Validation of the Deterministic Realistic Method Applied to CATHARE on LB LOCA Experiments

2002 ◽  
Author(s):  
Jean-Yves Sauvage ◽  
Ste´phane Laroche
Keyword(s):  
Experimental Data ◽  
Statistical Analysis ◽  
Deterministic Model ◽  
Evaluation Model ◽  
Flexible Tool ◽  
Model Based ◽  
Design Basis ◽  
Best Estimate ◽  
Design Basis Accident

Framatome-ANP and EDF have defined a generic approach for using a best-estimate code in design basis accident studies called Deterministic Realistic Method (DRM). It has been applied to elaborate a LB LOCA ECCS evaluation model based on the CATHARE code. From a prior statistical analysis of uncertainties, the DRM derives a conservative deterministic model, preserving the realistic nature of the simulation, to be used in the further applications. The conservatism of the penalized model is demonstrated comparing penalized calculations with relevant experimental data. The DRM proved to be a highly flexible tool and has been applied successfully to meet the specific French and specific Belgian requirements of Safety Authorities.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

scholarly journals Application of RELAP5/MOD3.2 Cladding Deformation Model for VVER-1000 Fuel in Design-Basis Accident Analysis

2016 ◽  
pp. 19-22
Author(s):  
Yu. Vorobyov ◽  
O. Zhabin ◽  
M. Frankova
Keyword(s):  
Experimental Data ◽  
Limited Range ◽  
Simplified Model ◽  
Model Parameters ◽  
Accident Analysis ◽  
Deformation Model ◽  
The Core ◽  
Design Basis ◽  
Channel Blockage ◽  
Design Basis Accident

The paper presents applicability of built-in RELAP5/MOD3.2 cladding deformation model for VVER-1000 fuel with cladding of Zr+1 % Nb alloy. Experimental data and simplified model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the hot channel blockage after cladding swelling and rupture in the interval of temperatures from 600 to 1200°С and interval of pressures from 1 to 12 MPa. It is demonstrated that RELAP5/MOD3.2 builtin model of cladding deformation can be applied to VVER-1000 cladding of Zr+1%Nb alloy rupture estimation only in the certain limited range of parameters. The analysis of RELAP5/MOD3.2 cladding deformation model parameters influence on the peak cladding temperature in double-ended cold leg break was performed. The paper presents recommendations on the use of RELAP5/MOD3.2 built-in cladding deformation model in the design basis accident analysis of VVER-1000 reactors.

scholarly journals Analysis of a severe beyond design basis accident for the EGP-6 reactor of the Bilibino NPP. Radioactive source term determination

2018 ◽  
Vol 4 (2) ◽  
pp. 135-142
Author(s):  
Ruben Mukhamadeev ◽  
Leonid Parafilo ◽  
Yury Baranaev ◽  
Albert Suvorov
Keyword(s):  
Experimental Data ◽  
Source Term ◽  
Fission Products ◽  
Severe Accident ◽  
Water Mixture ◽  
Design Basis ◽  
Short Period ◽  
Design Basis Accident ◽  
Fuel Material ◽  
Graphite Stack

Analysis was performed of dynamic phase of severe accident of the EGP-6 reactor of the Bilibino NPP, due to uncontrolled reactivity insertion initiated by withdrawal of two pare of automatic control rods with followed by full failure of reactor emergency protection system. This initial event leads to promt increasing of reactor core power up to 450% of nominal value with short period, coupled with rise of temperature of fuel, pressure and temperature of coolant. These factors lead to crisis of heat exchange with subsequent ruptures tubes of fuel assemblies and coolant blow down into graphite stack. All its lead to rise of pressure in reactor shell and damage of it, outflow of steam-water mixture through up-reactor area to ventilation system, communication corridors and reactor hall and further – to atmospheric release. Transient processes were calculated using code RELAP5/Mod3.2. It was considered stages of processes of fuel damage and evaluated dynamic of a number and degree of damaged fuel assembles. They were grouped on burn-up and for each group it was performed analysis of dynamic of damage values. Further it was considered processes of yield of fission products from damaged fuel with models, based on experimental data on yield of fission products from fuel material, used in assembles of Bilibino NPP fuel type (fuel tubes with steel cladding, where fuel material is grits of uranium dioxide in magnesium), under condition of severe accident, especially performed in SSC IPPE. Transport of fission products with steam and air up to release points was evaluated with models, based on experimental data of fission product transport through graphite stack under conditions of severe accident, also especially performed in SSC IPPE. Evaluation of source term was performed in accordance with accident dynamic and assumed modes of release for conservative and most possible approaches. It was noted good self-protection property of EGP-6 reactor under severe beyond design basis accident condition.

Thermoeconomic Analysis of Air Conditioning Systems

2014 ◽  
Vol 875-877 ◽  
pp. 1748-1753 ◽  
Author(s):  
Wei Liang Cheng ◽  
Hui Ji ◽  
An Di
Keyword(s):  
Experimental Data ◽  
Air Conditioning ◽  
Performance Enhancement ◽  
Evaluation Model ◽  
Air Conditioning System ◽  
Model Based ◽  
Operation Costs ◽  
And Performance ◽  
Air Conditioning Systems

In order to decrease the operation costs of air conditioning systems, an evaluation model based on unit thermoeconomic costs of thermoeconomic theory is presented in this paper. By using real components and fictitious components in an air conditioning system, the relationships between the fuel and product are established, and then the operation performances of the air conditioning system can be analyzed and evaluated. The unit thermoeconomic costs can be obtained with the experimental data. The results show that the unit thermoeconomic cost of the system is the lowest when the vaporizing temperature is at 16.3°C, and the unit thermoeconomic cost of the compressor component is the highest. Therefore, the direction and emphases of the technique improvement and performance enhancement are provided.

Load-balancing evaluation model based on pre-defined variability

2009 ◽  
Vol 29 (10) ◽  
pp. 2849-2851
Author(s):  
Li-lun ZHANG ◽  
Jian-ping WU ◽  
Jun-qiang SONG
Keyword(s):  
Load Balancing ◽  
Evaluation Model ◽  
Model Based

Fuzzy Modeling to Simulate Surface Roughness in CNC Turning of AISI 1045 Steel

2011 ◽  
Vol 486 ◽  
pp. 262-265
Author(s):  
Amit Kohli ◽  
Mudit Sood ◽  
Anhad Singh Chawla
Keyword(s):  
Experimental Data ◽  
Surface Roughness ◽  
Process Parameters ◽  
Fuzzy Model ◽  
Fuzzy Modeling ◽  
Cnc Machine ◽  
Model Based ◽  
Aisi 1045 Steel ◽  
Aisi 1045 ◽  
1045 Steel

The objective of the present work is to simulate surface roughness in Computer Numerical Controlled (CNC) machine by Fuzzy Modeling of AISI 1045 Steel. To develop the fuzzy model; cutting depth, feed rate and speed are taken as input process parameters. The predicted results are compared with reliable set of experimental data for the validation of fuzzy model. Based upon reliable set of experimental data by Response Surface Methodology twenty fuzzy controlled rules using triangular membership function are constructed. By intelligent model based design and control of CNC process parameters, we can enhance the product quality, decrease the product cost and maintain the competitive position of steel.

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