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

Analysis of the Containment Thermal-Hydraulic Behavior of a Small Reactor During Loss of Coolant Accident

2017 ◽  
Author(s):  
Qian Lin ◽  
Weizhong Zhang
Keyword(s):  
Cooling System ◽  
Three Dimensional ◽  
Dimensional Model ◽  
Three Dimensional Model ◽  
One Dimensional ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Loss Of Coolant ◽  
Calculation Results ◽  
The One

The containment thermal hydraulics of a small reactor during loss of coolant accident (LOCA) is studied by a lumped parameter one-dimensional model and a three-dimensional model. The capability of a kind of heat exchanger type passive containment cooling system (PCCS) is analyzed by the one-dimensional model. The calculation results show that, the decay heat can be removed and the containment pressure can be decreased by the proposed PCCS. The steam and non-condensable gas (the air) distribution in the containment is investigated, the mixing and stratification behaviors are analyzed for several different cases, in which the PCCS and condenser are located at higher, base or lower position. The sensitivity analysis of the PCCS elevation shows that, in despite of the different gas stratification, the containment pressures are nearly the same. Similar conclusions can be obtained by the one-dimensional model and three-dimensional model. The preliminary results may indicate that, the designed PCCS and condenser can be located at a lower part, which will be benefit for the economy of the small reactor or meet other requirements.

scholarly journals CFD Application to Hydrogen Risk Analysis and PAR Qualification

2009 ◽  
Vol 2009 ◽  
pp. 1-10 ◽  
Author(s):  
Jinbiao Xiong ◽  
Yanhua Yang ◽  
Xu Cheng
Keyword(s):  
Fluid Dynamics ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Risk Analysis ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Minor Effect ◽  
Hydrogen Distribution ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

A three dimensional computation fluid dynamics (CFD) code, GASFLOW, is applied to analyze the hydrogen risk for Qinshan-II nuclear power plant (NPP). In this paper, the effect of spray modes on hydrogen risk in the containment during a large break loss of coolant accident (LBLOCA) is analyzed by selecting three different spray strategies, that is, without spray, with direct spray and with both direct and recirculation spray. A strong effect of spray modes on hydrogen distribution is observed. However, the efficiency of the passive auto-catalytic recombiners (PAR) is not substantially affected by spray modes. The hydrogen risk is significantly increased by the direct spray, while the recirculation spray has minor effect on it. In order to simulate more precisely the processes involved in the PAR operation, a new PAR model is developed using CFD approach. The validation shows that the results obtained by the model agree well with the experimental results.

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.

Lower Plenum Voiding

1982 ◽  
Vol 104 (3) ◽  
pp. 479-486 ◽  
Author(s):  
D. Bharathan ◽  
G. B. Wallis ◽  
H. J. Richter
Keyword(s):  
Reactor Core ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Coolant Circuit ◽  
Two Phase ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

One of the phenomena involved in a loss-of-coolant accident in a pressurized water reactor may be lower plenum voiding. This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. Steam generated in the reactor core may flow out of the bottom of the reactor core, turn in the lower plenum of the vessel, in a direction countercurrent to the emergency core coolant flow, and escape via the break. If its velocity is high enough, this steam may sweep water from the bottom (lower plenum) of the reactor vessel. Emergency coolant added to the vessel may also be carried out by the escaping steam and thus the reflooding of the core would be delayed. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with data obtained by other researchers.

scholarly journals Application of Fractional Scaling Analysis for Development and Design of Integral Effects Test Facility

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Milorad B. Dzodzo ◽  
Francesco Oriolo ◽  
Walter Ambrosini ◽  
Marco Ricotti ◽  
Davor Grgic ◽  
...  
Keyword(s):  
Input Data ◽  
Time Sequence ◽  
Reactor Vessel ◽  
Test Facility ◽  
Scaling Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Consecutive Time ◽  
Time Decomposition ◽  
Time Sequences

Abstract The aim of this paper is to present a fractional scaling analysis (FSA) application for a system with interacting components where multiple figures of merits need to be respected during complex transient accident scenario with several consecutive time sequences. This paper presents FSA application to the International Reactor Innovative and Secure (IRIS) reactor and Simulatore Pressurizzato per Esperienze di Sicurezza 3 (SPES3) integral effects test (IET) facility. The FSA was applied for the small break loss of coolant accident (SBLOCA) on the direct vessel injection (DVI) line as the most challenging transient scenario. The FSA methodologies were applied for two figures of merits: (1) reactor and containment vessels pressure responses, and (2) reactor vessel water collapsed level response. The space decomposition was performed first. The reactor vessel and containment vessel were divided in components so that important phenomena and their consequences can be evaluated in each of them. After that, the time decomposition in consecutive time sequences was performed for the considered transient (DVI SBLOCA) based on the starts, or ends, of the defining events. The configuration of the system in each time sequence might be different and dependent on the control system actions connecting, or disconnecting, various components of the system due to the valves openings, or closings. This way, the important phenomena and their consequences can be evaluated for each component and time sequence. Also, this paper presents and discusses options for deriving nondimensional groups and calculation of distortions between prototype and model responses for complex transients containing multiple consecutive time sequences. The input data for scaling analysis are based on the results of RELAP/GOTHIC analysis performed for IRIS and RELAP analysis performed for SPES3. The scaling analysis was applied iteratively several times for different IRIS and SPES3 configurations. Based on the intermediate results, some components in the IRIS and SPES3 were redesigned so that the distortions between IRIS and SPES3 responses are decreased.

scholarly journals A Study of Prediction Model Improvement for Air-Oxidation Breakaway in a Postulated Spent Nuclear Fuel Pool Complete Loss of Coolant Accident

2021 ◽  
Vol 13 (3) ◽  
pp. 1442
Author(s):  
Sanggil Park ◽  
Jaeyoung Lee ◽  
Min Bum Park
Keyword(s):  
Kinetic Model ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Complete Loss ◽  
Air Oxidation ◽  
Loss Of Coolant Accident ◽  
Spent Nuclear Fuel Pool ◽  
Loss Of Coolant ◽  
Model Derivation ◽  
Model Improvement

The temperature of zirconium alloy cladding on the postulated spent nuclear fuel pool complete loss of coolant accident is abruptly increased at a certain time and the cladding is almost fully oxidized to weak ZrO2 in the air. This abrupt temperature escalation phenomenon induced by the air-oxidation breakaway is called a zirconium fire. Although an air-oxidation breakaway kinetic model correlated between time and temperature has been implemented in the MELCOR code, it is likely to bring about unexpected large errors because of many limitations of model derivation. This study suggests an improved time–temperature correlated kinetic model using the Johnson–Mehl equation. It is based on that the air-oxidation breakaway is initiated by the phase transformation from the tetragonal to monoclinic ZrO2 at the oxide–metal interface in the cladding. This new model equation is also evaluated with the Zry-4 air-oxidation literature data. This equation resulted in the almost similar air-oxidation breakaway timing to the actual experimental data at 800 °C. However, at 1000 °C, it showed an error of about 8 min. This could be inferred from the influence of the ZrN phase change due to the nitrogen existing in air.

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