Simulation of Nuclear Fuel Behavior in Accident Conditions With the DIONISIO Code

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Martín Lemes ◽  
Alicia Denis ◽  
Alejandro Soba
Keyword(s):  
Nuclear Fuel ◽  
Computer Code ◽  
Hydraulic Modeling ◽  
Application Range ◽  
Behavior Simulation ◽  
External Conditions ◽  
Capture And Release ◽  
Fuel Behavior ◽  
Accident Conditions ◽  
Thermal Hydraulic Modeling

DIONISIO is a computer code designed to simulate the behavior of one nuclear fuel rod during its permanence within the reactor. Starting from the power history and the external conditions to which the rod is subjected, the code predicts all the meaningful variables of the system. Its application range has been recently extended to include accidental conditions, in particular the so-called loss of coolant accidents (LOCA). In order to make realistic predictions, the conditions in the rod environment have been taken into account since they represent the boundary conditions with which the differential equations describing the fuel phenomena are solved. Without going into the details of the thermal-hydraulic modeling, which is the task of the specific codes, a simplified description of the conditions in the cooling channel during a LOCA event has been developed and incorporated as a subroutine of DIONISIO. This has led to an improvement of the fuel behavior simulation, which is evidenced by the considerable number of comparisons with experiments carried out, many of them reported in this paper. Moreover, this work describes a model of high temperature capture and release of hydrogen in the nuclear fuel cladding, in scenarios typical of LOCA events. The corresponding computational model is being separately tested and will be next included in the DIONISIO thermal-hydraulic module.

Comparison of Thermal Hydraulic Simulations of Beyond Design Basis Events Using the MAAP4 and CENTS Computer Codes

2009 ◽  
Author(s):  
N. Reed LaBarge ◽  
Barbara R. Baron ◽  
Raymond E. Schneider ◽  
Mathew C. Jacob
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Computer Code ◽  
Hydraulic Modeling ◽  
Severe Accident ◽  
Pressurized Water ◽  
Design Basis ◽  
Best Estimate ◽  
Water Reactors ◽  
Thermal Hydraulic Modeling

The MAAP4 computer code (Reference 1) is often used to perform thermal hydraulic simulations of severe accident sequences for nuclear power plant Probabilistic Risk Assessments (PRAs). MAAP4 can be used to simulate accidents for both Boiling Water Reactors (BWRs) as well as Pressurized Water Reactors (PWRs). This assessment employs MAAP 4.0.6a for PWRs (References 1 and 5), which incorporates explicit thermal hydraulic modeling of the Reactor Coolant System (RCS) and Steam Generators (SGs), along with a nodalized integrated containment model. In the PRA environment, MAAP4 has been used for applications such as the development of PRA Level 1 and Level 2 success criteria and human action timings. The CENTS computer code (Reference 2) is a simulation tool that is typically used to analyze non-Loss of Coolant Accident (non-LOCA) events postulated to occur in nuclear power plants incorporating Combustion Engineering (CE) and Westinghouse Nuclear Steam Supply System (NSSS) designs. It is licensed by the NRC perform design basis non-LOCA safety analyses. It is a best estimate code which uses detailed thermal hydraulic modeling of the RCS and SGs; however, it does not model the containment performance. It is used to perform a wide spectrum of licensing and best estimate non-LOCA event analysis and has the capability to simulate operator actions. The CENTS models are the basis for several full scope simulators in the industry. The purpose of the analyses described in this paper is to compare MAAP4 and CENTS predictions for the Station Blackout (SBO) and Total Loss of Feedwater (TLOFW) scenarios for a representative PWR in the Westinghouse fleet that employs a CE NSSS design. The results of this comparison are used to highlight postulated MAAP4 user challenges and assist in developing guidance on selecting MAAP4 parameters for use in these scenarios. The results of the analyses presented in this paper indicate several useful insights. Overall, this paper shows that when care is taken to normalize the MAAP4 and CENTS primary side natural circulation flowrate and SG modeling, the trends of the MAAP4 and CENTS predictions of core uncovery agree reasonably well.

Thermal–hydraulic modeling and transient analysis of helium-cooled tungsten target coupled with ADS core

2016 ◽  
Vol 87 ◽  
pp. 612-620
Author(s):  
Tianji Peng ◽  
Zhiwei Zhou ◽  
Sicong Xiao ◽  
Xuanyu Sheng ◽  
Long Gu
Keyword(s):  
Transient Analysis ◽  
Hydraulic Modeling ◽  
Tungsten Target ◽  
Thermal Hydraulic Modeling

Thermal Hydraulic Analysis of Severe Accident in PFBR

2010 ◽  
Author(s):  
K. Velusamy ◽  
P. Chellapandi ◽  
G. R. Raviprasan ◽  
P. Selvaraj ◽  
S. C. Chetal
Keyword(s):  
Atmospheric Pressure ◽  
Fluid Dynamic ◽  
Hydraulic Modeling ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Maximum Pressure ◽  
Dynamic Calculation ◽  
Thermal Hydraulic Modeling ◽  
Thermal Hydraulic Analysis ◽  
Design Pressure

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.

Thermal-hydraulic modeling and analysis of spool valve with sloping U-shape notch by bond graph

2015 ◽  
Vol 22 (11) ◽  
pp. 4205-4212
Author(s):  
Lei Lou ◽  
Wan-rong Wu ◽  
Zhao-Qiang Wang ◽  
Xiang-jing Liang
Keyword(s):  
Hydraulic Modeling ◽  
Bond Graph ◽  
Modeling And Analysis ◽  
Thermal Hydraulic Modeling ◽  
Spool Valve

Multiscale Modeling of Uranium Mononitride: Point Defects Diffusion, Self-Diffusion, Phase Composition

2017 ◽  
Vol 375 ◽  
pp. 101-113 ◽  
Author(s):  
Sergey Starikov ◽  
Alexey Kuksin ◽  
Daria Smirnova ◽  
Alexey Dolgodvorov ◽  
Vladimir Ozrin
Keyword(s):  
Nuclear Fuel ◽  
Point Defects ◽  
Atomistic Simulation ◽  
Nitrogen Pressure ◽  
Computational Approach ◽  
Self Diffusion ◽  
Partial Nitrogen Pressure ◽  
Diffusion Phase ◽  
Fuel Behavior ◽  
Uranium Mononitride

Multiscale computational approach is used to evaluate microscopic parameters for description of nitride nuclear fuel. The results of atomistic simulation and thermodynamic modeling allow to estimate diffusivity and concentrations of point defects at various stoichiometric ratios of UN1+x. The diffusivities of Xe atom were calculated in various equilibrium states. In addition, we obtained the dependence of partial nitrogen pressure on x and temperature. The results of atomistic simulation were used for modeling of nuclear fuel behavior with use of mechanistic fuel codes.

First Steps Towards the Development of a 3D Nuclear Fuel Behavior Solver With OpenFOAM

2018 ◽  
Author(s):  
Alessandro Scolaro ◽  
Ivor Clifford ◽  
Carlo Fiorina ◽  
Andreas Pautz
Keyword(s):  
Nuclear Fuel ◽  
Material Properties ◽  
Three Dimensional ◽  
Small Strain ◽  
Theoretical Background ◽  
Transfer Model ◽  
Temperature Dependent ◽  
Reactor Physics ◽  
Radial Temperature ◽  
Fuel Behavior

A new 3D fuel behavior solver is currently under collaborative development at the Laboratory for Reactor Physics and Systems Behaviour of the École Polytechnique Fédérale de Lausanne and at the Paul Scherrer Institut. The long term objective is to enable a more accurate simulation of inherently 3D safety-relevant phenomena which affect the performance of the nuclear fuel. The current implementation is a coupled three-dimensional heat conduction and linear elastic small strain solver, which models the effects of burnup- and temperature dependent material properties, swelling, relocation and gap conductance. The near future developments will include the introduction of a smeared pellet cracking model and of material inleasticities, such as creep and plasticity. After an overview of the theoretical background, equations and models behind the solver, this work focuses on the recent preliminary verification and validation efforts. The radial temperature and stress profiles predicted by the solver for the case of an infinitely long rod are compared against their analytical solution, allowing the verification of the thermo-mechanics equations and of the gap heat transfer model. Then, an axisymmetric model is created for 4 rods belonging to the Halden assembly IFA-432. These models are used to predict the fuel centerline temperature during power ramps recorded at the beginning of life, when the fuel rod performance is still not affected by more complex high burnup effects. Finally, the predictions are compared with the experimental measurements coming from the IFPE database. This first preliminary results allow a careful validation of the temperature-dependent material properties and of the gap conductance models.

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