Thermal hydraulic limits analysis using statistical propagation of parametric uncertainties
- Nuclear Science and Engineering Dept., Massachusetts Inst. of Technology, 77 Massachusetts Ave, Cambridge, MA 02139 (United States)
- Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA 02139 (United States)
The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design, a reactor power upgraded from 6 MW to 7 MW is proposed in order to maintain the same reactor performance of the HEU core. Previous approach in analyzing the impact of engineering uncertainties on thermal hydraulic limits via the use of engineering hot channel factors (EHCFs) was unable to explicitly quantify the uncertainty and confidence level in reactor parameters. The objective of this study is to develop a methodology for MITR thermal hydraulic limits analysis by statistically combining engineering uncertainties with an aim to eliminate unnecessary conservatism inherent in traditional analyses. This method was employed to analyze the Limiting Safety System Settings (LSSS) for the MITR, which is the avoidance of the onset of nucleate boiling (ONB). Key parameters, such as coolant channel tolerances and heat transfer coefficients, were considered as normal distributions using Oracle Crystal Ball to calculate ONB. The LSSS power is determined with 99.7% confidence level. The LSSS power calculated using this new methodology is 9.1 MW, based on core outlet coolant temperature of 60 deg. C, and primary coolant flow rate of 1800 gpm, compared to 8.3 MW obtained from the analytical method using the EHCFs with same operating conditions. The same methodology was also used to calculate the safety limit (SL) for the MITR, conservatively determined using onset of flow instability (OFI) as the criterion, to verify that adequate safety margin exists between LSSS and SL. The calculated SL is 10.6 MW, which is 1.5 MW higher than LSSS. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22107797
- Resource Relation:
- Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 25 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
- Country of Publication:
- United States
- Language:
- English
Similar Records
Transition Core Planning and Safety Analyses in Support of LEU Fuel Conversion of the University of Missouri Research Reactor (MURR)
Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion
Related Subjects
COOLANTS
DATA COVARIANCES
FLOW RATE
FUEL ELEMENTS
HEAT TRANSFER
HIGHLY ENRICHED URANIUM
HOT CHANNEL FACTOR
ISOTOPE SEPARATION
NUCLEATE BOILING
PRIMARY COOLANT CIRCUITS
RESEARCH REACTORS
SAFETY ENGINEERING
SAFETY MARGINS
SLIGHTLY ENRICHED URANIUM
THERMAL HYDRAULICS