scholarly journals A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

2011 ◽  
Vol 2011 ◽  
pp. 1-11 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D'Auria ◽  
Oscar Mazzantini
Keyword(s):  
Hydraulic System ◽  
Initial Conditions ◽  
Probabilistic Method ◽  
Three Dimensional ◽  
Computer Code ◽  
Calculation Methods ◽  
Neutron Physics ◽  
Fuel Rod ◽  
Fuel Pin ◽  
Computer Codes

Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

scholarly journals A Procedure to Address the Fuel Rod Failures During LB-LOCA Transient in Atucha-2 NPP

2010 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria ◽  
Oscar Mazzantini
Keyword(s):  
Hydraulic System ◽  
Initial Conditions ◽  
Probabilistic Method ◽  
Three Dimensional ◽  
Computer Code ◽  
Calculation Methods ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pin ◽  
Computer Codes

Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes. This may imply the use of a dedicated fuel rod thermo-mechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermo-mechanical code. The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions (e.g. pin power axial profiles) are provided by core physics and three dimensional neutron kinetic coupled thermal-hydraulic system codes (RELAP5-3D©) calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, with the aim of a better understanding of the uncertainties involved and their technological consequences on the behavior of the fuel rods, not addressed in the current paper.

Methodology for the Analysis of Fuel Behavior During LOCA and RIA for Licensing Purposes

2012 ◽  
Author(s):  
Martina Adorni ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Computer Code ◽  
Safety Analysis ◽  
Fuel Rod ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fuel Pin ◽  
Computer Codes ◽  
Flow Blockage

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and may imply the use of system thermal hydraulic computer codes. Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code, which can be coupled with thermal-hydraulic system and neutron kinetic codes to be used for the safety analysis. This paper describes the development and the application of a methodology for the analysis of the Large Break Loss of Coolant Accident (LB-LOCA) scenario in Atucha-2 Nuclear Power Plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis (Chapter 15 Final Safety Analysis Report, FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. A strong effort has been performed in order to enhance the fuel behaviour code capabilities and to improve the reliability of the code results.

scholarly journals International Course to Support Nuclear Licensing by User Training in the Areas of Scaling, Uncertainty, and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes: 3D S.UN.COP Seminars

2008 ◽  
Vol 2008 ◽  
pp. 1-16 ◽  
Author(s):  
Alessandro Petruzzi ◽  
Francesco D'Auria ◽  
Tomislav Bajs ◽  
Francesc Reventos ◽  
Yassin Hassan
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Effective Means ◽  
Computer Code ◽  
Nuclear Facilities ◽  
User Training ◽  
Research Organizations ◽  
Coupled Codes ◽  
Regulatory Bodies ◽  
Computer Codes

Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the “user effect” and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to the areas of the scaling, uncertainty, and 3D coupled code analysis.

Development and Verification of Multi-Physical Coupling Calculation Code for Typical PWR Core

2020 ◽  
Author(s):  
Zhigang Li ◽  
Ping An ◽  
Junjie Pan ◽  
Wei Liu ◽  
Wei Lu ◽  
...  
Keyword(s):  
Dimensional Space ◽  
Three Dimensional ◽  
Hydraulic Calculation ◽  
Relative Power ◽  
Calculation Methods ◽  
Neutron Physics ◽  
Dimensional Space Time ◽  
Coupling Calculation ◽  
Calculation Code ◽  
Three Dimensional Space

Abstract Strong feedback phenomenon between the neutronics physical and thermal-hydraulic has an important impact on the design and safety analysis of pressured water reactor (PWR). In order to accurately simulate the strong coupling effect of neutron physics and thermal hydraulic in PWR, a reactor multi-physical coupling calculation code (ARMcc) is developed, in which the three-dimensional space-time neutron dynamic equation is solved by nodal expansion method (NEM) and nodal Green's function method (NGFM), the coolant temperature and fuel temperature are solved by single channel model and the cylinder heat conduction model respectively. The 3D IAEA benchmark, the 3D Langenbuch Maurer Werner (LMW) benchmark and NEACRP-L-335 benchmark are used to verify the neutronics model and coupling calculation solution ability respectively. The results show that: 1) the NEM and NGFM have high accuracy in solving the three-dimensional space-time neutron dynamics equation; 2) the results of neutronics and thermal-hydraulic coupling steady/transient calculation such as core relative power and fuel Doppler temperature are in good agreement with those of the NEACRP-L-335 benchmark, and the calculation accuracy is equivalent to similar software such as PARCS. Four coupled neutron physics and thermal hydraulic calculation modes are used to analyze the influence of different neutron physics calculation methods and thermal hydraulic calculation methods on the key parameters of PWR transient process in this paper. The results show that the mode of NGFM + FVM can more accurately simulate the core relative power peak and Doppler temperature.

Enhancement of Heat Transfer Performance in Nuclear Fuel Rod Using Nanofluids and Surface Roughness Technique

2015 ◽  
Author(s):  
Kang Liu ◽  
Titan C. Paul ◽  
Leo A. Carrilho ◽  
Jamil A. Khan
Keyword(s):  
Heat Transfer ◽  
Surface Roughness ◽  
Heat Transfer Coefficient ◽  
Transfer Coefficient ◽  
Nuclear Fuel ◽  
Three Dimensional ◽  
Pressurized Water ◽  
Bulk Fluid ◽  
Fuel Rod ◽  
Dimensional Surface

The experimental investigations were carried out of a pressurized water nuclear reactor (PWR) with enhanced surface using different concentration (0.5 and 2.0 vol%) of ZnO/DI-water based nanofluids as a coolant. The experimental setup consisted of a flow loop with a nuclear fuel rod section that was heated by electrical current. The fuel rod surfaces were termed as two-dimensional surface roughness (square transverse ribbed surface) and three-dimensional surface roughness (diamond shaped blocks). The variation in temperature of nuclear fuel rod was measured along the length of a specified section. Heat transfer coefficient was calculated by measuring heat flux and temperature differences between surface and bulk fluid. The experimental results of nanofluids were compared with the coolant as a DI-water data. The maximum heat transfer coefficient enhancement was achieved 33% at Re = 1.15 × 105 for fuel rod with three-dimensional surface roughness using 2.0 vol% nanofluids compared to DI-water.

Investigation of Quench-16 Experiment With MELCOR Computer Code

2014 ◽  
Author(s):  
Petya Vryashkova ◽  
Pavlin Groudev ◽  
Antoaneta Stefanova
Keyword(s):  
Experimental Data ◽  
Flow Rate ◽  
Initial Temperature ◽  
Computer Code ◽  
Air Flow Rate ◽  
Fuel Rod ◽  
Fuel Bundle ◽  
Oxygen Starvation ◽  
Air Ingress ◽  
Good Agreement

This paper presents a comparison of MELCOR calculated results with experimental data for the QUENCH-16 experiment. The analysis for the air ingress experiment QUENCH-16 has been performed by INRNE. The calculations have been performed with MELCOR code. The QUENCH-16 experiment has been performed on 27-th of July 2011 in the frame of the EC-supported LACOMECO program. The experiments have focused on air ingress investigation into an overheated core following earlier partial oxidation in steam. QUENCH-16 has been performed with limited pre-oxidation and low air flow rate. One of the main objectives of QUENCH-16 was to examine the interaction between nitrogen and oxidized cladding during a prolonged period of oxygen starvation. The bundle is made from 20 heated fuel rod simulators arranged in two concentric rings and one unheated central fuel rod simulator, each about 2.5 m long. The tungsten heaters were surrounded by annular ZrO2 pellets to simulate the UO2 fuel. The geometry and most other bundle components are prototypical for Western-type PWRs. To improve the obtained results it has been made a series of calculations to select an appropriate initial temperature of the oxidation of the fuel bundle and modified correlation oxidation of Zircaloy with MELCOR computer code. The compared results have shown good agreement of calculated hydrogen and oxygen starvation in comparison with test data.

Numerical investigation of the propagation of shock waves in rigid porous materials: development of the computer code and comparison with experimental results

1996 ◽  
Vol 324 ◽  
pp. 163-179 ◽  
Author(s):  
A. Levy ◽  
G. Ben-Dor ◽  
S. Sorek
Keyword(s):  
Porous Materials ◽  
Initial Conditions ◽  
Computer Code ◽  
Experimental Results ◽  
Numerical Code ◽  
Materials Development ◽  
Numerical Predictions ◽  
Wide Range ◽  
Dimensional Version ◽  
The One

The governing equations of the flow field which is obtained when a thermoelastic rigid porous medium is struck head-one by a shock wave are developed using the multiphase approach. The one-dimensional version of these equations is solved numerically using a TVD-based numerical code. The numerical predictions are compared to experimental results and good to excellent agreements are obtained for different porous materials and a wide range of initial conditions.

Forklift Hydraulic System Simulation Based on ADAMS

2012 ◽  
Vol 424-425 ◽  
pp. 598-602 ◽  
Author(s):  
You Min Wang ◽  
Chun Zhao ◽  
Jian Hua Zhang
Keyword(s):  
Hydraulic System ◽  
Production Cost ◽  
Control Function ◽  
Three Dimensional ◽  
Improve Design ◽  
Kinematics Simulation ◽  
Simulation Based ◽  
Motion Speed ◽  
Work Size ◽  
Control Functional

In order to improve design performance, shorten development cycles, reduce production cost, we design and research the forklift hydraulic system, developed forklift hydraulic system diagram. Forklift virtual prototype’s 3-D solid modeling is designed by Pro / E three-dimensional software, and imported into the ADAMS environment. Add constraints and drivers exert the control function separately to the tilting cylinder and lifting cylinder, carry on the kinematics simulation. Through the analysis to the compound motion actuation control functional arrangement、the compound motion speed graph、the gate’s tilt angle graph、the tilting cylinder stress graph and the lifting cylinder stress graph, he simulation result indicated: each cylinder design is reasonable, the movement without interference，the reasonable work scope satisfied to the work size request

scholarly journals Computation of Three-Dimensional, Rotational Flow Through Turbomachinery Blade Rows for Improved Aerodynamic Design Studies

1986 ◽  
Author(s):  
S. V. Subramanian ◽  
R. Bozzola ◽  
Louis A. Povinelli
Keyword(s):  
Three Dimensional ◽  
Maximum Flow ◽  
Cost Effective ◽  
Computer Code ◽  
Integration Scheme ◽  
Aerodynamic Design ◽  
Design Studies ◽  
Flow Equations ◽  
Transonic Compressor ◽  
Numerical Predictions

The performance of a three dimensional computer code developed for predicting the flowfield in stationary and rotating turbomachinery blade rows is described in this study. The four stage Runge-Kutta numerical integration scheme is used for solving the governing flow equations and yields solution to the full, three dimensional, unsteady Euler equations in cylindrical coordinates. This method is fully explicit and uses the finite volume, time marching procedure. In order to demonstrate the accuracy and efficiency of the code, steady solutions were obtained for several cascade geometries under widely varying flow conditions. Computed flowfield results are presented for a fully subsonic turbine stator and a low aspect ratio, transonic compressor rotor blade under maximum flow and peak efficiency design conditions. Comparisons with Laser Anemometer measurements and other numerical predictions are also provided to illustrate that the present method predicts important flow features with good accuracy and can be used for cost effective aerodynamic design studies.

Integrity Assessment of Damaged Flexible Pipe Cross-Sections

2014 ◽  
Author(s):  
Guomin Ji ◽  
Bernt J. Leira ◽  
Svein Sævik ◽  
Frank Klæbo ◽  
Gunnar Axelsson ◽  
...  
Keyword(s):  
Cross Sections ◽  
Three Dimensional ◽  
Element Model ◽  
Computer Code ◽  
Flexible Pipe ◽  
Dynamic Stresses ◽  
Integrity Assessment ◽  
Residual Capacity ◽  
Nonlinear Finite Element Model

This paper presents results from a case study performed to evaluate the residual capacity of a 6″ flexible pipe when exposed to corrosion damages in the tensile armour. A three-dimensional nonlinear finite element model was developed using the computer code MARC to evaluate the increase in mean and dynamic stresses for a given number of damaged inner tensile armor wires. The study also includes the effect of these damages with respect to the associated stresses in the pressure spiral. Furthermore, the implications of a sequence of wire failures with respect to the accumulated time until cross-section failure in a probabilistic sense are addressed.

Sign in / Sign up

Export Citation Format

Share Document