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-; ISSN 0003-018X; TRN: 96:006307-0082
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 75; Conference: Winter meeting of the American Nuclear Society (ANS) and the European Nuclear Society (ENS), Washington, DC (United States), 10-14 Nov 1996; Other Information: PBD: 1996
- Country of Publication:
- United States
- Language:
- English
Similar Records
An analysis of the proposed MITR-III core to establish thermal-hydraulic limits at 10 MW. Final report
The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5
Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU
Technical Report
·
Sun Jun 01 00:00:00 EDT 1997
·
OSTI ID:426361
+2 more
The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5
Technical Report
·
Sun Jan 01 00:00:00 EST 2017
·
OSTI ID:426361
Validation of the MULCH-II code for thermal-hydraulic safety analysis of the MIT research reactor conversion to LEU
Conference
·
Tue Jul 15 00:00:00 EDT 2008
·
OSTI ID:426361