Quiescent double barrier regime in the DIII-D tokamak

A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the E x B shearing rate.     

Reaching high poloidal beta at Greenwald density with internal transport barrier close to full noninductive current drive

In the ASDEX Upgrade tokamak, high poloidal beta up to beta(pol) = 3 at the Greenwald density with H-mode confinement has been reached. Because of the high beta, the plasma current is driven almost fully noninductively, consisting of 51% bootstrap and 43% neutral beam driven current. To reach these conditions the discharge is operated at low plasma current ( I(P) = 400 kA) and high neutral beam heating power ( P(NBI) = 10 MW). The discharge combines an edge (H mode) and internal transport barrier at high densities without confinement-limiting MHD activities. The extrapolation to higher plasma currents may offer a promising way for an advanced scenario based fusion reactor.     

Experimental evidence for gradient length-driven electron transport in tokamaks

Energy transport by the electrons in a tokamak is examined in steady-state and power modulation experiments using electron cyclotron heating. The results are consistent with the assumption that temperature profiles are limited by a critical gradient length, leading to "stiff" profiles. The modulation experiments show that the stiffness factor increases with temperature. They strongly suggest that turbulence driven by the electron temperature gradient may be a dominant mechanism of electron transport. Although possibly not universal, these results are valid under various plasma conditions.     

Sensitivity of the plasma electric potential to the transport barriers formation

Heavy Ion Beam Probe diagnostics was used to study plasma potential and density in the regimes with improvement in confinement on the T-10 tokamak. Time-spatial potential distribution was analyzed during the formation of transport barriers. The time evolution of the potential profile shows simultaneous formation of the internal negative potential well and external layer with strong Er, during the simultaneous formation of the internal and external transport barriers. The time history of the plasma potential in the inner region (0.6<ρ<1) shows its clear correlation with theD _α intensity. Presented at the Workshop on the Role of Electric Fields in Plasma Confinement and Exhaust, Budapest, 18–19 June, 2000. This work was supported by Russian Basic Research Foundation, Grant No. 99-02-18457.     

Exponentially growing tearing modes in Rijnhuizen Tokamak Project plasmas

The local measurement of the island width w, around the resonant surface, allowed a direct test of the extended Rutherford model [P. H. Rutherford, PPPL Report-2277 (1985)], describing the evolution of radiation-induced tearing modes prior to disruptions of tokamak plasmas. It is found that this model accounts very well for the observed exponential growth and supports radiation losses as being the main driving mechanism. The model implies that the effective perpendicular electron heat conductivity in the island is smaller than the global one. Comparison of the local measurements of w with the magnetic perturbed field B showed that w proportional to B1/2 was valid for widths up to 18% of the minor radius.     

Effect of heating on the suppression of tearing modes in tokamaks

The suppression of (neoclassical) tearing modes is of great importance for the success of future fusion reactors like ITER. Electron cyclotron waves can suppress islands, both by driving noninductive current in the island region and by heating the island, causing a perturbation to the Ohmic plasma current. This Letter reports on experiments on the TEXTOR tokamak, investigating the effect of heating, which is usually neglected. The unique set of tools available on TEXTOR, notably the dynamic ergodic divertor to create islands with a fully known driving term, and the electron cyclotron emission imaging diagnostic to provide detailed 2D electron temperature information, enables a detailed study of the suppression process and a comparison with theory.     

剂量引导精确放射治疗中三维评价算法设计及实现

针对临床剂量学验证中二维评价方法只能反映每张平片上剂量分布情况, 并根据剂量引导精确放疗系统中三维剂量验证需求设计算法实现对三维剂量分布的比较。 根据公式, 在Visual C++控制台下实现剂量偏差、 位置偏差及γ分析三维评价算法。 设计两个计划进行算法验证: 计划1为方野照射等效水, 用于验证算法的正确性; 计划2为适形野照射仿真头模, 用于验证算法的实用性。 结果表明, 计划1中, 对剂量偏差,容许值范围内通过率为100%; 对位置偏差和γ分析, 容许范围内通过率为100%,容许边界值处为100±7.7%。 计划2中, 3种算法通过率分别为88.35%、100%和95.07%。 这说明该算法可用于剂量引导精确放疗系统中实现三维剂量分布的评价。     

ITER��һ��ģ��Be/Cu����ʧЧ����

为了分析ITER第一壁模块试制过程中Be/Cu界面经热等静压扩散连接后出现的不可接受的大尺寸缺陷,对Be/Cu连接失效的模块进行了破坏性试验.首先采用线切割对模块进行解剖,并对连接界面缺陷区域及非缺陷区域通过金相观察、SEM观察及能谱分析进行对比,初步确定了界面连接的失效机理.     

中国环流器二号A装置(HL-2A)工程研制

中国环流器二号A装置(HL 2A)(设计指标:大半径1.65m、小半径0 4m、环向磁场2 8T、环向等离子体电流480kA)是我国已建成的第一个偏滤器托卡马克实验型磁约束聚变装置。HL 2A装置的首要研究目标是利用其独特的大体积极向偏滤器在高参数等离子体条件下开展与偏滤器位形运行有关的研究。本文总结了HL 2A装置工程设计、制造与安装、工程调试等研制的主要内容和关键技术。工程调试和初步物理实验的结果表明:HL 2A装置的主要工程参数和性能已具备开展物理实验的条件,并已成功地运行于偏滤器位形。2003年11月底,HL 2A装置获得等离子体电流168kA,等离子体存在时间920ms,等离子体线平均密度1.7×1019m-3,环向磁场1 4T,极限真空度为4.6×10-6Pa。     

Estimation of 14 MeV neutron rate from triton burn-up in future W7-X deuterium plasma campaigns

Fast ion confinement is of major importance for the ignition of a burning fusion plasma. In future deuterium plasma campaigns of the Wendelstein 7-X stellarator, W7-X, the amount of triton burn-up is one possible measure for fast ion confinement. A well-established technique to observe triton burn-up is the 14 MeV neutron rate. In this paper, it is estimated whether an existing scintillating fibre neutron detector is also suited to measure triton burn-up in W7-X with sufficient accuracy. An estimation is presented, which can be applied to any tokamak or stellarator design and is one-dimensional in the minor radius. The inputs are profiles of density, temperature, and differential volume element as well as the triton slowing-down time. The estimation calculates the thermal deuteron fusion rate and the associated deuteron-triton fusion rate; thus, the triton burn-up generated 14 MeV neutron rate. It neither takes triton diffusion nor explicit losses into account. This thermally generated fusion rate is compared to the neutral beam injection heating induced beam-plasma fusion rate.