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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.
Meeting Spotlight
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.
Hyoung Kyu Cho, Yun Je Cho, Moon Oh Kim, Goon Cherl Park
Nuclear Technology | Volume 159 | Number 1 | July 2007 | Pages 39-58
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT07-A3855
Articles are hosted by Taylor and Francis Online.
In this study, a new concept in reactor cavity cooling systems (RCCSs) for high-temperature gas-cooled reactors (HTGRs) is proposed. The proposed RCCS consists of both water pools and active air-cooling systems, in order to overcome the disadvantages of the weak cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. The cooling capability of the RCCS during normal operation and under accident conditions was evaluated on the basis of a series of experiments that were performed in a scaled test facility. The reactor vessel of the test facility was a 1/10 linear scaled model of a 265-MW pebble bed modular reactor (PBMR), and the RCCS of the test facility was designed to limit the volumetric-averaged reactor vessel wall temperature below the maximum permissible wall temperature of the prototype reactor. The experiments were conducted by simulating the heat released from the reactor vessel wall to the RCCS. The power was reduced by 1/100 to preserve the heat flux, and the timescale was reduced by 1/10 to preserve the stored energy per volume. In the normal operation tests, detailed information on the temperature distribution and heat removal fraction of the upper pool and side pool was obtained. In the loss of all forced convection accident test, the passive afterheat removal capability of the RCCS was evaluated. These experimental results will be used to validate the reactor safety analysis codes and to evaluate the feasibility of the water pool-type RCCS.