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Human Factors, Instrumentation & Controls
Improving task performance, system reliability, system and personnel safety, efficiency, and effectiveness are the division's main objectives. Its major areas of interest include task design, procedures, training, instrument and control layout and placement, stress control, anthropometrics, psychological input, and motivation.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
Standards Program
The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Latest News
Commercial nuclear innovation "new space" age
In early 2006, a start-up company launched a small rocket from a tiny island in the Pacific. It exploded, showering the island with debris. A year later, a second launch attempt sent a rocket to space but failed to make orbit, burning up in the atmosphere. Another year brought a third attempt—and a third failure. The following month, in September 2008, the company used the last of its funds to launch a fourth rocket. It reached orbit, making history as the first privately funded liquid-fueled rocket to do so.
M. P. Sharma, A. K. Nayak
Nuclear Science and Engineering | Volume 188 | Number 2 | November 2017 | Pages 175-186
Technical Paper | doi.org/10.1080/00295639.2017.1339539
Articles are hosted by Taylor and Francis Online.
The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube–type, heavy water–moderated, and boiling light water–cooled natural-circulation–based reactor. The fuel bundle of AHWR contains 54 fuel rods arranged in three concentric rings of 12, 18, and 24 fuel rods. This fuel bundle is divided into a number of imaginary interacting flow passages called subchannels. Transition from a single-phase to a two-phase flow condition occurs in the reactor rod bundle with an increase in power. Predicting the thermal margin of the reactor has necessitated determining the diversion cross flow of coolant among these subchannels under two-phase flow. Thus, it is vital to evaluate cross flow between subchannels of the AHWR rod bundle. In this paper, experiments were carried out to investigate the diversion cross-flow phenomena for single- and two-phase flow in the simulated subchannels of the reactor. The size of the rod and the pitch in the test were the same as that of the actual rod bundle in the prototype. The cross-flow tests were carried out at atmospheric condition without adding heat. In addition, the capability of the existing correlation is also checked to predict the cross-flow resistance coefficient, and it is found that none of these models accurately predict the measured cross-flow resistance coefficient for the AHWR rod bundle. In view of this, a new model applicable to AHWR has been presented that predicts the cross-flow resistance coefficient quite accurately.