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Conference Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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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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ANS joins others in seeking to discuss SNF/HLW impasse
The American Nuclear Society joined seven other organizations to send a letter to Energy Secretary Christopher Wright on July 8, asking to meet with him to discuss “the restoration of a highly functioning program to meet DOE’s legal responsibility to manage and dispose of the nation’s commercial and legacy defense spent nuclear fuel (SNF) and high-level radioactive waste (HLW).”
B. Boer, D. Lathouwers, J. L. Kloosterman, T. H. J. J. Van Der Hagen, G. Strydom
Nuclear Technology | Volume 170 | Number 2 | May 2010 | Pages 306-321
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT10-A9485
Articles are hosted by Taylor and Francis Online.
The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5.The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature.For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Temperature Reactor-10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system.It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100° C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648° C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976° C) during normal operation.