Numerical Study on the Thermal Field and Heat Transfer Characteristics of a Hexagonal-Close-Packed Pebble Bed

Fuel elements in a high-temperature gas-cooled reactor (HTGR) core may be stacked with a hexagonal close-packed (HCP) structure; therefore, analyzing the temperature distribution and heat transfer efficiency in the HCP pebble bed is of great significance to the design and safety of HTGR cores. In this study, the heat transfer characteristics of an HCP pebble bed are studied using CFD. The thermal fields and convective heat transfer coefficients under different coolant inlet velocities are obtained, and the velocity fields in the gap areas are also analyzed in different planes. It is found that the strongest heat transfer is shown near the right vertices of the top and bottom spheres, while the weakest heat transfer takes place in areas near the contact points where no fluid flows over; in addition, the correlation of the overall heat transfer coefficient with the Reynolds number is proposed as havg = 0.1545(k/L)Re0.8 (Pr = 0.712, 1.6 × 104 ≤ Re ≤ 4 × 104). It is also found that the heat transfer intensity of the HCP structure is weaker than that of the face-centered-cubic structure. These findings provide a reference for reactor designers and will contribute to the development of safer pebble-bed cores.

Post-Test Numerical Analysis of a Helium-Cooled Breeding Blanket First Wall under LOFA Conditions with the MELCOR Fusion Code

The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user’s assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learnt.

Study on the effect of various burnable poisons on pebble bed reactor with OTTO fuelling schemes using Serpent 2 Monte Carlo code

Pebble bed reactor with a once-through-then-out fuelling scheme has the advantage of simplifying the refueling system. However, the core upper-level power density is relatively higher than the bottom, producing an asymmetric core axial power distribution. Several burnable poison (BP) configurations are used to flatten the peak power density and improve power distribution while suppressing the excess core reactivity at the beginning of the burnup cycle. This study uses HTR-PM, China’s pebble bed reactor core, to simulate several burnable poison (BP) configurations. Serpent 2 coupled with Octave and a discrete element method simulation is used to model and simulate the pebble bed reactor core. It is found that erbium needs a large volumetric fraction in either QUADRISO or distributed BP to perform well. On the other hand, gadolinium and boron need a smaller volumetric fraction but perform worse in radial power distribution criteria in the fuel sphere. This study aims to verify the effect of BP added fuel pebbles on an OTTO refueling scheme HTR-PM core axial power distribution and excess reactivity.

Double Heterogeneity Neutronic Simulation of HTR-10 for First Criticality and Full Power Initial Core with TORT-TD code

In order to support the verification and validation of computational methods and codes for the safety assessment of pebble bed High-Temperature Gas-cooled Reactors (HTGRs), the calculation of first criticality and full power initial core of the high-temperature pebble bed reactor 10 MWt (HTR-10) has been defined as one of the problems specified for both code-to-code and code-to-experiment benchmarking with a focus on neutronics. HTR-10 Experimental facility serves as the source of information for the currently designed high-temperature gas-cooled nuclear reactor. It is also desired to verify the existing codes against the data obtained in the facility. In HTR-10, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in the reactor consist of TRISO particles, which are embedded in the graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneity of this type of reactor. In this paper, the first criticality and the power distribution in full power initial core calculations of HTR-10 are used to demonstrate treatment of this double heterogeneity using TORT-TD and Serpent for cross-section generation. HTR-10 has unique characteristics in terms of the randomness in geometry, as in all pebble bed reactors. In this technique, the core structure is modeled by TORT-TD, and Serpent is used to provide the cross-section in a double heterogeneity approach. Results obtained by TORT-TD calculations are compared with available data. It is observed that TORT-TD calculation yield sufficiently accurate results in terms of initial criticality and power distribution in full power initial core of the HTR-10 reactor.