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.

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.

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.

scholarly journals On a Roadmap for Future Industrial Nuclear Reactor Core Simulation in the U.K. to Support the Nuclear Renaissance

Energies ◽  
2018 ◽  
Vol 11 (12) ◽  
pp. 3509 ◽  
Author(s):  
Bruno Merk ◽  
Mark Bankhead ◽  
Dzianis Litskevich ◽  
Robert Gregg ◽  
Aiden Peakman ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Basic Research ◽  
Modelling And Simulation ◽  
Software Environment ◽  
Specific Detail ◽  
Technical Requirements ◽  
Strategic Approach ◽  
Nuclear Reactor Core ◽  
Core Simulation

The U.K. has initiated the nuclear renaissance by contracting for the first two new plants and announcing further new build projects. The U.K. government has recently started to support this development with the announcement of a national programme of nuclear innovation. The aim of this programme with respect to modelling and simulation is foreseen to fulfil the demand in education and the build-up of a reasonably qualified workforce, as well as the development and application of a new state-of-the-art software environment for improved economics and safety. This document supports the ambition to define a new approach to the structured development of nuclear reactor core simulation that is based on oversight instead of looking at detail problems and the development of single tools for these specific detail problems. It is based on studying the industrial demand to bridge the gap in technical innovation that can be derived from basic research in order to create a tailored industry solution to set the new standard for reactor core modelling and simulation for the U.K. However, finally, a technical requirements specification has to be developed alongside the strategic approach to give code developers a functional specification that they can use to develop the tools for the future. Key points for a culture change to the application of modern technologies are identified in the use of DevOps in a double-strata approach to academic and industrial code development. The document provides a novel, strategic approach to achieve the most promising final product for industry, and to identify the most important points for improvement.

scholarly journals Lagrangian Approach for the Study of Heat Transfer in a Nuclear Reactor Core Using the SPH Methodology

2019 ◽  
pp. 108-124
Author(s):  
F. Pahuamba-Valdez ◽  
E. Mayoral-Villa ◽  
C. E. Alvarado-Rodríguez ◽  
J. Klapp ◽  
A. M. Gómez-Torres ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Lagrangian Approach ◽  
Nuclear Reactor Core

scholarly journals Gas-cooled nuclear reactor core shaping using heat exchange intensifiers

2019 ◽  
Vol 5 (1) ◽  
pp. 75-80
Author(s):  
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny
Keyword(s):  
Heat Exchange ◽  
Pressure Drop ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Temperature Fields ◽  
Ansys Fluent ◽  
Safety Requirements ◽  
Cooling Channels ◽  
The Core ◽  
Nuclear Reactor Core

The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping. The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed. The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields. The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels. Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop. The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).

Sign in / Sign up

Export Citation Format

Share Document