Determination of heat flux leading to the onset of flow instability in MTR reactors
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Salah El-Din El-Morshedy
Abstract
The prediction of heat flux leading to the Onset of Flow Instability (OFI) phenomena is an important consideration in the design of Material Testing Reactors (MTR) due to the possibility of flow excursion during postulated accident. From the thermal-hydraulic point of view, OFI is the critical phenomenon limiting MTR reactor power. In a previous work, an empirical correlation is developed to predict the subcooling at OFI in narrow vertical rectangular channels simulating a coolant channel of MTR. In the present work, an innovative model to determine the heat flux leading to OFI in MTR reactors is introduced based on the previous correlation. The developed model gives a very low deviation of only 1.65% from the experimental data of Whittle & Forgan that covers a wide range of MTR operating conditions. The heat flux leading to OFI is also predicted by both Whittle & Forgan and Fabrega correlations for comparison. The present model is then applied on the IAEA 10 MW MTR generic reactor to predict the Best-Estimate (BE) and Best-Estimate-Plus-Uncertainty (BEPU) Onset of Flow Instability Ratio (OFIR) and the power leading to OFI as well as the bubble detachment parameter under different coolant velocities and inlet temperatures. The model is also used to predict both the OFIR and bubble detachment parameter in the reactor under unprotected Loss-of-Flow transient for exponential flow decay with a time constant of 1.0 s (fast LOFA), 10, 15 and 25 s (slow LOFA) from a power level of 10 MW. For BEPU calculation, a combined statistical method with direct propagation of errors is adapted to treat the uncertainty factors for fuel fabrication and measured parameters in the BEPU calculation. The model results is analyzed and discussed.
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Author contributions: The author has accepted responsibility for the entire content of this submitted manuscript and approved submission.
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Research funding: None declared.
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Conflict of interest statement: The author declares no conflicts of interest regarding this article.
References
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Articles in the same Issue
- Frontmatter
- Effect of engraved concentric circles on pool boiling of water
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- Calendar of events
Articles in the same Issue
- Frontmatter
- Effect of engraved concentric circles on pool boiling of water
- Numerical determination of condensation pressure drop of various refrigerants in smooth and micro-fin tubes via ANN method
- Effect of metal layer height on heat transfer inside molten pool
- Transient analysis of MTR research reactor during fast and slow loss of flow accident
- Determination of heat flux leading to the onset of flow instability in MTR reactors
- Experimental study on direct contact condensation of saturated steam at low mass flux in subcooled quiescent water
- Evaluation and integral analysis of ADS and CMT failures during AP1000 SBLOCA with ASYST VER 3 simulation code
- Computational fluid dynamics simulation of material testing reactor spent fuel cooling in wet storage
- The role of advanced nuclear reactors in non-electrical and strategic applications, producing sustainable energy supplies and reducing the greenhouse gasses
- Optimization of PID controller for water level control of the nuclear steam generator using PSO and GA
- Assessment for nuclear security using Analytic Hierarchy Process (AHP) incorporated with Neural Networking Method in nuclear power plants (NPPs)
- Safety assessment and management of spent nuclear fuel for TRIGA mark II research reactor
- Calendar of events
