HELOKA experimental facility, aimed to test the test blanket modules (TBM) mock-ups and prototypes, is under construction at the Research Centre Karlsruhe. HELOKA is an “8-shape” helium loop (HL), which feeds the TBM test section with helium (1.4kg/s, 300°C, 8MPa). The maximum temperature at the test section outlet is 500°C. HELOKA data acquisition and control system (DACS) comprises the facility control system and instrumentation. In addition to DACS and working independently from it, the central interlock and safety system (CISS) ensures the facility machine protection and personnel safety, with specific interlock logic, acting upon off-normal events or conditions. The paper briefly describes HELOKA DACS and CISS and reports in detail on the current development status: the installation and commissioning of DACS stage 1 consisting of the supervisory control system (SCS) and the control and monitoring for the water cooling system (WCS) and the related power supply.

This paper presents the thermal-hydrodynamic model used to simulate the behavior of the Helium Loop Karlsruhe (HELOKA) facility and describes the mechanism used to control various loop parameters. An accurate control of the temperature during the warm-up and flat top phases is achieved solely by controlling the heater power. During testing campaign, when the helium flow has to be cooled, the power of the heater is set to zero and the temperature is controlled using a control valve installed in the by-pass of the heater. The adopted solution reduces the harmonic distortions when operating at reduced power while keeping the investment cost low.

Helium loop Karlsruhe (HELOKA) is a new test facility, in advanced status of design at the Forschungszentrum Karlsruhe, which will be used to test the helium cooled pebble bed (HCPB) breeder blanket for ITER under realistic pressure, temperature and He flow conditions. A redundant water cooling system, built in the 1980s, with a nominal heat load of 7MW, will be used to remove the heat from the He loop. The paper describes first the thermo-hydraulic analysis performed on the water cooling system and then new feedback and feed-forward control systems for the electric drivers, based on the actual heat load, to increase efficiency and save energy.

The testing of the ITER toroidal field model coil (TFMC) in the background field of the EURATOM-LCT coil took place in autumn 2002 at the TOSKA facility of the Forschungszentrum Karlsruhe in the framework of the ITER R&D programme. The maximum currents in the two coils, in combined operation, were 16kA in the LCT coil and 80kA in the TFMC, respectively. The heat load of both coils, including the eddy current losses in the passive structures and the joule losses due to the joint resistances, was removed by a secondary loop of forced flow supercritical He. About 2% of the stored energy was transferred to the cryogenic system after all the safety discharges of both coils together. Most of the energy (about 98%) was extracted and transferred to the dump resistors of both coils, located outside the vacuum vessel. A computer code, based on the full inductance and resistance matrices, has been developed with SIMULINK™. After validation with experimental data the code has been used to perform circuit analysis and to evaluate the power dissipation and energy transferred to the cryogenic plant and to the external power circuits.

The Bus Bar III (BBIII), fabricated within the Toroidal Field Model Coil Task of the International Thermonuclear Experimental Reactor (ITER), was tested at the Forschungszentrum Karlsruhe, Germany, in the spring of 2004. The BBIII consists of an approximately 7 m long NbTi dual-channel conductor with a thick square stainless steel jacket, cooled by forced flow supercritical He. It was energized with currents up to 80 kA and operates in its self magnetic field (up to /spl sim/0.8 T). The BBIII was instrumented with Hall-probe heads and arrays, voltage rings and longitudinal voltage taps for electro-magnetic measurements, in order to get experimental data to be used for the validation of a recently developed hybrid thermal-hydraulic electro-magnetic code (THELMA), as well as for the assessment of the possibility of performing a reliable reconstruction of the current distribution in the conductor cross section under controlled conditions. In the tests, current ramps at different rates were applied to characterize the conductor time constants, while two different resistive heaters (one upstream of the BBIII inlet, another one directly on the BBIII jacket) were separately operated in order to approach current sharing in the conductor and to observe the related current re-distribution. In this paper, a summary of the collected experimental results is presented, with particular emphasis on those aspects more relevant for the forthcoming THELMA analysis.

The tests of the toroidal field model coil (TFMC) were completed in 2002 in the TOSKA facility of Forschungszentrum Karlsruhe, Germany. Operation reached a combined 80 kA in the TFMC and 16 kA in the LCT coil, resulting in a peak electromechanical load very close to that expected in the full-size ITER TF coils (800 kN/m). Here we concentrate on the measurements of the current sharing temperature T/sub cs/ of the TFMC conductor, possibly the highlight of the whole test campaign. These tests were performed by increasing in steps the helium inlet temperature T/sub in/ in double pancake DP1, resulting in an increasing normal voltage V across the DP1.1 and DP1.2 conductors, and were repeated for several combinations of currents in the TFMC and in the LCT coil. The analysis of the V - T/sub in/ characteristic by means of the M&M code allows to self-consistently deriving an estimate of T/sub cs/, as well as an indirect assessment of the "average" strain state in the conductor. The TFMC isolated strand has also been very recently characterized at different applied uniaxial strain, and preliminary results indicate a stronger reduction of carrying capacity compared to the extrapolation from Summers scaling used in the analysis so far. As a consequence, the performance of the TFMC conductor, as preliminarily re-evaluated here, appears more in line with the strand performance than in previous analysis, although a BI-dependent "degradation" is still present.

The tests of the Toroidal Field Model Coil (TFMC) were completed in 2002 in the TOSKA facility of Forschungszentrum Karlsruhe, Germany. Operation reached a combined 80 kA in the TFMC and 16 kA in the LCT coil, resulting in a peak electromechanical load very close to that expected in the full-size ITER TF coils (800 kN/m). Here we concentrate on the measurements of the current sharing temperature (T-cs) of the TFMC conductor, possibly the highlight of the whole test campaign. These tests were performed by increasing in steps the helium inlet temperature T-in in double pancake DP1, resulting in an increasing normal voltage V across the DP1.1 and DP1.2 conductors, and were repeated for several combinations of currents in the TFMC and in the LCT coil. The analysis of the V - T-in characteristic by means of the M&M code allows to self-consistently deriving an estimate of T-cs, as well as an indirect assessment of the "average" strain state in the conductor. The TFMC isolated strand has also been very recently characterized at different applied uniaxial strain, and preliminary results indicate a stronger reduction of carrying capacity compared to the extrapolation from Summers scaling used in the analysis so far. As a consequence, the performance of the TFMC conductor, as preliminarily re-evaluated here, appears more in line with the strand performance than in previous analysis, although a BI-dependent "degradation" is still present.

A computational tool is being developed for the analysis of superconductive magnets, combining detailed descriptions of termination joints and cables to thermo-hydraulic (TH) models. In parallel, an experiment (Stability Experiment Upgrade—SexUp) has been designed with the target to study the current distribution in cable in conduit conductors (CICC). Finally, the establishment of a reliable method for the measurement of the current distribution profile on the cable cross section is being implemented on the ITER toroidal field model coil (TFMC).

The construction and testing of the Toroidal Field Model Coil (TFMC) is part of one of the ITER large R&D projects. The main goal was to demonstrate the feasibility and the mechanical integrity of the design. One of the highlights of the first test phase was to measure the current sharing temperature, T-CS, of the conductor by heating the helium entering from the inlet. Because neither temperature sensors nor voltage taps are positioned inside the coil, only the helium inlet temperature and the voltage along the whole conductor length can be used. for the evaluation of T-CS. In addition, an inner pancake joint is located at the inlet in a rather high magnetic field and the peak field region is only about 1.5 m apart from the joint. The determination of the T-CS relies on the exact knowledge of the thermohydraulics of both the joint and the conductor region. The paper describes and compares the different numerical models used for the evaluation of the T-CS. Nine T-CS tests at different coil currents were performed, all ending in a quench. The measured T-CS is in good agreement with the expectations.

As a joint European effort an ITER Toroidal Field Model Coil (TFMC) was manufactured in industry and has been assembled in the TOSKA test facility of the Forschungszentrum Karlsruhe. After cool down and acceptance tests of the racetrack shaped coil made of a Nb3Sn cable in conduit conductor the first test campaign started in July 2001 reaching the design current of 80 kA within one week. This paper describes the assembly in the test facility, summarizes the acceptance tests before and after cool down, and reports on the first test results.

The test of the Toroidal Field (TF) Model Coil of the International Thermonuclear Experimental Reactor (ITER) has been the opportunity to measure the DC resistances of all the joints of a real coil and to compare them to values previously measured on prototype full-size joint samples. This paper describes and discusses the different methods used for measuring all the joint resistances, and gives the results of joint resistance measurements (1-2 n/spl Omega/ range) as function of the coil current up to the maximum value of 80 kA. Comparisons with resistances measured on prototype joints in relevant field/current conditions are presented and discussed.

The vertical position of the JET plasma is normally unstable and feedback stabilisation is therefore needed. A new power amplifier (fast radial field amplifier-FRFA) based on GTO inverters has been procured to cope with configurations characterized by plasma having a high degree of vertical instability, beyond the stabilising capability of the phase controlled thyristor power converter used so far. The new power amplifier is composed of four identical subunits which can be connected in two different configurations to achieve output voltages up to 10 kV (at 2.5 kA) and output currents up to 5 kA (at 5 kV). The amplifier is characterised by a peak output power of 25 MW and by switching frequencies of the individual GTOs of 1 kHz. The series connection of more inverters allows more voltage levels (up to nine) to be made available on the load. The paper presents the results of the extensive series of tests on a dummy load performed on the amplifier both at factory, where a complete subunit was assembled, and at the JET sites. The tests included the achievement of the full performances and the assessment of the correct operation of all the required control modes

The first measurement of the T/sub cs/ at 80 kA was successfully performed in the ITER TFMC. Two resistive heaters are available on the inlet piping of the P1.1 and P1.2 pancakes, and can be independently operated. An "optimum" heating scenario, based on the multi-step (staircase) strategy developed before the tests, was determined and used for T/sub cs/ measurement. The test ended with the quench of the coil, followed by the dump. A normal zone was originated first in the P1.2 conductor, with the inlet helium temperature in P1.2 of about 8.7 to 8.9 K just before the quench, as expected from previous analysis. The results of the test are presented and analysis is performed for an accurate assessment of T/sub cs/, evaluating the effects of Joule heating in the joint, heat exchange through the joint between P1.2 and the slightly colder P1.1, and helium propagation from the heater to the peak field region in the conductor.

The TOSKA test facility of the Forschungszentrum Karlsruhe, which was especially built for testing large superconducting coils, has been upgraded for the test of the ITER Toroidal Field Model Coil (TFMC) in the framework of the ITER-EDA R&D programme. The upgrade includes the extension of the cryogenic supply system, a new 20 kA power supply for the background field generated by the existing LCT Coil as well as two 80 kA current leads and a 80 kA dump circuit for the TFMC. At this stage the power supply systems, including their dump circuits, can be tested only up to 10 kA with a 25 tons Cu coil. Therefore, suitable calculation tools have been developed for circuit analysis, control optimisation and simulation of possible scenarios at full current.

In 1997, the JET device was operated for an extensive campaign with deuterium–tritium (D–T) plasmas (the DTE1 campaign). A comprehensive network of machine protection systems was necessary so that this experimental campaign could be executed safely without damage to the machine or release of activated material. This network had been developed over many years of JET deuterium plasma operation and therefore the modifications for D–T operation was not a significant problem. The DTE1 campaign was executed successfully and safely and the machine protection systems proved reliable and robust and, in the limited cases where they were required to act, functioned correctly. The machine protection systems at JET are described and their categorisation and development over time are summarised. The management, commissioning and operational experience during DTE1 are discussed and some examples of fault scenarios are described. The experience with protection systems at JET highlights the importance of correct design and philosophy decisions being taken at an early stage. It is shown that this experience will be invaluable data input to the safe operation of future large fusion machines.

Recent experiments in D-T plasmas on the JET and TFTR tokamaks have evaluated a wide range of ITER relevant ion cyclotron heating scenarios. Absorption of fast waves at the second-harmonic tritium resonance has provided bulk ion heating in TFTR supershots and electron heating in JET H-mode discharges. In JET, deuterium minority heating has generated 1.7 MW of fusion power with 6 MW of radio frequency power giving a record steady-state Q-value of 0.22. Strong bulk ion heating has been achieved with He-3 minority heating with central ion temperatures up to 13 keV being produced in H-modes with a density of 3.6 x 10(19) m(-3) Hydrogen, deuterium and He-3 minority heating methods have produced plasmas with normalized confinement times greater than or equal to that required by ITER for ignition. These H-modes are characterized by small-amplitude, high-frequency ELMs, each of which transports less than 1.5% of the plasma energy content to the limiters. The heavy minority scheme of tritium in a deuterium plasma has been demonstrated both as a heating scheme and a generator of suprathermal neutrons. On TFTR mode conversion to an ion Bernstein wave has achieved central bulk ion heating in supershots with target ion temperatures greater than 20 keV.

The antinociceptive effect of 3 alpha-tropyl 2-(p-bromophenyl)propionate [(+/-)-PG-9] (10-40 mg kg-1 s.c.; 30-60 mg kg-1 p.o.; 10-30 mg kg-1 i.v.; 10-30 micrograms/mouse i.c.v.) was examined in mice, rats and guinea pigs by use of the hot-plate, abdominal-constriction, tail-flick and paw-pressure tests. (+/-)-PG-9 antinociception peaked 15 min after injection and then slowly diminished. The antinociception produced by (+/-)-PG-9 was prevented by the unselective muscarinic antagonist atropine, the M1-selective antagonists pirenzepine and dicyclomine and the acetylcholine depletor hemicholinium-3, but not by the opioid antagonist naloxone, the gamma-aminobutyric acidB antagonist 3-aminopropyl-diethoxy-methyl-phosphinic acid, the H3 agonist R-(alpha)-methylhistamine, the D2 antagonist quinpirole, the 5-hydroxytryptamine4 antagonist 2-methoxy-4-amino-5-chlorobenzoic acid 2-(diethylamino)ethyl ester hydrochloride, the 5-hydroxytryptamin1A antagonist 1-(2-methoxyphenyl)-4-[4-(2-phthalimido)butyl]piperazine hydrobromide and the polyamines depletor reserpine. Based on these data, it can be postulated that (+/-)-PG-9 exerted an antinociceptive effect mediated by a central potentiation of cholinergic transmission. (+/-)-PG-9 (10-40 mg kg-1 i.p.) was able to prevent amnesia induced by scopolamine (1 mg kg-1 i.p.) and dicyclomine (2 mg kg-1 i.p.) in the mouse passive-avoidance test. Affinity profiles of (+/-)-PG-9 for muscarinic receptor subtypes, determined by functional studies (rabbit vas deferens for M1, guinea pig atrium for M2, guinea pig ileum for M3 and immature guinea pig uterus for putative M4), have shown an M4/M1 selectivity ratio of 10.2 that might be responsible for the antinociception and the anti-amnesic effect induced by (+/-)-PG-9 through an increase in acetylcholine extracellular levels. In the antinociceptive and antiamnesic dose range, (+/-)-PG-9 did not impair mouse performance evaluated by the rota-rod test and Animex apparatus.

Two pumped diverters have been installed and tested in JET under ITER relevant conditions. A closed divertor is found to increase the particle and impurity exhaust rate in agreement with code modelling. Excellent power handling is demonstrated, allowing high current discharges with record stored energy (15 MJ) and quasi-steady-state discharges with high fusion triple product (4 x 10(20) m(-3) s keV). The ITERH93-P confinement scaling law is confirmed over abroad range and a more favourable beta scaling is found. No hysteresis is found in the II-mode power threshold. A data base of highly radiating discharges including impurity seeding describes the relationship between radiated power, impurity concentration and density. First results of high performance with optimised magnetic shear are presented. Strong internal confinement barriers develop despite the relatively small input of toroidal momentum and particle fuelling from ICRH and NBI.

The aim of coil protection system is to detect electrical faults and to protect the coils against mechanical or thermal over-stressing. The protection implemented are over-voltage and over-current for all the circuits, circuit equation simulation and comparison with the measured currents, ampere turn protection, and tensile, shear, and thermal stresses of poloidal and divertor coils. The protective actions include immediate removal of the voltage from the coils and circuit breaker trip. The tensile and shear stress of the poloidal coils and divertor coils is computed as a linear combination of the vertical force, radial force, and energy dissipated. The radial and vertical force of each coil is computed with flux loops and ampere turn measurements. During pulses, the TF coils expand in the radial direction and in the vertical directions. This motion can be approximated, on a slow time scale, by a linear combination of the in-plane magnetic force — due to the interaction of the current with the toroidal field — and the dissipated energy.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.