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The investigation of time-dependent power requirements for a future nuclear fusion reactor is part of the systems integration task for the European Fusion Programme. A pulsed DEMO will require intermittent and continuous electrical power to operate the various plant systems. For example, the heating and current drive system will require power during I p start-up, heating, plasma burn and I p ramp-down phases. This paper presents the modelling of these power requirements over the whole pulse cycle for all plant systems in the systems code PROCESS. In the EUROfusion baseline case the continuous power requirements are −290 MW and are dominated by the primary coolant pumping requirements. The total intermittent power requirements range from −50 to −375 MW. The time between pulses strongly influences the average plant net electrical power production and for a power plant would have to be minimised to stay commercially viable. The EUROfusion DEMO baseline is estimated to produce enough electrical power to meet the machine's aims given the current burn time, time between pulses and power requirements. [ABSTRACT FROM AUTHOR]

Highlights: [•] A three-stage compound cryopump could be used in fusion reactors. [•] Helium “ash” is adsorbed at 4.5K, and deuterium and tritium are adsorbed at 15–22K. [•] We used a free Direct Simulation Monte Carlo code (DS2 and DS2V). [•] 90% of the deuterium and tritium are successfully trapped at 15K. [•] Fuel is returned directly to the plasma, without isotopic separation or storage. [ABSTRACT FROM AUTHOR]

The official Spherical Tokamak for Energy Production mission aims to demonstrate the ability to generate net electricity from fusion with the STEP Prototype Power plant. One of the key technological and engineering challenges in fusion power plants is managing the loads on the first wall within acceptable limits. Therefore, the conceptual design development of the STEP Prototype Power plant needs to be based on load estimates derived using legitimate plasma physics assumptions through dynamic and flexible tools. The current design foresees the STEP main chamber first wall to withstand steady-state heat loads of up to ∼ 1 MW/m 2 , excluding critical regions expected to receive higher heat loads such as the baffle regions approaching the divertors. These critical areas will require ad hoc assessments and will be designed with the presence of limiters. This article focuses on the models and methodology adopted for designing the 2-D poloidal contour of the STEP first wall, based on the anticipated charged particle and radiation heat loads during normal operation. Firstly, the models adopted for calculating the charged particle and radiation heat loads are introduced. The first model is validated through benchmarking against the particle tracing code SMARDDA, while the second model is verified by comparing it with data from the MAST-U experiment. Secondly, the model used to design the 2-D first wall contour according to the heat loads is explained. We acknowledge that this preliminary design stage assumes certain simplifications, notably an axisymmetric geometry, for computational efficiency and clarity in presentation. It is understood that subsequent design phases will address the complexities of real-world engineering, including non-axisymmetric effects, transient plasma scenarios, and the impact of disruptions on the first wall design. Finally, an automatic procedure based on these models is presented for defining the 2-D poloidal contour of the STEP first wall to minimize heat loads, taking into account the need to radiate most of the alpha-particle and auxiliary heating power. By providing an overview of the models, methodology, and an automatic procedure, this paper contributes to the design process of the STEP first wall, addressing the engineering challenges associated with fusion power plant development. • Proposes a methodological approach for designing the initial first wall contour of the STEP Prototype Power plant. • Addresses significant engineering challenges associated with managing heat loads on the first wall in fusion power plants. • Introduces models and methodology for predicting heat loads resulting from charged particles and radiation. • Develops an automatic procedure for defining the 2D poloidal contour of the STEP first wall to minimize heat loads. • Contributes to the ongoing design process of the STEP first wall and addresses engineering challenges in fusion power plant development. [ABSTRACT FROM AUTHOR]

As power plant design programmes approach the transition between the pre-conceptual and conceptual design phases, the systems code PROCESS has been improved to incorporate more detailed plasma physics, engineering, and analysis modules. Unlike many systems codes, PROCESS combines the physics modelling with both technology and costs analysis. Some of the key topics in the conceptual design phase are toroidal field (TF) magnet design, divertor power handling, operational sensitivity, and economic uncertainty analysis. Models covering these areas have been integrated or improved in PROCESS. During pre-conceptual design, systems codes are an essential tool for exploring fusion power plant concepts. They allow one to model the interaction of the plant systems and quickly perform reactor optioneering. To be able to carry out these large scoping studies, the fidelity of the models can be restricted to reduce the computational time. For example, the EUROfusion EU-DEMO baseline designs are created using the systems code PROCESS and the ability to measure these trade-offs has led to important design choices being examined during the DEMO pre-conceptual design phase. Ruling out unfeasible designs allows one to efficiently identify where in the design space to carry out detailed design work. The paper describes how PROCESS has begun retooling for use in later stages of power plant conceptual design. Details are given for new additions to the PROCESS uncertainty quantification tools, high-temperature superconducting magnet model, TF coil model, and new models for spherical tokamaks (STEP). Additionally, the paper covers recent developments on the novel systems code BLUEPRINT, which is being used in both the EUROfusion and STEP projects. The paper concludes with an outlook on systems code activities at UKAEA and work with external partners. [ABSTRACT FROM AUTHOR]

Injection of electromagnetic waves in the electron-cyclotron (EC) frequency range is one of the most promising schemes to drive part of the plasma current in a tokamak fusion reactor. The theoretical calculation of the driven current, as usually performed by ray/beam tracing codes, relies on the knowledge of the magnetic equilibrium, the electron density, and temperature profiles on the one hand and of the wave injection parameters on the other. If the optimum current drive efficiency for a given scenario is sought, extensive parameter scans are usually performed to determine the best injection conditions. This is, however, not a viable approach in typical systems-code applications, where the plasma configuration is not provided in sufficient detail and parameter scans would be anyway too demanding from the computational point of view. In this case, a different approach is required. In this paper, a procedure for the evaluation of the optimum current driven by EC waves for given global parameters is proposed, which relies on a single numerical calculation of the current drive efficiency, based on the adjoint method (including momentum-conserving corrections). The results are shown to be in good agreement with the full numerical optimization of the EC current drive efficiency for a variety of reactor relevant scenarios. This simplified approach also helps clarify the physics underlying the optimum current-drive conditions and the limitations to the achievable current-drive efficiency in reactor-grade plasmas. [ABSTRACT FROM AUTHOR]

Abstract: The JET neutral beam injection (NBI) system is undergoing an upgrade of both beam power and pulse duration, which will be completed in 2011. In order to obtain an early assessment of the performance of the upgraded injectors, two positive ion neutral injectors (PINIs) with modified ion source and accelerator configuration were installed on Octant 8 Neutral Injector Box and successfully commissioned in summer 2009. Both PINIs were routinely delivering ∼2MW of deuterium neutral beam power during the JET experimental campaign in autumn 2009. These early tests allowed us to predict with confidence that the JET NBI upgrade objective of injecting 34MW of total deuterium neutral beam power into the JET plasma will be achieved. [Copyright &y& Elsevier]

Abstract: A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the magnetic configurations, and has included consideration of the roles that divertor closure and increasing magnetic connection length will play. [Copyright &y& Elsevier]