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.

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.

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.

Multiphysics Modeling of Pressurized Water Reactor Fuel Performance

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chungyu ◽  
Yuan Cenxi
Keyword(s):  
Three Dimensional ◽  
Fission Products ◽  
Axial Direction ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Rods ◽  
Reactivity Insertion

Nuclear fuel rods operate under complex radioactive, thermal, and mechanical conditions. Nowadays, fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three-dimensional (3D) module within the framework of a general-purpose finite element solver, i.e., abaqus, for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermomechanical behavior of fuel rods. The swelling of fission products causes a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress in the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.

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.

Three Dimensional Modeling of the Thermo-Mechanical Performance of the Fuel Rods of a PWR

2017 ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chunyu ◽  
Li Aolin ◽  
Yuan Cenxi
Keyword(s):  
Mechanical Performance ◽  
Three Dimensional ◽  
Fission Products ◽  
General Purpose ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Three Dimensional Modeling ◽  
Water Reactors

The fuel rods of pressurized water reactors operate under complex radioactive, thermal and mechanical conditions. Multiphysics has to be taken into account in order to evaluate their performance. Many existing fuel rod codes make great simplifications on analyzing the behavior of fuel rods. The present study develops a three dimensional module within the framework of a general-purpose finite element solver, i.e. ABAQUS, for modeling the thermo-mechanical performance of the fuel rods. A typical fuel rod is modeled and the temperature as well as the stress within the pellets are computed. The results show that the burnup levels have an important influence on the fuel temperature. The swelling of fission products cause dramatically increasing of pellet strain. The change of the cladding stress and radial displacement with the axial length can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress to the outer regime of the pellet and may cause further fragmentation to the pellets.

scholarly journals Engineering Margin in Calculations of Energy Release in VVER-1000 core

2018 ◽  
pp. 20-26
Author(s):  
A.M. Abdullayev ◽  
A.I. Zhukov ◽  
S.V. Maryokhin ◽  
S.D. Riabchykov
Keyword(s):  
Energy Release ◽  
Three Dimensional ◽  
Measurement Data ◽  
Reactor Core ◽  
Small Scale ◽  
Two Dimensional ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies

A method for calculating the engineering margin factor (EMF) in calculations of the energy release in the core of VVER-1000 reactors is proposed in the paper. The analysis of various approaches in the calculation of EMF is carried out and various factors influencing EMF and the ways of their consideration —deterministic and statistical — are determined. The main attention is paid to the influence of gaps between the fuel assemblies on the energy release of fuel rods and the contribution of this factor to the EMF. The limitations and conservatism of two-dimensional small-scale calculations of the energy release of fuel rods in case of deviation of the gap size between the fuel assemblies from the design one are shown. A three-dimensional approach to calculating the contribution of gaps to the EMF is proposed. The approach is based on detailed measurements of the shape of fuel assemblies removed from the core performed at Zaporizhzhya NPP [13]; simulation of the distribution of gaps in the reactor core [16] using measurement data; two-dimensional calculations of the energy release of fuel rods in separate fuel assemblies, surrounded by gaps of different widths, with mirroring boundary conditions; three-dimensional calculations of energy release of fuel rods in fuel assemblies in the reactor core. Two-dimensional and three-dimensional calculations are performed by the wellknown ALPHA-H/PHOENIX-H/ANC-H codes. The proposed approach allows considering not only the change in the fuel rod power, particularly of the peripheral rods, which is inherent in the currently used methods of calculating EMF, but also takes into account the change in the power of the fuel assemblies in the core, which makes the proposed method more realistic and removes the excessive conservatism of EMF calculations and, thereby, allows improving fuel efficiency. For fuel assemblies produced by Westinghouse, it is proposed to use full EMF: for fuel rod power (FΔH) 1.111 and for fuel rod linear power (FQ) 1.173. The use of the BEACONTM monitoring system makes it possible to further reduce the EMF: for fuel rod power (FΔH) - up to 1.084 and for fuel rod linear power (FQ) - up to 1.121.

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.

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

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.

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