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Title: Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.
Authors:
; ;  [1]
  1. Los Alamos National Lab., Albuquerque, NM (United States) [and others
Publication Date:
OSTI Identifier:
107766
Report Number(s):
NUREG/CP--0142-Vol.2; CONF-950904--Vol.2
ON: TI95017078; TRN: 95:021388
Resource Type:
Technical Report
Resource Relation:
Conference: 7. international topical meeting on nuclear reactor thermal-hydraulics (Nureth-7), Saratoga Springs, NY (United States), 10-15 Sep 1995; Other Information: PBD: Sep 1995; Related Information: Is Part Of Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11; Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)]; PB: 795 p.
Research Org:
Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
Country of Publication:
United States
Language:
English
Subject:
22 NUCLEAR REACTOR TECHNOLOGY; 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; SAFETY ANALYSIS; T CODES; R CODES; PWR TYPE REACTORS; LOSS OF COOLANT; HYDRAULICS; REACTOR SAFETY; DIAGRAMS; THEORETICAL DATA