[en] The density profile of the fusion plasmas can be investigated by the reflectometry diagnostics. The measurement principle is based on the radar techniques which calculate the phase shift of a millimeter wave propagating into the plasma and reflected at a cut-off layer. However, this propagation is perturbed by the plasma turbulence. These phenomena affect the phase delay measurement by not well understood a process. In this work we have tried to find the mechanisms and origin of the turbulence which is responsible for the phase disturbance. We point out the role of collisionality and plasma radiation in controlling the instability and also, demonstrate that the phase delay of the probing wave is very sensitive to the plasma MHD phenomena and is less affected by the micro-turbulence. The second part of this work is the development and the use of a new heterodyne reflectometer. The principal characteristics are given. Its heterodyne detection allows the separation of phase and amplitude information from the detected signal and then to study their contribution to the mechanism of signal perturbation. The use of this reflectometer allows us to point out the following points: - a high dynamic availability, required by the large amplitude drops, often greater than 30 db; - fast sweep operation requirement to 'freeze' the plasma turbulence; - multiple reflection effects which modulate the amplitude and phase of the probing wave if they are not suppressed by filtering the detected signal; - very good localisation of the measurement (of the order of millimeter). The heterodyne reflectometer developed during this work offers several advantages of different distinct reflectometry techniques (fast sweep, absolute and differential phase measurements, heterodyne detection). It could be developed to work over higher frequency range so as to measure density profile over larger radial extension with very high performances. (author)

[en] The WEST (Tungsten Environment in Steady-state Tokamak) project in France is crucial for reaching sustainable fusion energy by simulating reactor-like conditions. WEST provides a unique facility to integrated and test technologies for Long Pulse Operations. Previous operation on Tore Supra, and numerous studies showed that all components receiving convective heat flux from particles and thermal radiations from the plasma must be water-cooled. Key components such as the divertors and limiters were identified for their high thermal load. Their design has been studied for many projects and is well documented. The integration process involved rigorous design and manufacturing standards, particularly focusing on over 2000 Copper/Stainless Steel junctions critical for preventing water leaks during operation. Additionally dedicated models were developed to compute thermal loads on all components within WEST Vacuum Vessel (VV), including the VV inner shell, ports walls, diagnostics… It was shown that depending on the plasma scenario and the reflectivity hypothesis of the surfaces, temperatures can significantly vary, forcing the design teams to take into account conservative envelop cases. Dedicated water cooled protection panels were integrated to shield the larger surfaces of the VV. Actively cooled protections on some of the most loaded diagnostics inside the ports were developed. These upgrades were installed during several annual shutdowns. The current WEST configuration includes more than 98% of the VV surface with actively cooled components. Innovative technologies are not indispensable, however all components must be manufactured and installed under high quality requirements, and stringent tightness control methods to prevent any risk of leak over the years from the numerous welds and hundreds of meters of cooling channels. However, some areas were too complex to cool down, for two main reasons: either it was too difficult to integrate water cooled components, or it was too complex to route water pipes to feed these components and to perform all welds with the required quality. Those aspects cannot be overlooked since they can severely affect the operational domain for long pulse operations. The remaining inertial structures are monitored with thermocouples and IR diagnostics. Flowmeter and thermocouples are installed in all water loops and a dedicated calorimetry diagnostic was developed. The energy balance is closed with an imbalance of about 10% of the total injected energy for most of the campaigns. Those diagnostics are mandatory to monitor the behaviour of poorly cooled components, to check that the cooling remains efficient during long pulse operation (no flow perturbation or critical flux…) and allow assessing the lifetime of inertial components. Yet, degassing occurred during long pulse operation, leading to an increase of plasma density and eventually disruptions. This is likely caused by overheating of surfaces. Indeed, thermocouples show that the temperature is steeply rising on some areas around the WEST upper divertor, reaching values above 300°C after 300s pulses and 1GJ injected. Some inertial areas not controlled by thermocouple temperature measurement may also have caused this outgassing. The path to high power continuous operation in WEST still requires improvements of protection of its internal components.

[en] Purpose: To evaluate the therapeutic efficacy of moderate-dose total abdominopelvic irradiation (TAI) in a retrospective series of pretreated non-Hodgkin's lymphomas (NHL). Methods and Materials: From 1977 to 1994, 45 patients received TAI after failure of chemotherapy (CT). According to the Working Formulation, 10 patients were diagnosed with class A (group I), 19 with class B, C, or D (follicular) (group II), and 16 with class E or more severe (group III) NHL. Irradiation consisted of two daily fractions of 0.80 Gy each for a total dose of 20 Gy. Results: Mean follow-up after TAI was 102 months (range 8-156). For the entire group, the complete response (CR) rate was 66%, the partial response (PR) rate 29%, 10-year overall survival (OS) 35%, 10-year disease-free survival (DFS) 29%, and median survival 32 months. When results between subgroups were compared, CR was 70% in group I, 84% in group II, and 44% in group III; and survival was statistically higher in group II than in groups I and III: 10-year OS 52% vs. 10% (p < 0.01) and 31% (p < 0.05), respectively, 10-year DFS 37% vs. 10% (p < 0.03) and 19% (p < 0.05), respectively. Grade III or IV complications were gastrointestinal in 27% of patients and hematologic in 25%. Conclusion: Large-field irradiation in moderate doses could provide an alternative to bone marrow transplantation in refractory NHL, especially in cases showing a follicular growth pattern

[en] Remote experiment with WEST tokamak in CEA Cadarache France from the ITER Remote Experiment Centre (REC) in Japan was successfully carried out on November 2018. The construction of the REC was carried out at the Fusion Energy Research Centre (IFERC) under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in the field of fusion energy research. First, WEST and REC sites were connected via broad band networks, that are SINET5 (Tokyo-London: 20 Gbps), GÉANT (London-Paris: 100 Gbps) and RENATER (Paris-Cadarache: 10 Gbps), over about 10000 km of transmission path length. The partners network (Partners Zone) for remote participants was used on the WEST side. It is accessible through a firewall and is separated from WEST internal networks by a second firewall. The WEST Portal in the Partners Zone allows accessing a set of information and tools such as the machine status and the experiment logbooks. The Pulse schedule editor is also available by login onto the Partner Zone server (Altair). The remote experiment was devoted to test the ITER plasma-facing components and their shaping under high heat flux. It started by the pulse preparation in REC, where the remote session leader edited the pulses schedule based on the communication between REC and WEST control room via VC system. After the approval of the Pulse Schedule by Engineer in Charge in WEST, discharges were successfully executed repeatedly with modified scenario based on intershot analysis. The large video wall (LVW) in the REC room, displayed the near-real time countdown of the plasma discharge, a video of plasma generated in WEST tokamak, the live data of the time trace of the main parameters of the plasma, such as the current and the plasma density, in addition to the view of the video conference system. In the intershot period, data analyses were carried out on Applications server located in the Partner Zone.

[en] Highlights: ► We propose a generic method to detect and correct in real time faults on magnetic sensor. ► This method is applied to Tore Supra and tested offline with real data. ► Then the method is modified to be applied to ITER ex-vessel sensor configuration. ► The method is tested on the ITER case with simulated data. - Abstract: In tokamaks, magnetic field probe sensors are used to measure the plasma position. If a sensor provides a wrong data, the error may propagate through the control loop and cause undesirable contact between the vessel wall and the plasma. In the case of a tokamak with water cooled walls, these types of event may be very serious. Despite of these unlikely faults, the potential damages call for a real time check of magnetic sensor data before using them for control. In this paper a simple and generic method based on the comparison of each sensor to a weighted sum of its neighbors is proposed. From the analysis of the residue (the result of the comparison), the fault can be detected and compensated. The method is tuned and tested against Tore Supra experimental data. Then, the method is adapted to ITER and assessed on a reference ITER scenario using simulated magnetic sensor data.

[en] Purpose: To evaluate the efficacy of total abdominopelvic (TAI) and total body irradiation (TBI) in heavily pretreated follicular non-Hodgkin's lymphoma (NHL). Patients and Methods: From 1983 to 1998, 34 patients received TAI (n = 22) or TBI (n = 12). All had Stage III or IV, Class B, C, D NHL in the working formulation and failed after receiving 1-5 regimens of chemotherapy. TAI was given at 20 Gy over a 3-week period. TBI was delivered in two successive half-body irradiations of 15 Gy over a 2-week period with a 4-week interval between each. Results: Mean follow-up from TAI or TBI was 120 months (range, 6-180). Seventy-six percent of patients achieved complete response and 24% partial response. Median survival was 62 months, 5-year and 10-year overall survival was 59% and 41%, and disease-free survival was 56% and 30%, respectively. Grade III or IV toxicity was gastrointestinal in 38% of patients and hematologic in 30%. No toxic death or delayed complications were observed. Conclusion: Extended-field irradiation is feasible and efficient after failure of chemotherapy in follicular NHL

[en] WEST is a full metallic environment tokamak, with an X-point divertor configuration. It is targeted at testing ITER like divertor prototypes made of actively cooled bulk tungsten units, in tokamak conditions during long pulse operation. Several partners around the world contribute by providing various elements like parts of the divertor, ICRH antenna, and diagnostics, or by contributing to the scientific program. Like a few other fusion devices, the experimental program is open to the participation of these WEST partners. Since the C1 campaign, two years ago, the WEST experimental sessions are prepared and conducted in a collaborative way with remote access to several IT tools developed for this purpose. Each user needs a unique account to access all these tools: · A Web Portal giving access to various web applications.

[en] Plasma control is recognized to be a crucial issue for the achievement of ITER objectives. One of the most challenging tasks for the preparation of the ITER operation will therefore be the design and qualification of a variety of control algorithms. This highlights the need for a simulation platform capable of supporting the design, integration and test of advanced control algorithms on complex physics models. With this aim, a generic multi-purpose 'flight' Simulator (GMFS) is being developed at IRFM (Institut de Recherche sur la Fusion par confinement Magnetique), CEA Cadarache, France. The GMFS is based on Kepler, a free interdisciplinary open-source Java software. Kepler will be used as a simulation platform to test and improve control algorithms before their actual use in the real control system. The physics and engineering codes complementary to the control algorithms will be supplied by the EFDA Integrated Tokamak Modelling Task Force (ITM-TF). The GMFS will be benchmarked, at the beginning, on the Tore Supra Tokamak. In this paper we will report on a test case suitable to demonstrate the feasibility of a part of GMFS, namely the development of workflows where to create and verify ITER plasma boundary feedback control algorithms. lt consists of: a) derivation of a linear plasma response model; b) design of a control diagram under the ScicosLab/Scicos open-source software; c) porting of the diagram under Kepler; d) substitution of the Kepler controller with a controller generated by a special Scicos extension; e) substitution of the simplified static linear model with the free-boundary equilibrium code CEDRES++. The test case demonstrated the feasibility of employing Kepler, ScicosLab/Scicos and other expressly made codes in view of the conception of valuable instruments for the active control of ITER and it can be considered as a first step in this direction. (authors)

[en] We present a method based on the use of toroidal harmonics and on a modelization of the poloidal field coils and divertor coils for the 2D interpolation and extrapolation of discrete magnetic measurements in a tokamak. The method is generic and can be used to provide the Cauchy boundary conditions needed as input by a fixed domain equilibrium reconstruction code like Equinox (Blum et al 2012 J. Comput. Phys. 231 960–80). It can also be used to extrapolate the magnetic measurements in order to compute the plasma boundary itself. The proposed method and algorithm are detailed in this paper and results from numerous numerical experiments are presented. The method is foreseen to be used in the real-time plasma control loop on the WEST tokamak (Bucalossi et al 2011 Fusion Eng. Des. 86 684–8). (paper)

[en] Highlights: • On Tokamaks, analogue integrators are key instruments. • Our analogue integrators is reliable and fast, thus relevant for fusion facilities. • Our integrator fulfils requirements for long time pulses of WEST Tokamak. • The performances are amongst the best never obtained for analogue integrators. - Abstract: The WEST project consists of implementing a divertor configuration in the Tore Supra Tokamak. This modification requires to renew the magnetic diagnostics and especially adding new magnetic sensors in the divertor region. This has provided an opportunity to consider some enhancements to the existing integrator modules of the magnetic diagnostics. Two prototypes have been produced using different combinations of technical solutions, aiming to achieve lower drift and higher common mode rejection ratio (CMRR). This paper presents the explored and tested ways we followed in order to create enhanced analogue integrators for long pulses. The results of extensive and comprehensive tests are summarized. The achieved performances are analyzed regarding the WEST project specifications and furthermore, the latest ITER requirements for magnetic diagnostics integrators are also considered.