CFD Analysis of PWR Reactor Vessel Upper Plenum Sections: Flow Simulation in Control Rods Guide Tubes

2010 ◽  
Author(s):  
Min-Tsung Kao ◽  
Chung-Yun Wu ◽  
Ching-Chang Chieng ◽  
Yiban Xu ◽  
Kun Yuan ◽  
...  
Keyword(s):  
Stokes Equations ◽  
Reactor Vessel ◽  
Unit Cell ◽  
Guide Tube ◽  
Control Rod ◽  
Geometry Model ◽  
Mesh Sensitivity ◽  
Representative Unit Cell ◽  
Sensitivity Studies ◽  
Domain Models

The AP1000™ PWR reactor vessel upper plenum contains numerous control rod guide tubes and support columns. Below the upper plenum are the upper core plate and the top core region of the fuel assemblies. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions are performed, conducting local simulations for smaller sections of the domain can provide crucial and detailed physical aspects of the flow. These sub-domain models can also be used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. The study discussed in this paper focuses on the sections of the domain related to the control rod guide tubes. The commercial CFD code STAR-CCM+ is employed to generate the computational mesh, to solve the Reynolds-averaged Navier-Stokes equations for incompressible flow with a Realizable k-epsilon turbulence model, and to post-process the results. Two sub-domains are modeled and analyzed: (1) a 1/4 section of one control rod guide tube by itself and (2) a representative unit cell containing two sections of adjacent control rod guide tubes and one 1/4 section of a neighboring support column. For the 1/4 guide tube model (sub-domain 1), trimmed meshes of up to 16 million cells are generated to compute the flow and pressure fields in both complete and simplified (without chamfers and narrow gaps) models. Comparisons of the results lead to the conclusion that the simplified geometry model might be used when developing larger domain models in the future. The representative unit cell (sub-domain 2) is assumed to be positioned in the center of the upper plenum where the global lateral flow effects are minimal. At this position, the lateral flows are generated mainly by the flow as it exits the guide tubes. After flow enters the unit cell from the bottom, there are three potential locations for flow to leave the unit cell: (1) lower locations near the support column and the upper core plate, (2) side windows in the lower portion of the guide tubes, and (3) upper locations near the guide plates positioned inside the guide tubes. Both trimmed and polyhedral meshes are generated as part of the mesh sensitivity studies. Predicted flow rates at each of the outflow locations in conjunction with results from the mesh sensitivity studies provide guidance on (1) what geometry to preserve or remove, (2) what geometry can be simplified to reduce the required mesh, and (3) an estimate of the total mesh required to model the control rod guide tube and support column geometry in the much larger simulation of the entire upper plenum and top fuel domain.

Effect of CRGT Cooling on Modes of Global Vessel Failure of a BWR Lower Head

2012 ◽  
Author(s):  
Walter Villanueva ◽  
Chi-Thanh Tran ◽  
Pavel Kudinov
Keyword(s):  
Heat Transfer ◽  
Vessel Wall ◽  
Reactor Vessel ◽  
Creep Model ◽  
Guide Tube ◽  
Mechanical Loads ◽  
Control Rod ◽  
Melt Pool ◽  
Creep Analysis ◽  
Lower Head

An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head. A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis. It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) ‘ballooning’ of the vessel bottom for smaller pools, and (b) ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.

Numerical Simulation of Flow Field Inside Reactor Upper Plenum for PWR With Code_Saturne

2017 ◽  
Author(s):  
Jiawei Liu ◽  
Puzhen Gao ◽  
Tingting Xu ◽  
Jiesheng Min ◽  
Guofei Chen
Keyword(s):  
Stokes Equations ◽  
Fluid Dynamic ◽  
Structure Design ◽  
Navier Stokes ◽  
Guide Tube ◽  
Flow Induced Vibration ◽  
Local Pressure ◽  
Pressurized Water ◽  
Local Flow ◽  
Control Rod

Flow characteristic in upper plenum has a strong influence on reactor functional margin and rod cluster control assembly (RCCA) guide tube wear. Upper plenum flow governs loops flow rate measurement via hot leg temperature which has also an influence on the reactor protection system. For RCCA guide tube wear, it appears in operation with RCCA flow-induced vibration, leading to its replacement. It is important to know the flow condition in the upper plenum, and in particular the outlet. Existing Generation III reactors have their own specialties on the design. Comparison between current technologies is a good way for better understanding on the key structure design for the upper plenum. In this paper, simplified models based on upper plenum structure of Korean advanced pressurized water reactor (PWR) and Westinghouse design AP1000 are constructed and meshed with a volume around 6 million cells to obtain a 3-dimensional global and local flow distributions inside the upper plenum and to characterize the vital flow features for reactor safety. The Navier-Stokes equations are solved with standard k-ε turbulence model by using EDF in-house open source computational fluid dynamic (CFD) software: Code_Saturne. Through calculations, pressure and velocity distributions are obtained, axial and lateral variations have been analyzed. Compared with APR1400, it can be observed that for the design of AP1000, the rotational flow entrained in the edge of upper plenum and high velocity area due to the hot leg suction effect contribute to the relatively lower local pressure, and may have an impact on the drop velocity of control rod.

scholarly journals Thermo-Viscoelastic Response of 3D Braided Composites Based on a Novel FsMsFE Method

Materials ◽  
2021 ◽  
Vol 14 (2) ◽  
pp. 271
Author(s):  
Jun-Jun Zhai ◽  
Xiang-Xia Kong ◽  
Lu-Chen Wang
Keyword(s):  
Finite Element ◽  
Cell Model ◽  
Structural Parameters ◽  
Thermal Relaxation ◽  
Viscoelastic Behavior ◽  
Unit Cell ◽  
Time Dependent ◽  
3D Braided Composites ◽  
Equivalent Time ◽  
Representative Unit Cell

A homogenization-based five-step multi-scale finite element (FsMsFE) simulation framework is developed to describe the time-temperature-dependent viscoelastic behavior of 3D braided four-directional composites. The current analysis was performed via three-scale finite element models, the fiber/matrix (microscopic) representative unit cell (RUC) model, the yarn/matrix (mesoscopic) representative unit cell model, and the macroscopic solid model with homogeneous property. Coupling the time-temperature equivalence principle, multi-phase finite element approach, Laplace transformation and Prony series fitting technology, the character of the stress relaxation behaviors at three scales subject to variation in temperature is investigated, and the equivalent time-dependent thermal expansion coefficients (TTEC), the equivalent time-dependent thermal relaxation modulus (TTRM) under micro-scale and meso-scale were predicted. Furthermore, the impacts of temperature, structural parameters and relaxation time on the time-dependent thermo-viscoelastic properties of 3D braided four-directional composites were studied.

Preliminary Design of Unloading Device for MNSR LEU Conversion

2018 ◽  
Author(s):  
Hao Qian ◽  
Li Yiguo ◽  
Peng Dan ◽  
Wu Xiaobo ◽  
Lu Jin ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Reactor Vessel ◽  
Cover Plate ◽  
Spent Fuel ◽  
Long Distance ◽  
Control Rod ◽  
Small Cover ◽  
Sealing Performance ◽  
Safety And Reliability ◽  
Application Requirements

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.

SEA model for structural acoustic coupling by means of periodic finite element models of the structural subsystems

2021 ◽  
Vol 263 (1) ◽  
pp. 5301-5309
Author(s):  
Luca Alimonti ◽  
Abderrazak Mejdi ◽  
Andrea Parrinello
Keyword(s):  
Finite Element ◽  
Numerical Approach ◽  
Unit Cell ◽  
Analytical Models ◽  
Finite Element Models ◽  
Fe Model ◽  
Acoustic Coupling ◽  
Representative Unit Cell ◽  
Definition Of ◽  
Acoustic Treatment

Statistical Energy Analysis (SEA) often relies on simplified analytical models to compute the parameters required to build the power balance equations of a coupled vibro-acoustic system. However, the vibro-acoustic of modern structural components, such as thick sandwich composites, ribbed panels, isogrids and metamaterials, is often too complex to be amenable to analytical developments without introducing further approximations. To overcome this limitation, a more general numerical approach is considered. It was shown in previous publications that, under the assumption that the structure is made of repetitions of a representative unit cell, a detailed Finite Element (FE) model of the unit cell can be used within a general and accurate numerical SEA framework. In this work, such framework is extended to account for structural-acoustic coupling. Resonant as well as non-resonant acoustic and structural paths are formulated. The effect of any acoustic treatment applied to coupling areas is considered by means of a Generalized Transfer Matrix (TM) approach. Moreover, the formulation employs a definition of pressure loads based on the wavenumber-frequency spectrum, hence allowing for general sources to be fully represented without simplifications. Validations cases are presented to show the effectiveness and generality of the approach.

Predicting Thermal Mixing and Fatigue Inside Control Rod Guide Tubes

2013 ◽  
Author(s):  
Eric Lillberg
Keyword(s):  
Thermal Fatigue ◽  
Turbulent Mixing ◽  
Low Frequency ◽  
Experimental Measurements ◽  
Guide Tube ◽  
Control Rod ◽  
Thermal Mixing ◽  
Purge Flow ◽  
Control Rods ◽  
Good Agreement

The cracked control rods shafts found in two Swedish NPPs were subjected to thermal fatigue due to mixing of cold purge flow with hot bypass water in the upper part of the top tube on which the control rod guide tubes rests. The interaction between the jets formed at the bypass water inlets is the main source of oscillation resulting in low frequency downward motion of hot bypass water into the cold purge flow. This ultimately causes thermal fatigue in the control rod shaft in the region below the four lower bypass water inlets. The transient analyses shown in this report were done to further investigate this oscillating phenomenon and compare to experimental measurements of water temperatures inside the control rod guide tube. The simulated results show good agreement with experimental data regarding all important variables for the estimation of thermal fatigue such as peak-to-peak temperature range, frequency of oscillation and duration of the temperature peaks. The results presented in this report show that CFD using LES methodology and the open source toolbox OpenFOAM is a viable tool for predicting complex turbulent mixing flows and thermal loads.

Numerical Analysis of the Corium Behavior Within the Fuel Support Piece by MPS

2016 ◽  
Author(s):  
Ronghua Chen ◽  
Lie Chen ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Eutectic Reaction ◽  
Flow Behavior ◽  
Three Dimensional ◽  
Severe Accident ◽  
Guide Tube ◽  
Melt Flow ◽  
Control Rod ◽  
Mps Method ◽  
Moving Particle ◽  
Accident Conditions

In the typical boiling water reactor (BWR), each control rod guide tube supports four fuel assemblies via an orificed fuel support piece in which a channel is designed to be a potential corium relocation path from the core region to the lower head under severe accident conditions. In this study, the improved Moving Particle Semi-implicit (MPS) method was adopted to analyze the melt flow and ablation behavior in this region during a severe accident of BWR. A three-dimensional particle configuration was constructed for analyzing the melt flow behavior within the fuel support piece. Considering the symmetry of the fuel support piece, only one fourth of the fuel support was simulated. The eutectic reaction between Zr (the material of the corium) and stainless steel (the material of the fuel support piece) was taken into consideration. The typical melt flow and freezing behaviors within the fuel support piece were successfully reproduced by MPS method. In all the simulation cases, the melt discharged from the hole of the fuel support piece instead of plugging the fuel support piece. The results indicate that MPS method has the capacity to analyze the melt flow and solidification behavior in the fuel support piece.

Flow Analyses Using RELAP/MOD3.3 Code for the Simulant Melt Experiments (LAVA-ERVC) Under the External Reactor Vessel Cooling

2004 ◽  
Author(s):  
Kyoung-Ho Kang ◽  
Rae-Joon Park ◽  
Sang-Baik Kim ◽  
Hee-Dong Kim
Keyword(s):  
Heat Transfer ◽  
Free Volume ◽  
Boiling Heat Transfer ◽  
Nucleate Boiling ◽  
Flow Path ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Nucleate Boiling Heat Transfer ◽  
Sensitivity Studies

Flow analyses using RELAP5/ MOD3.3 code have been performed to investigate the occurrence and the effects of steam binding for the LAVA-ERVC experiments. The main objectives of the LAVA-ERVC experiments are investigations of coolability through external reactor vessel cooling according to RPV insulation design. It could be found from the sensitivity studies for the flow path in the annulus of insulation that steam binding could occur in case of the limited steam venting capacity, which is definitely coincident with the LAVA-ERVC experimental results. In case of sufficient flow path for the steam venting, the vessel experienced effective cooling by nucleate boiling heat transfer. And existence of the upper free volume had little effect on occurrence of steam binding in the LAVA-ERVC experiments.

Experimental Analyses by SIMMER-III on Duct-Wall Failure and Fuel Discharge/Relocation Behavior

2013 ◽  
Author(s):  
Hidemasa Yamano ◽  
Yoshiharu Tobita
Keyword(s):  
Three Dimensional ◽  
Computer Code ◽  
Failure Behavior ◽  
Guide Tube ◽  
Control Rod ◽  
Duct Wall ◽  
Experimental Analyses ◽  
Multi Phase ◽  
Pool Formation ◽  
Good Agreement

This paper describes experimental analyses using SIMMER-III/IV, which are two/three-dimensional multi-component multi-phase Eulerian fluid-dynamics codes, for the purpose of the code validation. Two topics of key phenomena in core disruptive accidents were presented in this paper: duct-wall failure and fuel discharge/relocation behavior. To analyze the duct-wall failure behavior, the SCARABEE BE+3 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall event progression; which was characterized by coolant boiling, clad melting, fuel failure, molten pool formation, duct-wall failure, etc.; observed in the experiment. The CAMEL C6 experiment investigated the fuel discharge and relocation behavior through a simulated control rod guide tube, which is important in evaluating the neutronic reactivity. SIMMER-IV well simulated fuel-coolant interaction, sodium voiding, fuel relocation behavior observed in the experiment. These experimental analyses indicated the validity of the SIMMER-III/IV computer code for the duct wall failure and fuel discharge/relocation behavior.

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