United Kingdom Atomic Energy Authority

Abstract At JET, Deuterium-Helium-3 (D- 3 He) experiments using ion cyclotron resonance heating (ICRH) with the three-ion scheme were conducted, involving thermal Deuterium (D), thermal Helium-3 ( 3 He), and a fast D beam injected via Neutral Beam Injection (NBI) and further heated using ICRH. In these plasmas, intense neutron emission at the level of ≈ 10 16 n/s was observed, and a non-thermal neutron spectrum was measured with the JET time of flight neutron spectrometer. Furthermore, the spatial profile of ≈ 16.4 MeV gamma rays originating from the 3 He(D,γ) 5 Li reaction was observed. This is a weak branch (≈ 10 -5 relative probability) of the more common 3 He(D,p)α reaction and it can be used to assess the production of alpha particles in these experiments. In this work we analyse a set of data from nuclear diagnostics in D-D NBI - 3 He experiments at JET and compare them with synthetic diagnostics based on TRANSP simulations of the fast deuterium distribution function, using the so-called "kick operator" to model power transfer from the wave to the beam fast deuterons. The neutron rate predicted by the TRANSP simulations broadly agrees with data measured with fission chambers. The shape of the neutron emission spectrum is also well understood, once finite Larmor radius effects are introduced into calculations. Data from the gamma ray camera are used to infer the alpha birth profile through tomographic reconstructions based on the TREVISO code. These were compared with corresponding calculations starting from TRANSP results and showing an overall good agreement with the experimental tomographies. The validated TRANSP results are finally used to discuss the modifications of the fast deuterium distribution function and of the alpha birth profile in a set of discharges.

The EU-DEMO design is exploring several blanket models, including the Water Cooled Lithium Lead blankets. In this blanket model, liquid LiPb will be used as both the tritium breeder and multiplier. This liquid will experience a spatially varying neutron field as it moves across the tokamak. The production of radioisotopes would depend on the flow parameters of LiPb across this spatially varying neutron field. In this study, we have explored the production of radioisotopes in such a field and compared it with an equivalent, static, non-flowing simulation. While static calculations are simpler and computationally less expensive, we find that the predicted production of a radionuclide deviates from flowing fluid simulations to varying degrees, depending on the radionuclide’s half-life. We also demonstrate how the computational methodology developed can be used to explore online tritium extraction parameters and their impact on the End-of-Life tritium inventory. • Simulations of flowing fluid demonstrate that the production of a radionuclide can deviate from equivalent predictions without flow, depending on its half-life. • A computational methodology has been developed to explore the impact of tritium extraction parameters on the end-of-life inventory. • In LiPb, 210 Po is produced continuously via beta decay, and so its concentration is strongly influenced by the flow path • The short half-life of 207m Pb means that its predicted concentration is the same in both the static and flowing scenarios.

Abstract This paper reports on the generic prediction capability of full electromagnetic plasma initiation modelling with DYON, which was carried out for the first time in fusion research by the joint modelling of the International Tokamak Physics Activity—Integrating Operation Scenario group. The following devices were included in the experiment database: VEST (spherical torus, copper coils, Stainless steel wall, R / a = 0.3 m/0.2 m, V v = 3.7 m 3 ), MAST-U (spherical torus, copper coils, C wall, R / a = 0.7 m/0.5 m, V v = 55 m 3 ), EAST (conventional tokamak, superconducting coils, metallic wall, R / a = 1.85 m/0.5 m, V v = 38 m 3 ), DIII-D (conventional tokamak, copper coils, C wall, R / a = 1.67 m/0.65 m, V v = 35 m 3 ), and KSTAR (conventional tokamak, superconducting coils, C wall, R / a = 1.8 m/0.5 m, V v = 55 m 3 ). Despite the different hardware features of the devices, the required operating spaces of the loop voltage induction and prefill gas pressure for inductive plasma initiation in each device were successfully reproduced by the predictive simulations with DYON using only the individual hardware design and the control room input data for each discharge. This successful validation across multiple machines demonstrates that the full electromagnetic DYON modelling can capture the essential physics of inductive plasma initiation. The simulation settings commonly employed for all modelling and the modifications necessary to account for the discrepancies between individual devices are reported. Predictions for ITER based on the multi-machine validation indicate that a wide range of prefill gas pressures exists for the Townsend breakdown and the plasma burn-through (0.01–1.5 mPa).

Abstract Understanding the tolerance of REBa 2 Cu 3 O <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mrow> <mml:msub> <mml:mi/> <mml:mrow> <mml:mn>7</mml:mn> <mml:mo>−</mml:mo> <mml:mi>x</mml:mi> </mml:mrow> </mml:msub> </mml:mrow> </mml:math> (REBCO) high-temperature superconductors to neutron damage is essential for compact fusion reactor design. Here we report the first in situ measurements of the superconducting performance of REBCO coated conductor (CC) during neutron irradiation. The CC was exposed to 14 MeV neutrons at fluxes up to 7.3 × 10 7 n cm −2 s −1 while carrying current at 40 K. In situ voltage measurements revealed no instantaneous disruption of superconductivity, but the critical current ( <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mrow> <mml:msub> <mml:mi>I</mml:mi> <mml:mrow> <mml:mi mathvariant="normal">c</mml:mi> </mml:mrow> </mml:msub> </mml:mrow> </mml:math> ) decreased linearly with fluence at 0.03 A per 10 12 n cm −2 . Careful analysis excluded thermal effects as the origin of this degradation. Consistent with a point-defect damage mechanism, superconducting performance was recovered fully following room temperature annealing. This observed onset of <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mrow> <mml:msub> <mml:mi>I</mml:mi> <mml:mrow> <mml:mi mathvariant="normal">c</mml:mi> </mml:mrow> </mml:msub> </mml:mrow> </mml:math> reduction occurred several orders of magnitude below the critical neutron fluences anticipated by previous ex situ studies, indicating that service lifetimes of REBCO-based magnets may be substantially shorter than current projections.

Abstract This work introduces a reduced model to predict the ‘resonance overlap threshold’ governing transport of fast ions (FIs) by toroidal Alfvén eigenmodes (TAEs) in tokamak plasmas. TAE-FI resonance occurs within distinct ‘resonance regions’ of particle phase space, which grow wider when the TAE mode amplitude increases. If these resonance regions are separate then FI transport is limited and localised; if multiple resonance regions overlap then large-scale stochastic transport of FIs can occur, jeopardising confinement. Predicting the resonance overlap threshold between these two scenarios is an important problem in the field of FI transport modelling. Current workflows rely on computationally expensive orbit-following codes; the new reduced resonance overlap model provides a much simpler and faster alternative, for ease of implementation in integrated models. In this paper, we apply this approach to passing particles and find good agreement with more detailed numerical modelling, including HALO code simulations. Models based on this approach could provide a useful step in increasing efficiency of predictive modelling for next-generation fusion reactors.

The STEP prototype power plant (SPP) will utilize a double null diverted (DND) plasma topology to manage very high divertor heat fluxes within a compact design. The DND configuration provides advantages for divertor power balancing and H-mode access while it is connected, i.e., <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\delta R_{\text {sep}}$</tex-math> </inline-formula> is small. Due to vertical instabilities and limitations of active vertical control, the plasma may deviate from a connected DND resulting in imbalanced divertor heat loads and potentially loss of H-mode, both compromising the viability of the plasma scenario. To mitigate this, we propose a novel double null balance (DNB) control scheme to control <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\delta R_{\text {sep}}$</tex-math> </inline-formula> in conjunction with plasma vertical position control, preventing disconnection. In this study, we test an implementation of DNB control on TCV (tokamak à configuration variable). This controller demonstrates greatly improved control of <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\delta R_{\text {sep}}$</tex-math> </inline-formula>, maintaining connection during vertical steps of up to +/−3 cm. The controller also demonstrates effective balancing of radiated power during these excursions, preventing radiated power spikes to the inner divertor regions. These positive results are a very encouraging proof of concept for the DNB control, warranting further study and consideration for potential integration into the SPP and other future double-null power plant designs.

TCV is a medium sized tokamak equipped with a large suite of plasma diagnostics, a versatile array of poloidal shaping coils, an electron cyclotron heating system, and a two source neutral beam injection (NBI) system. The NBI system is capable of a simultaneous injection of energetic neutrals in the co-and counter-current directions. The resulting fast ion (FI) populations are used to study a multitude of FI-driven instabilities. While plasma instabilities in the below 1 MHz frequency range can be observed via standard diagnostics such as soft x-ray detectors, magnetics, fast ion loss detectors, and reflectometers, fluctuations in the range of &gt;10 MHz require a specialized diagnostic. For this purpose the TCV tokamak has been equipped with a dedicated ion cyclotron emission (ICE) diagnostic, following a similar approach to AUG [1] and W7-X [2]. The diagnostic consists of a pair of magnetic coils (8 turns, 177 nH, 16 mm long, 7 mm in diameter), oriented orthogonally to each other to detect magnetic field fluctuations in the toroidal and poloidal directions. The coils are housed in a stainless steel electrostatic shield with a slit and are installed on the torus low field side at the midplane location, behind graphite protection tiles. The coil electric outputs are routed to the vacuum feedthroughs via a pair of coaxial cables, at which point one of the outputs is grounded at the feedthrough, from the airside. The second output is then routed to a rectifying radio frequency wave detector and a fast digitizer in the diagnostics rack. The rectifying detector is sensitive to signals in the 10-100 MHz frequency range and can measure the signal amplitude and the phase difference between the two probes, while the frequency information of the signal is lost. The benefit of this detection method is that the output signal can be digitized at a “slow” speed (200 kHz in the case of TCV) with a low cost digitizer and the data volume per plasma discharge is low. The fast digitizer, on the other hand, preserves the frequency information, albeit with a larger data volume. For the case of TCV, a 250 MHz sampling rate digitizer has been temporarily used, while a dedicated 1 GHz digitizer is currently being implemented. First results of high frequency (&gt;10 MHz) instabilities detected in the presence of energetic ions will be presented.