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

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.

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Thermal hydraulic analysis of small nuclear reactor core by using comparative approach of THEATRe and Relap5 code

International Journal of Nuclear Energy Science and Technology ◽

10.1504/ijnest.2011.045097 ◽

2011 ◽

Vol 6 (4) ◽

pp. 271

Author(s):

Salah Ud Din Khan ◽

Minjun Peng ◽

Shahab Ud Din Khan

Keyword(s):

Nuclear Reactor ◽

Reactor Core ◽

Comparative Approach ◽

Hydraulic Analysis ◽

Nuclear Reactor Core ◽

Thermal Hydraulic Analysis ◽

Relap5 Code

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

Energies ◽

10.3390/en11123509 ◽

2018 ◽

Vol 11 (12) ◽

pp. 3509 ◽

Cited By ~ 3

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.

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CROSSFLOW CALCULATION FOR THE FORT ST. VRAIN NUCLEAR REACTOR CORE.

10.2172/4111753 ◽

1970 ◽

Author(s):

R Kiefer

Keyword(s):

Nuclear Reactor ◽

Reactor Core ◽

Nuclear Reactor Core

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Experimental study of heat transfer through cooling water circuit in a reactor vault by using Al2O3 nano fluid

Thermal Science ◽

10.2298/tsci170527207a ◽

2018 ◽

Vol 22 (2) ◽

pp. 1149-1161 ◽

Cited By ~ 3

Author(s):

Maria Anish ◽

Balakrishnan Kanimozh

Keyword(s):

Heat Transfer ◽

Temperature Distribution ◽

Base Fluid ◽

Nuclear Reactor ◽

Characteristic Curve ◽

Reactor Core ◽

Cooling Water ◽

Heat Transfer Process ◽

Nano Fluid ◽

Al2o3 Nanoparticle

The heat produced in the nuclear reactor due to fission reaction must be kept in control or else it will damage the components in the reactor core. Nuclear plants are using water for the operation dissipation of heat. Instead, some chemical substances which have higher heat transfer coefficient and high thermal conductivity. This experiment aims to find out how efficiently a nanofluid can dissipate heat from the reactor vault. The most commonly used nanofluid is Al2O3 nanoparticle with water or ethylene as base fluid. The Al2O3 has good thermal property and it is easily available. In addition, it can be stabilized in various PH levels. The nanofluid is fed into the reactor?s coolant circuit. The various temperature distribution leads to different characteristic curve that occurs on various valve condition leading to a detailed study on how temperature distribution carries throughout the cooling circuit. As a combination of Al2O3 as a nanoparticle and therminol 55 as base fluid are used for the heat transfer process. The Al2O3 nanoparticle is mixed in therminol 55 at 0.05 vol.% concentration. Numerical analysis on the reactor vault model was carried out by using ABAQUS and the experimental results were compared with numerical results.

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