Effect of reactor coolant pumps following a small break in a pressurized water reactor

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Small-break loss-of-coolant accidents were calculated to help determine whether to trip the reactor-coolant pumps early in the accident when the reactor scrams or to delay the pump trip (pump trip times ranged from 450 s to no trip at all). Four-in.-diam (approximate) cold-leg breaks in Babcock and Wilcox (B and W) and Westinghouse (W) pressurized-water reactors were investigated using the Transient Reactor Analysis Code, TRAC-PD2. The results indicated that for a 4-in.-diam cold-leg break the optimum mode of pump operation is design dependent. In terms of primary system mass depletion, the case with no pump trip was preferable for the ... continued below

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Pages: 12

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Elliott, J.L.; Lime, J.F. & Willcutt, G.J.E. Jr. January 1, 1982.

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Description

Small-break loss-of-coolant accidents were calculated to help determine whether to trip the reactor-coolant pumps early in the accident when the reactor scrams or to delay the pump trip (pump trip times ranged from 450 s to no trip at all). Four-in.-diam (approximate) cold-leg breaks in Babcock and Wilcox (B and W) and Westinghouse (W) pressurized-water reactors were investigated using the Transient Reactor Analysis Code, TRAC-PD2. The results indicated that for a 4-in.-diam cold-leg break the optimum mode of pump operation is design dependent. In terms of primary system mass depletion, the case with no pump trip was preferable for the W plant, whereas an early pump trip was preferable for the B and W plant. When the pumps were not operating in the W plant, the loop seals plugged with liquid, leading to a pressure buildup in the upper plenum and, consequently, a high liquid flow through the break. The vent valves in the B and W plant mitigated the consequences of the loop seals plugging; the effect was enough to favor an early pump trip.

Physical Description

Pages: 12

Notes

NTIS, PC A02/MF A01.

Source

  • International meeting on thermal nuclear reactor safety, Chicago, IL, USA, 29 Aug 1982

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  • Other: DE82021765
  • Report No.: LA-UR-82-2521
  • Report No.: CONF-820802-11
  • Grant Number: W-7405-ENG-36
  • Office of Scientific & Technical Information Report Number: 5047635
  • Archival Resource Key: ark:/67531/metadc1056198

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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  • January 1, 1982

Added to The UNT Digital Library

  • Jan. 22, 2018, 7:23 a.m.

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  • Feb. 1, 2018, 7:04 p.m.

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Elliott, J.L.; Lime, J.F. & Willcutt, G.J.E. Jr. Effect of reactor coolant pumps following a small break in a pressurized water reactor, article, January 1, 1982; New Mexico. (digital.library.unt.edu/ark:/67531/metadc1056198/: accessed September 23, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.