Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

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Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations ... continued below

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Gougar, H. D. & Davis, C. B. April 1, 2006.

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Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

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  • Report No.: INL/EXT-06-11057
  • Grant Number: DE-AC07-99ID-13727
  • DOI: 10.2172/911272 | External Link
  • Office of Scientific & Technical Information Report Number: 911272
  • Archival Resource Key: ark:/67531/metadc890855

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Office of Scientific & Technical Information Technical Reports

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  • April 1, 2006

Added to The UNT Digital Library

  • Sept. 22, 2016, 2:13 a.m.

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  • Nov. 7, 2016, 4:45 p.m.

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Gougar, H. D. & Davis, C. B. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs, report, April 1, 2006; [Idaho Falls, Idaho]. (digital.library.unt.edu/ark:/67531/metadc890855/: accessed November 17, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.