P. Lomas

Forschungszentrum Jülich, Jülich, North Rhine-Westphalia, Germany

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Publications (293)369.52 Total impact

  • [Show abstract] [Hide abstract]
    ABSTRACT: The JET ITER-like Wall (ILW) provides the same plasma facing component configuration as ITER during its active phase: beryllium in the main chamber and tungsten in the divertor. Moving from a carbon-based wall to an all metals wall requires some operational adjustment. The reduction in radiation at the plasma edge and in the divertor can lead to high power loads on the plasma facing components both in steady state and in transients and requires the development of radiative scenarios and the use of massive gas injection to mitigate disruptions. These tools are even more important now because an all metal wall is much less forgiving to thermal overloading the carbon based wall used to be. Here the impact of the first 11 months of operation on the ILW plasma facing components is discussed.
    Fusion Engineering and Design 10/2014; · 0.84 Impact Factor
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    ABSTRACT: The JET scrape-off layer (SOL) has been characterized with a reciprocating probe in inner wall (IW), and outer wall (OW), limited plasmas. Experiments revealed that SOL profiles are substantially broader (by a factor of ∼5–7.5 in the power e-folding length) for IW limited than in OW limited plasmas. Results are consistent with the larger radial turbulent transport found for IW limited plasmas. Major differences are observed between IW and OW limited plasmas on the density and electron temperature e-folding lengths, parallel flow, radial turbulent transport as well as on the temporal and spatial characteristics of the fluctuations. Experimental findings on JET suggest that the differences in the SOL characteristics for both configurations are due to a combination of a poloidal asymmetry in radial transport with a reduced cross-field transport across the last closed flux surface associated with the confinement improvement observed for OW limited plasmas.The dependence of the SOL power e-folding length on the main plasma parameters was also investigated for IW limited plasmas and a modest negative dependence on both the plasma current and the line-averaged density found. Finally, it is shown that the SOL radial transport and the amplitude of the fluctuations increase with plasma current and decrease with line-averaged density for IW limited plasmas.
    Nuclear Fusion 08/2014; 54(8). · 2.73 Impact Factor
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    ABSTRACT: The former all-carbon wall on JET has been replaced with beryllium in the main torus and tungsten in the divertor to mimic the surface materials envisaged for ITER. Comparisons are presented between Type I H-mode characteristics in each design by examining respective scans over deuterium fuelling and impurity seeding, required to ameliorate exhaust loads both in JET at full capability and in ITER.
    06/2014;
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    ABSTRACT: In order to preserve the integrity of large tokamaks such as ITER, the number of disruptions has to be limited. JET has operated previously with a low frequency of disruptions (i.e., disruption rate) of 3.4% [P. C. de Vries et al., Nucl. Fusion 51, 053018 (2011)]. The start of operations with the new full-metal ITER-like wall at JET showed a marked rise in the disruption rate to 10%. A full survey was carried out to identify the root causes, the chain-of-events and classifying each disruption, similar to a previous analysis for carbon-wall operations. It showed the improvements made to avoid various disruption classes, but also indicated those disruption types responsible for the enhanced disruption rate. The latter can be mainly attributed to disruptions due to too high core radiation but also due to density control issues and error field locked modes. Detailed technical and physics understanding of disruption causes is essential for devising optimized strategies to avoid or mitigate these events.
    04/2014; 21(5).
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    ABSTRACT: The ITER-like wall (ILW) at JET provides a unique opportunity to study the combination of material (beryllium and tungsten) that will be used for the plasma facing components (PFCs) in ITER. Both the limiters (Be) and divertor (CFC W coated and bulk W) have been designed to maximize their power handling capability. During the last experimental campaign (October 2010–July 2011) this capability has been assessed and even challenged in the case of the Be wall. The Be limiters' power handling capability (19 MW m−2 s−1/2), predicted with a simple model, has been proven to be robust by the experiments despite an unexpected power load pattern. This capability has been pushed to its limit leading to Be melt events, which revealed that the power load is toroidally asymmetric. The protection system of the ILW did not prevent melt events mainly because the protection strategy relies on the assumption that the power load is toroidally symmetric. The bulk W divertor target performed as predicted. Operations were constrained by: (i) an energy load limit (60 MJ m−2); (ii) the limited number of cycles of the surface temperature above 1200 °C in order to prevent thermal fatigue. This latter limit has been exceeded about 300 times and no signs of damage or thermal fatigue have been observed by the photogrammetric survey.
    Physica Scripta 04/2014; 2014(T159):014009. · 1.03 Impact Factor
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    ABSTRACT: This paper presents the electromagnetic modeling of the plasma current breakdown phase of the JET tokamak. The first part of this paper models the presence of the JET iron core up-down asymmetry and the effects of the eddy currents in the reconstruction of the magnetic topology needed for the plasma start. The second part describes the approach used to evaluate the ionized particle connection length inside the vacuum chamber at breakdown. The results obtained are validated using JET experimental measurements.
    IEEE Transactions on Magnetics 01/2014; 50(2):937-940. · 1.42 Impact Factor
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    ABSTRACT: The aim of the JET ITER-like Wall Project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.
    Fusion Engineering and Design 11/2013; 88(6-8). · 0.84 Impact Factor
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    ABSTRACT: This paper reports the impact on confinement and power load of the high-shape 2.5MA ELMy H-mode scenario at JET of a change from an all carbon plasma facing components to an all metal wall. In preparation to this change, systematic studies of power load reduction and impact on confinement as a result of fuelling in combination with nitrogen seeding were carried out in JET-C and are compared to their counterpart in JET with a metallic wall. An unexpected and significant change is reported on the decrease of the pedestal confinement but is partially recovered with the injection of nitrogen.
    10/2013;
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    ABSTRACT: Disruptions are a major concern for next-generation tokamaks, including ITER. Heat loads, electromag-netic forces and runaway electrons generated by disruptions have to be mitigated for a reliable operationof future machines. Massive gas injection is one of the methods proposed for disruption mitigation. Thisarticle reports the first use of massive gas injection as an active disruption protection system at JET. Duringthe 2011–2012 campaigns, 67 disruptions have been mitigated by the disruption mitigation valve (DMV)following a detection by mode lock amplitude and loop voltage changes. Most of disruptions where thevalve was intended to be used were successfully mitigated by the DMV, although at different stages ofthe typical slow disruptions of the ITER-like wall. The fraction of magnetic and thermal energy radiatedduring the disruption was found to be increased by the action of the DMV. Vertical forces dispersion wasalso reduced. No non-sustained breakdown was observed following pulses terminated by the disruptionmitigation valve.
    Fusion Engineering and Design 10/2013; 88:1101. · 0.84 Impact Factor
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    ABSTRACT: The build-up of plasma parameters following the H-mode transition in JET has been analysed in view of its consequences for the alpha power evolution in the access to burning plasma conditions in ITER. JET experiments show that the build-up of plasma temperature both at the plasma core and the plasma edge occurs in timescales comparable to the energy confinement time. In contrast, the evolution of the edge and core densities differs strongly depending on the level of plasma current in the discharge and of the associated NBI penetration. For higher plasma current H-mode discharges (Ip > 2.0–2.5 MA, depending on plasma shape), with naturally higher plasma densities for which NBI penetration is poorer, the core density evolves in much longer timescales than the edge density leading to the formation of rather hollow density profiles. These hollow density profiles persist for timescales of several energy confinement times until they are usually terminated by a sawtooth. Modelling of the JET experiments with JETTO shows that the density build-up following the H-mode transition can be described with a purely diffusive model, despite the low collisionalities of high current H-mode plasmas at JET. The consequences of these JET experimental/modelling findings for the access to burning plasma conditions in the ITER QDT = 10 scenario are presented.
    Nuclear Fusion 07/2013; 53(8):083031. · 2.73 Impact Factor
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    ABSTRACT: We present the results from a new fuelling scan database consisting of 14 high triangularity (δ ~ 0.41), type I ELMy H-mode JET plasmas. As the fuelling level is increased from low, (ΓD ~ 0.2 × 1022 el s−1, ne,ped/nGW = 0.7), to high dosing (ΓD ~ 2.6 × 1022 el s−1, ne,ped/nGW = 1.0) the variation in ELM behaviour is consistent with a transition from 'pure type I' to 'mixed type I/II' ELMs (Saibene et al 2002 Plasma Phys. Control. Fusion 44 1769). However, the pulses in this new database are better diagnosed in comparison to previous studies and most notable have pedestal measurements provided by the JET high resolution Thomson scattering (HRTS) system. We continue by presenting, for the first time, the role of pedestal structure, as quantified by a least squares mtanh fit to the HRTS profiles, on the performance across the fuelling scan. A key result is that the pedestal width narrows and peak pressure gradient increases during the ELM cycle for low fuelling plasmas, whereas at high fuelling the pedestal width and peak pressure gradient saturates towards the latter half of the ELM cycle. An ideal MHD stability analysis shows that both low and high fuelling plasmas move from stable to unstable approaching the ideal ballooning limit of the finite peeling–ballooning stability boundary. Comparison to EPED predictions show on average good agreement with experimental measurements for both pedestal height and width however when presented as a function of pedestal density, experiment and model show opposing trends. The measured pre-ELM pressure pedestal height increases by ~20% whereas EPED predicts a decrease of 25% from low to high fuelling. Similarly the measured pressure pedestal width widens by ~55%, in poloidal flux space, whereas EPED predicts a decrease of 20% from low to high fuelling. We give two possible explanations for the disagreement. First, it may be that EPED under predicts the critical density, which marks the transition from kink-peeling to ballooning-limited plasmas. Second, the stronger broadening of the experimental pedestal width than predicted by EPED is an indication that other transport related processes contribute to defining the pedestal width such as enhanced inter-ELM transport as observed at high fuelling, for mixed type I/II ELMy pulses.
    Nuclear Fusion 07/2013; 53(8):083028. · 2.73 Impact Factor
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    ABSTRACT: The JET scrape-off layer has been characterized with a reciprocating probe in inner wall, IW, and outer wall, OW, limited plasmas. Broad SOL profiles are observed for IW limited plasmas with power e-folding length substantially larger (by a factor of ∼5–7.5) than in OW limited plasmas. The properties of the fluctuations in the SOL parameters indicate larger turbulent transport for IW limited plasmas. The striking differences observed between IW and OW limited plasmas on the power e-folding length, parallel flow, turbulent transport as well as the characteristics of the fluctuations support the existence of a poloidally localized region of enhanced radial transport near the outboard midplane. The dependence of the SOL power e-folding length on the main plasma parameters was also investigated for IW limited plasmas and a modest negative dependence on both the plasma current and the line-averaged density found.
    Journal of Nuclear Materials 07/2013; 438:S189–S193. · 2.02 Impact Factor
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    ABSTRACT: As part of the ILW project, new diagnostics have been installed in order to protect the plasma-facing components (PFCs). Here we present the diagnostics used to monitor the PFC temperature, thermocouples and cameras, and assess the consistency of their measurements. In dedicated limited L-mode plasmas, the surface of the limiter tiles are heated up to 900 °C. The comparison of surface temperature measurements from IR and near IR cameras, which have been calibrated against a black-body source, leads to a Be emissivity of 0.18, comparable with the theoretical one. Energy calculation derived from thermocouples, which are embedded in both limiters and divertor target plates (W-coated CFC), is compared to a 1D model based on thermal quadrupole approach (benchmarked with an ANSYS model) associated to an inversion computation. The analysis of 20 pulses shows that a good energy balance is achieved within the error bar of the model, assessed to be of 30%.
    Journal of Nuclear Materials 07/2013; 438:1208-1211. · 2.02 Impact Factor
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    ABSTRACT: Mitigation of type-I edge-localized modes (ELMs) was observed with the application of an n = 2 field in H-mode plasmas on the JET tokamak with the ITER-like wall (ILW). Several new findings with the ILW were identified and contrasted to the previous carbon wall (C-wall) results for comparable conditions. Previous results for high collisionality plasmas with the C-wall saw little or no influence of either n = 1 or n = 2 fields on the ELMs. However, recent observations with the ILW show large type-I ELMs with a frequency of ~45 Hz were replaced by high-frequency (~200 Hz) small ELMs during the application of the n = 2 field. With the ILW, splitting of the outer strike point was observed for the first time during the strong mitigation of the type-I ELMs. The maximal surface temperature (Tmax) on the outer divertor plate reached a stationary state and has only small variations of a few degrees due to the small mitigated ELMs. In moderate collisionality H-mode plasmas, similar to previous results with the C-wall, both an increase in the ELM frequency and density pump-out were observed during the application of the n = 2 field. There are two new observations compared with the C-wall results. Firstly, the effect of ELM mitigation with the n = 2 field was seen to saturate so that the ELM frequency did not further increase above a certain level of n = 2 magnetic perturbations. Secondly splitting of the outer strike point during the ELM crash was seen, resulting in mitigation of the maximal ELM peak heat fluxes on the divertor region.
    Nuclear Fusion 06/2013; 53(7):073036. · 2.73 Impact Factor
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    ABSTRACT: The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.
    Fusion Engineering and Design 06/2013; 88(5):400–407. · 0.84 Impact Factor
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    ABSTRACT: To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be up to 30% higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be re-established only when using gas puff levels of a few 1021 es−1. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at βN ≈ 3, hybrids) have been achieved with W concentrations well below the maximum acceptable level.
    Physics of Plasmas 05/2013; 20(5). · 2.38 Impact Factor
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    ABSTRACT: The recent installation of a full metal, ITER-like, first wall provided the opportunity to study the impact of the plasma-facing materials on plasma initiation or breakdown. This study for the first time presents a full experimental characterisation of tokamak breakdown at JET, using all discharges since 2008, covering both operations with a main chamber carbon and a beryllium ITER-like main chamber wall. It was found that the avalanche phase was unaffected by the change in wall material. However, changes in out-gassing by the wall and lower carbon levels resulted in better controlled density and significantly lower radiation during the burn-through phase with the ITER-like wall. Breakdown failures, that usually developed with a carbon wall during the burn-through phase (especially after disruptions) were absent with the ITER-like wall. These observations match with the results obtained from a new model of plasma burn-through that includes plasma-surface interactions (Kim et al 2012 Nucl. Fusion 52 103016). This shows that chemical sputtering of carbon is the determining factor for the impurity content, and hence also radiation, during the burn-through phase for operations with a carbon wall. As seen experimentally, with a beryllium main wall, the plasma surface effects predicted by the model do not raise the radiation levels much above those expected for pure deuterium plasmas. With the ITER-like wall, operation with higher pre-fill pressures, and thus higher breakdown densities, was possible, which helped maintaining the density after breakdown.
    Nuclear Fusion 05/2013; 53(5):3003-. · 2.73 Impact Factor
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    ABSTRACT: The initial conditioning cycle of JET ILW is analysed and compared with restart and operation in 2008 with a carbon dominated wall. Comparable water and oxygen decay times are observed during bake-out in both cases. Despite a 2 × 10−3 mbar l/s leak rate during plasma operation, no further wall conditioning has been necessary after plasma restart in ILW, which dramatically contrasts with 2008. Plasma O content is lower with the ILW. Higher O levels are measured after nights or week-ends, BeO layers being formed and re-eroded, but do not impact plasma operation and performance. First results on isotopic wall changeover by GDC on the ILW six months of the first D2 campaign evidence a reservoir of about 3 × 1022 atoms, i.e. ten time lower than in carbon PFCs. A study in JET of the glow discharge current distribution for different ratios of the ionization mean free paths to the vessel dimensions seems to indicate sufficient toroidal and poloidal homogeneity in ITER.
    Journal of Nuclear Materials 01/2013; 438, Supplement:S1172 - S1176. · 2.02 Impact Factor
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    ABSTRACT: In the past, the Joint European Torus (JET) has operated with a first-wall composed of Carbon Fibre Composite (CFC) tiles. The thermal properties of the wall were monitored in real-time during plasma operations by the WALLS system. This software routinely performed model-based thermal calculations of the divertor and Inner Wall Guard Limiter (IWGL) tiles calculating bulk temperatures and strike-point positions as well as raising alarms when these were beyond operational limits. Operation with the new ITER-like wall presents a whole new set of challenges regarding machine protection. One example relates to the new beryllium limiter tiles with a melting point of 1278 °C, which can be achieved during a plasma discharge well before the bulk temperature rises to this value. This requires new and accurate power deposition and thermal diffusion models. New systems were deployed for safe operation with the new wall: the Real-time Protection Sequencer (RTPS) and the Vessel Thermal Map (VTM). The former allows for a coordinated stop of the pulse and the latter uses the surface temperature map, measured by infra-red (IR) cameras, to raise alarms in case of hot-spots. Integration of WALLS with these systems is required as RTPS responds to raised alarms and VTM, the primary protection system for the ITER-like wall, can use WALLS as a vessel temperature provider. This paper presents the engineering design, implementation and results of WALLS towards D-T operation, where it will act as a primary protection system when the IR cameras are blinded by the fusion reaction neutrons. The first operational results, with emphasis on its performance, are also presented.
    Fusion Engineering and Design 01/2013; · 0.84 Impact Factor
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    ABSTRACT: Recent experiments at JET combining reciprocating probe measurements upstream and infrared thermography at the plasma-facing components (PFC) on plasmas in limiter configurations show that the common approach to predicting the power load on the limiter underestimates the heat flux at the contact point by a factor 1.5?3. The current model and scaling laws used for predicting the power load onto the first wall during limiter current ramp-up/down in ITER are uncertain and a better understanding of the heat transport to the PFCs is required. The heat loads on PFCs are usually predicted by projecting the parallel heat flux associated with scrape-off layer (SOL) properties at the outer mid-plane (upstream) along the magnetic field lines to the limiter surface and deducing the surface heat flux through a cosine law, thus ignoring any local effect of the PFC on transport within the SOL. The underestimate of the heat flux is systematic in inner wall limiter configurations, independent of the plasma parameters, whereas in outer limiter configuration this is not observed, probably because of the much shorter SOL power decay length. Models that can explain this enhanced heat flux around the contact point are proposed and discussed although no definitive conclusion can be drawn.
    Nuclear Fusion 01/2013; 53(7):073016. · 2.73 Impact Factor

Publication Stats

1k Citations
369.52 Total Impact Points

Institutions

  • 2013
    • Forschungszentrum Jülich
      • Zentralabteilung für Chemische Analysen (ZCH)
      Jülich, North Rhine-Westphalia, Germany
  • 2002–2012
    • Culham Centre for Fusion Energy
      Abingdon-on-Thames, England, United Kingdom
  • 2010
    • MIT Portugal
      Porto Salvo, Lisbon, Portugal
  • 2007
    • Instituto Técnico y Cultural
      Santa Clara de Portugal, Michoacán, Mexico
  • 2002–2004
    • Max Planck Institute for Plasma Physics
      • Max Planck Institute for Plasma Physics, Greifswald
      Arching, Bavaria, Germany
  • 1989–2004
    • General Atomics
      San Diego, California, United States
  • 1999
    • University of Toronto
      Toronto, Ontario, Canada