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Cross-section of a typical SN reactor design generated by BLUEPRINT. Legend: 1-plasma 2-breeding blankets 3-vacuum vessel 4-TF coil case 5-TF coil winding pack 6-CS coils 7-PF coils 8-cryostat vacuum vessel 9-radiation shield 10-divertor 11-cryostat thermal shield 12-vacuum vessel thermal shield
Source publication
The European DEMO fusion power reactor (EU-DEMO) is still in the pre-conceptual design phase. The design strategy for the EU-DEMO hinges on investigating multiple reactor designs and technologies in parallel, progressively down-selecting these in the mid-2020's, in preparation for the conceptual design phase. The present implementation of the strat...
Citations
... To successfully design and deploy blanket and first wall components, a powerful and accessible nuclear software ecosystem is necessary to expedite the learning process and shorten design iteration timescales from the current norm. The assessment of nuclear and radiological effects in a reactor design, known as neutronics, significantly impacts downstream requirements during design and exhibits a number of computational challenges that frequently result in serious analysis bottlenecks [5][6][7][8][9][10]. ...
We present the first fully open-source capabilities for shutdown dose rate
(SDR) calculations of fusion energy facilities based on the Rigorous 2-Step
(R2S) methodology. These capabilities have been implemented in the OpenMC
Monte Carlo particle transport code, building on its existing capabilities
while also leveraging new features that have been added to the code to support
SDR calculations, such as decay photon source generation. Each of the
individual physics components in the R2S workflow---neutron transport,
activation, decay photon source generation, and photon transport---have been
verified through code-to-code comparisons with MCNP6.2 and FISPACT-II 4.0. These
comparisons generally demonstrate excellent agreement between
codes for each of the physics components. The full cell-based R2S workflow was
validated by performing a simulation of the first experimental campaign from
the Frascati Neutron Generator (FNG) ITER dose rate benchmark problem from
the Shielding INtegral Benchmark Archive and Database (SINBAD). For short cooling
times, the dose calculated by OpenMC agrees with the experimental measurements
within the stated experimental uncertainties. For longer cooling times, an
overprediction of the shutdown dose was observed relative to experiment, which
is consistent with previous studies in the literature. Altogether, these features constitute
a combination of capabilities in a single, open-source codebase to provide the
fusion community with a readily-accessible option for SDR
calculations and a platform for rapidly analyzing the performance of fusion technology.
... Existing inverse free-boundary equilibrium codes [10][11][12][13][14][15] generally solve this problem by minimising the distance of the plasma boundary from a set of control points, which may extend into the divertor region. The second-order non-linear Grad-Shafranov equation (GSE), which determines the plasma shape, is iteratively solved concurrently with the optimisation. ...
... • Intuitive: rather than requiring the user to define multiple boundary control points and weights, a single input -the number of SHs constraineddetermines the quality of core fit. A palette of exact constraints provide a natural way to create the desired divertor geometry; • Rapid: the problem is a simple constrained least squares optimisation and its solution (10) amounts to multiplication by a matrix with fewer elements than the number of PF coils squared, i.e. instantaneous on a human timescale and an excellent candidate for real-time application; • Robust: as shown in Figure 1 a good core fit is practically guaranteed for moderate ℓ max , yet avoiding iteration of the non-linear GSE means there are no issues with convergence. ...
Power exhaust is a critical challenge for spherical tokamak reactors, making the design, optimisation and control of advanced divertor configurations crucial. These tasks are greatly simplified if the poloidal magnetic fields in the core and divertor regions can be varied independently. We present a novel method which facilitates decoupling of the core plasma equilibrium from the divertor geometry optimisation and control, using vacuum spherical harmonic constraints. This has the advantage that it avoids iterative solution of the Grad-Shafranov equation, making it easy to use, rapid and reliable. By comparing a large number of MAST-U equilibrium reconstructions against their approximations using spherical harmonics, a small number (~4) of harmonics is found to be sufficient to closely reproduce the plasma boundary shape. We show experimentally that poloidal field changes designed to leave harmonics unaffected indeed have no effect on the core plasma shape. When augmented with divertor geometry constraints, this approach gives a powerful tool for creating advanced magnetic configurations, and its simplicity brings improvements in speed and robustness when solving coil position optimisation problems. We discuss the clear benefits to real-time feedback control, feed-forward scenario design and coilset optimisation with a view to future reactors.
... It also isolates trends, regions of interest in the design space and principal design parameters affecting the cost of the reactor and Cost of Electricity (COE). This will serve as a basis for further, 3-D studies and refinements in a smaller parameter-space by systemdesign codes such as PROCESS [4,21], TREND [22], BLUEPRINT [23] or ASC [19]. Due to the limits of 0D analyses and the technological assumptions, absolute estimates are therefore only indicative, but relative arguments are reliable, e.g. in cost-savings with field increase, reduced aspect ratio, improved confinement, etc. ...
High Temperature Superconductors expand the design space of stellarator power-plants toward high magnetic fields B, enabling compact major radii $R$. The present paper scans the space of B, R and other design parameters, finding solutions that are promising from a physics and engineering standpoint, while minimizing the capital cost of the power-plant and the levelized cost of fusion electricity. Similarly, it identifies minimum-cost design points for next-step burning plasma stellarator experiments of fusion gain 1 < Q < 10. The study assumes advanced stellarator configurations of reduced aspect ratio, heated by Neutral Beam Injection. Plasma-facing, flowing liquid metal walls protect it from high heat and neutron fluxes. The study relies on analytical first-principle calculations, and established zero-dimensional empirical scaling laws. Power flows are illustrated by Sankey diagrams. Plasma operating contours are used to determine the reactor’s start-up path. Sensitivity analyses are conducted to identify the most critical reactor parameters within physics, engineering and costing, quantifying their influence on the economics of the power plants. Such 0D study suggests that the assumed next generation HTS, flowing liquid metal walls, and advances in compact plasma configurations could lead to an ignited stellarator power-plant of aspect ratio A ~ 4, R ≤ 4 m, B > 9 T, and normalized plasma pressure β ~ 5 % would minimize both the cost of electricity and capital cost while achieving a net electric power of about 1 GW.
... It is also important to assess how accurate we believe current predictive models are when extrapolating to the compact high field FPP scenario parameters. Estimates of various FPP parameters and metrics (such as fusion energy gain, wall neutron loading or divertor heat flux width) can be made using combinations of systems codes (Dragojlovic et al. 2010;Stambaugh et al. 2011;Kessel et al. 2015;Kovari et al. 2016;Menard et al. 2016;Coleman & McIntosh 2019;Morris et al. 2021) and scaling laws for properties like global energy confinement (ITER Physics Expert Group on Confinement et al. 1999;Petty 2008) or proximity to the Greenwald density limit (Greenwald 2002). However, many (if not all) of these scaling laws are empirically derived from collections of limited experimental data, and their extrapolation to future burning plasmas with significantly different parameters, actuators and operating scenarios is inherently fraught with large uncertainty. ...
The OMFIT STEP (Meneghini et al. , Nucl. Fusion , vol. 10, 2020, p. 1088) workflow has been used to develop inductive and steady-state H-mode core plasma scenario use cases for a $B_0 = 8 \, {\rm T}$ , $R_0 = 4 \, {\rm m}$ machine to help guide and inform future higher-fidelity studies of core transport and confinement in compact tokamak reactors. Both use cases are designed to produce 200 MW or more of net electric power in an up-down symmetric plasma with minor radius $a = 1.4 \, {\rm m}$ , elongation $\kappa = 2.0$ , triangularity $\delta = 0.5$ and effective charge $Z_{{\rm eff}} \simeq 2$ . Additional considerations based on the need for compatibility of the core with reactor-relevant power exhaust solutions and external actuators were used to guide and constrain the use case development. An extensive characterization of core transport in both scenarios is presented, the most important feature of which is the extreme sensitivity of the results to the quantitative stiffness level of the transport model used as well as the predicted critical gradients. This sensitivity is shown to arise from different levels of transport stiffness exhibited by the models, combined with the gyroBohm-normalized fluxes of the predictions being an order of magnitude larger than other H-mode plasmas. Additionally, it is shown that although heating in both plasmas is predominantly to the electrons and collisionality is low, the plasmas remain sufficiently well coupled for the ions to carry a significant fraction of the thermal transport. As neoclassical transport is negligible in these conditions, this situation inherently requires long-wavelength ion gyroradius-scale turbulence to be the dominant transport mechanism in both plasmas. These results are combined with other basic considerations to propose a simple heuristic model of transport in reactor-relevant plasmas, along with simple metrics to quantify coupling and core transport properties across burning and non-burning plasmas.
... Unfortunately, CAD geometries tend to introduce a significant runtime penalty as opposed to the more standard constructive solid geometry representation. Additionally, preparing and cleaning the CAD model of a full fusion reactor geometry is a lengthy procedure [5]. Cleaning CAD models consists mainly of removing the small geometrical cells that would slow down MC simulations excessively, and it is normally done manually. ...
... Cleaning CAD models consists mainly of removing the small geometrical cells that would slow down MC simulations excessively, and it is normally done manually. Because of these requirements, MC fusion neutronics calculations are currently one of the design bottlenecks of fusion reactors [5]: from building the model to having the full results, presently a single design iteration of ITER requires two to three months of neutronics modelling. ...
Multi-group (MG) Monte Carlo (MC) shares many of the advantages of continuous energy (CE) Monte Carlo, while being significantly faster. Historically, all fusion neutronics calculations have been done with CE MC. This paper investigates the applicability of the faster but less accurate MG MC to a fusion-like spherical geometry by isolating and quantifying the effects of the approximations introduced. These approximations are: the energy and spatial discretisation used, scattering anisotropy, and the flux separability approximation. To reduce the maximum flux error obtained with MG MC below 10%, more than 700 energy groups would be necessary. Similarly, thick materials would have to be discretised into multiple fine layers. The error introduced by the flux separability approximation would be significant if angle-dependent cross sections or an equivalent treatment were not used. Finally, a P3 Legendre expansion must be used to represent the scattering anisotropy properly, as standard P1 anisotropy results in large errors.
... Coated conductors (CCs) made with a biaxially textured layer of the anisotropic high-temperature superconductor (HTS) REBa 2 Cu 3 O 7 (RE = rare-earth, REBCO) compounds are seen as a key-enabling technology for small fusion reactors, 8,9 owing to their large engineering critical current densities (J e ) in high magnetic fields and at temperatures well above 4.2 K. However, in service, the electromagnets in a tokamak are exposed to fluxes of fast neutrons 10 even with the addition of shielding materials, 11,12 and it is known that the performance of superconducting materials degrades as a result of the damage created by neutron irradiation. [13][14][15] THE EFFECT OF IN SITU IRRADIATION ON THE SUPERCONDUCTING PERFORMANCE OF REBA 2 CU 3 O 7−δ -COATED CONDUCTORS As with all Type II superconductors, the critical current density (J c ) of the REBCO layer in a CC is determined by the flux pinning landscape, * produced in part by microstructural defects and is influenced by temperature (T), applied magnetic field vector (B), and strain ( ε ). ...
Commercial fusion power plants will require strong magnetic fields that can only be achieved using state-of-the-art high-temperature superconductors in the form of REBa 2 Cu 3 O 7−δ -coated conductors. In operation in a fusion machine, the magnet windings will be exposed to fast neutrons that are known to adversely affect the superconducting properties of REBa 2 Cu 3 O 7−δ compounds. However, very little is known about how these materials will perform when they are irradiated at cryogenic temperatures. Here, we use a bespoke in situ test rig to show that helium ion irradiation produces a similar degradation in properties regardless of temperature, but room-temperature annealing leads to substantial recovery in the properties of cold-irradiated samples. We also report the first attempt at measuring the superconducting properties while the ion beam is incident on the sample, showing that the current that the superconductor can sustain is reduced by a factor of three when the beam is on.
Impact statement
REBa 2 Cu 3 O 7−δ high-temperature superconductors are an enabling technology for plasma confinement magnets in compact commercial fusion power plants, owing to their ability to carry very high current densities when processed as quasi-single crystals in the form of coated conductors. In service in a fusion device, the magnet windings will be exposed to a flux of fast neutrons that will induce structural damage that will adversely affect the superconducting performance, but very little data are currently available on the effect of irradiation at the cryogenic temperatures relevant for superconducting magnets. Moreover, even room-temperature annealing substantially affects superconducting properties after irradiation, so to obtain key technical data for fusion magnet designers, it is important to measure these properties in situ , under irradiation. This work shows that for the first time, it is important to consider how energetic particles directly influence superconductivity during irradiation because we observe a reduction in zero-resistance current by a factor of as much as three when an ion beam is incident on the sample. Although neutrons will not interact with the material in the same way as charged ions, primary knock-on ions from neutron damage are expected to have a similar effect to the He ⁺ ions used in our study.
Graphical abstract
... In this context, it is desirable to construct a simulation tool that can predict and assess the performance of these fusion reactor concepts while accelerating and improving their design cycle. Research groups in Germany (the PROCESS code 5 ), the United Kingdom (the BLUEPRINT and SYCOMORE codes 4 ), and China (the CFETR code 6 ) have recognized the need for integrated simulations. Coleman and McIntosh 4 state that with their integrated simulation environment BLUEPRINT, "the typical activities required to generate a DEMO design point can be sped up by four orders of magnitude (from months to minutes) paving the way for a rigorous and broad exploration of the design space." ...
... 3. Using PyNE and ALARA, use the flux calculated from Shift to activate material inside the mesh and obtain high-energy resolution photon emission rates from activated materials. 4. Use photon emission rates from ALARA as the source for forward photon transport to calculate the dose rate in areas of interest. ...
The Fusion Energy Reactor Models Integrator (FERMI) is an integrated simulation environment under development for the coupled simulation of the plasma, first wall, and blanket of fusion reactor designs. The FERMI goals are to shorten the overall design cycle while guaranteeing unprecedented accuracy, thus integrating fusion design activities, facilitating an optimal reactor design, and reducing development risks. These goals are achieved by coupling single-physics solvers into a multiphysics simulation environment (FERMI). The Integrated Plasma Simulator (IPS)–FASt TRANsport (IPS-FASTRAN) simulation framework is used for the following: plasma physics, MCNP/Shift codes for neutron and photon transport, OpenFoam for computational fluid dynamics and magnetohydrodynamics (MHD), HyPerComp Incompressible MHD solver for Arbitrary Geometry (HIMAG) for dual-coolant lead-lithium (DCLL) blankets, and DIABLO for structural mechanics simulations. These codes are coupled using the open-source library named precise Code Interaction Coupling Environment (preCICE). FERMI’s features are tested with the analysis of the liquid immersion blanket (LIB) [proposed in the Affordable Robust Compact (ARC)–class tokamak design], the DCLL blanket [proposed in the Fusion Pilot Plant (FPP) design], and other benchmark cases. The calculated figures of merit are the tritium breeding ratio, material activation, displacements per atom, shutdown dose rate, heat deposited in the vacuum vessel and blanket, temperature hot spots, and displacements caused by swelling and creep. A critical technical problem is multiphysics code coupling, which is tackled here, and the first three-dimensional (3D) simulations of the DCLL-FPP and LIB-ARC blankets are presented. To the authors’ knowledge, FERMI represents the first effort to perform 3D simulations of nuclear fusion first wall and blankets in a fully coupled multiphysics manner.
... However, the study thus does not directly elucidate the assignment of time-dependent loads of each of the active auxiliary systems to the described states. The indicated reactor power balance in the BLUEPRINT design code study [14] is exclusively intended to demonstrate the capability of the proposed design process. It is developed to reduce the design point definition time but does not propose specific reactor designs [14]. ...
... The indicated reactor power balance in the BLUEPRINT design code study [14] is exclusively intended to demonstrate the capability of the proposed design process. It is developed to reduce the design point definition time but does not propose specific reactor designs [14]. Integrated system codes like [14,15] generally tend to improve the calculation algorithms while being capable of comparing and optimizing concepts for a DEMO reactor. ...
... It is developed to reduce the design point definition time but does not propose specific reactor designs [14]. Integrated system codes like [14,15] generally tend to improve the calculation algorithms while being capable of comparing and optimizing concepts for a DEMO reactor. From the energy system engineering perspective, we thus identified a lack of a definition of operating states of a fusion power plant from a start-up to the shut-down accompanied with a concise summary of respective (presumed for future commercial power plants) power consumption and production behavior as well as underlying assumptions' set on the main system components. ...
Fusion power plants are not yet considered specifically in European long-term energy system studies. In order to include them in such studies a corresponding and valid parametrization of their operating performance has to be established despite the fact that fusion reactor design is still an ongoing effort.
The goal of the present paper is to specify and energetically represent the prospect of feasible operation and dynamics of tokamak and stellarator type fusion power plants from an energy system perspective. Special focus is given on time and operation mode dependent self-consumption. The basis of the parametrization is a one GWel power output plant. As a result, we propose the representation of fusion power plants as a system with three main components (fusion reactor, thermal energy storage (TES) and power conversion system), followed by a set of parameters for both tokamak and stellarator type devices. Five different operating states are defined for a fusion plant, depending on the required and active auxiliary subsystems. The comparison between operational dynamics of conventional and fusion power plants showed no tremendous differences due to the TES utilization. However, fusion plants had a lower full-load operation efficiency due to higher self-consumption as well as extensive pre-production losses.
... The ADCs are the output of a procedure that builds a 3D model starting from a 2D shape. The 2D optimised geometries come from CREATE_NL [16] software and NOVA [17] code aiming at plasma control. The electromagnetic analysis performed for the 2D axisymmetric cases focuses on the vertical stability of the plasma and on the assessment of plasma behaviour in case of disruptions. ...
The DEMO tokamak exhibits extraordinary complexity due to the constraints and requirements pertaining to different fields of physics and engineering. The multidisciplinary nature of the DEMO system makes its design phase extremely challenging since different and often opposite requirements need to be accounted for. Toroidal field (TF) coils generate the toroidal magnetic field required to magnetically confine the plasma particles and support at the same time the poloidal field coils. They must bear tremendous loads deriving from electromagnetic interactions between the coil currents and the generated magnetic field. An efficient tokamak design aims at minimizing the energy stored in its magnetic field and hence at reducing the toroidal volume within the TF coils whose shape would hence ideally mimic co-centrically the shape of the plasma. In order to bear the enormous forces a D-shape is most suitable for the TF coils as it allows them to resist the very large compression on the inner side and to carry the electro-magnetic (EM) pressure mainly by membrane stresses preventing large bending to occur on the outer side. At the same time the divertor structures must fit within the TF coils and this requires adaptations of the TF coil shape in the case of so-called advanced divertor configurations (ADCs), which require larger divertor structures. This article shows the TF coils adapted to ADCs using a structural optimisation procedure applied to the reference shape. The introduced strategy takes as structural optimum the iso-stress profile associated to each coil. A continuous transformation, based on radial basis functions mesh morphing, turns the baseline finite element (FE) model into its iso-stress counterpart, with a series of intermediate configurations available for electromagnetic and structural investigations as output. The adopted strategy allowed to determine, for each of the ADC cases, a candidate shape. Static membrane stress levels during magnetization could be reduced significantly from more than 700 MPa to below 450 MPa.
... In fact, simulating CAD geometries instead of CSG generally slows down MC simulations. Additionally, preparing and cleaning the CAD model of a full fusion reactor geometry is a lengthy procedure [76]. Cleaning CAD models consists mainly of removing the small geometrical cells that would slow down MC simulations excessively, and it is normally done manually. ...
... Cleaning CAD models consists mainly of removing the small geometrical cells that would slow down MC simulations excessively, and it is normally done manually. Because of these requirements, MC fusion neutronics calculations are currently one of the design bottlenecks of fusion reactors [76]: from building the model to having a full set of results, presently a single design iteration of ITER requires two to three months of neutronics modelling. ...
Monte Carlo (MC) neutron transport simulations can be performed with several types of nuclear data representations; the most common ones are continuous energy (CE) and multi-group (MG) data. CE is the highest fidelity option, since it allows each cross section resonance to be represented precisely. However, due to the need to perform lengthy binary searches and expensive memory look-ups, CE MC is normally computationally expensive. Some alternatives, such as MG MC, introduce approximations but are computationally cheaper. The benefits of the two data types can be combined to speed up MC calculations while retaining as much accuracy as possible. This thesis proposes and investigates such acceleration methods for MC, based on the use of variable fidelity nuclear data.
The first method proposed consists of approximating thermal cross sections with simple analytical functions, like high order polynomials and exponentials. During a CE MC calculation, whenever a cross section in the approximated energy range is needed, a cheap function evaluation can be done instead of more expensive operations. The method, applied to thermal reactor types, produced a speed-up of up to 15%, and is extremely simple to implement. While a small bias can be introduced in the results, several ways to minimise this have been successfully tested.
The second method proposed can accelerate fission source convergence in eigenvalue calculations by simulating most of the inactive cycles with MG nuclear data. The MG cross sections needed can be generated online at the beginning of a calculation, during a few initial CE cycles. On 3D reactor models with burnt fuel composition, the acceleration provided is the largest and it is up to a factor of 4. On the other hand, convergence to a slightly different fission source can introduce a small error in the results and an increase in the standard deviation between independent runs.
Finally, a new application area for MG MC was investigated, namely neutron transport in fusion reactors: the sources of error introduced by MG MC when simulating fusion neutronics models were quantified. A simplified spherical model was simulated; the results indicated that, to contain the error introduced within acceptable limits, high order scattering anisotropy must be accounted for, and angle-dependent MG cross sections must be used to compensate for the flux separability approximation. To produce a low error with MG MC fine energy group structures and material discretisations are necessary too. This part of the thesis did not focus directly on a variable fidelity nuclear data method, but paved the way for future research in that direction.
