Optimization of a helicon plasma source for maximum density with minimal ion heating

Measurements of electron density and perpendicular ion temperatures in an argon helicon plasma are presented for five different antennas: A Nagoya type III antenna, a double-saddle antenna, a 19 cm long m = +1 helical antenna, a 30 cm long m = +1 helical antenna and a 19 cm m = +1 helical antenna with wide straps. Electromagnetic wave measurements in the range from 100 kHz to 50 MHz are also presented for a wide range of plasma parameters. The data show a clear transition between RF power coupling to the plasma to create density and coupling to heat ions. The transition from plasma production to ion heating indicates that the mechanism responsible for heating the ions is distinct from the mechanism responsible for ionizing the plasma in a helicon source. The primary objective of the experiments described here is to identify the operational conditions for a helicon source such that the intrinsic ion heating is minimized without sacrificing density production. Secondary objectives of this project include: identifying the optimal antenna configuration for density production and/or ion heating, investigating the mechanism responsible for ion heating through measurements of the fluctuating magnetic field at the edge of the source, and determining if downstream density measurements can be used as a quantitative measure of the electron density in the helicon source

COMMON COMPONENTS OF HUMAN ERROR IN DESIGN, MAINTENANCE, OR OPERATIONS

ABSTRACT Often, public reports of accidents only identify the last, obvious failure or immediate cause of the accident. If human error is the immediate cause or final failure, further assessment of accident contributors may stop, and an enhanced training program is often determined to be the primary solution for preventing further accidents of this type. However, in many cases, the accident is the final result of many inputs, decisions, actions and inactions. To demonstrate this characteristic of accidents, the 20 stories in a publication titled "Set Phasers on Stun" have been categorized into action errors and planning errors that involve designers, mechanics, or operators. For each story, the hazard and the number of simultaneous failures are listed. Then two of the 20 stories are assessed in detail; one story involves an action error and the other one involves a planning error. In each of these two stories, the system is first described as it should operate and then its risk is quantitatively assessed to identify findings, lessons learned, recommendations, analogies to the other 18 stories, and applications. This paper has three immediate goals. One, to recognize the difference between an action error and a planning error. Two, to recognize that most accidents involve 2 to 4 simultaneous failures. Three, to appreciate that quantifying the failure frequency serves two benefits. Because it is usually difficult to find out exactly what happened after an accident, the calculated frequency can help confirm what actually happened. When various alternatives are recommended, it can also help to select the most economic ones. This paper has two long term goals. One, consider assessing the failure rates of near misses. By reducing near misses, larger accidents will be reduced. Two, consider assessing the failure rates of personal near misses because you know what actually happened

ASME development of risk-based inspection guidelines for nuclear power plants

A methodology has been developed for use in preparing guidelines for the in-service inspection of nuclear power plant pressure boundary and structural components. This methodology is a further development of a general methodology previously published by the task force for application to any industry. It is unique in that it utilizes probabilistic risk assessment (PRA) information to improve the quantification of risks associated with component ruptures. A procedure has also been recommended for using the resulting quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Characteristics which an inspection strategy must possess in order to maintain target rupture probabilities are being determined by structural risk and reliability analysis (SRRA) calculations. The major features of this methodology are described in this paper, and example applications are briefly discussed. Plans for additional work in developing and applying the methodology are described

Risk based ASME Code requirements

The objective of this ASME Research Task Force is to develop and to apply a methodology for incorporating quantitative risk analysis techniques into the definition of in-service inspection (ISI) programs for a wide range of industrial applications. An additional objective, directed towards the field of nuclear power generation, is ultimately to develop a recommendation for comprehensive revisions to the ISI requirements of Section XI of the ASME Boiler and Pressure Vessel Code. This will require development of a firm technical basis for such requirements, which does not presently exist. Several years of additional research will be required before this can be accomplished. A general methodology suitable for application to any industry has been defined and published. It has recently been refined and further developed during application to the field of nuclear power generation. In the nuclear application probabilistic risk assessment (PRA) techniques and information have been incorporated. With additional analysis, PRA information is used to determine the consequence of a component rupture (increased reactor core damage probability). A procedure has also been recommended for using the resulting quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Structural risk and reliability analysis (SRRA) calculations are then used to determine characteristics which an inspection strategy must posess in order to maintain component rupture probabilities below target values. The methodology, results of example applications, and plans for future work are discussed

Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results