Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code, CUPID

2010 ◽  
Vol 133 (5) ◽  
Author(s):  
Hyoung Kyu Cho ◽  
Byong Jo Yun ◽  
Ik Kyu Park ◽  
Jae Jun Jeong
Keyword(s):  
Computational Analysis ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Reactor Vessel ◽  
Physical Models ◽  
Low Flow ◽  
Two Phase ◽  
Rate Condition ◽  
Flow Models ◽  
Loss Of Coolant Accident

For the analysis of transient two-phase flows in nuclear reactor components such as a reactor vessel, a steam generator, and a containment, KAERI has developed a three-dimensional thermal hydraulic code, CUPID. It adopts a three-dimensional, transient, two-phase and three-field model and includes various physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, the CUPID code and its two-phase flow models were assessed against the downcomer boiling experiment, which was performed to simulate the downcomer boiling phenomena. They may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. The computational analysis using the CUPID code showed that it can appropriately predict the multidimensional boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size model.

Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code CUPID

2010 ◽  
Author(s):  
H. K. Cho ◽  
B. J. Yun ◽  
I. K. Park ◽  
J. J. Jeong
Keyword(s):  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Physical Models ◽  
Injection System ◽  
Hydraulic Analysis ◽  
Two Phase ◽  
Rate Condition ◽  
The Core ◽  
Benchmark Calculation ◽  
Thermal Hydraulic Analysis

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.

scholarly journals Assessment of the MARS Code Using the Two-Phase Natural Circulation Experiments at a Core Catcher Test Facility

2017 ◽  
Vol 2017 ◽  
pp. 1-13
Author(s):  
Dong Hun Lee ◽  
Su Ryong Choi ◽  
Kwang Soon Ha ◽  
Han Young Yoon ◽  
Jae Jun Jeong
Keyword(s):  
Computational Analysis ◽  
Nuclear Reactor ◽  
Natural Circulation ◽  
Three Dimensional ◽  
Heating Surface ◽  
Severe Accident ◽  
Test Facility ◽  
Two Phase ◽  
Core Catcher ◽  
Parametric Effects

A core catcher has been developed to maintain the integrity of nuclear reactor containment from molten corium during a severe accident. It uses a two-phase natural circulation for cooling molten corium. Flow in a typical core catcher is unique because (i) it has an inclined cooling channel with downwards-facing heating surface, of which flow processes are not fully exploited, (ii) it is usually exposed to a low-pressure condition, where phase change causes dramatic changes in the flow, and (iii) the effects of a multidimensional flow are very large in the upper part of the core catcher. These features make computational analysis more difficult. In this study, the MARS code is assessed using the two-phase natural circulation experiments that had been conducted at the CE-PECS facility to verify the cooling performance of a core catcher. The code is a system-scale thermal-hydraulic (TH) code and has a multidimensional TH component. The facility was modeled by using both one- and three-dimensional components. Six experiments at the facility were selected to investigate the parametric effects of heat flux, pressure, and form loss. The results show that MARS can predict the two-phase flow at the facility reasonably well. However, some limitations are obviously revealed.

scholarly journals CFD Code Validation against Stratified Air-Water Flow Experimental Data

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
F. Terzuoli ◽  
M. C. Galassi ◽  
D. Mazzini ◽  
F. D'Auria
Keyword(s):  
Experimental Data ◽  
Numerical Simulations ◽  
Water Flow ◽  
Nuclear Reactor ◽  
Cold Water ◽  
Three Dimensional ◽  
Free Surface Flow ◽  
Surface Flow ◽  
Two Phase ◽  
Code Validation

Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code (NEPTUNE CFD). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.

Comparison of Three-Dimensional Fine and Coarse Nodalization of the Reactor Vessel for a Small-Break Loss-of-Coolant Accident

1991 ◽  
Vol 94 (1) ◽  
pp. 28-43
Author(s):  
Bahram Nassersharif ◽  
James S. Peery ◽  
Evelyn M. Mullen ◽  
Stephen R. Behling
Keyword(s):  
Three Dimensional ◽  
Reactor Vessel ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

VVER-440 Steam Generator’s Two-Phase Flow Analysis

2016 ◽  
Vol 821 ◽  
pp. 57-62
Author(s):  
Lukas Joch ◽  
Roman Krautschneider
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Flow Analysis ◽  
Three Dimensional ◽  
Reactor Power ◽  
Physical Models ◽  
Two Phase ◽  
Ansys Fluent

The subject of this report is creation of three-dimensional thermal hydraulic model of horizontal steam generator for Dukovany nuclear power plant. A procedure is presented for simulation and analysis of secondary side of PGV-440 steam generator for nominal and increased reactor power. A two-fluid approach is applied for modeling physical processes inside the steam generator. Physical models were implemented in ANSYS Fluent CFD environment using User Defined Functions (UDFs). Results from this thermal hydraulic numerical model can be used for various other subsequent nuclear power plant operations and safety analysis.

Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2

2002 ◽  
Author(s):  
Chang Hwan Park ◽  
Doo Yong Lee ◽  
Ik Jeong ◽  
Un Chul Lee ◽  
Kune Y. Suh ◽  
...  
Keyword(s):  
Flow Model ◽  
Correction Term ◽  
Sensitivity Study ◽  
Severe Accident ◽  
Primary System ◽  
Two Phase ◽  
Flow Rates ◽  
Flow Models ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.

Sensitivity Analysis of the Response of PWR Containment During a Loss of Coolant Accident Using RELAP5-3D and MELCOR

2014 ◽  
Author(s):  
Rodolfo Vaghetto ◽  
Andrew Franklin ◽  
Alessandro Vanni ◽  
Yassin A. Hassan
Keyword(s):  
Sensitivity Analysis ◽  
Discharge Coefficient ◽  
Physical Models ◽  
Geometrical Parameters ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Flow Models ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Selection Of

The prediction of specific parameters for the reactor containment, such as pressure and sump pool temperature, is of paramount importance when studying the thermal-hydraulic phenomena involved in the debris generation, transport, and accumulation during Loss of Coolant Accidents (LOCA). The response of the reactor containment during these events may significantly vary depending of several factors such as break size and location, and other plant-specific features. When modeling the reactor and containment response using systems codes, the predictions may also depend on the selection of physical models, correlations and their coefficients. A sensitivity analysis of the response of a typical Pressurized Water Reactor (PWR) 4-loop reactor system and associated containment during a large break LOCA was conducted using RELAP5-3D and MELCOR to investigate the influence of geometrical parameters (break location), physical models (chiked flow models), and related coefficients (discharge coefficient at the break), on the containment response. The simulation results showed how the containment response changed by varying the selected parameters and confirmed the importance of identifying and studying the factors triggering the containment engineered features (containment sprays) when simulation the containment response.

Analysis of a PHWR Outlet Feeder Vibration Due to Random Turbulence-Induced Excitation

2007 ◽  
Author(s):  
Y. Han ◽  
V. P. Janzen ◽  
B. A. W. Smith ◽  
T. Godet
Keyword(s):  
Finite Element ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Nuclear Industry ◽  
Finite Element Code ◽  
Two Phase ◽  
Accelerated Corrosion ◽  
Beam Elements ◽  
Vibration Tests ◽  
Flow Accelerated Corrosion

A nuclear-reactor feeder pipe has been analyzed to assess its vibration and stresses due to internal coolant random turbulence-induced excitation. The structural models were created from both isoparametric beam elements and continuum elements in the finite-element code H3DMAP, a general three-dimensional mechanics analysis package developed to solve a wide variety of problems encountered in the nuclear industry. The feeder was also modelled with beam elements in VIBIC, a finite-element code designed for the vibration assessment of beam-like structures. The excitation forces were based on recent experimental results obtained from piping-vibration tests that used two-phase air/water flow, performed at E´cole Polytechnique at Montreal. Feeder vibration levels, in terms of normalized response amplitude and velocity, and normalized vibratory stresses along the feeder are predicted for the as-designed feeder and for feeders thinned by flow-accelerated corrosion.

Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Mukesh Kumar ◽  
A. K. Nayak ◽  
Sumit V. Prasad ◽  
P. K. Verma ◽  
R. K. Singh ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Low Flow ◽  
Experimental Demonstration ◽  
Integral Test ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Channel Conditions ◽  
Channel Type

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

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