Dryout Heat Flux During Penetration of Water Into Solidifying Rock

2006 ◽  
Vol 128 (8) ◽  
pp. 847-850 ◽  
Author(s):  
Michael Epstein
Keyword(s):  
Heat Flux ◽  
Porous Medium ◽  
Steady State ◽  
Nuclear Reactor ◽  
Change Theory ◽  
Reactor Core ◽  
Core Material ◽  
One Dimensional ◽  
Nuclear Reactor Core ◽  
Good Agreement

A model for the dryout heat flux during penetration of water into solidifying rock is developed by combining steady-state one-dimensional phase change theory with available semiempirical equations for (i) the dryout heat flux in a porous medium and (ii) the permeability of hot rock cooled by water. The model is in good agreement with measurements made during the pouring of water onto molten magma. The implication of the model with respect to stabilizing molten-nuclear-reactor-core material by flooding from above is discussed.

scholarly journals UNCERTAINTY CHARACTERIZATION FRAMEWORK FOR STEADY-STATE AND TRANSIENT NEUTRONICS SIMULATIONS OF A CANDU REACTOR

2021 ◽  
Vol 247 ◽  
pp. 15002
Author(s):  
Hany S. Abdel-Khalik ◽  
Alexandre Trottier ◽  
Dumitru Serghiuta ◽  
Dongli Huang
Keyword(s):  
Steady State ◽  
Cross Section ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Modeling Uncertainty ◽  
Group Format ◽  
Nuclear Reactor Core ◽  
Nodal Diffusion ◽  
Complete Framework

This paper reports on the development and testing of a comprehensive few-group cross section input uncertainty library for the NESTLE-C nodal diffusion-based nuclear reactor core simulator. This library represents the first milestone of a first-of-a-kind framework for the integrated characterization of uncertainties in steady-state and transient CANDU reactor simulations. The objective of this framework is to propagate, prioritize and devise a mapping capability for uncertainties in support of model validation of best-estimate calculations. A complete framework would factor both input and modeling uncertainty contributions. The scope of the present work is limited to the propagation of multi-group cross-section uncertainties through lattice physics calculations down to the few-group format, representing the input to the NESTLE-C core simulator, and finally to core responses of interest.

CONSUL Code Package Application for LMFR Core Calculations

2008 ◽  
Author(s):  
A. V. Chibinyaev ◽  
P. S. Teplov ◽  
M. V. Frolova
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Operational Parameters ◽  
Fast Power ◽  
Fast Power Reactor ◽  
Void Effect ◽  
Nuclear Reactor Core ◽  
Core Characteristics ◽  
Development Step ◽  
Good Agreement

CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90-th. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test “Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect” and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculations.

scholarly journals Effect Of U-9mo/Al Fuel Densities On Neutronic And Steady State Thermal Hydraulic Parameters Of Mtr Type Research Reactor

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Surian PINEM
Keyword(s):  
Heat Flux ◽  
Steady State ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Research Work ◽  
Reactor Core ◽  
Hydraulic Parameters ◽  
Limit Value ◽  
Maximum Temperatures ◽  
Coolant Velocity

<p>The objectives of this research work are to carry out a detailed neutronic and steady state thermal hydraulics analysis for a MTR research reactor fuelled with the low enrichment U-9Mo/Al dispersion fuels of various uranium densities. The high density uranium fuel will increase the cycle length of the reactor operation and the heat flux in the reactor core. The increasing heat flux at the fuel will causing increase the temperature of the fuel and cladding so that the coolant velocity has to be increased. However, the coolant velocity in the fuel element has a limit value due to the thermal hydraulic stability considerations in the core.  Therefore, the neutronic and the steady state thermal hydraulic analysis are important in the design and operation of nuclear reactor safety.  The calculations were performed using WIMS-D5 and MTRDYN codes. The WIMS-D5 code used for generating the group constants of all core materials as well as the neutronic and steady state thermal hydraulic parameters   were determined by using the MTRDYN code. The calculation results showed that the excess reactivity increases as the uranium density increases since the mass of fuel in the reactor core is increased.  Using the critical velocity concept, the maximum coolant velocity at fuel channel is 11.497 m/s.  The maximum temperatures of the coolant, cladding and fuel meat with the uranium density of 3,66 g/cc are 70.85°C, 150.79°C and 153.24°C, respectively.  The maximum temperatures are fulfilled the design limit so reactor has a safe operation at the nominal power.</p>

An Evaluation of the RPI Model for the Prediction of the Wall Heat Flux Partitioning in Subcooled Boiling Flows

2014 ◽  
Author(s):  
Ronak Thakrar ◽  
Janani S. Murallidharan ◽  
Simon P. Walker
Keyword(s):  
Heat Flux ◽  
Atmospheric Pressure ◽  
Nuclear Reactor ◽  
State Of The Art ◽  
Wall Heat Flux ◽  
Subcooled Boiling ◽  
One Dimensional ◽  
Flux Partitioning ◽  
Boiling Flow ◽  
Good Agreement

Subcooled boiling flows are encountered in most nuclear reactor configurations. Wall heat flux partitioning models form an integral part of the subcooled boiling formulations in CFD codes. These models attempt to describe the flow of heat from the wall into the fluid by dividing it according to several mechanisms of heat transfer. This work presents a one-dimensional evaluation of the wall heat flux partitioning model of Kurul and Podowski, also referred to commonly as the RPI model, which is used in the state-of-the-art codes of today. This model was assessed against the measurements of Okawa et al. for a vertically upward subcooled boiling flow of water at near atmospheric pressure. Although the predictions showed good agreement with the measured wall temperatures, significant discrepancies were observed in the predictions of the constituent sub-models that comprised the overall model. Prospects for improvement are discussed.

scholarly journals Localizing Perturbations in Pressurized Water Reactors Using One-Dimensional Deep Convolutional Neural Networks

Sensors ◽  
2021 ◽  
Vol 22 (1) ◽  
pp. 113
Author(s):  
Laurent Pantera ◽  
Petr Stulík ◽  
Antoni Vidal-Ferràndiz ◽  
Amanda Carreño ◽  
Damián Ginestar ◽  
...  
Keyword(s):  
Neural Networks ◽  
Steady State ◽  
Convolutional Neural Networks ◽  
Nuclear Reactor ◽  
Power Level ◽  
Reactor Core ◽  
Simulated Data ◽  
Deep Convolutional Neural Networks ◽  
Pressurized Water ◽  
One Dimensional

This work outlines an approach for localizing anomalies in nuclear reactor cores during their steady state operation, employing deep, one-dimensional, convolutional neural networks. Anomalies are characterized by the application of perturbation diagnostic techniques, based on the analysis of the so-called “neutron-noise” signals: that is, fluctuations of the neutron flux around the mean value observed in a steady-state power level. The proposed methodology is comprised of three steps: initially, certain reactor core perturbations scenarios are simulated in software, creating the respective perturbation datasets, which are specific to a given reactor geometry; then, the said datasets are used to train deep learning models that learn to identify and locate the given perturbations within the nuclear reactor core; lastly, the models are tested on actual plant measurements. The overall methodology is validated on hexagonal, pre-Konvoi, pressurized water, and VVER-1000 type nuclear reactors. The simulated data are generated by the FEMFFUSION code, which is extended in order to deal with the hexagonal geometry in the time and frequency domains. The examined perturbations are absorbers of variable strength, and the trained models are tested on actual plant data acquired by the in-core detectors of the Temelín VVER-1000 Power Plant in the Czech Republic. The whole approach is realized in the framework of Euratom’s CORTEX project.

Identification of a nuclear reactor core (VVER) using recurrent neural networks

2002 ◽  
Vol 29 (10) ◽  
pp. 1225-1240 ◽  
Author(s):  
Mehrdad Boroushaki ◽  
Mohammad B. Ghofrani ◽  
Caro Lucas
Keyword(s):  
Neural Networks ◽  
Recurrent Neural Networks ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Nuclear Reactor Core

The analysis of emergency situations related to fires at nuclear power plants

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Emergency Situations ◽  
The Usa ◽  
Large Fires ◽  
Short Circuits ◽  
Nuclear Reactor Core

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.

Handling of Collimated Irradiation on a Plane-Parallel Participating Medium

2000 ◽  
Author(s):  
Christian Proulx ◽  
Daniel R. Rousse ◽  
Rodolphe Vaillon ◽  
Jean-François Sacadura
Keyword(s):  
Heat Flux ◽  
Monte Carlo ◽  
Monte Carlo Method ◽  
Radiative Heat ◽  
Radiative Heat Flux ◽  
Scattering Media ◽  
One Dimensional ◽  
Plane Parallel ◽  
Collimated Beam ◽  
Good Agreement

Abstract This article presents selected results of a study comparing two procedures for the treatment of collimated irradiation impinging on one boundary of a participating one-dimensional plane-parallel medium. These procedures are implemented in a CVFEM used to calculate the radiative heat flux and source. Both isotropically and anisotropically scattering media are considered. The results presented show that both procedures provide results in good agreement with those obtained using a Monte Carlo method, when the collimated beam impinges normally.

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