Interprocess communication within the DIII-D plasma control system

The DIII-D tokamak fusion research experiment`s real-time digital plasma control system (PCS) is a complex and ever evolving system. During a plasma experiment, it is tasked with some of the most crucial functions at DIII-D. Key responsibilities of the PCS involve sub-system control, data acquisition/storage, and user interface. To accomplish these functions, the PCS is broken down into individual components (both software and hardware), each capable of handling a specific duty set. Constant interaction between these components is necessary prior, during and after a standard plasma cycle. Complicating the matter even more is that some components, mostly those which deal with user interaction, may exist remotely, that is to say they are not part of the immediate hardware which makes up the bulk of the PCS. The four main objectives of this paper are to (1) present a brief outline of the PCS hardware/software and how they relate to each other; (2) present a brief overview of a standard DIII-D plasma cycle (a shot); (3) using three sets of PCS sub-systems, describe in more detail the communication processes; and (4) evaluate the benefits and drawbacks of said systems

Current status of DIII-D real-time digital plasma control

This paper describes the current status of real-time digital plasma control for the DIII-D tokamak. The digital plasma control system (PCS) has been in place at DIII-D since the early 1990s and continues to expand and improve in its capabilities to monitor and control plasma parameters for DIII-D fusion science experiments. The PCs monitors over 200 tokamak parameters from the DIII-D experiment using a real-time data acquisition system that acquires a new set of samples once every 60 {micro}s. This information is then used in a number of feedback control algorithms to compute and control a variety of parameters including those affecting plasma shape and position. A number of system related improvements has improved the usability and flexibility of the DIII-D PCS. These include more graphical user interfaces to assist in entering and viewing the large and ever growing number of parameters controlled by the PCS, increased interaction and accessibility from other DIII-D applications, and upgrades to the computer hardware and vended software. Future plans for the system include possible upgrades of the real-time computers, further links to other DIII-D diagnostic measurements such as real-time Thomson scattering analysis, and joint collaborations with other tokamak experiments including the NSTX at Princeton

Advances in the operation of the DIII-D neutral beam computer systems

The DIII-D neutral beam system routinely provides up to 20 MW of deuterium neutral beam heating in support of experiments on the DIII-D tokamak, and is a critical part of the DIII-D physics experimental program. The four computer systems previously used to control neutral beam operation and data acquisition were designed and implemented in the late 1970`s and used on DIII and DIII-D from 1981--1996. By comparison to modern standards, they had become expensive to maintain, slow and cumbersome, making it difficult to implement improvements. Most critical of all, they were not networked computers. During the 1997 experimental campaign, these systems were replaced with new Unix compliant hardware and, for the most part, commercially available software. This paper describes operational experience with the new neutral beam computer systems, and new advances made possible by using features not previously available. These include retention and access to historical data, an asynchronously fired ``rules`` base, and a relatively straightforward programming interface. Methods and principles for extending the availability of data beyond the scope of the operator consoles will be discussed

Differential expression of HSPA1 and HSPA2 proteins in human tissues; tissue microarray-based immunohistochemical study

In the present study we determined the expression pattern of HSPA1 and HSPA2 proteins in various normal human tissues by tissue-microarray based immunohistochemical analysis. Both proteins belong to the HSPA (HSP70) family of heat shock proteins. The HSPA2 is encoded by the gene originally defined as testis-specific, while HSPA1 is encoded by the stress-inducible genes (HSPA1A and HSPA1B). Our study revealed that both proteins are expressed only in some tissues from the 24 ones examined. HSPA2 was detected in adrenal gland, bronchus, cerebellum, cerebrum, colon, esophagus, kidney, skin, small intestine, stomach and testis, but not in adipose tissue, bladder, breast, cardiac muscle, diaphragm, liver, lung, lymph node, pancreas, prostate, skeletal muscle, spleen, thyroid. Expression of HSPA1 was detected in adrenal gland, bladder, breast, bronchus, cardiac muscle, esophagus, kidney, prostate, skin, but not in other tissues examined. Moreover, HSPA2 and HSPA1 proteins were found to be expressed in a cell-type-specific manner. The most pronounced cell-type expression pattern was found for HSPA2 protein. In the case of stratified squamous epithelia of the skin and esophagus, as well as in ciliated pseudostratified columnar epithelium lining respiratory tract, the HSPA2 positive cells were located in the basal layer. In the colon, small intestine and bronchus epithelia HSPA2 was detected in goblet cells. In adrenal gland cortex HSPA2 expression was limited to cells of zona reticularis. The presented results clearly show that certain human tissues constitutively express varying levels of HSPA1 and HSPA2 proteins in a highly differentiated way. Thus, our study can help designing experimental models suitable for cell- and tissue-type-specific functional differences between HSPA2 and HSPA1 proteins in human tissues

PTEN as a Prognostic and Predictive Marker in Postoperative Radiotherapy for Squamous Cell Cancer of the Head and Neck

BACKGROUND: Tumor suppressor PTEN is known to control a variety of processes related to cell survival, proliferation, and growth. PTEN expression is considered as a prognostic factor in some human neoplasms like breast, prostate, and thyroid cancer. METHODOLOGY/PRINCIPAL FINDINGS: In this study we analyzed the influence of PTEN expression on the outcome of a randomized clinical trial of conventional versus 7-days-a-week postoperative radiotherapy for squamous cell cancer of the head and neck. The patients with cancer of the oral cavity, oropharynx, and larynx were randomized to receive 63 Gy in fractions of 1.8 Gy given 5 days a week (CF) or 7 days a week (p-CAIR). Out of 279 patients enrolled in the study, 147 paraffin blocks were available for an immunohistochemical assessment of PTEN. To evaluate the prognostic value of PTEN expression and the effect of fractionation relative to PTEN, the data on the outcome of a randomized clinical trial were analyzed. Tumors with a high intensity of PTEN staining had significant gain in the loco-regional control (LRC) from p-CAIR (5-year LRC 92.7% vs. 70.8%, for p-CAIR vs. CF, p = 0.016, RR = 0.26). By contrast, tumors with low intensity of PTEN did not gain from p-CAIR (5-year LRC 56.2% vs. 47.2%, p = 0.49, RR = 0.94). The intensity of PTEN highly affected the LRC in a whole group of 147 patients (5-year LRC 80.9% vs. 52.3% for high vs. low PTEN, p = 0.0007, RR = 0.32). In multivariate Cox analysis, including neck node involvement, EGFR, nm23, Ki-67, p53, cyclin D1, tumor site and margins, PTEN remained an independent predictor of LRC (RR = 2.8 p = 0.004). CONCLUSIONS/SIGNIFICANCE: These results suggest that PTEN may serve as a potent prognostic and predictive marker in postoperative radiotherapy for high-risk squamous cell cancer of the head and neck

Progress Towards High Performance, Steady-state Spherical Torus

Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} 500 kA. In parallel, start-up using radio-frequency current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor