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At least 217 records · Page 12

Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing “expensive” experiments, is used for numerical integration of {sup 235}U and {sup 238}U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with {sup 235}U chambers and calibrate {sup 238}U chambers by responses of the {sup 235}U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor based on the average responses of the horizontal and vertical neutron chambers remains spatially stable as the source moves and can be used in addition to the staff monitor at neutron fluxes in the detectors four orders of magnitude lower than on the first wall, where staff detectors are located. Owing to low background, detectors of neutron chambers do not need calibration in the reactor because it is actually determination of the absolute detector efficiency for 14-MeV neutrons, which is a routine out-of-reactor procedure.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Design, Manufacture, and Experimental Serviceability Validation of ITER Blanket Components

In 2014, the Russian Federation and the ITER International Organization signed two Procurement Arrangements (PAs) for ITER blanket components: 1.6.P1ARF.01 “Blanket First Wall” of February 14, 2014, and 1.6.P3.RF.01 “Blanket Module Connections” of December 19, 2014. The first PA stipulates development, manufacture, testing, and delivery to the ITER site of 179 Enhanced Heat Flux (EHF) First Wall (FW) Panels intended for withstanding the heat flux from the plasma up to 4.7MW/m{sup 2}. Two Russian institutions, NIIEFA (Efremov Institute) and NIKIET, are responsible for the implementation of this PA. NIIEFA manufactures plasma-facing components (PFCs) of the EHF FW panels and performs the final assembly and testing of the panels, and NIKIET manufactures FW beam structures, load-bearing structures of PFCs, and all elements of the panel attachment system. As for the second PA, NIKIET is the sole official supplier of flexible blanket supports, electrical insulation key pads (EIKPs), and blanket module/vacuum vessel electrical connectors. Joint activities of NIKIET and NIIEFA for implementing PA 1.6.P1ARF.01 are briefly described, and information on implementation of PA 1.6.P3.RF.01 is given. Results of the engineering design and research efforts in the scope of the above PAs in 2015–2016 are reported, and results of developing the technology for manufacturing ITER blanket components are presented.

42 ENGINEERING↗

Hydrogen lines in the infrared region and spectral background for the thomson scattering diagnostics of the iter divertor plasma

Calculations are made of the plasma spectral background, which is important for the Thomson scattering diagnostics in the ITER divertor. Theoretical grounds have been elaborated for computing the hydrogen spectral line shapes in the infrared spectral region for a divertor plasma in ITER. The shape of the P-7 Paschen line (transition n = 7 {yields} n = 3) located near the laser scattering signal has been calculated for various lines of sight in the ITER divertor. Contributions from different mechanisms of broadening the P-7 line have been examined. The spectral intensities of bremsstrahlung and photorecombination continuum have been calculated. All calculations use data on the spatial distribution of temperatures and densities of all species of plasma particles computed with the SOLPS4.3 code for basic operation regimes of the ITER divertor.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Improvements in Iterative Convergence of FUN3D Solutions

This paper presents a hierarchical adaptive nonlinear iteration method (HANIM) implemented in the NASA computational fluid dynamics code, FUN3D, to improve robustness and computational efficiency. In contrast to the legacy FUN3D iterative solver that relies on an approximate Jacobian, a simple multicolor Gauss-Seidel point-implicit iteration scheme, and linear Courant-Friedrichs-Lewy number (CFL) ramping, HANIM is based upon a hierarchy of modules including preconditioner, generalized conjugate residual, realizability check, nonlinear control, and CFL adaption modules. HANIM performance is systematically compared with the performance of the legacy solver of FUN3D and a baseline solver based on a preconditioner alone. Iterative solutions are compared for three benchmark cases: a subsonic separated flow around a hemisphere cylinder, a supersonic flow through a long duct, and a subsonic flow over the NASA wing-fuselage juncture model. Two Reynolds-averaged Navier-Stokes turbulence models are used in these computations, namely, the negative variant of the linear one-equation Spalart-Allmaras model and its nonlinear extension based on quadratic constitutive relations.

CFD↗

Fast Acting Eddy Current Driven Valve for Massive Gas Injection on ITER

Tokamak plasma disruptions present a significant challenge to ITER as they can result in intense heat flux, large forces from halo and eddy currents, and potential first-wall damage from the generation of multi-MeV runaway electrons. Massive gas injection (MGI) of high Z material using fast acting valves is being explored on existing tokamaks and is planned for ITER as a method to evenly distribute the thermal load of the plasma to prevent melting, control the rate of the current decay to minimize mechanical loads, and to suppress the generation of runaway electrons. A fast acting valve and accompanying power supply have been designed and first test articles produced to meet the requirements for a disruption mitigation system on ITER. The test valve incorporates a flyer plate actuator similar to designs deployed on TEXTOR, ASDEX upgrade, and JET [1 3] of a size useful for ITER with special considerations to mitigate the high mechanical forces developed during actuation due to high background magnetic fields. The valve includes a tip design and all-metal valve stem sealing for compatibility with tritium and high neutron and gamma fluxes.

Fast Acting Eddy current↗

The Roles and Developments needed for Diagnostics in the ITER Fusion Device

Harnessing the power from Fusion on earth is an important and challenging task. Excellent work has been carried out in this area over the years with several demonstrations of the ability to produce power. Now, a new large device is being constructed in the south of France. This is called ITER. ITER is a large-scale scientific experiment that aims to demonstrate a possibility to produce commercial energy from fusion. This project is now well underway with the many teams working on the construction and completing various aspects of the design. This device will carry up to 15 MA of plasma current and produce about 500 MW of power, 400 MW approximately in high energy neutrons. The typical temperatures of the electrons inside this device are in the region of a few hundred million Kelvin. It is maintained using a magnetic field. This device is pushing several boundaries from those currently existing. As a result of this, several technologies need to be developed or extended. This is especially true for the systems or diagnostics that measure the performance and provide the control signals for this device. A diagnostic set will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include amongst others, measurements of the plasma shape, temperature, density, impurity concentration, and particle and energy confinement times. The system will comprise about 45 individual measuring systems drawn from the full range of modern plasma diagnostic techniques, including magnetics, lasers, X-rays, neutron cameras, impurity monitors, particle spectrometers, radiation bolometers, pressure and gas analysis, and optical fibres. These devices will have to be made to work in the new and challenging environment inside the vacuum vessel. These systems will have to cope with a range of phenomena that extend the current knowledge in the Fusion field. One amongst them is the parasitic effect of the neutrons on the while all the performing with great accuracy and precision. The levels of neutral particle flux, neutron flux and neutron fluence will be respectively about 5, 10 and 10,000 times higher than the harshest experienced in today's machines. The pulse length of the fusion reaction-or the amount of time the reaction is sustained-will be about 100 times longer. (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

ITER Port Interspace Pressure Calculations

The ITER ('the way' in Latin) project in Cadarache, France, will operate the largest tokamak ever built to demonstrate the physical and technical feasibility of nuclear fusion as an energy source. The ITER vacuum vessel is equipped with 50 access ports. Each port has an opening in the bio-shield that communicates with a dedicated port cell. During tokamak operation, the bio-shield opening must be closed with a concrete plug to shield the radiation coming from the plasma. This port plug separates the port cell into a port interspace (between vacuum vessel closure lid and port plug) on the inner side and a port cell on the outer side. This paper presents calculations of pressures and temperatures in the ITER port interspace after a double-ended guillotine break (DEGB) of a pipe in the tokamak cooling water system (TCWS). It is assumed that the DEGB occurs during the worst possible conditions-during water baking operation, with water at 523 K (250 deg. C) and 4.4 MPa. These conditions are more severe than during normal tokamak operation with the water at 398 K (125 deg. C) and 2 MPa. Two computer codes are employed in these calculations: RELAP5-3D Version 4.2.1 to calculate blowdown releases from the pipe break, and MELCOR, Version 1.8.6 to calculate pressures and temperatures in the port interspace. A sensitivity study has been performed to optimize the maximum peak pressure in the port interspace volume of ITER. The optimum case was for a labyrinth area (FL131) of 0.5 m{sup 2} and a relief panel (FL135) kept open after the first cycle. This case keeps the pressure in the port interspace under 150 kPa and the temperatures under 385 K. (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Near Surface Disposal Facility Iterative Assessment - 19199

Canadian Nuclear Laboratories (CNL) is proposing the construction of a Near Surface Disposal Facility (NSDF) on the Chalk River site, in order to dispose of legacy low level waste and anticipated building debris from future decommissioning. A Performance Assessment is being conducted, using an iterative process, to assess the short and long term effects of this facility. Three iterations have already been completed, each building on the last, and a further iteration is underway to improve on the current results. This paper discusses the iterative process taken in these assessments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Neutronics for diagnostics in ITER - Paper 121

ITER, currently under construction in the south of France, will be the first tokamak producing large amounts of neutrons and having a fusion energy gain factor, which is the ratio between the fusion power produced and the input power necessary to maintain the plasma, greater than 10. ITER will be able to generate 500 MW of fusion power, from which 400 MW are 14.1 MeV neutrons from the Deuterium-Tritium fusion reactions in the thermonuclear plasma. For such nuclear machine neutronics are of paramount importance, being a design driver for many of its components, in particular its Ports and its numerous Diagnostics, for which the influence of neutronics is especially important in three main areas: - Shielding: there is a strong need to provide nuclear shielding to areas where human access is foreseen for maintenance. Also, to minimize any impact in the primary shielding components (blankets) in order to avoid significant increases in the heating of the Vacuum Vessel (VV) and Superconductive Coils. - Sensors and diagnostic components: most of the diagnostics have sensitive components that need to be shielded against radiation to survive. Efficient shielding is not always possible, and the radiation levels are usually the decisive factor for selecting the most appropriate component. - Neutron measurements: a large set of neutron diagnostics will be implemented to measure the neutron production and its emissivity profile. Although the scale and performance of ITER will not be very far from a commercial reactor, it is ultimately an experiment to improve the understanding of reactor-grade tokamak plasmas and their effect on plasma facing components. To accomplish this task, large efforts are being dedicated to design over 45 different diagnostics systems. This paper gives an overview of the main activities related to neutronics within the ITER Diagnostics Division. They include the neutron streaming problems in the Ports and the common neutronics issues of the diagnostics housed inside them, as well as the challenges of the in vessel diagnostics. Also the special calculations needed for the neutron diagnostics. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

SOLEDGE3X full vessel plasma boundary simulations of ITER non-active phase plasmas

The onset of detachment in the ITER machine is analyzed in this work through the help of 2D-axisymmetric boundary plasma simulations with the SOLEDGE3X-EIRENE code, which features a numerical domain for the plasma solver extending up to the first wall. The plasma boundary is computed in scenarios from the first non-active phase of ITER, in pure H and at 20 MW. This set of simulations is used in two aspects: first, to study the plasma detachment in the divertor, and second, the plasma conditions, fluxes, and beryllium erosion at the first wall. Here, the code results are also compared to those obtained with the well-established SOLPS-ITER code, which includes a plasma numerical domain only covering the main SOL. Results show an increase in the SOL width λ q with increasing density, and a detailed analysis is carried out, for the first time, on each of the different plasma-neutral interactions in the code’s physics model in EIRENE. The gross beryllium erosion rates of first wall panels are estimated from 2D simulations, with the aim of assessing their sensitivity to two parameters: the divertor density regime, and the presence of density shoulders in the far-SOL formed by enhanced perpendicular transport at this location. The erosion contributions from neutrals and ions are considered in each case, and the charge-exchange atoms fluxes and energy distributions are provided, highlighting the two atom populations (cold and charge-exchange).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Physics‐based iterative reconstruction for dual‐source and flying focal spot computed tomography

Purpose For single‐source helical Computed Tomography (CT), both Filtered‐Back Projection (FBP) and statistical iterative reconstruction have been investigated. However, for dual‐source CT with flying focal spot (DS‐FFS CT), a statistical iterative reconstruction that accurately models the scanner geometry and acquisition physics remains unknown to researchers. Therefore, our purpose is to present a novel physics‐based iterative reconstruction method for DS‐FFS CT and assess its image quality. Methods Our algorithm uses precise physics models to reconstruct from the native cone‐beam geometry and interleaved dual‐source helical trajectory of a DS‐FFS CT. To do so, we construct a noise physics model to represent data acquisition noise and a prior image model to represent image noise and texture. In addition, we design forward system models to compute the locations of deflected focal spots, the dimension, and sensitivity of voxels and detector units, as well as the length of intersection between x‐rays and voxels. The forward system models further represent the coordinated movement between the dual sources by computing their x‐ray coverage gaps and overlaps at an arbitrary helical pitch. With the above models, we reconstruct images by an advanced Consensus Equilibrium (CE) numerical method to compute the maximum a posteriori estimate to a joint optimization problem that simultaneously fits all models. Results We compared our reconstruction with Siemens ADMIRE, which is the clinical standard hybrid iterative reconstruction (IR) method for DS‐FFS CT, in terms of spatial resolution, noise profile, and image artifacts through both phantoms and clinical scan datasets. Experiments show that our reconstruction has a higher spatial resolution, with a Task‐Based Modulation Transfer Function (MTF task ) consistently higher than the clinical standard hybrid IR. In addition, our reconstruction shows a reduced magnitude of image undersampling artifacts than the clinical standard. Conclusions By modeling a precise geometry and avoiding data rebinning or interpolation, our physics‐based reconstruction achieves a higher spatial resolution and fewer image artifacts with smaller magnitude than the clinical standard hybrid IR.

Wang, Xiao↗

Multistage mixed precision iterative refinement

Abstract Low precision arithmetic, in particular half precision (16‐bit) floating point arithmetic, is now available in commercial hardware. Using lower precision can offer significant savings in computation and communication costs with proportional savings in energy. Motivated by this, there has been a renewed interest in mixed precision iterative refinement schemes for solving linear systems , and new variants of GMRES‐based iterative refinement have been developed. Each particular variant with a given combination of precisions leads to different condition number‐based constraints for convergence of the backward and forward errors, and each has different performance costs. The constraints for convergence given in the literature are, as an artifact of the analyses, often overly strict in practice, and thus could lead a user to select a more expensive variant when a less expensive one would have sufficed. In this work, we develop a multistage mixed precision iterative refinement solver which aims to combine existing mixed precision approaches to balance performance and accuracy and improve usability. For a user‐specified initial combination of precisions, the algorithm begins with the least expensive approach and convergence is monitored via inexpensive computations with quantities produced during the iteration. If slow convergence or divergence is detected using particular stopping criteria, the algorithm switches to use a more expensive, but more reliable variant. A novel aspect of our approach is that, unlike existing implementations, our algorithm first attempts to use “stronger” GMRES‐based solvers for the solution update before resorting to increasing the precision(s). In some scenarios, this can avoid the need to refactorize the matrix in higher precision. We perform extensive numerical experiments on a variety of random dense problems and problems from real applications which confirm the benefits of the multistage approach.

Oktay, Eda↗

Elliptic polylogarithms and iterated integrals on elliptic curves. Part I: general formalism

We introduce a class of iterated integrals, defined through a set of linearly independent integration kernels on elliptic curves. As a direct generalization of multiple polylogarithms, we construct our set of integration kernels ensuring that they have at most simple poles, implying that the iterated integrals have at most logarithmic singularities. We study the properties of our iterated integrals and their relationship to the multiple elliptic polylogarithms from the mathematics literature. On the one hand, we find that our iterated integrals span essentially the same space of functions as the multiple elliptic polylogarithms. On the other, our formulation allows for a more direct use to solve a large variety of problems in high-energy physics. As a result, we demonstrate the use of our functions in the evaluation of the Laurent expansion of some hypergeometric functions for values of the indices close to half integers.

97 MATHEMATICS AND COMPUTING↗

A simulation environment for ITER PCS development

A simulation environment known as the Plasma Control System Simulation Platform (PCSSP), specifically designed to support development of the ITER Plasma Control System (PCS), is currently under construction by an international team encompassing a cross-section of expertise in simulation and exception handling for plasma control. The proposed design addresses the challenging requirements of supporting the PCS design. This work provides an overview of the PCSSP project and a discussion of some of the major features of its design. Plasma control for the ITER tokamak will be significantly more challenging than for existing fusion devices. An order of magnitude greater performance is needed for some types of control, which together with limited actuator authority, implies that optimized individual controllers and nonlinear saturation logic are required. At the same time, consequences of control failure are significantly more severe, which implies a conflicting requirement for robust control. It also implies a requirement for comprehensive and robust exception handling. Coordinated control of multiple competing objectives with significant interactions, together with many shared uses of actuators to control multiple variables, implies that highly integrated control logic and shared actuator management will be required. It remains a challenge for the integrated technologies to simultaneously address these multiple and often competing requirements to be demonstrated on existing fusion devices and adapted for ITER in time to support its operational schedule. We describe ways in which the PCSSP will help address these challenges to support design of both the ITER PCS itself and the algorithms that will be implemented therein. and at the same time greatly reduce the cost of that development. We summarize the current status of the PCSSP design task, including system requirements and preliminary design documents already delivered as well as features of the ongoing detailed architectural design. The methods being incorporated in the detailed design are based on prior experience with control simulation environments in fusion and on standard practices prevalent in development of control-intensive industrial product designs.

43 PARTICLE ACCELERATORS↗

Application of linear multifrequency-grey acceleration to preconditioned Krylov iterations for thermal radiation transport

The thermal radiative transfer (TRT) equations comprise a radiation equation coupled to the material internal energy equation. Linearization of these equations produces effective, thermally-redistributed scattering through absorption-reemission. In this paper, we investigate the effectiveness and efficiency of Linear-Multi-Frequency-Grey (LMFG) acceleration that has been reformulated for use as a preconditioner to Krylov iterative solution methods. We introduce two general frameworks, the scalar flux formulation (SFF) and the absorption rate formulation (ARF), and investigate their iterative properties in the absence and presence of true scattering. SFF has a group-dependent state size but may be formulated without inner iterations in the presence of scattering, while ARF has a group-independent state size but requires inner iterations when scattering is present. We compare and evaluate the computational cost and efficiency of LMFG applied to these two formulations using a direct solver for the preconditioners. Finally, this work is novel because the use of LMFG for the radiation transport equation, in conjunction with Krylov methods, involves special considerations not required for radiation diffusion.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

A low-rank power iteration scheme for neutron transport criticality problems

Computing effective eigenvalues for neutron transport often requires a fine numerical resolution. Here, the main challenge of such computations is the high memory effort of classical solvers, which limits the accuracy of chosen discretizations. In this work, we derive a method for the computation of effective eigenvalues when the underlying solution has a low-rank structure. This is accomplished by utilizing dynamical low-rank approximation (DLRA), which is an efficient strategy to derive time evolution equations for low-rank solution representations. The main idea is to interpret the iterates of the classical inverse power iteration as pseudo-time steps and apply the DLRA concepts in this framework. In our numerical experiment, we demonstrate that our method significantly reduces memory requirements while achieving the desired accuracy. Analytic investigations show that the proposed iteration scheme inherits the convergence speed of the inverse power iteration, at least for a simplified setting.

97 MATHEMATICS AND COMPUTING↗

A family of independent Variable Eddington Factor methods with efficient preconditioned iterative solvers

We present a family of discretizations for the Variable Eddington Factor (VEF) equations that have high-order accuracy on curved meshes and efficient preconditioned iterative solvers. The VEF discretizations are combined with the Discontinuous Galerkin transport discretization from to form effective high-order, linear transport methods. The VEF discretizations are derived by extending the unified analysis of Discontinuous Galerkin methods for elliptic problems presented by Arnold et al. to the VEF equations. This framework is used to define analogs of the interior penalty, second method of Bassi and Rebay, minimal dissipation local Discontinuous Galerkin, and continuous finite element methods. The analysis of subspace correction preconditioners, which use a continuous operator to iteratively precondition the discontinuous discretization, is extended to the case of the non-symmetric VEF system. Numerical results demonstrate that the VEF discretizations have arbitrary-order accuracy on curved meshes, preserve the thick diffusion limit, and are effective on a proxy problem from thermal radiative transfer in both outer transport iterations and inner preconditioned linear solver iterations. We demonstrate that the VEF solution converges to the S N transport solution as the mesh is refined on both problems with smooth and non-smooth behavior in angle. Parallel performance studies show that the interior penalty VEF discretization's linear solve weak scales out to 1024 processors and strong scales well on a single node. Particular attention is paid to the parallel performance of the VEF algorithm when used in combination with a parallel block Jacobi transport sweep.

97 MATHEMATICS AND COMPUTING↗

Gas species, their evolution and segregation through the ITER vacuum systems

This paper takes the ITER fueling requirements and current knowledge of gas balance and exhaust from operating tokamaks to predict all likely gas inputs into the ITER Vacuum systems. Areas where gas dynamics modeling is relevant to the ITER design are highlighted. The design and operation of the ITER vacuum system gives an element of segregation of different gas flows and species. This paper analyses the time dependent gas segregation in the vacuum system resulting from different temperature dependences of cryogenic sorption and condensation processes of different gas species. As a specific example, the optimal transfer of Ar-41 through the vacuum system is studied with respect to its decay and the resulting effects on the design of system components.

43 PARTICLE ACCELERATORS↗