Development of a multichannel analysis code for the MITR-III safety analysis
Journal Article
·
· Transactions of the American Nuclear Society
OSTI ID:426361
- Massachusetts Inst. of Technology, Cambridge, MA (United States)
This paper describes the development of a MULti-CHannel analysis (MULCH-II) code to be used for the safety analysis of the Massachusetts Institute of Technology Research Reactor (MITR). The code models the primary and the secondary coolant systems with special emphasis on analysis of detailed thermal-hydraulic conditions in the core region. The hot channel is modeled in parallel with the average channels to predict conditions in the core during a flow excursion instability. Fuel and cladding temperatures are calculated under all conditions so that the margin to fuel failure is given in addition to the thermal-hydraulic conditions.
- OSTI ID:
- 426361
- Report Number(s):
- CONF-961103--
- Journal Information:
- Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 75; ISSN 0003-018X; ISSN TANSAO
- Country of Publication:
- United States
- Language:
- English
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