Thermal-hydraulic analysis for the upgraded MIT nuclear research reactor

Author

Hu, Lin Wen ; Bernard, John A.

Author_Institution

MIT, Cambridge, MA, USA

Volume

2

fYear

1997

fDate

9-15 Nov 1997

Firstpage

908

Abstract

The thermal-hydraulic analysis performed for a planned upgrade of the MIT Research Reactor from 5 to 6-7 MW using the existing heat removal equipment is reported. This upgrade is one of several options under consideration for the relicensing of the MIT Research Reactor in 1999. Included in this paper is a description of the calculations of thermal-hydraulic limits which include limiting safety system settings (LSSS) and safety limits (SL). The LSSS and SL were respectively chosen to avoid onset of nucleate boiling and flow instability during steady-state operation. These analyses were performed using the MULti-CHannel-II or MULCH-II code which was developed and benchmarked at MIT for design and safety analysis of the MIT Research Reactor

Keywords

boiling; fission reactor cooling; fission reactor safety; fission research reactors; flow instability; MULCH-II code; flow instability; heat removal equipment; limiting safety system settings; nucleate boiling; steady-state operation; thermal-hydraulic analysis; thermal-hydraulic limits; upgraded MIT nuclear research reactor; Educational technology; Fuels; Inductors; Licenses; Performance analysis; Safety; Senior members; Steady-state; Transient analysis; Valves;

fLanguage

English

Publisher

ieee

Conference_Titel

Nuclear Science Symposium, 1997. IEEE

Conference_Location

Albuquerque, NM

ISSN

1082-3654

Print_ISBN

0-7803-4258-5

Type

conf

DOI

10.1109/NSSMIC.1997.670451

Filename

670451