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Improvement of core plasma power and current balance models for tokamak systems code considering H-mode plasma profiles

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FUSION ENGINEERING AND DESIGN 2025年 216卷

作者： Sugiyama, Shota Goto, Takuya Utoh, Hiroyasu Sakamoto, Yoshiteru Natl Inst Quantum Sci & Technol QST Rokkasho Aomori 0393212 Japan Hel Fus Co Ltd Chuo Ku Tokyo 1040061 Japan

We propose the models for the systems codes to calculate the bootstrap current fraction, neutral beam-driven current, and synchrotron radiation loss power, considering the plasma profiles with the H-mode pedestals. The scaling formulae of the bootstrap current fraction are developed for the parabolic and H-mode plasma profiles by fitting the data prepared using a magnetohydrodynamic equilibrium calculation code including a detailed model of the bootstrap current. The developed simple model of the neutral beam-driven current is applicable to arbitrary plasma profiles and shapes. We show the geometric approximation of the flux surface averaged magnetic field strength, plasma surface area, and plasma volume to calculate the synchrotron radiation loss power using the CYTRAN module with only the parameters used in the systems codes. The validity of these models is confirmed by comparing the results to those calculated by the detailed models. These models enable the systems codes to consistently consider the plasma profiles with the pedestals in the core plasma power and current balances because the other quantities in the balance equations can be evaluated by calculating the profiles of the quantities and integrating them numerically within computing time acceptable for the systems analysis.

关键词： systems code Bootstrap current fraction Neutral beam-driven current

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Development, Verification, and Validation of an Advanced systems code KP-SAM for Kairos Power Fluoride Salt-Cooled High-Temperature Reactor (KP-FHR)

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NUCLEAR SCIENCE AND ENGINEERING 2023年第5期197卷 813-839页

作者： Zhao, Haihua Fick, Lambert Heald, Alexander Zhou, Quan Richesson, Samuel Sutton, Noah Haugh, Brandon Kairos Power LLC Alameda CA 94501 USA

To meet the Kairos Power (KP) Fluoride Salt-Cooled High-Temperature Reactor (FHR) (KP-FHR) development and commercialization schedules, the System Analysis Module (SAM), which is an advanced systems code for Generation IV liquid-cooled reactors developed at Argonne National Laboratory (ANL), has been selected as the basis for the development of the KP-FHR systems code KP-SAM. This allows for an accelerated joint development effort between the KP and ANL teams. This paper presents a general overview of the KP-SAM development process, its current status, completed verification, and ongoing validation efforts. KP-SAM development follows the U.S. Nuclear Regulatory Commission Evaluation Model Development and Assessment Process framework. SAM is a high-order fully implicit transient systems code written in C++. The SAM software design, major physical models, and Jacobian Free Newton Krylov-based numerical methods are briefly discussed. KP-SAM has matured enough to be used for the unvalidated demonstration safety analysis for the low-power KP-FHR test reactor (Hermes) as part of Preliminary Safety Analysis Report work. By following the guidance of an internal KP-FHR thermal fluid Phenomena Identification and Ranking Table report, some of the most important separate-effects-test validations were completed for the first iteration. A scaled integral-effects test is under detailed design and will be built in 2022 to provide key data to validate KP-SAM for licensing safety analysis.

关键词： KP-FHR KP-SAM systems code validation

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systems code studies on the optimization of design parameters for a pulsed DEMO tokamak reactor

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FUSION ENGINEERING AND DESIGN 2017年第Nov.期123卷 206-211页

作者： Biel, W. Beckers, M. Kemp, R. Wenninger, R. Zohm, H. Forschungszentrum Julich GmbH Inst Energie & Klimaforsch D-52425 Julich Germany Univ Ghent Dept Appl Phys Ghent Belgium Culham Sci Ctr CCFE Abingdon OX14 3DB Oxon England EUROfus Power Plant Phys & Technol PPPT Dept Garching Germany Max Planck Inst Plasma Phys Boltzmannstr 2 D-85748 Garching Germany

In the European strategy towards fusion power, a demonstration tokamak fusion reactor (DEMO) is foreseen as the next single step between ITER and a power plant. The current baseline concept is a tokamak reactor with net electrical output power of P-el similar to 500 MW and plasma pulse duration of tpuise similar to 2 h. systems codes are commonly used in the design process as numerical tools for optimization studies. The key performance data of the reactor such as P-el and tpuise are depending on a variety of design and plasma parameters. In the application of systems codes within this multi-dimensional parameter space, a clear quantitative understanding of the most suitable optimization criteria has to be developed, and various physics and technology limits should be obeyed to obtain meaningful' results. In this work we use a fusion reactor systems code to perform parameter variations for a pulsed DEMO tokamak reactor. Various output quantities are presented as a basis for the quantitative assessment of the numerical results, and options for a further development of the current DEMO baseline design are proposed and briefly discussed. (C) 2017 The Author(s). Published by Elsevier B.V.

关键词： Tokamak Fusion reactor systems code

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Implementation and verification of a HELIAS module for the systems code PROCESS

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FUSION ENGINEERING AND DESIGN 2015年 98-99卷 2227-2230页

作者： Warmer, F. Beidler, C. D. Dinklage, A. Egorov, K. Feng, Y. Geiger, J. Kemp, R. Knight, P. Schauer, F. Turkin, Y. Ward, D. Wolf, R. Xanthopoulos, P. Max Planck Inst Plasma Phys D-17491 Greifswald Germany Culham Ctr Fus Energy Abingdon OX14 3DB Oxon England

In order to study design points of next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. The code package PROCESS is such a tool, widely used for tokamak systems studies. In this work, the implementation and verification of a HELIAS module into PROCESS is addressed. These HELIAS models include: a plasma geometry model based on Fourier coefficients, a basic island divertor model, as well as a coil model which combines scaling aspects based on the Helias 5-B reactor design in combination with analytic inductance and field calculations. The models are verified firstly with respect to W7-X. Secondly, the generality of the models is used to represent the tokamak which is compared against the original tokamak PROCESS models using a DEMO design as reference case. Both approaches show very good agreement. (C) 2015 Elsevier B.V. All rights reserved.

关键词： HELIAS PROCESS systems code Verification

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Definition of the basic DEMO tokamak geometry based on systems code studies

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FUSION ENGINEERING AND DESIGN 2015年第Oct.期98-99卷 1556-1560页

作者： Meszaros, Botond Bachmann, Christian Kemp, Richard Federici, Gianfranco EFDA Power Plant Phys & Technol Garching Germany Culham Sci Ctr CCFE Abingdon OX14 3DB Oxon England

This paper describes the methodology that has been developed and applied to derive the principal geometry of the main DEMO tokamak systems, in particular the radial and vertical cross section based on the systems code output parameters, while exact parameters are described elsewhere [1]. This procedure reviews the analysis of the radial and vertical build provided by the system code to verify critical integration interfaces, e.g. missing or too large gaps and/or insufficient thickness of components, and updates these dimensions based on results of more detailed analyses (e.g. neutronics, plasma scenario modelling, etc.) that were carried out outside of the system code in the past years. As well as providing a 3D configuration model of the DEMO tokamak for integrated engineering analysis, the results can also be used to refine the systems code model. This method, subject to continuous refinement, controls the derivation of the main machine parameters and ensures their coherence vis-a-vis a number of agreed controlled physics and engineering assumptions. (C) 2015 The Authors. Published by Elsevier B.V.

关键词： Nuclear fusion DEMO Configuration Geometry PROCESS systems code

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SAM: A Modern System code for Advanced Non-LWR Safety Analysis

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NUCLEAR TECHNOLOGY 2024年

作者： Hu, Rui Zou, Ling O'Grady, Daniel Mui, Travis Ooi, Zhiee Jhia Hu, Guojun Cervi, Eric Yang, Gang Andrs, David Lindsay, Alex Permann, Cody Salko, Robert Zhou, Quan Fick, Lambert Heald, Alexander Zhao, Haihua Argonne Natl Lab 9700 S Cass Ave Argonne IL 60439 USA Idaho Natl Lab 1955 Fremont Ave Idaho Falls ID 83415 USA Oak Ridge Natl Lab 1 Bethel Valley Rd Oak Ridge TN 37830 USA Kairos Power LLC 707 W Tower Ave Suite A Alameda CA 94501 USA

The System Analysis Module (SAM), developed at Argonne National Laboratory and by collaborators at other organizations, is for advanced non-light water reactor safety analysis. SAM aims to provide fast-running, modest-fidelity, whole-plant transient analysis capabilities that are essential for fast-turnaround design scoping and engineering analyses of advanced reactor concepts. To facilitate code development, SAM utilizes the MOOSE object-oriented application framework, its underlying finite element library, and linear and nonlinear solvers to leverage modern advanced software environments and numerical methods. SAM aims to solve tightly coupled physical phenomena, including fission reaction, heat transfer, fluid dynamics, and thermal-mechanical responses in advanced reactor structures, systems, and components with high accuracy and *** paper gives an overview of the SAM code development, including goals and functional requirements, physical models, current capabilities, verification and validation, software quality assurance, and examples of simulations for advanced nuclear reactor applications.

关键词： SAM systems code safety analysis advanced reactors

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A reduced-order model to estimate first wall particle and heat fluxes for systems codes

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FUSION ENGINEERING AND DESIGN 2024年 204卷

作者： Lobo, Tiago Pomella Pestchanyi, Sergey Maione, Ivan Alessio Karlsruhe Inst Technol Karlsruhe Germany

Current systems codes (SCs) do not assess the impact of plasma filament parameters and dynamics in the SOL when performing systems-level analyses. Simulation tools based on simplified turbulent transport models could address such need, such as the TOKES code, but these tend to run in timescales that are prohibitive for incorporation in SCs. This work presents a technique to build surrogate models for SCs, obtained with the modes of Reduced-Order Models (ROMs), and is exemplified with results from TOKES. These surrogates were developed with Principal Components Analysis (PCA), multivariate regression against TOKES inputs (filament temperatures, densities, and ejection speeds) and k-fold cross-validation, applied to the results after transformations based on rational powers of the inputs to optimize the prediction capabilities of the surrogates. Selection was performed with a modified Kullback-Leibler Divergence (KLD), and validation, with withheld cases. Main results include validation of the methodology due to the relatively low number of modes needed to represent more than 90% of the data variance, and the data transformation exponents that optimize the regression.

关键词： Plasma filament Far scrape-off layer (far-SOL) Reduced order model (ROM) TOKES EU-DEMO systems code

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Definition of an EU-DEMO design point robust to epistemic plasma physics uncertainties

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NUCLEAR FUSION 2025年第3期65卷 036039-036039页

作者： Coleman, M. Zohm, H. Bourdelle, C. Maviglia, F. Pearce, A. J. Siccinio, M. Spagnuolo, A. Wiesen, S. EUROfusion Boltzmannstr 2 D-85748 Garching Bavaria Germany United Kingdom Atom Energy Author Culham Sci Ctr Abingdon OX14 3DB Oxon England Max Planck Inst Plasma Phys Boltzmannstr 2 D-85748 Garching Bavaria Germany Alternat & Atom Energies Agcy F-13108 St Paul Les Durance Bouches Du Rhon France Associazone EURATOM ENEA Fus C R Frascati 65 I-00044 Rome Italy Karlsruhe Inst Technol Hermann Von Helmholtz Pl 1 D-76344 Eggenstein Leopoldshafen Baden Wurttembe Germany Dutch Inst Fundamental Energy Res Zaale 20POB 6336 NL-5600 HH Eindhoven North Bravant Netherlands

A novel approach for explicitly and robustly accounting for epistemic uncertainties in plasma confinement in the 0D design of tokamaks is presented, in which the plasma safety factor at the 95th percentile flux surface, q95, is used to apply a design margin on confinement in terms of the H-factor, H. The PROCESS code is used to explore the EU-DEMO design space in terms of aspect ratio, major radius, q95, and net electric power, Pel,net using the above approach. A comprehensive account of the input assumptions and underlying models used is given. We formulate a proposal for a new EU-DEMO design point which addresses some key shortcomings of the previous design point (EU-DEMO 2017 (Wenninger and Kembleton 2017 Technical Report EFDA_D_2NDSKT (EUROfusion);Siccinio et al 2022 Fusion Eng. Des. 176 113047)), in particular by the inclusion of an explicit and quantified design margin for epistemic confinement uncertainties, and a considerable reduction in divertor heat loads during re-attachment. We find that self-consistently including a size-able margin on confinement using our approach considerably reduces the potential Pel,net of a reactor, compared with no margin. However, we argue that by doing so, the engineering design of future tokamak fusion power plants can effectively be insulated from epistemic confinement uncertainties insofar as possible, facilitating the design process.

关键词： fusion reactor DEMO systems code uncertainties confinement

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Parameter Space Constraints for Compact Spherical Tokamak Fusion Reactors

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FUSION SCIENCE AND TECHNOLOGY 2024年第6期80卷 741-765页

作者： Maya, P. N. Deshpande, S. P. Inst Plasma Res Gandhinagar 382428 Gujarat India Homi Bhabha Natl Inst Training Sch Complex Mumbai 400094 India

Parameter space for spherical tokamak reactors is explored quantitatively to elucidate the main constraints for spherical tokamak design choices. Using a constant plasma current Ip search constraint, a set of four Ip scenarios (5, 10, 15, and 20 MA) is first explored in a wide parameter space. Considering modest but gradually increasing auxiliary power, a set of four machine configurations (major radius $$R$$R = 1.25, 1.75, 2.25, and 3.5 m) is explored next, optimizing the Ip and the bootstrap fraction. Constraints that narrow down the vast parameter space are elaborated along with critical assumptions, such as current drive efficiency, H-mode enhancement factor, nuclear shielding efficiency, and confinement scaling. Limits on the current density of the center post and how it affects the shielding are quantitatively indicated, thereby setting a lower limit on the aspect ratio.

关键词： Fusion power plant spherical tokamak reactor design systems code scoping studies

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The Influence of Confinement Scaling on Spherical Tokamak Power Plant Design and Performance

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IEEE TRANSACTIONS ON PLASMA SCIENCE 2024年第9期52卷 3985-3991页

作者： Ashe, Christopher Muldrew, Stuart I. United Kingdom Atom Energy Author Culham Campus Abingdon OX14 3DB England

Spherical tokamaks (STs) exhibit significant promise as the foundation for compact fusion power plants, offering reduced aspect ratios and enhanced plasma performance that can potentially lower capital costs compared to conventional tokamak designs. The key to achieving an optimal design lies in understanding the sensitivity of the fusion power plant to plasma energy confinement times. However, due to the intricate nature of transport physics and the scarcity of data on highly radiative plasmas required for power plants, extrapolating performance from existing machines introduces substantial uncertainties. In this study, we employed the world-leading fusion power plant systems code, PROCESS, to explore the effects of different energy confinement time scalings on scoping and determining the design of a 1-GW(e) net electric ST power plant. By comparing various commonly used scalings, we highlight the design impact of employing ST scalings versus those typically applied to conventional aspect ratios, considering both size and performance aspects. Our findings demonstrate that when allowed to freely optimize the choice of confinement scaling has negligible impact on the optimally found design point and is instead driven highly by engineering constraints. In a highly constrained scenario, the conventional IPB98(y,2) scaling consistently shows conservative values across a range of ST plasma performance scenarios. We recommend its utilization for future large design space exploration studies as a low-risk choice due to its intermediary performance between the broad scope of ST scalings and also as a proxy for addressing complex transport considerations in configuring initial ST concept designs.

关键词： Energy confinement spherical tokamak (ST) systems code tokamak

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