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Ström, P., Petersson, P., Rubel, M., Bergsåker, H., Bykov, I., Frassinetti, L., . . . et al., . (2019). Analysis of deposited layers with deuterium and impurity elements on samples from the divertor of JET with ITER-like wall. Journal of Nuclear Materials, 516, 202-213
Open this publication in new window or tab >>Analysis of deposited layers with deuterium and impurity elements on samples from the divertor of JET with ITER-like wall
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2019 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 516, p. 202-213Article in journal (Refereed) Published
Abstract [en]

Inconel-600 blocks and stainless steel covers for quartz microbalance crystals from remote corners in the JET-ILW divertor were studied with time-of-flight elastic recoil detection analysis and nuclear reaction analysis to obtain information about the areal densities and depth profiles of elements present in deposited material layers. Surface morphology and the composition of dust particles were examined with scanning electron microscopy and energy-dispersive X-ray spectroscopy. The analyzed components were present in JET during three ITER-like wall campaigns between 2010 and 2017. Deposited layers had a stratified structure, primarily made up of beryllium, carbon and oxygen with varying atomic fractions of deuterium, up to more than 20%. The range of carbon transport from the ribs of the divertor carrier was limited to a few centimeters, and carbon/deuterium co-deposition was indicated on the Inconel blocks. High atomic fractions of deuterium were also found in almost carbon-free layers on the quartz microbalance covers. Layer thicknesses up to more than 1 micrometer were indicated, but typical values were on the order of a few hundred nanometers. Chromium, iron and nickel fractions were less than or around 1% at layer surfaces while increasing close to the layer-substrate interface. The tungsten fraction depended on the proximity of the plasma strike point to the divertor corners. Particles of tungsten, molybdenum and copper with sizes less than or around 1 micrometer were found. Nitrogen, argon and neon were present after plasma edge cooling and disruption mitigation. Oxygen-18 was found on component surfaces after injection, indicating in-vessel oxidation. Compensation of elastic recoil detection data for detection efficiency and ion-induced release of deuterium during the measurement gave quantitative agreement with nuclear reaction analysis, which strengthens the validity of the results.

Keywords
Fusion, Tokamak, Plasma-wall interactions, ToF-ERDA, NRA, SEM
National Category
Fusion, Plasma and Space Physics
Research subject
Physics
Identifiers
urn:nbn:se:kth:diva-240616 (URN)10.1016/j.jnucmat.2018.11.027 (DOI)000458897100020 ()2-s2.0-85060313456 (Scopus ID)
Note

QC 20190125

Available from: 2018-12-20 Created: 2018-12-20 Last updated: 2019-10-29Bibliographically approved
Jepu, I., Matthews, G. F., Widdowson, A., Rubel, M., Fortuna-Zalesna, E., Zdunek, J., . . . Lungu, C. P. (2019). Beryllium melting and erosion on the upper dump plates in JET during three ITER-like wall campaigns. Nuclear Fusion, 59(8), Article ID 086009.
Open this publication in new window or tab >>Beryllium melting and erosion on the upper dump plates in JET during three ITER-like wall campaigns
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2019 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, no 8, article id 086009Article in journal (Refereed) Published
Abstract [en]

Data on erosion and melting of beryllium upper limiter tiles, so-called dump plates (DP), are presented for all three campaigns in the JET tokamak with the ITER-like wall. High-resolution images of the upper wall of JET show clear signs of flash melting on the ridge of the roofshaped tiles. The melt layers move in the poloidal direction from the inboard to the outboard tile, ending on the last DP tile with an upward going waterfall-like melt structure. Melting was caused mainly by unmitigated plasma disruptions. During three ILW campaigns, around 15% of all 12376 plasma pulses were catalogued as disruptions. Thermocouple data from the upper dump plates tiles showed a reduction in energy delivered by disruptions with fewer extreme events in the third campaign, ILW-3, in comparison to ILW-1 and ILW-2. The total Be erosion assessed via precision weighing of tiles retrieved from JET during shutdowns indicated the increasing mass loss across campaigns of up to 0.6 g from a single tile. The mass of splashed melted Be on the upper walls was also estimated using the high-resolution images of wall components taken after each campaign. The results agree with the total material loss estimated by tile weighing (similar to 130 g). Morphological and structural analysis performed on Be melt layers revealed a multilayer structure of re-solidified material composed mainly of Be and BeO with some heavy metal impurities Ni, Fe, W. IBA analysis performed across the affected tile ridge in both poloidal and toroidal direction revealed a low D concentration, in the range 1-4 x 10(17) D atoms cm(-2).

Place, publisher, year, edition, pages
Institute of Physics (IOP), 2019
Keywords
JET, ITER-like wall, beryllium, erosion, melt layer motion
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-255301 (URN)10.1088/1741-4326/ab2076 (DOI)000472804000005 ()2-s2.0-85070830243 (Scopus ID)
Note

QC 20190807

Available from: 2019-08-07 Created: 2019-08-07 Last updated: 2019-10-04Bibliographically approved
Litnovsky, A., Voitsenya, V. S., Reichle, R., Walsh, M., Razdobarin, A., Dmitriev, A., . . . Mertens, P. (2019). Diagnostic mirrors for ITER: research in the frame of International Tokamak Physics Activity. Nuclear Fusion, 59(6), Article ID 066029.
Open this publication in new window or tab >>Diagnostic mirrors for ITER: research in the frame of International Tokamak Physics Activity
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2019 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, no 6, article id 066029Article in journal (Refereed) Published
Abstract [en]

Mirrors will be used as first plasma-viewing elements in optical and laser-based diagnostics in ITER. Deterioration of the mirror performance due to e.g. sputtering of the mirror surface by plasma particles or deposition of impurities will hamper the entire performance of the affected diagnostic and thus affect ITER operation. The Specialists Working Group on First Mirrors (FM SWG) in the Topical Group on Diagnostics of the International Tokamak Physics Activity (ITPA) plays an important role in finding solutions for diagnostic first mirrors. Sound progress in research and development of diagnostic mirrors in ITER was achieved since the last overview in 2009. Single crystal (SC) rhodium (Rh) mirrors became available. SC rhodium and molybdenum (Mo) mirrors survived in conditions corresponding to similar to 200 cleaning cycles with a negligible degradation of reflectivity. These results are important for a mirror cleaning system which is presently under development. The cleaning system is based on sputtering of contaminants by plasma. Repetitive cleaning was tested on several mirror materials. Experiments comprised contamination/cleaning cycles. The reflectivity SC Mo and Rh mirrors has changed insignificantly after 80 cycles. First in situ cleaning using radiofrequency (RF) plasma was conducted in EAST tokamak with a mock-up plate of ITER edge Thomson Scattering (ETS) with five inserted mirrors. Contaminants from the mirrors were removed. Physics of cleaning discharge was studied both experimentally and by modeling. Mirror contamination can also be mitigated by protecting diagnostic ducts. A deposition mitigation (DeMi) duct system was exposed in KSTAR. The real-time measurement of deposition in the diagnostic duct was pioneered during this experiment. Results evidenced the dominating effect of the wall conditioning and baking on contamination inside the duct. A baffled cassette with mirrors was exposed at the main wall of JET for 23,6 plasma hours. No significant degradation of reflectivity was measured on mirrors located in the ducts. Predictive modeling was further advanced. A model for the particle transport, deposition and erosion at the port-plug was used in selecting an optical layout of several ITER diagnostics. These achievements contributed to the focusing of the first mirror research thus accelerating the diagnostic development. Modeling requires more efforts. Remaining crucial issues will be in a focus of the future work of the FM SWG.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2019
Keywords
ITER, diagnostic mirrors, in situ mirror cleaning, recovery of reflectivity, single crystal mirrors, mirror protection, ITPA
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-252366 (URN)10.1088/1741-4326/ab1446 (DOI)000467461600008 ()2-s2.0-85067584703 (Scopus ID)
Note

QC 20190718

Available from: 2019-07-18 Created: 2019-07-18 Last updated: 2019-07-18Bibliographically approved
Brezinsek, S., Kirschner, A., Mayer, M., Baron-Wiechec, A., Borodkina, I., Borodin, D., . . . Widdowson, A. (2019). Erosion, screening, and migration of tungsten in the JET divertor. Nuclear Fusion, 59(9), Article ID 096035.
Open this publication in new window or tab >>Erosion, screening, and migration of tungsten in the JET divertor
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2019 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, no 9, article id 096035Article in journal (Refereed) Published
Abstract [en]

The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influx of W into the confined region. The screening of W by the divertor and the transport of W in the plasma determines largely the W content in the plasma core, but the W source strength itself has a vital impact on this process. The JET tokamak experiment provides access to a large set of W erosion-determining parameters and permits a detailed description of the W source in the divertor closest to the ITER one: (i) effective sputtering yields and fluxes as function of impact energy of intrinsic (Be, C) and extrinsic (Ne, N) impurities as well as hydrogenic isotopes (H, D) are determined and predictions for the tritium (T) isotope are made. This includes the quantification of intra- and inter-edge localised mode (ELM) contributions to the total W source in H-mode plasmas which vary owing to the complex flux compositions and energy distributions in the corresponding phases. The sputtering threshold behaviour and the spectroscopic composition analysis provides an insight in the dominating species and plasma phases causing W erosion. (ii) The interplay between the net and gross W erosion source is discussed considering (prompt) re-deposition, thus, the immediate return of W ions back to the surface due to their large Larmor radius, and surface roughness, thus, the difference between smooth bulk-W and rough W-coating components used in the JET divertor. Both effects impact on the balance equation of local W erosion and deposition. (iii) Post-mortem analysis reveals the net erosion/deposition pattern and the W migration paths over long periods of plasma operation identifying the net W transport to remote areas. This W transport is related to the divertor plasma regime, e.g. attached operation with high impact energies of impinging particles or detached operation, as well as to the applied magnetic configuration in the divertor, e.g. close divertor with good geometrical screening of W or open divertor configuration with poor screening. JET equipped with the ITER-like wall (ILW) provided unique access to the net W erosion rate within a series of 151 subsequent H-mode discharges (magnetic field: B-t = 2.0 T, plasma current: I-p = 2.0 MA, auxiliary power P-aux = 12 MW) in one magnetic configuration accumulating 900 s of plasma with particle fluences in the range of 5-6 x 10(26) D(+ )m(-2) in the semi-detached inner and attached outer divertor leg. The comparison of W spectroscopy in the intra-ELM and inter-ELM phases with post-mortem analysis of W marker tiles provides a set of gross and net W erosion values at the outer target plate. ERO code simulations are applied to both divertor leg conditions and reproduce the erosion/deposition pattern as well as confirm the high experimentally observed prompt W re-deposition factors of more than 95% in the intra- and inter-ELM phase of the unseeded deuterium H-mode plasma. Conclusions to the expected divertor conditions in ITER as well as to the JET operation in the DT plasma mixture are drawn on basis of this unique benchmark experiment.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2019
Keywords
tungsten divertor, JET, erosion and deposition, ERO modelling, W spectroscopy, ITER divertor
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-256241 (URN)10.1088/1741-4326/ab2aef (DOI)000478620300002 ()2-s2.0-85072087051 (Scopus ID)
Note

QC 20191022

Available from: 2019-10-22 Created: 2019-10-22 Last updated: 2019-12-11Bibliographically approved
Moon, S., Petersson, P., Rubel, M., Fortuna-Zalesna, E., Widdowson, A., Jachmich, S., . . . Contributors, J. E. (2019). First mirror test in JET for ITER: Complete overview after three ILW campaigns. Nuclear Materials and Energy, 19, 59-66
Open this publication in new window or tab >>First mirror test in JET for ITER: Complete overview after three ILW campaigns
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2019 (English)In: Nuclear Materials and Energy, E-ISSN 2352-1791, Vol. 19, p. 59-66Article in journal (Refereed) Published
Abstract [en]

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015–2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011–2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20–80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5 × 10 17 cm −2 , W is up to 4.3 × 10 17 cm −2 , while the content of Ni is the greatest in the outer divertor: 3.8 × 10 17 cm −2 . Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 × 10 16 cm −2 . The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s − 1 ) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
52.40 Hf, Diagnostic mirrors, Erosion-deposition, First Mirror Test, ITER-like Wall, JET
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-246438 (URN)10.1016/j.nme.2019.02.009 (DOI)2-s2.0-85061529095 (Scopus ID)
Note

QC 20190401

Available from: 2019-04-01 Created: 2019-04-01 Last updated: 2019-04-01Bibliographically approved
Rubel, M. (2019). Fusion Neutrons: Tritium Breeding and Impact on Wall Materials and Components of Diagnostic Systems. Journal of fusion energy, 38(3-4), 315-329
Open this publication in new window or tab >>Fusion Neutrons: Tritium Breeding and Impact on Wall Materials and Components of Diagnostic Systems
2019 (English)In: Journal of fusion energy, ISSN 0164-0313, E-ISSN 1572-9591, Vol. 38, no 3-4, p. 315-329Article in journal (Refereed) Published
Abstract [en]

A concise overview is given on the impact of fusion neutrons on various classes of materials applied in reactor technology: plasma-facing, structural and functional tested for tritium production and for diagnostic systems. Tritium breeding in the reactor blanket, fuel cycle and separation of hydrogen isotopes are described together with issues related to primary (tritium) and induced radioactivity. Neutron-induced damage and degradation of material properties are addressed. Material testing under neutron fluxes and safety issues associated with handling components in the radioactive environment are described. A comprehensive list of references to monographs and research papers is included to help navigation in literature.

Place, publisher, year, edition, pages
Springer, 2019
Keywords
Controlled fusion, Tritium, Radiation damage, Transmutation, Fuel cycle
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-255727 (URN)10.1007/s10894-018-0182-1 (DOI)000476509600006 ()2-s2.0-85053238998 (Scopus ID)
Note

QC 20190813

Available from: 2019-08-13 Created: 2019-08-13 Last updated: 2019-08-13Bibliographically approved
Murari, A., Bekris, N., Figueiredo, J., Kim, H.-T., Von Thun, C. P., Balboa, I., . . . Widdowson, A. (2019). Implementation and exploitation of JET enhancements at different fuel mixtures in preparation for DT operation and next step devices. Paper presented at 30th Symposium on Fusion Technology (SOFT), SEP 16-21, 2018, Messina, ITALY. Fusion engineering and design, 146, 741-744
Open this publication in new window or tab >>Implementation and exploitation of JET enhancements at different fuel mixtures in preparation for DT operation and next step devices
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2019 (English)In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 146, p. 741-744Article in journal (Refereed) Published
Abstract [en]

In the framework of the ITER Physics Department of the EUROfusion Consortium, JET mission is focused on preparing for the next step devices by developing high performance scenarios and testing reactor relevant techniques and technologies. In terms of scenario development, a complete scan in isotopic composition has already started and, according to the present schedule, should include full T operation and culminate in a 50/50 DT campaign by 2020. In DD, significant increases in input power and various adjustments in fueling have allowed reaching the IPBp8(y,2) scaling up to 3 MA. A complete set of diagnostic upgrades is being implemented to support operation and to guarantee adequate scientific exploitation of the experiments. With regard to the technology, the main enhancements and specific tests are focused on maximizing the returns from the unprecedented 14 MeV neutron field and the tritium fuel cycle.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
JET, DT operation, Burning plasma, Diagnostics, Fuel cycle, Tritium blanket module
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-262988 (URN)10.1016/j.fusengdes.2019.01.068 (DOI)000488307400165 ()2-s2.0-85060354974 (Scopus ID)
Conference
30th Symposium on Fusion Technology (SOFT), SEP 16-21, 2018, Messina, ITALY
Note

QC 20191031

Available from: 2019-10-31 Created: 2019-10-31 Last updated: 2019-10-31Bibliographically approved
Coad, J. P., Rubel, M., Likonen, J., Bekris, N., Brezinsek, S., Matthew, G. F., . . . Widdowson, A. M. (2019). Material migration and fuel retention studies during the JET carbon divertor campaigns. Fusion engineering and design, 138, 78-108
Open this publication in new window or tab >>Material migration and fuel retention studies during the JET carbon divertor campaigns
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2019 (English)In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 138, p. 78-108Article in journal (Refereed) Published
Abstract [en]

The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing C-13-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Fusion, JET, Divertor, Carbon, Plasma-facing components
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-244549 (URN)10.1016/j.fusengdes.2018.10.002 (DOI)000457663100013 ()2-s2.0-85056661344 (Scopus ID)
Note

QC 20190313

Available from: 2019-03-13 Created: 2019-03-13 Last updated: 2019-03-13Bibliographically approved
Meyer, H., Frassinetti, L., Garcia Carrasco, A., Ratynskaia, S. V., Rubel, M., Thorén, E., . . . et al., . (2019). Overview of physics studies on ASDEX Upgrade. Nuclear Fusion, 59(11), Article ID 112014.
Open this publication in new window or tab >>Overview of physics studies on ASDEX Upgrade
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2019 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, no 11, article id 112014Article in journal (Refereed) Published
Abstract [en]

The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q(95) = 5.5, beta(N) <= 2.8) at low density. Higher installed electron cyclotron resonance heating power <= 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with P-sep/R <= 11 MW m(-1) with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently H-H98(y,H-2) <= 1.1. This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of T-i and E-r allow for inter ELM transport analysis confirming that E-r is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of 'natural' no ELM regimes have been extended. Stable I-modes up to n/n(GW) <= 0.7 have been characterised using beta-feedback. Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle-measured for the first time-or the cross-phase angle of T-e and n(e) fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow T-e profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvenic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2019
Keywords
nuclear fusion, magnetic confinement, tokamak physics, ITER, DEMO
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-263334 (URN)10.1088/1741-4326/ab18b8 (DOI)000490603100002 ()2-s2.0-85072124840 (Scopus ID)
Note

QC 20191106

Available from: 2019-11-06 Created: 2019-11-06 Last updated: 2019-11-06Bibliographically approved
Joffrin, E., Bergsåker, H., Bykov, I., Frassinetti, L., Fridström, R., Garcia Carrasco, A., . . . et al., . (2019). Overview of the JET preparation for deuterium-tritium operation with the ITER like-wall. Nuclear Fusion, 59(11), Article ID 112021.
Open this publication in new window or tab >>Overview of the JET preparation for deuterium-tritium operation with the ITER like-wall
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2019 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, no 11, article id 112021Article in journal (Refereed) Published
Abstract [en]

For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D-T mixtures since 1997 and the first ever D-T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D-T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D-T preparation. This intense preparation includes the review of the physics basis for the D-T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D-T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfven eigenmode antennas, neutral gauges, radiation hard imaging systems...) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D-T campaign provides an incomparable source of information and a basis for the future D-T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2019
Keywords
fusion power, JET, tritium, isotope
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-260157 (URN)10.1088/1741-4326/ab2276 (DOI)000484122200001 ()2-s2.0-85070875113 (Scopus ID)
Note

QC 20190926

Available from: 2019-09-26 Created: 2019-09-26 Last updated: 2019-10-04Bibliographically approved
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Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0001-9901-6296

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