Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

2006 ◽  
Author(s):  
Fabio Moretti ◽  
Daniele Melideo ◽  
Fulvio Terzuoli ◽  
Francesco D’Auria
Keyword(s):  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Three Dimensional ◽  
Sensitivity Analyses ◽  
Model Parameters ◽  
Pressurized Water ◽  
Relevant Role ◽  
Nuclear Reactor Safety ◽  
Mixing Problem

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by deboration or boron dilution events, followed by transport of a deborated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality of such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the “swirl” effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a deborated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions.

Study on the Influence of Safety Injection Rate on the Process of SBLOCA

2013 ◽  
Author(s):  
He Hui ◽  
Liang-ming Pan
Keyword(s):  
Heat Transfer ◽  
Analytical Model ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Injection Rate ◽  
Relief Valve ◽  
Flow And Heat Transfer ◽  
Long Time ◽  
Nuclear Reactor Safety ◽  
Relap5 Code

After the nuclear accident of TMI-2, more and more researchers devote to the researches of SBLOCA. It is almost impossible for the large break LOCA depending on the probability of the accident during the reactor life cycle. But it is possible for SBLOCA, such as opening a relief valve for a long time. After reactor shutdown, reactor takes the form of directly safety injection, which has great effects on the flow and heat transfer of the reactor pressure vessel and it has influence on the process of the SBLOCA finally. In present work, an SBLOCA analytical model has been developed with RELAP5 code to analyze special design features at SBLOCA accident. The results are compared with the calculations from Westinghouse NOTRUMP code. This analytical model has also used to analyze the effects of different safety injection rate on the process of accident. The transient variation about some key parameters have been obtained. e.g. the temperature, pressure variations of core, void fraction of core. The results show that the rate of safety injection has significant influence on the process of SBLOCA and the characteristics about the heat transfer of pressure vessel. Different safety injection rates influence the nuclear reactor safety differently.

FLUENT Analysis of a ROSA Two-Phase Stratification Test

2009 ◽  
Author(s):  
Tatiana Farkas ◽  
Iva´n To´th
Keyword(s):  
Best Practice ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Water Injection ◽  
Water Layer ◽  
Two Phase ◽  
Injection Point ◽  
Best Practice Guidelines ◽  
Vof Model ◽  
Nuclear Reactor Safety

One of the OECD ROSA project tests, investigating temperature stratification in the cold legs and the downcomer during ECCS water injection under two-phase natural circulation conditions was analysed with the FLUENT code. The guidance given in the “Best Practice Guidelines for the Use of CFD in Nuclear Reactor Safety Applications” of the OECD GAMA group was followed. The standard k-ε turbulence model of FLUENT was applied along with the VOF (Volume of Fluid) description of the steam/water phases. In order to model the interface of the two phases more closely, transient calculations were performed. Based on a comparison of calculated and measured results, it is found that the standard k-ε turbulence along with the VOF model gave acceptable description of the temperature distribution within the water layer. However, the model under-estimates mixing at the injection point, while it is over-estimated further downstream in the stratified water flow. The condensation model applied in FLUENT strongly under-estimated the subcooling of the steam phase. Insufficient condensation might explain why the downcomer level drops below the cold leg, which was not the case in the test.

scholarly journals Finite Element Based Full-Life Cyclic Stress Analysis of 316 Grade Nuclear Reactor Stainless Steel Under Constant, Variable, and Random Fatigue Loading

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Bipul Barua ◽  
Subhasish Mohanty ◽  
Joseph T. Listwan ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Analysis ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Fatigue Loading ◽  
Cyclic Stress ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fatigue Evaluation

Although S∼N curve-based approaches are widely followed for fatigue evaluation of nuclear reactor components and other safety critical structural systems, there is a chance of large uncertainty in estimated fatigue lives. This uncertainty may be reduced by using a more mechanistic approach such as physics based three-dimensional (3D) finite element (FE) methods. In a recent paper (Barua et al., 2018, ASME J. Pressure Vessel Technol., 140(1), p. 011403), a fully mechanistic fatigue modeling approach which is based on time-dependent stress–strain evolution of material over the entire fatigue life was presented. Based on this approach, in this work, FE-based cyclic stress analysis was performed on 316 nuclear grade reactor stainless steel (SS) fatigue specimens, subjected to constant, variable, and random amplitude loading, for their entire fatigue lives. The simulated results are found to be in good agreement with experimental observation. An elastic-plastic analysis of a pressurized water reactor (PWR) surge line (SL) pipe under idealistic fatigue loading condition was performed and compared with experimental results.

Buoyancy Driven Coolant Mixing Studies of Natural Circulation Flows at the ROCOM Test Facility Using ANSYS CFX

2006 ◽  
Author(s):  
Thomas Ho¨hne ◽  
So¨ren Kliem ◽  
Ulrich Rohde ◽  
Frank-Peter Weiß
Keyword(s):  
Pressure Vessel ◽  
Natural Circulation ◽  
Reactor Pressure Vessel ◽  
Density Difference ◽  
Test Facility ◽  
Mean Flow ◽  
Pressurized Water ◽  
The Core ◽  
Ansys Cfx ◽  
Reactor Pressure

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Thermal hydraulics analyses showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show a stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

Two and Three-Dimensional Simulations of Enhanced Heat Transfer in Nuclear Fuel Rod Bundles

2009 ◽  
Author(s):  
Leo A. Carrilho ◽  
Jamil Khan ◽  
Michael E. Conner ◽  
Abdel Mandour ◽  
Milorad B. Dzodzo
Keyword(s):  
Heat Transfer ◽  
Finite Element ◽  
Turbulent Flow ◽  
Stokes Equations ◽  
Three Dimensional ◽  
Element Model ◽  
Sensitivity Analyses ◽  
Artificial Roughness ◽  
Pressurized Water ◽  
Fuel Rod

The effects of artificial roughness for the purpose of thermal performance improvement in pressurized water nuclear reactors are investigated. The artificial roughness consists of two-dimensional ribs parallel to the turbulent flow. The fuel rod bundle subchannel is preliminarily modeled as an annulus using the finite element method in ANSYS/FLOTRAN. The Navier-Stokes equations are solved from the SST (Shear Stress Transport) turbulence model for the simulated annulus thermal-flow. The analyses are performed for ribs dimensions and pitch provided by published previous work. It is found that, heat transfer and differential pressure have similar behavior with highest heat transfer occurring at the reattachment point. The finite element model describes well the characteristics of turbulent flow in smooth and rough rod when compared to previous semi-empirical models. Next paper extends the analysis by comparing numerical results with experimental test data and sensitivity analyses for different roughness configurations.

Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach

2015 ◽  
Author(s):  
Subhasish Mohanty ◽  
William K. Soppet ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Nuclear Reactor ◽  
Axial Stress ◽  
Cyclic Plasticity ◽  
Model Parameters ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Test Conditions ◽  
Over Time

At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S∼N) curves and Coffin-Manson type empirical relations. In most cases, the S∼N curves are generated from uniaxial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S∼N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.

The Prospect of the Modeling Simulation for the Passive Containment Cooling System of the Advanced PWR in China

2011 ◽  
Vol 88-89 ◽  
pp. 82-87 ◽  
Author(s):  
Yan Wang
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Code Design ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Water Reactors ◽  
Nuclear Reactor Safety ◽  
Main Steam Line ◽  
Main Steam ◽  
Main Component

The passive containment cooling system (PCCS) is one main component of the various passive safety systems in the advanced 3rd generation pressurized water reactor. Several containment analysis codes are modified for the simulation and analysis on the PCCS by researchers. These codes for the PCCS were validated by comparison with transient test data from some separate effect tests and integral tests, and used to evaluate the heat-removed capability of the PCCS under the postulated events, such as a loss-of-coolant accident (LOCA) and a main steam line break (MSLB). The advanced 3rd generation pressurized water reactors of AP1000 project were build in China, and the future advanced PWR with higher power will be researched and designed. So it is very important to develop a new code with self-owned intellectual property rights, which is used for the simulation and analysis of the PCCS performance. In this paper, the main ideas of the code design, including the framework of the code, the important mechanism models and the code validation, etc., will be introduced. In addition, the computational fluid Dynamics (CFD) codes also had been applied to solve some thermal-hydraulic problems in Nuclear Reactor Safety (NRS) analysis.

Comparison of Pressure-Temperature Limits for a Pressurized Water Reactor Pressure Vessel Considering Beltline and Extended Beltline Regions

2018 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Yu-Yu Shen ◽  
Chin-Cheng Huang
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Pressure Vessels ◽  
Reference Temperature ◽  
Regulatory Body ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

To ensure the structural integrity of the embrittled reactor pressure vessels (RPVs) during startup or shutdown operation, the pressure-temperature (P-T) limits are mainly determined by the fracture toughness of beltline region material with the highest level of neutron embrittlement. However, other vessel parts such as nozzles with structural discontinuities may affect the limits due to the higher stress concentration, even though the neutron embrittlement is insignificant. Therefore, not only beltline material with the highest reference temperature, but also other components with structural discontinuities have to be considered for the development of P-T limits of RPV. In the paper, the pressure-temperature operational limits of a Taiwan domestic pressurized water reactor (PWR) pressure vessel considering beltline and extended beltline regions are established per the procedure of ASME Code Section XI-Appendix G. The three-dimensional finite element models of PWR inlet and outlet nozzles above the beltline region are also built to analyze the pressure and thermal stress distributions for P-T limits calculation. The analysis results indicate that the cool-down P-T limit of the domestic PWR vessel is still dominated by the beltline region, but the heat-up limit is partially controlled by the extended beltline region. On the other hand, the relations of reference temperature between nozzles and beltline region on the P-T limits are also discussed. Present work could be a reference for the regulatory body and is also helpful for safe operation of PWRs in Taiwan.

Study on Structural Integrity Assessment of Reactor Pressure Vessel Based on Three-Dimensional Thermal-Hydraulics and Structural Analyses

2014 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Tadashi Watanabe ◽  
Yutaka Nishiyama
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Coolant Water ◽  
Reactor Pressure

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.

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