Deuterium Operation Research Articles

Assessment of radiation dose rate resulting from activated corrosion products in the PRIMA facility cooling loops

PRIMA (Padova Research Injector Megavolt Accelerated) is the name of the ITER neutral beam facility currently under construction at the RFX Consortium located in Padua – Italy. PRIMA consists of two injectors which will test at the same time the ion source (SPIDER) and the whole system (MITICA). Both injectors accelerate negative deuterium and hydrogen ions with a maximum energy of 1 MeV for MITICA and 100 keV for SPIDER, with a maximum beam current of 40 A. Accelerated ions are stopped on a CuCrZr alloy calorimeter. Operation in deuterium leads to production of an intense neutron flux through D–D and D–T reactions on the calorimeter panels which is likely to activate materials and components around the injectors, among which the corrosion products formed in the cooling loops; the radioactivity of the activated corrosion products (ACPs) may persist even after injector shutdown, representing a relevant radiation protection issue during repair and inspection activities. In the present paper the occupational radiation exposure (ORE) is evaluated. The total amount of radioactivity produced inside the cooling loops is calculated under different operative conditions and the radiation dose received by workers carrying out maintenance activities on the injector is also assessed for both experiments.

3D Monte Carlo vacuum modeling of the neutral beam injection system of ITER

ITER plans to implement several heating neutral beam injection (HNB) systems, which shoot the energetic neutral beams of hydrogen or deuterium atoms into the plasma. For efficient neutralization of the accelerated hydrogen or deuterium ions, a particular gas density profile along the beamline is required in the HNB system. A novel Monte Carlo simulation code has been developed which is able to simulate not only the beamline density profile, but also the 3D density distribution inside the vacuum vessel. This paper describes the main ideas behind and the essential features of the new code and shows how the resulting density distribution depends upon the installed cryopumping system. The numbers given in this paper refer to the reference case with deuterium operation.

Optimizing UNU/ICTP PFF Plasma Focus for Neon Soft X-ray Operation

The United Nations University/International Centre for Theoretical Physics Plasma Focus Facility (UNU/ICTP PFF), a 3.3-kJ plasma focus, was designed for operation in deuterium with a speed factor S such that the axial run-down time matches the current rise time at an end axial speed of nearly 10 cm/mus. For operation in neon, we first consider that a focus pinch temperature between 200 and 500 eV may be suitable for a good yield of neon soft X-rays, which corresponds to an end axial speed of 6-7 cm/mus. On this basis, for operation in neon, the standard UNU/ICTP PFF needs to have its anode length z0 reduced by some 30%-40% to maintain the time matching. Numerical experiments using the Lee model code are carried out to determine the optimum configuration of the electrodes for the UNU/ICTP PFF capacitor system. The results show that an even more drastic shortening of anode length z0 is required, from the original 16 to 7 cm, at the same time, increasing the anode radius ldquoardquo from 0.95 to 1.2 cm, to obtain an optimum yield of Y sxr = 9.5 J. This represents a two- to threefold increase in the Y sxr from that computed for the standard UNU/ICTP PFF.

Experimental study of neutron emission characteristics in a compact sub-kilojoule range miniature plasma focus device

In order to gain insight into the possible neutron production mechanism in sub-kilojoule miniature plasma foci, the anisotropy in neutron and hard x-ray emissions from a newly developed fast miniature plasma focus device (2.4 µF, 27 nH, T/4 ∼ 400 ns, 12–15 kV, 170–270 J, total mass ∼25 kg) is investigated. The system produces an average neutron yield of (1.2±0.2) × 106 neutrons/shot into 4π sr at ∼80 kA peak discharge current and 5.5 mbar deuterium operation, as measured by a 3He proportional counter. Relative measurements of radiation yields, to gauge the emission anisotropy, have been performed using a pair of fast scintillator–photomultiplier detectors placed along the axial (0°) and radial (90°) directions. The average forward to radial neutron yield anisotropy is found to be 1.3±0.2. The average peak neutron energy for the axial and radial directions is estimated to be (2.9±0.3) MeV and (2.6±0.1) MeV, respectively. The observed higher fluence along the anode axis and the coincidence between occurrence of maximum neutron yield and anisotropy at 5.5 mbar deuterium filling gas pressure, suggest that the neutron production mechanism in the sub-kilojoule range miniature plasma focus device FMPF-1 may be predominantly beam–target in nature.

Progress of the development of the IPP RF negative ion source for the ITER neutral beam system

IPP Garching has successfully developed a RF-driven negative ion source for the ITER neutral beam injection system. The RF source is now an interesting alternative to the reference design with filamented sources due to its in principle maintenance-free operation. Current densities of 330 A m−2 and 230 A m−2 have been achieved for hydrogen and deuterium, respectively, at a pressure of 0.3 Pa and an electron/ion ratio of 1 for a small extraction area (7.0 × 10−3 m2) and short pulses (<4 s). Reliable deuterium operation with more than 150 pulses in the required parameter range was obtained by an improved caesium operation utilizing a control of all source temperatures (grid as well as source body) and monitoring the amount of caesium in the source. The development now concentrates on extending the pulse length to up to 1 h and extending the size of the source. The long pulse test bed went into operation last year; pulses of up to some 100 s with more or less stable conditions in terms of extracted currents and Cs dynamics have been achieved with current densities in the range of 150–200 A m−2 in hydrogen operation. The pulse length, however, is still limited by non-sufficient cooling of some parts of the source and the RF circuit. The commissioning of the so-called half-size source test facility started recently with the first plasma pulses; this large RF source with the width and half the height of the ITER NNBI source is dedicated to demonstrate the homogeneity of a large RF plasma—extraction is foreseen in a later phase—and the tests of an ITER-relevant RF circuit.

European contributions to the beam source design and R&D of the ITER neutral beam injectors

The article reports on the progress made by the ITER European Home Team in strong interaction with the ITER Joint Central Team and the Japan Atomic Energy Research Institute regarding several key aspects of the beam source for the ITER injectors: (1) Integration of the SINGAP accelerator into the ITER injector design. This is a substantially simpler concept than the multiaperture, multigap (MAMuG) accelerator of the ITER NBI reference design that has the potential for significant cost savings and that avoids some of the weaknesses of the reference design such as the need for intermediate high voltage potentials from the high voltage power supply and pressurized gas insulation. (2) High energy negative ion acceleration using a SINGAP accelerator. (3) Long pulse (i.e. >1000 s) negative ion source operation in deuterium. (4) RF source development, which could reduce the scheduled maintenance of the ITER injectors (as it uses no filaments), and simplify the transmission line and the auxiliary power supplies for the ion source.

Confinement analysis in low-confinement mode of hydrogen isotope experiments on the Tokamak Fusion Test Reactor

The effect of isotope on confinement in high-recycling, L-mode plasmas is studied on the Tokamak Fusion Test Reactor (TFTR) [see D. M. Meade, J. Fusion Energy 7, 107 (1988)] by comparing hydrogen and deuterium plasmas with the same magnetic field and similar electron densities and heating power, with both Ohmic and deuterium-neutral-beam heating. Following a long operational period in deuterium, nominally hydrogen plasmas were created through hydrogen glow discharge and hydrogen gas puffing in Ohmic plasmas, which saturated the exposed limiter surface with hydrogen and raised the H/(H+D) ratio from 10±3% to 65±5%. Ohmic deuterium discharges obtained higher stored energy and lower loop voltage than hydrogen discharges with similar limiter conditions. Neutral-beam power scans were conducted in L-mode plasmas at minor radii of 50 and 80 cm, with plasma currents of 0.7 and 1.4 MA. To minimize transport differences from the beam deposition profile and beam heating, deuterium neutral beams were used to heat the plasmas of both isotopes. Total stored energy increased approximately 20% from nominally hydrogen plasmas to deuterium plasmas during auxiliary heating. Of this increase about half can be attributed to purely classical differences in the energy content of unthermalized beam ions. Kinetic measurements indicate a consistent but small increase in central electron temperature and total stored electron energy in deuterium relative to hydrogen plasmas, but no change in total ion stored energy. No significant differences in particle transport, momentum transport, and sawtooth behavior are observed. Overall, only a small improvement (∼10%) in global energy confinement time of the thermal plasma is seen between operation in hydrogen and deuterium.

Isotope effects in an electron beam excited negative ion source

Results which focus on comparing operation in hydrogen and deuterium are presented. Measurements of plasma parameters, including electron density, electron temperature, electron energy distribution function and extracted negative ion current, have been made in the source and extractor regions of an electron beam excited negative ion source. For the same operating conditions, negative ion current is found to be higher in deuterium than in hydrogen.

Operations with tritium neutral beams on the tokamak fusion test reactor

In November of 1993 the tokamak fusion test reactor began operating with a deuterium-tritium fuel mixture instead of the pure deuterium which it had used heretofore. The major portion of this tritium has been supplied as energetic neutral particles injected by the neutral beams. After an initial run in which some ion sources used a mixture of 2% T and 98% D to test tokamak systems, full tritium beam operations commenced, with some of the ion sources run on pure tritium and some on deuterium to optimize the fuel mixture in the core plasma. Hundreds of tritium source shots have now occurred, with reliability which is better than that typical of deuterium operation. The maximum power injected with deuterium and tritium beams was 39.6 MW. D-T fusion power levels of up to 10.7 MW have been produced. Energy confinement in D-T plasmas of the “supershot” variety appears to be better than in similar deuterium plasmas.

Performance of the neutron diagnostic system for Alcator C-Mod

Measurement of the global neutron production from the Alcator C-Mod tokamak has been done for all experimental operation in deuterium, which commenced in August 1993. Up to 3×1012 neutrons/s have been measured for diverse plasmas, including high current (1 MA), pellet injected, and rf heated discharges. Measurements during the initial run period were obtained using two sets of neutron detectors housed in polyethylene loaded moderator stations. Two more moderator stations containing additional neutron detectors have been obtained from LLNL and have been installed in the Alcator C-Mod experimental cell. The system has been calibrated using Cf252 placed at discrete points inside the Alcator C-Mod vacuum vessel. This paper presents a description of the diagnostic, followed by details of the calibration method and results. A discussion of the derivation of ion temperature from the global measurements is included with first results from the integrated system.

Four-channel ZnS scintillator measurements of escaping tritons in TFTR

A four-channel scintillation detector capable of measuring tritons, protons, and alphas escaping from a tokamak plasma was operated during the 1986 run period of the Tokamak Fusion Test Reactor (TFTR). Signals consistent with the expected 1-MeV triton behavior have been observed during deuterium operation. Backgrounds associated with neutrons/gammas and soft x rays have been evaluated in situ and are shown to be at the ≊10%–20% level for the present detector. Such a detector is capable of measuring escaping alphas in deuterium-tritium-fueled tokamaks such as TFTR.

Hydrogen isotope retention in the JET limiters

In order to evaluate the deuterium and hydrogen distributions in JET after long term operation in deuterium in 1985, as well as after a few weeks with deuterium at the end of the hydrogen-based operations in 1984, limiters have been analysed using nuclear reaction techniques and laser induced mass analysis (LIMA). The results show the presence of very high amounts of deuterium and 1H (up to more than 10 22 atoms/m 2 on the sides and edges of the limiters). If this amount were the result of the implantation of deuterium into the limiter graphite, it would require ion energies far higher than those expected in the boundary layer. The deuterium concentration shows a broad minimum near the centre of the limiter, where the heat load from the plasma is a maximum. The depth distributions, which are obtained by the 3He(d, p) 4He nuclear reaction, indicate that deuterium is present to a considerable depth on the sides of the limiters. Similar amounts have been seen at the inner wall of the vessel — either on small samples mounted on the wall, or on the carbon tiles which covered the wall in 1985 — and on the RF antennae. It appears that co-deposition of carbon and deuterium enables significant trapping of hydrogen isotopes on all exposed surfaces.

The influence of impurity and particle control on Tandem Mirror Experiment Upgrade plasma operation

A variety of techniques are used in Tandem Mirror Experiment-Upgrade (TMX-U) to control impurities and reflux: repeated plasma pulses, glow discharge cleaning (GDC), and gettering. We performed a series of experiments under three different plasma-wall conditions: no wall conditioning after a machine maintenance cycle, glow-discharge-cleaned wall, and a gettered wall. We used several plasma diagnostics to determine the effect of these procedures on TMX-U plasma parameters. Spectroscopic measurements show that GDC reduces impurities and increases the electron temperature, enabling full-duration beam-sustained plasma operation without a large number of repeated plasma pulses. Gettering further reduces the impurities and the neutral pressure; this improved condition persists for several shots after gettering is stopped. Measurements from residual gas analyzers and an end-loss ion spectrometer show that hydrogen is present in the plasma during the initial deuterium operation after pumpdown; the hydrogen level decreases after plasma operation with gettering, indicating reduced wall recycling.

Partially Catalyzed Deuterium And Tritium-Assisted Plasma Characteristics

Partially catalyzed deuterium (PCD) fuel cycles and their sensitivity to ash buildup, radiation losses, T /SUB i/ /T /SUB e/ and /sup 3/He consumed are investigated. The study is machine independent, using a simple zero-dimensional steady-state model. The PCD fuel cycles include semi-catalyzed-deuterium where only a fraction of the /sup 3/He fuses and tritium-catalyzed-deuterium where /sup 3/He extracted from the plasma is converted into tritium, which is reinjected. Also considered is tritium-assisted operation where a fraction of the fusion neutrons is used to produce tritium, which is added to the PCD plasma. The PCD and tritium-assisted operation is shown to be attractive for certain nonelectrical applications. They avoid /sup 3/He recirculation required for catalyzed deuterium (Cat-D) operation and enable simplified blanket designs. The ignition temperature, n /SUB epsilon/ /tau/, and power density of PCD plasmas are very energy balance sensitive, but under certain conditions these properties can be comparable or superior to those of Cat-D. Cyclotron radiation losses can significantly impair, whereas tritium assistance can strongly improve, PCD performance.