Analysis of hypothetical channel flow blockage of a tubular fuel assembly in a 10 MWth nuclear research reactor
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Said M. A. Ibrahim
and
Hossam I. Yousif
Abstract
Heat removal from the core of nuclear reactors is a crucial safety requirement that needs much analysis and attention. Coolant channel flow blockage is one of the unintentional cases that may influence the required heat removal rate. As a result, unsustainable over heating of the fuel cladding could occur resulting eventually in affecting the fuel integrity. This paper deals with simulating the effect of total channel flow blockage in a single fuel assembly on its steady state thermal hydraulic parameters. Thermal hydraulic calculations are carried out for the much less studied IRT-4M ducted fuel assembly type in a 10 MWth WWR-S tank in pool research reactor. The simulated IRT-4M fuel assembly comprises six rounded rectangular fuel elements in which seven channels are used for cooling. A mathematical model is developed in which all steady state thermal hydraulic parameters and their distributions could be predicted for the blocked and unblocked flow channel cases. Calculations were performed for two separate hypothetical flow channel blockage cases in the innermost channel and the middle channel of the IRT-4M fuel assembly. The results showed that the maximum clad temperatures for the two-flow blockage cases are nearly 127.4 °C, and 132.2 °C, respectively. The temperature increases resulting from a single channel blockage are not high enough to cause fuel damage, although some local coolant boiling might occur. Model validation and results support the model outcomes and its applications in the safety of the research reactor during abnormal operation.
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Research ethics: Not applicable.
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Informed consent: Not applicable.
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Author contributions: S. Ibrahim: Methodology, Validation, Visualization, Formal analysis, Writing (review & editing), Conceptualization, Supervision, Project administration; M. Esawy: Data curation, Software, Methodology, Writing (original draft), Validation, Supervision; H. Yousif: Data curation, Software. All authors have accepted responsibility for the entire content of this manuscript and approved its submission.
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Use of Large Language Models, AI and Machine Learning Tools: None declared.
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Conflict of interest: The authors state no conflict of interest.
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Research funding: None declared.
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Data availability: Not applicable.
Nomenclature
| Symbols | Subscript | ||
|---|---|---|---|
| A | Surface area | A | Axial or average |
| Cp | Specific heat of water | Acc | Acceleration |
| D | Diameter | B | Bulk |
| DNB | Departure from nucleate boiling | Ch | Channel |
| DNBR | Departure from nucleate boiling ratio | Cl | Clad |
| F | Friction coefficient | Co | Coolant |
| G | Mass flux | Con | Convection |
| G | Gravitational acceleration | Crit | Critical |
| H | Active length of fuel tubes | F | Fuel |
| H | Heat transfer coefficient | FA | Fuel assembly |
| I | Enthalpy | Fric | Friction |
| K | Thermal conductivity | Grav | Gravitational |
| m˚ | Mass flow rate | H | Hydraulic or hot |
| N | Number of fuel | I | Inside |
| OFI | Onset of flow instability | In | Inlet |
| ONB | Onset of nucleate boiling | L | Liquid |
| P | Thermal power | M | Mixing |
| P | Pressure or perimeter | Max | Maximum nucleate boiling |
| PPF | Power peaking factor | NCB | Contribution |
| T | Thickness | O | Outside |
| T | Temperature | S | Surface |
| V | Volume | Sat | Saturation |
| V | Velocity | Sp | Single phase contribution |
| Z | Axial fuel direction | Sub | Sub-cooled boiling |
| Greek symbols | Tot | Total | |
| ∆ | Difference | Tp | Total contribution |
| µ | Dynamic viscosity (kg/m.s) | Tube | Tube |
| Nu | Nusselt number | V | Vapor |
| Pr | Prandtl number | ||
| Re | Reynold’s number | ||
| ⱱ | Kinematic viscosity (m2/s) | ||
| Φ | Heat flux | ||
| ρ | Density | ||
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Articles in the same Issue
- Frontmatter
- Study of reliability prediction method and uncertainty analysis of the pump and valve of a floating nuclear power plant
- Thermal-hydraulic and safety analysis of alternative ceramic fuels in next generation of VVER-1000 reactor
- Transient thermal analysis for optimal armor material in upgraded mockup for PST using MOOSE Framework
- Analysis of hypothetical channel flow blockage of a tubular fuel assembly in a 10 MWth nuclear research reactor
- Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs
- Improved intelligent model predictive controller for the nuclear power reactor system
- Monte Carlo simulation of photo-peak efficiency response function and calculation of total efficiency for a NaI (Tl) scintillator detector
- Calculation and application research on real-time source term diffusion in NPP auxiliary building under accident circumstances
- Calendar of events – issue 6 of KERN
