Search engine for discovering works of Art, research articles, and books related to Art and Culture
ShareThis
Javascript must be enabled to continue!

Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor

View through CrossRef
Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical energy generation. The largest group of operating nuclear power plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) has gross thermal efficiencies ranging from 30–36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8  MPa/257–293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fueled by natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., supercritical water-cooled reactors (SCWRs) have to be designed. This path of increasing thermal efficiency is considered as a conventional way that coal-fired power plants followed more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel pressure-channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulics code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the computational fluid dynamics (CFD) Fluent code has been used for better understanding of the specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical water loop and developing passive safety systems.
Title: Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor
Description:
Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical energy generation.
The largest group of operating nuclear power plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) has gross thermal efficiencies ranging from 30–36%.
Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.
5–7.
8  MPa/257–293°C).
However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fueled by natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively.
Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants.
A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.
e.
, supercritical water-cooled reactors (SCWRs) have to be designed.
This path of increasing thermal efficiency is considered as a conventional way that coal-fired power plants followed more than 50 years ago.
Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel pressure-channel (PCh) SCWR.
Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulics code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR.
In addition, the computational fluid dynamics (CFD) Fluent code has been used for better understanding of the specifics of heat transfer in supercritical water.
Future studies will be dedicated to materials and fuels testing in an in-pile supercritical water loop and developing passive safety systems.

Related Results

Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical-energy generation. The largest group of operating Nuclear Pow...
Studies of the Thermalhydraulics Subchannel Code ASSERT-PV 3.2-SC for Supercritical Applications
Studies of the Thermalhydraulics Subchannel Code ASSERT-PV 3.2-SC for Supercritical Applications
Abstract Over the last decade, several international thermalhydraulics benchmarking efforts have been carried out to support the development of the Generation IV sup...
En skvatmølle i Ljørring
En skvatmølle i Ljørring
A Horizontal Mill at Ljørring, Jutland.Horizontal water-mills have been in use in Jutland since the beginning of the Christian era 2). But the one here described shows so close a c...
Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor
Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor
The Molten Salt Reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for production of electricity, actinide burning, production of hydrogen, and production of ...
ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM
ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM
ANALISISPENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE IN...
Thermally and acoustically driven transport in supercritical fluids
Thermally and acoustically driven transport in supercritical fluids
Supercritical fluids are fluids at temperature and pressure above their respective critical values. Such fluids are increasingly being used in power generation, refrigeration and c...
Sub-Channel Analysis of Pb-Bi-Cooled Reactor With Modified COBRA-EN
Sub-Channel Analysis of Pb-Bi-Cooled Reactor With Modified COBRA-EN
Lead-bismuth eutectic (LBE) cooled fast reactor, one of the six types of reactors in Gen-IV, has very good inherent safety and significant advantages in reducing and burning nuclea...

Back to Top