basis for plasma and neutral transport in its two magnetic field (B-field) geometry setups-(1)the cusp and (2) non-cusp or linear B-field. The main focus of this study is to understand (1) radial plasma transport (2) heat and particle loads on the upstream dump and downstream target plate, and (3) the physics of plasma-neutral interactions in PISCES-RF. The simulation setup adheres to typical PISCES-RF experimental conditions, with a 2D helicon power deposition profile as an input heating source. SOLPS-ITER simulations reproduce experimental conditions with the Bohm diffusion model for both B-field configurations of the PISCES-RF experiment. Major energy loss channels include neutral radiation and power deposited on the wall and dump plate, with only 1% of the input power reaching the target. The ionization front is well confined near the dump plate due to the heating and puffing regions. Additionally, SOLPS-ITER simulation results are also found to be in very good agreement with the particle-in-cell calculations using the code-PICOS++ which supports the validity of SOLPS in low collisionality regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Runaway electrons and ITER

The potential for damage, the magnitude of the extrapolation, and the importance of the atypical—incidents that occur once in a thousand shots—make theory and simulation essential for ensuring that relativistic runaway electrons will not prevent ITER from achieving its mission. Most of the theoretical literature on electron runaway assumes magnetic surfaces exist. ITER planning for the avoidance of halo and runaway currents is focused on massive gas or shattered-pellet injection of impurities. In simulations of experiments, such injections lead to a rapid large-scale magnetic-surface breakup. Surface breakup, which is a magnetic reconnection, can occur on a quasi-ideal Alfvénic time scale when the resistance is sufficiently small. Nevertheless, the removal of the bulk of the poloidal flux, as in halo-current mitigation, is on a resistive time scale. The acceleration of electrons to relativistic energies requires the confinement of some tubes of magnetic flux within the plasma and a resistive time scale. The interpretation of experiments on existing tokamaks and their extrapolation to ITER should carefully distinguish confined versus unconfined magnetic field lines and quasi-ideal versus resistive evolution. The separation of quasi-ideal from resistive evolution is extremely challenging numerically, but is greatly simplified by constraints of Maxwell’s equations, and in particular those associated with magnetic helicity. Thus, the physics of electron runaway along confined magnetic field lines is clarified by relations among the poloidal flux change required for an e-fold in the number of electrons, the energy distribution of the relativistic electrons, and the number of relativistic electron strikes that can be expected in a single disruption event.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Prediction of Alfvén eigenmode energetic particle transport in ITER scenarios with a critical gradient model

A reduced 1D, local, critical-gradient model of energetic particle (EP) transport by Alfvén eigenmodes (AEs)—the TGLF-EP+Alpha model—is applied to a much-studied ITER base case and variations with lower plasma current and lower current penetration. The TGLF-EP+Alpha model is a greatly reduced and computationally inexpensive model of EP transport. Such a reduced critical gradient model, while inapplicable to transport driven by strongly nonlinear or non-local abrupt events, is a valuable tool for scoping studies needed in scenario optimization for ITER and beyond. Furthermore, it relies on the assumption of critical-gradient AE transport with the critical EP density gradient determined by linear AE stability calculations in the TGLF gyro-Landau fluid code automated with the parallel-processed TGLF-EP wrapper. EP transport is treated with simultaneous drive of AEs by fusion-born alpha particles and fast ions born from a 1 MeV neutral beam injection (NBI) heating. The effect of simultaneous drive creates about 50% increased particle transport in both EP channels. High magnetic safety factor q and low shear $\hat{s}$ are generally destabilizing to AEs, but low shear tends to be more important. A tailored q-profile, steady-staterelevant scenario can reduce AE-induced EP redistribution by better than 25% over the ITER base case despite having half the total current.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Drift orbit islands of energetic particles due to 3D fields in ITER

In this study, we report a comparative study of the orbit islands formed by the guiding center drift motion of test energetic particles (EPs), including both high-energy deuterium ions and fusion-born α-particles, with respect to the magnetic islands formed by tracing the field lines, for an ITER plasma representing the 15 MA baseline scenario. The particle drift orbit is modified by the presence of static resonant magnetic perturbations (RMPs) of different toroidal mode numbers n (n = 1 and 3 in this study), forming orbit islands in the Poincaré plane by passing EPs. The key findings are as follows. (i) With an n = 1 RMP field, the size of the orbit islands combined with the total RMP field, including the plasma response, is about three times smaller than that obtained by assuming the corresponding vacuum field. (ii) The orbit island size is not sensitive to the EP energy, and is comparable to those of the magnetic islands. (iii) Passing EPs with outward radial orbital drift form orbit islands that shift inward with respect to the corresponding magnetic islands, and vice versa. This radial shift, ΔΨ, measured in the normalized equilibrium poloidal flux, is quantified as ΔΨ ≃ 0.03X for the ITER baseline plasma and assuming n = 1 RMPs, with $X\equiv {(M/{M}_{\mathrm{p}})}^{1/2}{(E[\mathrm{M}\mathrm{e}\mathrm{V}])}^{1/2}{Z}^{-1}$, where M is the particle mass, Mp is the proton mass, E is the particle energy, and Z is the particle charge number. (iv) Trapped EPs do not form drift orbit islands, even in the presence of the 3D fields produced by the 90 kAt ELM control coil current in ITER, and thus possess good confinement properties. Nevertheless, the deformation of the trapped particle drift orbits in the Poincaré plane shows that the canonical toroidal angular momentum of EPs is no longer conserved in the presence of the RMP fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Non-linear MHD modelling of edge localized modes suppression by resonant magnetic perturbations in ITER

Edge localized modes (ELMs) suppression by resonant magnetic perturbations (RMPs) was studied with the non-linear magneto-hydro-dynamic (MHD) code JOREK for the ITER H-mode scenarios at 15 MA, 12.5 MA, 10 MA/5.3 T. In this study, the main aim was to demonstrate that ELMs can be suppressed by RMPs while the divertor 3D footprints of heat and particle fluxes remain within divertor material limits. The unstable peeling–ballooning modes responsible for ELMs without RMPs were modelled first for each scenario using numerically accessible parameters for ITER. Then the stabilization of ELMs by RMPs was modelled with the same parameters. RMP spectra, optimized by the linear MHD MARS-F code, with main toroidal harmonics N = 2, N = 3, N = 4 have been used as boundary conditions of the computational domain of JOREK, including realistic RMP coils, main plasma, scrape off layer (SOL) divertor and realistic first wall. The model includes all relevant plasma flows: toroidal rotation, two fluid diamagnetic effects and neoclassical poloidal friction. With RMPs, the main toroidal harmonic and the non-linearly coupled harmonics remain dominant at the plasma edge, producing saturated modes and a continuous MHD turbulent transport thereby avoiding ELM crashes in all scenarios considered here. The threshold for ELM suppression was found at a maximum RMP coils current of 45 kAt–60 kAt compared to the coils maximum capability of 90 kAt. In the high beta poloidal steady-state 10 MA/5.3 T scenario, a rotating QH-mode without ELMs was observed even without RMPs. In this scenario with RMPs N = 3, N = 4 at 20 kAt maximum current in RMP coils, similar QH-mode behaviour was observed however with dominant edge harmonic corresponding to the main toroidal number of RMPs. The present MHD modelling was limited in time by few tens of ms after RMPs were switched on until the magnetic energy of the modes saturates. As a consequence the thermal energy was still evolving on this time scale, far from the ITER confinement time scale and hence only the form of 3D footprints on the divertor targets can be indicated within this set-up. Also note, that the divertor physics was missing in this model, so realistic values of fluxes are out of reach in this modelling. However the stationary 3D divertor and particle fluxes could be simply extrapolated from these results to the stationary situation considering that a large power fraction should be radiated in the core and SOL and only about 50 MW power is going to the divertor, which is an arbitrary, but reasonable number used here. The 3D footprints with RMPs show the characteristic splitting with the main RMP toroidal symmetry. The maximum radial extension of the footprints typically was ~20 cm in inner divertor and ~40 cm in outer divertor with stationary heat fluxes decreasing further out from the initial strike point from ~5 MW m ₋2 to ~1 MW m ₋2 assuming a total power in the divertor and walls is 50 MW. The heat fluxes remain within the divertor target and baffle areas, however with rather small margin in the outer divertor which could be an issue for the first wall especially in transient regimes when part of the plasma thermal energy is released due to switching on the RMP coils. This fact should be considered when RMPs are applied with a more favorable application before or soon after the L–H transition, although optimization is required to avoid increasing the L–H power threshold with RMPs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Reducing the L-H transition power threshold in ITER-similar-shape DIII-D hydrogen plasmas

Recent dedicated DIII-D experiments in low-torque, ITER-similar-shape (ISS) hydrogen plasmas (at a plasma current I p ~ 1.5 MA and ITER-similar edge safety factor q 95 ~ 3.6) show that the L-H transition power threshold P LH can be reduced substantially (~30%) with L-mode helium admixtures n He /n e $\leqslant$ 25%. In the ensuing H-mode, helium ion fractions n He /n H remain below 25%. H-mode normalized pressure and confinement quality are only slightly affected by helium seeding, and Z eff $\leqslant$ 2.15 (including helium and carbon content). The plasmas investigated here are electron-heat dominated, with temperatures T e (0)/T i (0) $\geqslant$ 1 and edge heat flux ratio Q e /Q i (ρ = 0.95) ~ 1.2–1.5. Without mitigation, P LH is higher by a factor of 2–3 in comparison to similar ISS deuterium plasmas. ISS hydrogen plasmas with lower plasma current I p ~ 1 MA (increased edge safety factor q 95 ~ 5.1) exhibit a substantially lower power threshold. This plasma current dependence, also observed previously on ASDEX-U and in JET, is not accounted for by the commonly used 2008 ITPA multi-machine threshold scaling, but could potentially allow H-mode access at marginal heating power during the initial plasma current ramp-up. Attempts to reduce P LH with low-field- and high-field-side hydrogen pellet injection, using 1.7 mm diameter pellets, have not demonstrated a robust threshold reduction, in contrast to successful earlier experiments with larger 2.7 mm pellets. Finally, techniques for reducing P LH are very important for ITER, in particular for accessing H-mode in hydrogen plasmas during the Pre-Fusion Power Operation-1 (PFPO-1) campaign with marginal auxiliary heating power (20–30 MW of ECH).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Trends in runaway electron plateau partial recombination by massive H 2 or D 2 injection in DIII-D and JET and first extrapolations to ITER and SPARC

Abstract Experimental trends in thermal plasma partial recombination resulting from massive D 2 injection into high-Z (Ar) containing runaway electron (RE) plateaus in DIII-D and JET are studied for the purpose of achieving sufficiently low electron density ( n e ≈ 10 18 m − 3 ) to increase RE final loss MHD levels. In both DIII-D and JET, thermal electron density n e is found to drop by ∼100 × when the thermal plasma partially recombines, with a minimum at a vacuum vessel-averaged D 2 density in the range 10 20 − 10 21 m − 3 . RE effective resistivity also drops after partial recombination, indicating expulsion of the Ar content. The n e level after partial recombination is found to increase as RE current is increased. The amount of initial Ar in the RE plateau is not observed to have a strong effect on partial recombination. Partial recombination timescales of order 5 ms in DIII-D and 15 ms in JET are observed. These basic trends and timescales are matched with a 1D diffusion model, which is then used to extrapolate to ITER and SPARC tokamaks. Within the approximations of this model, it is predicted that ITER will be able to achieve sufficiently low n e values on time scales faster than expected RE plateau vertical drift timescales (of order 100 ms), provided sufficient D 2 or H 2 is injected. In SPARC, it is predicted that achieving significant n e recombination will be challenging, due to the very high RE current density. In both ITER and SPARC, it is predicted that achieving low n e will be easier with Ar as a background impurity (rather than Ne).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Radiation induced non-linear oscillations in ITER baseline scenario plasmas in DIII-D

Abstract This work shows how the radiation brought about by metals or metal-equivalent radiators such as Kr and Xe produces non-linear dynamics on otherwise stationary β N flattops of DIII-D ITER Baseline Scenario demonstration discharges. The Kr and Xe gases are used to reproduce the radiative loss rates of W in present machines that operate at core temperatures much lower than the expected ITER temperature. Experiments on DIII-D with injection of Kr and Xe, as well as with sources of intrinsic metals reach the range of radiated fraction values expected in the ITER core and experience slow oscillations in temperature and radiated power. In many cases of high radiated fraction, the core temperature decreases enough for the safety factor profile to rise above the 1/1 rational surface, naturally eliminating sawteeth and occasionally producing a persistent helical core. The oscillations can be reproduced by a modified Lotka–Volterra system for temperature and radiated fraction if diffusion and noise are included, which indicates that the interplay between temperature and radiation can be the main cause of the cyclic nature of the system. A new physics based model which includes equations for temperature, density and input power can also reproduce the oscillations observed in the experiments. The present results suggest that the non-linearity of the system can be increased by the inclusion of the inherently non-linear alpha heating term, which is proportional to ∼ n e 2 T i 2 , and obtains oscillations in the model when added to an otherwise more stationary system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Modelling and experiment to stabilize disruptive tearing modes in the ITER baseline scenario in DIII-D

Abstract The achievement of high gain, stationary conditions in a tokamak scenario aimed at producing fusion energy in the ITER Project is crucial to the demonstration that this form of energy can be used in future reactors to provide cheap and clean energy globally. Disruptions are a challenge for the fusion energy field, in particular for the ‘ITER Baseline Scenario’ (IBS), as reproduced in the DIII-D tokamak. This work shows that a solution has been found for the m = 2/ n = 1 tearing modes that have consistently caused disruptions in the IBS: stable operation down to zero input torque was achieved by modifying the current density profile at the beginning of the pressure flattop and the ELM character later in the discharges, guided by previous results showing that the most likely cause of these instabilities is the current density profile. The coupling between sawteeth, n>2 modes and the 2/1 TMs is shown to not be statistically significant, nor the leading origin for the evolution towards instability. Ideal and resistive MHD modeling provide positive verification that a steeper ‘well’ in the region of the q = 2 rational surface leads to worse ideal stability, higher tearing index Δ’ and lower threshold Δ’ c for resistive instabilities, consistent with the experimental results. This provides confidence that the methods used in this work can be extrapolated to other devices and applied to avoid disruptions in ITER and pulsed fusion devices worldwide.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Validation of the ERO2.0 code using W7-X and JET experiments and predictions for ITER operation

Abstract The paper provides an overview of recent modelling of global material erosion and deposition in the fusion devices Wendelstein 7-X (W7-X), JET and ITER using the Monte-Carlo code ERO2.0. For validating the modelling tool in a three-dimensional environment, W7-X simulations are performed to describe carbon erosion from the graphite test divertor units, which were equipped in operational phase OP 1.2 and analysed post-mortem. Synthetic spectroscopy of carbon line emission is compared with experimental results from the divertor spectrometer measurement system, showing a good agreement in the e-folding lengths in the radial intensity profiles of carbon. In the case of metallic wall materials, earlier modelling of the Be/W environment in JET and ITER is revisited and extended with an updated set of sputtering and reflection data, as well as including the mixing model for describing the Be/W dynamics in the divertor. Motivated by recent H/D/T isotope experiments in JET, limited and diverted configuration pulses are modelled, showing the expected trend of both Be and W erosion increasing with isotope mass. For the JET diverted configuration pulses, it is shown that Be migrates predominantly to the upper part of the inner divertor where it initially leads to strong W erosion. With longer exposure time, the growth of a Be deposited layer leads to a reduction of W erosion in that region. A similar trend is observed in simulations of the ITER baseline Q = 10 scenario, however with a more symmetric Be migration pattern leading to deposition also on the outer divertor.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Non-disruptive error field measurement in DIII-D low safety factor plasmas and projection to ITER

Previous experiments in DIII-D (Paz-Soldan et al 2022 Nucl. Fusion 62 126007) introduced a method to identify intrinsic error fields (EFs) in tokamaks with minimal disruption risk by promptly healing driven magnetic islands during the conventional 'compass scan'. This paper presents recent experimental and numerical advancements in extending this approach to low q 95 plasmas, and projects its applicability to ITER. Non-disruptive EF measurement is achieved at q 95 = 4.5 and 3.9 without any initial EF correction (EFC) by reducing the time between the occurrence of the locked mode (LM) and control action to 10 ms and increasing the density 50%–100%. However, 50% correction of the intrinsic EF is required to achieve island healing at q 95 = 3.2 with 10 ms delay for the control action. Nonlinear two-fluid modeling with the TM1 code reproduces the DIII-D experimental observations, indicating that promptly turning off the 3D coil current reduces both magnetic island width and electromagnetic force, while raising the density increases plasma viscosity, facilitating magnetic island healing. The simulations show that for scenarios with q 95 = 3.2, lowering the control action time to 5 ms will lead to island healing without EFC. TM1 simulations are extended to future ITER scenarios with 5 MA and 7.5 MA plasma currents, predicting the dependence of required density rise on action time and EF amplitude. These simulations indicate that, benefiting from the much longer resistive time, island healing can be successfully achieved in ITER when taking control action 100–500 ms after a LM occurrence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Modelling the limiter ramp-up of WEST for addressing the future challenges of ITER

This paper presents a joint experimental and numerical investigation into the physics of long limited plasma ramp-up in tokamaks with tungsten (W) first walls, a critical phase for ITER operation. The comparison between the average plasma quantities simulated using the SolEdge-HDG code and the measurements taken during three successive WEST discharges after boronisation shows how challenging it is to predict this phase. While simulations reproduce the general trends at the midplane, with reasonable match in density profiles, they consistently underestimate core temperatures possibly due to too large perpendicular heat conductivity. On the contrary, at the high field side (HFS) limiter, simulations overestimate the measured quantities, and highlights the limitation of using Bohm boundary conditions at grazing magnetic angles. Experimental measurements reveal that the boron layer is rapidly eroded, on a timescale comparable to a single ITER discharge. The subsequent transition from a boron-coated to a tungsten wall increases recycling and significantly degrades the core plasma, reducing the electron temperature by nearly half due to W contamination, despite wall parameters remaining stable. Furthermore, comparisons with Langmuir probes, bolometry, reflectometry, and spectroscopy indicate that the experimental far scrape-Off layer (SOL) is significantly wider than simulated. This wide SOL implies that boron erosion extends along the entire HFS limiter rather than being confined to the contact point. This work highlights some characteristics of the plasma during this phase of the discharge and emphasizes the current modelling issues that need to be resolved in order to obtain reliable predictions concerning the ITER ramp-up.

ITER↗
DOE OSTI2020

Iterative Learning Control for Video-rate Atomic Force Microscopy

We present a control scheme for video-rate atomic force microscopy with rosette pattern. The controller structure involves a feedback internal-model-based controller and a feedforward iterative learning controller. The iterative learning controller is designed to improve tracking performance of the feedback-controlled scanner by rejecting the repetitive disturbances arising from the system nonlinearities. We investigate the performance of two inversion techniques for constructing the learning filter. We conduct tracking experiments using a two-degree-of-freedom microelectromechanical system (MEMS) nanopositioner at frame rates ranging from 5 to 20 frames per second. Furthermore, the results reveal that the algorithm converges rapidly and the iterative learning controller significantly reduces both the transient and steady-state tracking errors. We acquire and report a series of high-resolution time-lapsed video-rate AFM images with the rosette pattern.

42 ENGINEERING↗
DOE OSTI2018

Performance Model of MapReduce Iterative Applications for Hybrid Cloud Bursting

Hybrid cloud bursting (i.e., leasing temporary off-premise cloud resources to boost the overall capacity during peak utilization) can be a cost-effective way to deal with the increasing complexity of big data analytics, especially for iterative applications. However, the low throughput, high latency network link between the on-premise and off-premise resources (“weak link”) makes maintaining scalability difficult. While several data locality techniques have been designed for big data bursting on hybrid clouds, their effectiveness is difficult to estimate in advance. Yet such estimations are critical, because they help users decide whether the extra pay-as-you-go cost incurred by using the off-premise resources justifies the runtime speed-up. To this end, the current paper presents a performance model and methodology to estimate the runtime of iterative MapReduce applications in a hybrid cloud-bursting scenario. The paper focuses on the overhead incurred by the weak link at fine granularity, for both the map and the reduce phases. This approach enables high estimation accuracy, as demonstrated by extensive experiments at scale using a mix of real-world iterative MapReduce applications from standard big data benchmarking suites that cover a broad spectrum of data patterns. As a result, not only are the produced estimations accurate in absolute terms compared with experimental results, but they are also up to an order of magnitude more accurate than applying state-of-art estimation approaches originally designed for single-site MapReduce deployments.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

SU-D-17A-03: 5D Respiratory Motion Model Based Iterative Reconstruction Method for 4D Cone-Beam CT

Purpose: The purpose of this work is to develop a new iterative reconstruction method for 4D cone-beam CT (CBCT) based on a published time-independent 5D respiratory motion model. The proposed method will offer a single high-resolution image at a user-selected breathing phase and the 5D motion model parameters, which could be used to generate the breathing pattern during the CT acquisition. Methods: 5D respiratory motion model was proposed for accurately modeling the motion of lung and lung tumor tissues. 4D images are then parameterized by a reference image, measured breathing amplitude, breathing rate, two time-independent vector fields that describe the 5D model parameters, and a scalar field that describes the change in HU as a function of breathing amplitude. In contrast with the traditional method of reconstructing multiple temporal image phases to reduce respiratory artifact, 5D model based method simplify the problem into the reconstruction of a single reference image and the 5D motion model parameters. The reconstruction formulation of the reference image and scalar and vector fields is a nonlinear least-square optimization problem that consists of solving the reference image and fields alternately, in which the reference image is regularized with the total variation sparsity transform and the vector fields are solved through linearizations regularized by the H1 norm. 2D lung simulations were performed in this proof-of-concept study. Results: The breathing amplitude, its rate, and the corresponding scalar and vector fields were generated from a patient case. Compared with filtered backprojection method and sparsity regularized iterative method for the phase-by-phase reconstruction, the proposed 5D motion model based method yielded improved image quality. Conclusion: Based on 5D respiratory motion model, we have developed a new iterative reconstruction method for 4D CBCT that has the potential for improving image quality while providing needed on-table motion information.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

TU-F-18A-02: Iterative Image-Domain Decomposition for Dual-Energy CT

Purpose: Dual energy CT (DECT) imaging plays an important role in advanced imaging applications due to its material decomposition capability. Direct decomposition via matrix inversion suffers from significant degradation of image signal-to-noise ratios, which reduces clinical value. Existing de-noising algorithms achieve suboptimal performance since they suppress image noise either before or after the decomposition and do not fully explore the noise statistical properties of the decomposition process. We propose an iterative image-domain decomposition method for noise suppression in DECT, using the full variance-covariance matrix of the decomposed images. Methods: The proposed algorithm is formulated in the form of least-square estimation with smoothness regularization. It includes the inverse of the estimated variance-covariance matrix of the decomposed images as the penalty weight in the least-square term. Performance is evaluated using an evaluation phantom (Catphan 600) and an anthropomorphic head phantom. Results are compared to those generated using direct matrix inversion with no noise suppression, a de-noising method applied on the decomposed images, and an existing algorithm with similar formulation but with an edge-preserving regularization term. Results: On the Catphan phantom, our method retains the same spatial resolution as the CT images before decomposition while reducing the noise standard deviation of decomposed images by over 98%. The other methods either degrade spatial resolution or achieve less low-contrast detectability. Also, our method yields lower electron density measurement error than direct matrix inversion and reduces error variation by over 97%. On the head phantom, it reduces the noise standard deviation of decomposed images by over 97% without blurring the sinus structures. Conclusion: We propose an iterative image-domain decomposition method for DECT. The method combines noise suppression and material decomposition into an iterative process and achieves both goals simultaneously. The proposed algorithm shows superior performance on noise suppression with high image spatial resolution and low-contrast detectability. This work is supported by a Varian MRA grant.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗