• Nuclear Engineering and Technology
• /
• v.46 no.5
• /
• pp.595-604
• /
• 2014

SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the $7^{th}$ Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the $6^{th}$ Framework Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the non-isolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.

• Nuclear Engineering and Technology
• /
• v.56 no.4
• /
• pp.1125-1134
• /
• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• Nuclear Engineering and Technology
• /
• v.48 no.1
• /
• pp.43-54
• /
• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Nuclear Engineering and Technology
• /
• v.17 no.1
• /
• pp.16-24
• /
• 1985

A computer model is developed capable of simulating the transient behavior of a pool-type liquid metal-cooled fast breeder reactor (LMFBR). The model, SIMFARP, is a fast running computer code which may be used to simulate the loss of power to any pump(s), a complete loss-of-forced cooling, and the natural circulation behavior. Eight governing equations are derived and a Runge-Kutta algorithm is applied to integrate the eight differential equations. The developed computer program is applied to two cases; loss of electric power to any pump(s), and loss of all external electric supply power without scram in Super-Phenix-I.

• Nuclear Engineering and Technology
• /
• v.40 no.3
• /
• pp.213-224
• /
• 2008

The pumping of coolant in a liquid metal fast reactor may be performed with an annular linear induction electro-magnetic (EM) pump. Linear induction pumps use a traveling magnetic field wave created by poly-phase currents, and the induced currents and their associated magnetic field generate a Lorentz force, whose effect can be the pumping of the liquid metal. The flow behaviors in the pump are very complex, including a time-varying Lorentz force and pressure pulsation, because an induction EM pump has time-varying magnetic fields and the induced convective currents that originate from the flow of the liquid metal. These phenomena lead to an instability problem in the pump arising from the changes of the generated Lorentz forces along the pump's geometry. Therefore, a magneto-hydro-dynamics (MHD) analysis is required for the design and operation of a linear induction EM pump. We have developed a time-harmonic 2-dimensional axisymmetry MHD analysis method based on the Maxwell equations. This paper describes the analysis and numerical method for obtaining solutions for some MHD parameters in an induction EM pump. Experimental test results obtained from an induction EM pump of CLIP-150 at the STC "Sintez," D.V. Efremov Institute of Electro-physical Apparatus in St. Petersburg were used to validate the method. In addition, we investigated some characteristics of a linear induction EM pump, such as the effect of the convective current and the double supply frequency (DSF) pressure pulsation. This simple model overestimated the convective eddy current generated from the sodium flow in the pump channel; however, it had a similar tendency for the measured data of the pump performance through a comparison with the experimental data. Considering its simplicity, it could be a base model for designing an EM pump and for evaluating the MHD flow in an EM pump.

• Nuclear Engineering and Technology
• /
• v.54 no.7
• /
• pp.2728-2735
• /
• 2022

The research on the flow and heat transfer characteristics of lead bismuth(LBE) is significant for the thermal-hydraulic calculation, safety analysis and practical application of lead-based fast reactors(LFR). In this paper, a new CFD model is proposed to solve the thermal-hydraulic analysis of LBE. The model includes two parts: turbulent model and turbulent Prandtl, which are the important factors for LBE. In order to find the best model, the experiment data and design of 19-pin hexagonal rod bundle with spacer grid, undertaken at the Karlsruhe Liquid Metal Laboratory (KALLA) are used for CFD calculation. Furthermore, the turbulent model includes SST k - 𝜔 and k - 𝜀; the turbulent Prandtl includes Cheng-Tak and constant (Pr_t =1.5,2.0,2.5,3.0). Among them, the combination between SST k - 𝜔 and Cheng-Tak is more suitable for the experiment. But in the low Pe region, the deviation between the experiment data and CFD result is too much. The reason may be the inlet-effect and when Pe is in a low level, the number of molecular thermal diffusion occupies an absolute advantage, and the buoyancy will enhance. In order to test and verify versatility of the model, the NCCL performed by the Nuclear Thermal-hydraulic Laboratory (Nuthel) of Xi'an Jiao tong University is used for CFD to calculate. This paper provides two verification examples for the new universal model.