Two-Phase Flow Field Simulation of Horizontal Steam Generators

Ataollah Rabiee,Amir Hossein Kamalinia,KAMAL HADAD 한국원자력학회 2017 Nuclear Engineering and Technology Vol.49 No.1

The analysis of steam generators as an interface between primary and secondary circuitsin light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailedthermal hydraulics investigation in order to predict their behavior during normal andtransient operational conditions. Two phase flow field simulation on adjacent tube bundlesis important in obtaining logical numerical results. However, the complexity of the tubebundles, due to geometry and arrangement, makes it complicated. Employment of porousmedia is suggested to simplify numerical modeling. This study presents the use of porousmedia to simulate the tube bundles within a general-purpose computational fluid dynamicscode. Solved governing equations are generalized phase continuity, momentum,and energy equations. Boundary conditions, as one of the main challenges in thisnumerical analysis, are optimized. The model has been verified and tuned by simpletwo-dimensional geometry. It is shown that the obtained vapor volume fraction near thecold and hot collectors predict the experimental results more accurately than in previousstudies.

Two-way fluid-structure interaction simulation for steady-state vibration of a slender rod using URANS and LES turbulence models

Tooraj Nazari,Ataollah Rabiee,Hossein Kazeminejad 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.2

Anisotropic distribution of the turbulent kinetic energy and the near-field excitations are the main causesof the steady state Flow-Induced Vibration (FIV) which could lead to fretting wear damage in verticallyarranged supported slender rods. In this article, a combined Computational Fluid Dynamics (CFD) andComputational Structural Mechanic (CSM) approach named two-way Fluid-Structure Interaction (FSI) isused to investigate the modal characteristics of a typical rod's vibration. Performance of an UnsteadyReynolds-Average Navier-Stokes (URANS) and Large Eddy Simulation (LES) turbulence models onasymmetric fluctuations of the flow field are investigated. Using the LES turbulence model, any largedeformation damps into a weak oscillation which remains in the system. However, it is challenging to useLES in two-way FSI problems from fluid domain discretization point of view which is investigated in thisarticle as the innovation. It is concluded that the near-wall meshes whiten the viscous sub-layer is ofgreat importance to estimate the Root Mean Square (RMS) of FIV amplitude correctly as a significantfretting wear parameter otherwise it merely computes the frequency of FIV.

Assessment of the core-catcher in the VVER-1000 reactor containment under various severe accidents

Salari Farhad,Rabiee Ataollah,Faghihi Farshad 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.1

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO þ LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO þ SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

Safety margin and fuel cycle period enhancements of VVER-1000 nuclear reactor using water/silver nanofl uid

Hassan Saadati,KAMAL HADAD,Ataollah Rabiee 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.5

In this study, the effects of selecting water/silver nanofluid as both a coolant and a reactivity controllerduring the first operating cycle of a light water nuclear reactor are investigated. To achieve this, coupledneutronicethermo-hydraulic analysis is employed to simulate the reactor core. A detailed VVER1000/446reactor core is modeled in monte carlo code (MCNP), and the model is verified using the porous mediaapproach. Results show that the maximum required level of silver nanoparticles is 1.3 Vol.% at thebeginning of the cycle; this value drops to zero at the end of cycle. Due to substitution of water/boric acidwith water/Ag nanofluid, reactor operation time at maximum power extends to 357.3 days, and theenergy generation increases by about 27.3%. The higher negative coolant temperature coefficient ofreactivity in the presence of nanofluid in comparison with the water/boric acid indicates that the reactoris inherently safer. Considering the safety margins in the presence of the nanofluid, minimum departurefrom nucleate boiling ratio is calculated to be 2.16 (recommendation is 1.75).

Degradation analysis of horizontal steam generator tube bundles through crack growth due to two-phase flow induced vibration

Kamalinia Amir Hossein,Rabiee Ataollah 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.12

A correct understanding of vibration-based degradation is crucial from the standpoint of maintenance for Steam Generators (SG) as crucial mechanical equipment in nuclear power plants. This study has established a novel approach to developing a model for investigating tube bundle degradation according to crack growth caused by two-phase Flow-Induced Vibration (FIV). An important step in the approach is to calculate the two-phase flow field parameters between the SG tube bundles in various zones using the porous media model to determine the velocity and vapor volume fraction. Afterward, to determine the vibration properties of the tube bundles, the Fluid-Solid Interaction (FSI) analysis is performed in eighteen thermal-hydraulic zones. Tube bundle degradation based on crack growth using the sixteen most probable initial cracks and within each SG thermal-hydraulic zone is performed to calculate useful lifetime. Large Eddy Simulation (LES) model, Paris law, and Wiener process model are considered to model the turbulent crossflow around the tube bundles, simulation of elliptical crack growth due to the vibration characteristics, and estimation of SG tube bundles degradation, respectively. The analysis shows that the tube deforms most noticeably in the zone with the highest velocity. As a result, cracks propagate more quickly in the tube with a higher height. In all simulations based on different initial crack sizes, it was observed that zone 16 experiences the greatest deformation and, subsequently, the fastest degradation, with a velocity and vapor volume fraction of 0.5 m/s and 0.4, respectively.

Conceptual design of a high neutron flux research reactor core with low enriched uranium fuel and low plutonium production

Rahimi, Ghasem,Nematollahi, MohammadReza,Hadad, Kamal,Rabiee, Ataollah Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.3

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.