Power Distribution and Fuel Centerline Temperature in a Pressure-Tube Supercritical Water-Cooled Reactor (PT SCWR)

2012 ◽  
Author(s):  
W. Peiman ◽  
Eu. Saltanov ◽  
L. Grande ◽  
I. Pioro ◽  
B. Rouben ◽  
...  
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Pressure Tube ◽  
Thermochemical Processes ◽  
Flux Profile ◽  
Centerline Temperature

SuperCritical Water-cooled nuclear Reactor (SCWR) designs are one of six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic pressure-tube SCWR consists of distributed fuel channels with coolant inlet and outlet temperatures of 350 and 625°C at 25 MPa, respectively. Such reactor coolant outlet conditions allow for high thermal efficiencies of SCW Nuclear Power Plant (NPP) of about 45–50%. In addition to high thermal efficiencies, SCWR designs provide the means for co-generation of hydrogen through thermochemical processes such as the Cu–Cl cycle. The main objective of this paper is to determine the power distribution inside the core of an SCWR by using a lattice code - DRAGON and a diffusion code - DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermal-hydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature for UO2 and UC nuclear fuels. Results of an analysis showed that the fuel centerline temperature of UC was significantly lower than that of UO2. This paper also studies effects of energy groups on multi-group diffusion calculations and proposes nine energy groups for further neutronic studies related to SCWRs.

Power Distribution in a Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)

2013 ◽  
Author(s):  
W. Peiman ◽  
I. Pioro ◽  
K. Gabriel
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Power Distribution ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Outlet Temperature ◽  
Fuel Bundle ◽  
Flux Profile ◽  
Pressure Channel ◽  
Centerline Temperature

SuperCritical Water-cooled nuclear Reactor (SCWR) is one of the six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic 1200-MWel pressure-channel SCWR operates at a pressure of 25 MPa with coolant inlet and outlet temperatures of 350°C and 625°C, respectively. High coolant outlet temperature allows for high thermal efficiencies within the range of 45–50%. On the other hand, the high operating temperature of SCWR in turn results in high fuel centerline and sheath temperatures. Hence, it is necessary to determine a power distribution inside a core of a reactor in order to ensure that a fuel and a fuel-bundle design comply with their corresponding temperature limits. The main objective of this paper is to determine a power distribution inside the core of a generic SCWR by using a lattice code DRAGON and a diffusion code DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermalhydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature of UO2 and UC nuclear fuels and a sheath temperature of a new fuel-bundle design. Results of this analysis showed that the fuel centerline temperature of the UC fuel was significantly lower than that of the UO2. This paper also proposes four energy groups for further neutronic studies related to SCWRs.

Sensitivity Analysis of Fuel Centerline Temperature in SCWRs

2012 ◽  
Author(s):  
Ayman Abdalla ◽  
Wargha Peiman ◽  
Igor Pioro ◽  
Kamiel Gabriel
Keyword(s):  
Sensitivity Analysis ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Beryllium Oxide ◽  
Bulk Fluid ◽  
Energy Demands ◽  
Fuel Bundle ◽  
Centerline Temperature

The Generation IV International Forum (GIF) is intended to encourage the world’s leading nuclear countries to develop nuclear energy systems that can supply future energy demands. There are six nuclear reactor concepts under research and development as part of the GIF. The SuperCritical Water-cooled Reactor (SCWR) is one of these six nuclear-reactor concepts. The proposed SCWRs operate at high temperatures and pressures at around 625°C and 25 MPa, respectively. These high operating parameters are essential in order to achieve a thermal efficiency of around 45–50%, which is significantly higher than those of the current conventional nuclear power plant (NPPs) which operate at a thermal efficiency in the range of 30–35%. The SCWRs high operating temperatures and pressures impose many challenges. One of these challenges is the heating of the fuel to temperatures that can cause fuel melting. The main objective of this paper is to conduct a sensitivity analysis in order to determine the factors mostly affecting the fuel centerline temperature. In this process, different thermal conductivity fuels such as Mixed Oxide Fuel (MOX), Uranium Oxide + Beryllium Oxide (UO2+BeO), and Uranium Carbide (UC) will be examined enclosed in a 54-element fuel bundle. Other factors such as the sheath material and the Heat Transfer Coefficient (HTC) might also affect the fuel centerline temperature. The HTC will be increased by a multiple of two and the fuel centerline temperature will be calculated. Therefore, in this paper the HTC, bulk-fluid, sheath and fuel centerline temperature will be calculated along the heated length of a generic SCWR fuel channel at an average channel thermal power of 8.5 MWth.

General Layouts of Supercritical-Water NPPs

2010 ◽  
Author(s):  
I. Pioro ◽  
M. Naidin ◽  
S. Mokry ◽  
Eu. Saltanov ◽  
W. Peiman ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Electrical Energy ◽  
Development Costs ◽  
Flow Circuit ◽  
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) Increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs and, in doing so, decrease electrical-energy costs. SuperCritical Water (SCW) NPPs will have much higher operating parameters compared to current NPPs (i.e., steam pressures of about 25 MPa and steam 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 an SCW NPP, to increase its reliability, and to achieve similar high thermal efficiencies as the advanced fossil-fired steam cycles, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil-fired thermal power plants (including their SC-turbine technology). The state-of-the-art SC-steam cycles at fossil-fired power plants are designed with a single-steam reheat and regenerative feedwater heating. Due to this, they reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43–50% on a Higher Heating Value (HHV) basis). This paper presents several possible general layouts of SCW NPPs, which are based on a regenerative-steam cycle. To increase the thermal efficiency and to match current SC-turbine parameters, the cycle also includes a single steam-reheat stage. Since these options include a nuclear steam-reheat stage, the SCWR is based on a pressure-tube design.

Thermal Aspects of Using Uranium Mononitride Fuel in a SuperCritical Water-Cooled Reactor at Maximum Heat Flux Conditions

2010 ◽  
Author(s):  
Ashley Milner ◽  
Caleb Pascoe ◽  
Hemal Patel ◽  
Wargha Peiman ◽  
Graham Richards ◽  
...  
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Thermal Efficiency ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Maximum Heat ◽  
Light Water ◽  
Maximum Heat Flux ◽  
Centerline Temperature ◽  
Uranium Mononitride

Generation IV nuclear reactor technology is increasing in popularity worldwide. One of the six Generation-IV-reactor types are SuperCritical Water-cooled Reactors (SCWRs). The main objective of SCWRs is to increase substantially thermal efficiency of Nuclear Power Plants (NPPs) and thus, to reduce electricity costs. This reactor type is developed from concepts of both Light Water Reactors (LWRs) and supercritical fossil-fired steam generators. The SCWR is similar to a LWR, but operates at a higher pressure and temperature. The coolant used in a SCWR is light water, which has supercritical pressures and temperatures during operation. Typical light water operating parameters for SCWRs are a pressure of 25 MPa, an inlet temperature of 280–350°C, and an outlet temperature up to 625°C. Currently, NPPs have thermal efficiency about of 30–35%, whereas SCW NPPs will operate with thermal efficiencies of 45–50%. Furthermore, since SCWRs have significantly higher water parameters than current water-cooled reactors, they are able to support co-generation of hydrogen. Studies conducted on fuel-channel options for SCWRs have shown that using uranium dioxide (UO2) as a fuel at supercritical-water conditions might be questionable. The industry accepted limit for the fuel centerline temperature is 1850°C and using UO2 would exceed this limit at certain conditions. Because of this problem, there have been other fuel options considered with a higher thermal conductivity. A generic 43-element bundle for an SCWR, using uranium mononitride (UN) as the fuel, is discussed in this paper. The material for the sheath is Inconel-600, because it has a high resistance to corrosion and can adhere to the maximum sheath-temperature design limit of 850°C. For the purpose of this paper, the bundle will be analyzed at its maximum heat flux. This will verify if the fuel centerline temperature does not exceed 1850°C and that the sheath temperature remains below the limit of 850°C.

scholarly journals Thermo-hydraulic Simulation of AP1000 Nuclear Reactor Fuel Assembly

2021 ◽  
Vol 31 (1) ◽  
pp. 60-71
Author(s):  
Leonardo Acosta Martínez ◽  
Carlos Rafael García Hernández ◽  
Jesus Rosales García ◽  
Annie Ortiz Puentes
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Capital Investment ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Geometric Model ◽  
Modular Construction ◽  
Construction Time ◽  
Hydraulic Simulation ◽  
In Series

One of the challenges of future nuclear power is the development of safer and more efficient nuclear reactor designs. The AP1000 reactor based on the PWR concept of generation III + has several advantages, which can be summarized as: a modular construction, which facilitates its manufacture in series reducing the total construction time, simplification of the different systems, reduction of the initial capital investment and improvement of safety through the implementation of passive emergency systems. Being a novel design it is important to study the thermohydraulic behavior of the core applying the most modern tools. To determine the thermohydraulic behavior of a typical AP1000 fuel assembly, a computational model based on CFD was developed. A coupled neutronic-thermohydraulic calculation was performed, allowing to obtain the axial power distribution in the typical fuel assembly. The geometric model built used the certified dimensions for this type of installation that appear in the corresponding manuals. The thermohydraulic study used the CFD-based program ANSYS-CFX, considering an eighth of the fuel assembly. The neutronic calculation was performed with the program MCNPX version 2.6e. The work shows the results that illustrate the behavior of the temperature and the heat transfer in different zones of the fuel assembly. The results obtained agree with the data reported in the literature, which allowed the verification of the consistency of the developed model.

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

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).

Supercritical Water-Cooled Nuclear Reactors (SCWRs): Current and Future Concepts — Steam Cycle Options

2008 ◽  
Author(s):  
R. B. Duffey ◽  
I. Pioro ◽  
X. Zhou ◽  
U. Zirn ◽  
S. Kuran ◽  
...  
Keyword(s):  
Steam Turbine ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Electrical Energy ◽  
Reactor Outlet ◽  
Cycle Efficiency ◽  
Direct Cycle

One of the six Generation IV nuclear reactor concepts is a SuperCritical Water-cooled nuclear Reactor (SCWR), which is currently under development. The main objectives for developing and utilizing SCWRs are to increase the thermal efficiency of Nuclear Power Plants (NPPs), to decrease electrical energy costs, and possibility for co-generation, including hydrogen generation. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing pressure-tube SCWR concepts. The targeted steam parameters at the reactor outlet are approximately 25 MPa and 625°C. This paper presents a survey on modern SuperCritical (SC) steam turbine technology and a study on potential steam cycles for the SCWR plants. The survey reveals that by the time the Gen IV SCWRs are market-ready, the required steam turbine technology will be well proven. Three potential steam cycles in an SCWR plant are presented: a dual-cycle with steam reheat, a direct cycle with steam reheat, and a direct cycle with a Moisture Separator and Reheater (MSR). System thermal-performance simulations have been performed to determine the overall cycle efficiency of the proposed cycles. The results show that the direct cycle with steam reheat has the highest efficiency. The direct cycle with MSR is an alternative option, which will simplify the reactor design at the penalty of a slightly lower cycle efficiency.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

scholarly journals Nuclear Hydrogen Production: Modeling and Preliminary Optimization of a Helical Tube Heat Exchanger

Energies ◽  
2021 ◽  
Vol 14 (11) ◽  
pp. 3113
Author(s):  
Lorenzo Bolfo ◽  
Francesco Devia ◽  
Guglielmo Lomonaco
Keyword(s):  
Heat Exchanger ◽  
Hydrogen Production ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Transport Sector ◽  
Production Plant ◽  
Primary Coolant ◽  
Tube Heat Exchanger ◽  
Cfd Models ◽  
Thermochemical Processes

Hydrogen production is a topical issue in an energy scenario where decarbonization is a priority, especially with reference to the transport sector that has a great weight on global emissions. Starting from this consideration, GIF (Generation-IV International Forum) investigated the possibility to produce hydrogen by nuclear energy. The “classic” strategy is based on the use of nuclear as energy source for the electrolysis; but on the medium-long term, a more sustainable and smart approach could be founded on the use of thermochemical processes (e.g., I-S) that require a direct coupling of a chemical plant to a nuclear reactor. In order to develop this strategy, it is mandatory to design and optimize all the key components involved in this complex plant. In this study, we developed the 3D-CAD and CFD models of the intermediate heat exchanger (IHX) installed in the Japanese HTTR nuclear power plant. This component is extremely important for both the safety of the two plants and the stability of the whole hydrogen production plant. Initially, our model (developed by AutoCAD 3D and implemented in Star CCM+) was validated on the basis of experimental data available in literature; then, an initial optimization of the IHX testing innovative materials, such as Alloy 617 and ODS–MA754, and a different primary coolant (supercritical CO2) was performed.

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