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.

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.

scholarly journals Design Option of Heat Exchanger for the Next Generation Nuclear Plant

2008 ◽  
Author(s):  
Chang H. Oh ◽  
Eung S. Kim
Keyword(s):  
Heat Exchanger ◽  
System Integration ◽  
Power Conversion ◽  
Nuclear Plant ◽  
Design And Technology ◽  
Production Plant ◽  
Next Generation ◽  
Tube Heat Exchanger ◽  
Intermediate Heat Exchanger ◽  
Hydrogen Plant

The Next Generation Nuclear Plant (NGNP), a very High temperature Gas-Cooled Reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit (PCU) for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger is very important. This paper will include analysis of one stage versus two stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical coil heat exchanger, and shell/tube heat exchanger.

scholarly journals Design Option of Heat Exchanger for the Next Generation Nuclear Plant

2009 ◽  
Vol 132 (3) ◽  
Author(s):  
Chang H. Oh ◽  
Eung S. Kim ◽  
Mike Patterson
Keyword(s):  
Heat Exchanger ◽  
System Integration ◽  
Power Conversion ◽  
Nuclear Plant ◽  
Design And Technology ◽  
Production Plant ◽  
Next Generation ◽  
Tube Heat Exchanger ◽  
Intermediate Heat Exchanger ◽  
Hydrogen Plant

The next generation nuclear plant (NGNP), a very high temperature gas-cooled reactor (VHTR) concept, will provide the first demonstration of a closed-loop Brayton cycle at a commercial scale, producing a few hundred megawatts of power in the form of electricity and hydrogen. The power conversion unit for the NGNP will take advantage of the significantly higher reactor outlet temperatures of the VHTRs to provide higher efficiencies than can be achieved with the current generation of light water reactors. Besides demonstrating a system design that can be used directly for subsequent commercial deployment, the NGNP will demonstrate key technology elements that can be used in subsequent advanced power conversion systems for other Generation IV reactors. In anticipation of the design, development, and procurement of an advanced power conversion system for the NGNP, the system integration of the NGNP and hydrogen plant was initiated to identify the important design and technology options that must be considered in evaluating the performance of the proposed NGNP. As part of the system integration of the VHTRs and the hydrogen production plant, the intermediate heat exchanger is used to transfer the process heat from VHTRs to the hydrogen plant. Therefore, the design and configuration of the intermediate heat exchanger are very important. This paper describes analyses of one stage versus two-stage heat exchanger design configurations and simple stress analyses of a printed circuit heat exchanger (PCHE), helical-coil heat exchanger, and shell-and-tube heat exchanger.

scholarly journals Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant

2008 ◽  
Author(s):  
E. A. Harvego ◽  
M. G. McKellar ◽  
M. S. Sohal ◽  
J. E. O’Brien ◽  
J. S. Herring
Keyword(s):  
High Temperature ◽  
Hydrogen Production ◽  
Economic Analysis ◽  
Nuclear Reactor ◽  
Primary System ◽  
Production Plant ◽  
Operating Pressure ◽  
System Pressure ◽  
Reference Design ◽  
High Temperature Electrolysis

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current (AC) to direct-current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

Dynamic Modeling of Nuclear Hydrogen Production Using Methane Steam Reforming

2021 ◽  
Author(s):  
Junyi Li ◽  
Zhe Dong ◽  
Bowen Li
Keyword(s):  
Hydrogen Production ◽  
Dynamic Modeling ◽  
Steam Reforming ◽  
Nuclear Power ◽  
Thermal Power ◽  
Open Loop ◽  
Membrane Reactors ◽  
Methane Steam Reforming ◽  
Production Plant ◽  
Methane Steam

Abstract Methane steam reforming (MSR) technology is one of the promising methods of hydrogen production and already available at an industrial scale, in which steam is added to methane to generate hydrogen. MSR carries out at a temperature of 500°C when catalysts and Pd-based membrane reactors are used. The nuclear steam supply system (NSSS) of a modular high-temperature gas-cooled reactor (MHTGR) can provide high-quality steam of around 570°C, which is an excellent heat source for MSR. MHTGR is a typical small modular reactor (SMR), of which the coolant is helium, and the moderator and structural material are graphite. The number of the MHTGR can be decided based on the thermal power required for MSR and electricity generation. In this paper, a six-modular MHTGR nuclear power plant with 1500MW thermal power coupled with the MSR process is designed. The hydrogen production rate is 9.72 tons per hour. The dynamic modeling is based on conservation laws of mass and energy. To examine the dynamic characteristics of the nuclear hydrogen production plant, open-loop responses of the model under different disturbances are presented.

Risk Assessment of Hydrogen Production System by Dimethyl Ether Steam Reforming With the Use of Nuclear Power

2009 ◽  
Author(s):  
Kazuo Koguchi ◽  
Shigeo Kasai ◽  
Makoto Takahashi ◽  
Toshio Wakabayashi
Keyword(s):  
Risk Assessment ◽  
Heat Exchanger ◽  
Hydrogen Production ◽  
Dimethyl Ether ◽  
Steam Reforming ◽  
Nuclear Power ◽  
Average Frequency ◽  
Event Tree ◽  
Event Tree Analysis ◽  
Tree Analysis

Hydrogen is regarded as a clean fuel because it does not pollute when burned with air. In the case of commercial use, there is a need to research how to produce hydrogen more efficiency and large scale. Although there are some methods of hydrogen production, it can be considered that the heat of the nuclear reactor is promising method. In the recent studies on the hydrogen production with nuclear power, there has focused on the technical issues. Therefore, the object in this paper is to perform the risk assessment for a system of hydrogen production plant by Dimethyl Ether (DME) steam reforming with the use of nuclear power. First, one of the suitable systems with the DME steam reforming plant was selected PWR. A FMEA (Failure Mode and Effects analysis) was performed to identify initiating events. After identifying initiating events, event tree analysis (ETA) was performed to quantify the average frequency of an accident at this system. The result of the PSA, the safety of DME steam reforming plant with nuclear power depends on a rupture of reformer and heat exchanger between hydrogen and DME by the result of FMEA. Event tree analysis shows that the average frequency of hydrogen or DME explosion is 7.7×10−7 year−1 in the case of the rupture of the reformer and 1.9×10−8 year−1 in the case of the break of the heat exchanger.

Investigations on RELAP5-3D to RELAP5-3D Coupling Methodology by PVMEXEC

2016 ◽  
Author(s):  
Valeria Parrinello ◽  
Marco Lanfredini ◽  
Alessandro Petruzzi ◽  
Marco Cherubini
Keyword(s):  
Steady State ◽  
Heat Exchanger ◽  
Nuclear Power ◽  
Power Plants ◽  
Coupled Model ◽  
Nuclear System ◽  
Tube Heat Exchanger ◽  
Parallel Virtual Machine ◽  
Solution Algorithms ◽  
Accident Scenarios

In the framework of a BEPU (Best Estimate Plus Uncertainty) approach within the licensing process of a nuclear power plant, the need to extend the resources of nuclear system thermal-hydraulics codes, such as RELAP5-3D, arises to allow more detailed simulations of the complex 3D reality of Nuclear Power Plants (NPPs), either under normal steady-state or during various accident scenarios. Currently, it is not possible to achieve the same degree of detail for a whole nuclear system when it is simulated with RELAP5-3D and this is due to the inherent limitations in the number of components and volumes to be used for the analysis. For this reason, it is of extreme interest the use of tools for codes coupling that enable the use of different codes for the simulation of different portions of a system in a unified analysis. In this paper the attention will be focused on the decomposition of the thermal-hydraulic domain of a system into subsystems to be simulated by different instances of the same code (e.g. RELAP5-3D) coupled together by means of PVMEXEC program and parallel virtual machine (PVM) technology. Explicit and semi-implicit solution algorithms were used for the analyses. Among the analyzed cases, the following will be discussed in detail with the aim to provide additional guidelines for the use of the PVMEXEC tool: (i) the Edward’s pipe blowdown test, (ii) a simplified countercurrent heat exchanger, (iii) different hydraulics and heat structure coupling schemes for a shell-tube heat exchanger and (iv) a three-task coupled model of a simplified BWR model.

Sign in / Sign up

Export Citation Format

Share Document