Thermal Design Options Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformly-Heated Fuel Channel

2009 ◽  
Author(s):  
Leyland J. Allison ◽  
Lisa Grande ◽  
Sally Mikhael ◽  
Adrianexy Rodriguez Prado ◽  
Bryan Villamere ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Operating Conditions ◽  
Thermal Design ◽  
Nuclear Fuels ◽  
Viable Option ◽  
Bulk Fluid ◽  
Fuel Channel ◽  
Uranium Carbide

SuperCritical Water-cooled nuclear Reactor (SCWR) options are one of the six reactor options identified in Generation IV International Forum (GIF). In these reactors the light-water coolant is pressurized to supercritical pressures (up to approximately 25 MPa). This allows the coolant to remain as a single-phase fluid even under supercritical temperatures (up to approximately 625°C). SCW Nuclear Power Plants (NPPs) are of such great interest, because their operating conditions allow for a significant increase in thermal efficiency when compared to that of modern conventional water-cooled NPPs. Direct-cycle SCW NPPs do not require the use of steam generators, steam dryers, etc. allowing for a simplified NPP design. This paper shows that new nuclear fuels such as Uranium Carbide (UC) and Uranium Dicarbide (UC2) are viable option for the SCWRs. It is believed they have great potential due to their higher thermal conductivity and corresponding to that lower fuel centerline temperature compared to those of conventional nuclear fuels such as uranium dioxide, thoria and MOX. Two conditions that must be met are: 1) keep the fuel centreline temperature below 1850°C (industry accepted limit), and 2) keep the sheath temperature below 850°C (design limit). These conditions ensure that SCWRs will operate efficiently and safely. It has been determined that Inconel-600 is a viable option for a sheath material. A generic SCWR fuel channel was considered with a 43-element bundle. Therefore, bulk-fluid, sheath and fuel centreline and HTC profiles were calculated along the heated length of a fuel channel.

Thermal-Design Options for Pressure-Channel SCWRS With Cogeneration of Hydrogen

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
Maria Naidin ◽  
Sarah Mokry ◽  
Farina Baig ◽  
Yevgeniy Gospodinov ◽  
Udo Zirn ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Electrical Energy ◽  
Thermal Design ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Development Costs ◽  
Pressure Channel ◽  
Steam Cycle

Currently there are a number of Generation IV supercritical water-cooled nuclear reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are (1) to increase the gross thermal efficiency of current nuclear power plants (NPPs) from 33–35% to approximately 45–50% and (2) to decrease the capital and operational costs and, in doing so, decrease electrical-energy costs (approximately US$ 1000∕kW or even less). SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25MPa and outlet temperatures of up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical cogeneration of hydrogen through thermochemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of a SCW NPP and to increase its reliability, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature supercritical (SC) fossil power plants (including their SC turbine technology). On this basis, several conceptual steam-cycle arrangements of pressure-channel SCWRs, their corresponding T‐s diagrams and steam-cycle thermal efficiencies are presented in this paper together with major parameters of the copper-chlorine cycle for the cogeneration of hydrogen. Also, bulk-fluid temperature and thermophysical properties profiles were calculated for a nonuniform cosine axial heat-flux distribution along a generic SCWR fuel channel, for reference purposes.

Conceptual Thermal-Design Options for Pressure-Tube SCWRs With Thermochemical Co-Generation of Hydrogen

2008 ◽  
Author(s):  
Sarah Mokry ◽  
Maria Naidin ◽  
Farina Baig ◽  
Yevgeniy Gospodinov ◽  
Udo Zirn ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Electrical Energy ◽  
Thermal Design ◽  
Pressure Tube ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Development Costs ◽  
Steam Cycle

Currently there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are: 1) To increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 33–35% to approximately 45–50%, and 2) To decrease the capital and operational costs and, in doing so, decrease electrical-energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25 MPa and outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of a SCW NPP and to increase its reliability, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil power plants (including their SC turbine technology). The state-of-the-art SC steam cycles in fossil power plants are designed with a single-steam reheat and regenerative feedwater heating and reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43% on a Higher Heating Value Basis). It would be beneficial if SCWRs could involve a regenerative feedwater heating and nuclear steam reheat to be able to adapt the current SC turbine technology and to achieve similar high thermal efficiencies as the advanced fossil steam cycles. The nuclear steam reheat is easier to implement inside pressure-tube or pressure-channel reactors compared to pressure-vessel reactors. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing concepts of the pressure-tube SCWRs. Therefore, no-reheat, single-reheat, and double-reheat cycles of future SCW NPPs were analyzed in terms of their thermal efficiencies. On this basis, several conceptual steam-cycle arrangements of pressure-tube SCWRs, their corresponding T-s diagrams and steam-cycle thermal efficiencies are presented in this paper together with major parameters of the copper-chlorine cycle for the co-generation of hydrogen. Also, bulk-fluid temperature and thermophysical properties profiles were calculated for a non-uniform cosine Axial Heat-Flux Distribution (AHFD) along a generic SCWR fuel channel, for reference purposes.

Study of the Influence of Microstructure and Intergranular Carbides on the Oxidation Behavior of a Nickel Base Alloy 690 TT in Supercritical Water Nuclear Reactor Conditions

2020 ◽  
Author(s):  
Alberto Sáez-Maderuelo ◽  
María Luisa Ruiz-Lorenzo ◽  
Francisco Javier Perosanz ◽  
Patricie Halodová ◽  
Jan Prochazka ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Oxidation Behavior ◽  
Nuclear Reactor ◽  
Base Alloy ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Alloy 690 ◽  
Reactor Operating

Abstract Alloy 690, which was designed as a replacement for the Alloy 600, is widely used in the nuclear industry due to its optimum behavior to stress corrosion cracking (SCC) under nuclear reactor operating conditions. Because of this superior resistance, alloy 690 has been proposed as a candidate structural material for the Supercritical Water Reactor (SCWR), which is one of the designs of the next generation of nuclear power plants (Gen IV). In spite of this, striking results were found [1] when alloy 690 was tested without intergranular carbides. These results showed that, contrary to expectations, the crack growth rate is lower in samples without intergranular carbides than in samples with intergranular carbides. Therefore, the role of the carbides in the corrosion behavior of Alloy 690 is not yet well understood. Considering these observations, the aim of this work is to study the effect of intergranular carbides in the oxidation behavior (as a preliminary stage of degenerative processes SCC) of Alloy 690 in supercritical water (SCW) at two temperatures: 400 °C and 500 °C and 25 MPa. Oxide layers of selected specimens were studied by different techniques like Scanning Electron Microscope (SEM) and Auger Electron Spectroscopy (AES).

Outline of current analysis procedure for CANDU 600 MW reactor buildings

1985 ◽  
Vol 12 (4) ◽  
pp. 796-804
Author(s):  
J. C. Mamet ◽  
O. Moselhi
Keyword(s):  
Prestressed Concrete ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Operating Conditions ◽  
System Control ◽  
Current Analysis ◽  
Analysis Process ◽  
Reactor Building

Reactor buildings of 600 MW CANDU nuclear power plants consist of a prestressed concrete containment structure, cylindrical in shape with a double spherical dome, and of a reinforced concrete internal structure with heavy walls and slabs that support the nuclear reactor, the primary heat transport system, control and safety mechanisms, etc. Both structures are supported on a common circular slab.In this paper, an outline of the static and seismic response analyses performed for these buildings is presented. Several computer models and codes are used and advantage is taken of the symmetry of revolution of part of the structure.By combining the results produced by the various models and accounting for discontinuities caused by openings, etc., a complete picture of the forces, displacements, or accelerations existing in the reactor building under operating conditions and during postulated accidents or seismic events may be drawn.This process has been partly automated by the development of relevant software. A flow chart of the whole analysis process is given. Key words: nuclear power plants, reactor building, containment, analysis, static, seismic, finite elements.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

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 Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

scholarly journals Structural Modeling and Dynamic Analysis of a Nuclear Reactor Building

2020 ◽  
Author(s):  
Evrim Oyguc ◽  
Abdul Hayır ◽  
Resat Oyguc
Keyword(s):  
Nonlinear Dynamic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Energy Sources ◽  
Finite Element Program ◽  
Design Spectrum ◽  
Reactor Building ◽  
Set Up

Increasing energy demand urge the developing countries to consider different types of energy sources. Owing the fact that the energy production capacity of renewable energy sources is lower than a nuclear power plant, developed countries like US, France, Japan, Russia and China lead to construct nuclear power plants. These countries compensate 80% of their energy need from nuclear power plants. Further, they periodically conduct tests in order to assess the safety of the existing nuclear power plants by applying impact type loads to the structures. In this study, a sample third-generation nuclear reactor building has been selected to assess its seismic behavior and to observe the crack propagations of the prestressed outer containment. First, a 3D model has been set up using ABAQUS finite element program. Afterwards, modal analysis is conducted to determine the mode shapes. Nonlinear dynamic time history analyses are then followed using an artificial strong ground motion which is compatible with the mean design spectrum of the previously selected ground motions that are scaled to Eurocode 8 Soil type B design spectrum. Results of the conducted nonlinear dynamic analyses are considered in terms of stress distributions and crack propagations.

scholarly journals Assessment of water hammer effects on boiling water nuclear reactor core dynamics

2007 ◽  
Vol 22 (1) ◽  
pp. 18-33 ◽  
Author(s):  
Anis Bousbia-Salah
Keyword(s):  
Pressure Wave ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Water Hammer ◽  
Pressure Wave Amplitude ◽  
Sensitivity Studies ◽  
Rapid Transients ◽  
Complex Phenomena

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

scholarly journals PRELIMINARY DESIGN OF RDE FEEDWATER PUMP IMPELLER

2018 ◽  
Vol 20 (1) ◽  
pp. 1 ◽  
Author(s):  
Sri Sudadiyo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Performance Characteristics ◽  
Preliminary Design ◽  
Triangle Diagram ◽  
Pump Impeller ◽  
Impeller Blade ◽  
Feedwater Pump

Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Eksperimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MWth, and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm.Keywords: Blade, impeller, pump, RDEDESAIN AWAL IMPELER POMPA AIR UMPAN RDE. Saat ini, pompa digunakan secara luas dalam pembangkit tenaga termal termasuk pembangkit listrik tenaga nuklir. Reaktor Daya Eksperimental (RDE) merupakan konsep reaktor nuklir yang diusulkan untuk tipe PLTN di Indonesia. RDE ini memiliki daya termal 10 MWth, dan menggunakan pompa air umpan dalam siklus uapnya. Kinerja pompa air umpan bergantung pada ukuran dan geometri model impeller, seperti jumlah sudu dan sudut sudu. Tujuan dari penelitian ini adalah untuk membuat rancangan awal impeller pompa air umpan untuk RDE dan untuk mensimulasikan karakteristik kinerjanya. Kode Fortran digunakan sebagai bantuan dalam penghitungan data untuk untuk mengkalkulasi secara cepat bentuk sudu impeller pompa air umpan, terutama pada kasus RDE. Analisis perhitungan dipecahkan menggunakan korelasi empiris yang terkait dengan ukuran dan geometri model impeller pompa, sedangkan analisis karakteristik kinerja dilakukan berdasarkan diagram segitiga kecepatan. Pengaruh bocoran, melalui impeler akibat celah yang diperlukan antara impeller pompa air umpan dan saluran volute, juga dipertimbangkan. Perbandingan antara pompa air umpan HTR-10 dan RDE menunjukkan kemiripan dalam garis tren bentuk kurva. Kurva karakteristik ini akan sangat berguna untuk perkiraan nilai kinerja pompa air umpan RDE. Desain awal pompa air umpan memberikan ukuran dan geometri model sudu impeller dengan 5-sudu, sudut masuk 14,5 derajat, sudut keluar 25 derajat, diameter dalam 81,3 mm, diameter luar 275,2 mm, ketebalan 4,7 mm, dan tinggi 14,1 mm. Selain itu, nilai optimal karakteristik kinerja diperoleh ketika kapasitas aliran 4,8 kg/s, head fluida 29,1 m, tenaga mekanik poros 2,64 kW, dan efisiensi 52 % pada kecepatan putaran 1750 rpm.Kata kunci: Sudu, impeler, pompa, RDE

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