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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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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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.
J-Ch. Sublet, D. E. Cullen, R. E. MacFarlane
Nuclear Technology | Volume 168 | Number 2 | November 2009 | Pages 293-297
Neutron Measurements | Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection | doi.org/10.13182/NT09-A9197
Articles are hosted by Taylor and Francis Online.
The results produced by a variety of currently available pointwise Monte Carlo neutron transport codes for the relatively simple problem of modeling a fast source of neutrons slowing down and thermalizing in water are compared. Initial comparisons showed rather large differences in the calculated flux: up to 80% differences. By working together to improve the results, iterations were done by (a) ensuring that all codes were using the same data, (b) improving the models used by the codes, and (c) correcting errors in the codes - no code is perfect. Even after a number of iterations, we still found differences, demonstrating that our Monte Carlo and supporting codes are far from perfect. In particular, we found that the often overlooked nuclear data-processing codes can be the weakest link in our systems of codes. The results presented here represent today's state of the art in the sense that all of the Monte Carlo codes are modern, widely available, and used codes. They all can use the most up-to-date nuclear data, and the results are recent; these are the results that current users of these codes should expect to obtain from them. As such, the accuracy and limitations of the codes presented here should serve as guidelines to code users in interpreting their results for similar problems. Results for the improved thermal scattering model now available, using advanced versions of NJOY-99.259, TRIPOLI-4.5, and MCNPX-2.6.f Beta, are presented. For comparisons among experimentally measured water cross sections and the unique JEFF-3.1 and ENDF/B-VII thermal scattering law, S(,) data are exemplified.