FUTURE OF NUCLEAR REACTORS TECHNOLOGY (IV Generation Nuclear Energy Systems)

As the global population continues to rise, the demand for energy increases exponentially, providing immense benefits to society by democratizing access to power, improving living standards, and contributing to better health outcomes and longer life expectancy. However, the widespread use of energy also presents significant environmental challenges, particularly concerning global climate change. Therefore, humanity needs to prioritize clean, safe, and cost-effective energy sources. Alongside technologies like Carbon Capture and Storage (CCS), photovoltaic (PV) systems, and wind power, nuclear energy stands out due to its unparalleled efficiency, producing the highest energy output per unit of fuel compared to all other options.

Let me share my passion for engineering and technology as well as put on the table new technologies and business opportunities, such as the SMALL REACTORS with clear applications for small consumers and remote locations as well as off-shore.

A) EVOLUTION OF NUCLEAR REACTOR TECHNOLOGY UP TO NOW.

Beginning in 2000, the countries constituting the GIF began meeting to discuss the research necessary to support next-generation reactors. From those initial meetings, a technology roadmap to guide the Generation IV effort was begun. The organization and execution of the roadmap became the responsibility of a Roadmap Integration Team that is advised by the Subcommittee on Generation IV Technology Planning of the U.S. Department of Energy’s Nuclear Energy Research Advisory Committee (NERAC). Road mapping is a methodology used to define and manage the planning and execution of Large-scale R&D Efforts. Each GIF country, including Canada and the USA, will focus on specific systems and subsystems of the R&D activities of the new IV Generation Nuclear Reactors.

The GOALs of the new IV Generation of Nuclear Reactors could be summarized as follows:

B) FUEL CYCLES AND SUSTAINABILITY

The following are the SIX (6) technologies selected by GIF as the most promising to develop the new Generation IV Reactors:

1. VHTR- Very-High-Temperature Reactor System.
2. GFR - Gas-Cooled Fast Reactor System
3. SFR - Sodium-cooled Fast Reactory System
4. LFR - Lead-cooled Fast tReactor System
5. MSR - Molten Salt Reactor System
6. SCWR - Supercritical-Water-cooled Reactor System
7. VHTR- Very-High-Temperature Reactor System.

B.1) VHTR - Very High Temperature Reactor (option modular)

The VHTR is a further step in the evolutionary design of High Temperature (HTR) Reactors. VHTR is helium gas-cooled, graphited moderated, thermal neutron spectrum reactor with a core outlet temperature higher than 900 Celsius and a goal temperature of 1000 Celsius, enough to support high temperature process such as the production of hydrogen by thermomechanical processes.

This Reactor is very useful for the cogeneration of electricity and hydrogen, as well as other industrial process such as: heat applications, iron industries, Oil, thermochemical, etc. In this design the CO2 emission are very reduced. At first, a once-through low-enriched uranium (<20% 235U) fuel cycle will be adopted, but a closed fuel cycle will be assessed, as well as potential symbiotic fuel cycles with other types of reactors (especially light water reactors (LWRs)) for waste reduction purposes. The system is expected to be available for commercial deployment by 2020s.

The technical basis for VHTR is the TRI-ISOtropic (TRISO)-coated particle fuel. The VHTR has potential for inherent safety, high thermal efficiency, process heat application capability, low operation and maintenance costs, and modular construction

KEY CHARACTERISTICS:

1. Helium Gas Cooling:

• Helium is used as the coolant because it is chemically inert, non-reactive with fuel or structural materials, and can withstand high temperatures without degrading.
• The helium gas circulates through the reactor core, absorbing heat from the nuclear fission process. Helium’s high heat capacity and thermal conductivity allow it to efficiently transport heat out of the reactor core to downstream systems.
• Graphite Moderation:

• The reactor uses graphite as a moderator to slow down fast neutrons produced during fission to thermal energy levels.
• Graphite helps sustain a controlled chain reaction by increasing the likelihood of neutrons interacting with the nuclear fuel, maintaining reactor efficiency.

1. Thermal Neutron Spectrum:

• In a thermal neutron spectrum reactor, neutrons are slowed down to low (thermal) energies, making the fission process more efficient with certain types of nuclear fuel, particularly uranium-235 or low-enriched uranium.
• This allows for a steady and controlled energy output while minimizing fuel waste.

1. Core Outlet Temperature (900-1000°C):

• The core outlet temperature is designed to be very high, around 900-1000°C, which is significantly higher than typical nuclear reactors.
• This high temperature enhances the thermal efficiency of the reactor, allowing more energy to be converted into electricity or other useful forms of power.
• The high temperature also supports industrial applications, like hydrogen production, which requires substantial heat.

B.2) GFR - Gas-cooled Fast Reactor

The high-temperature helium-cooled fast spectrum reactor, operating with a closed fuel cycle, is designed with a core outlet temperature around 850°C. This system leverages the benefits of fast spectrum reactors, offering long-term sustainability of uranium resources and minimizing waste through multiple fuel reprocessing and the fission of long-lived actinides. It also incorporates high-temperature technology, enabling high thermal cycle efficiency and industrial applications like hydrogen production. The system requires the development of durable refractory fuel elements and a robust safety architecture. Using dense fuels such as carbide or nitride enhances its performance in plutonium breeding and minor actinide burning

KEY CHARACTERISTICS:

Key Features of GFR:

1. Fast Spectrum Operation: The GFR operates in the fast neutron spectrum, allowing for efficient use of uranium resources and enhancing the ability to transmute long-lived actinides. This contributes to minimizing nuclear waste.
2. Closed Fuel Cycle:The system employs a closed fuel cycle, enabling the recycling and reprocessing of fuel. This approach extends the sustainability of nuclear fuel resources by enabling the reuse of fissile and fertile materials.
3. High-Temperature Operation:With core outlet temperatures reaching 850°C, the GFR achieves high thermal efficiency. This high-temperature operation also allows for industrial applications, such as hydrogen production, making it a key player in clean energy technologies.
4. Fuel Type:he reactor uses dense fuels like carbide or nitride, which are well-suited for fast reactors. These fuels provide superior performance in plutonium breeding and minor actinide burning, crucial for resource sustainability and waste reduction.
5. Safety Architecture:The GFR requires a robust safety design, given the challenges associated with high-temperature and high-energy-density systems. This includes the development of refractory fuel elements that can withstand extreme conditions within the reactor core.
6. Applications: The combination of fast spectrum and high-temperature operation positions the GFR as a key technology for both power generation and hydrogen production, contributing to a sustainable and carbon-free energy future.
7. Reactor Parameters/Reference Value::

B.3) SFR - Sodium-cooled Fast Reactor

The Sodium-Cooled Fast Reactor (SFR) utilizes liquid sodium as its coolant and operates with a closed fuel cycle, supporting both fuel breeding and actinide management. There are two primary fuel recycling methods under consideration: advanced aqueous processing, where MOX (Mixed Oxide Fuel) is preferred, and pyrometallurgical processing, which favors mixed metal alloy fuels. Given the extensive operational experience with sodium-cooled reactors across several countries, the deployment of SFR systems is targeted for the 2020s. Liquid sodium as a coolant offers high power density, a low coolant volume fraction, and low-pressure operation. Although sodium’s oxygen-free environment prevents corrosion, its reactivity with air and water necessitates a sealed coolant system.

SFR plant sizes range from small modular reactors producing 50–300 MWe to larger reactors of up to 1500 MWe. With an outlet temperature of 500–550°C, these reactors can leverage materials developed and proven in earlier fast reactor programs, ensuring reliability and operational efficiency.

Three possible options:

• Large size (600-1500 MWel):
• Intermediate size (300-1500 MWel):
• Small size (50-150 MWel): t with uranium plutonium minor actinide zirconium metal alloy fuel, supported by a fuel cycle based on pyrometallurgical processing in facilities integrated with the reactor.

B.4) LFR - Lead-cooled Fast Reactor System

The LDR is characterized by a fast neutron spectrum, being a close-cycle with full actinide recycling, possible in central or regioinal fuel cycle facilities, and high temperature operation at low pressure. Two optional for coolant, either lead (preferred) or Lead-bismuth eutectic (LBE). This kind of reactor could be operated as a breeder or a burner of actinides (see the notes) from spent fuel using iner matrix fuel, or a burner/breeder using Thorium matrix.

Optional size for these reactors:

• small 50e150 MWel, transportable system with a very long core life;
• medium 300e600 MWel system.
• In the long term, a large system of 1200 MWel may be envisaged

The LFR would have multiple applications including production of electricity, hydrogen, and process heat.

B.5) MSR - Molten Salt Reactor

This is a very special desing of reactor because of the inclusion of molten salt embracing the liquid fuel. MRS could be dually used as both breeder or burner. In the case of being as a burner, maybe used as efficient burner of transuranic elements from spent LWR fuel. On the other hand, it also has a breeding capability in any kind of neutron spectrum ranging from thermal (in a Thorium fuel cycle). It is a very promissing solution to reduce the radiotoxic nucleare waste.

The Molten Salt Reactor (MSR) is characterized by its innovative core, wherein fuel is dissolved in molten fluoride salt. This technology was initially investigated over five decades ago. Contemporary interest focuses on fast reactor concepts, considering them a long-term alternative to solid-fueled fast neutron reactors. The onsite fuel reprocessing unit employing pyrochemical methods facilitates the breeding of plutonium or uranium-233 from thorium.

?In the realm of nuclear reactors, uranium and plutonium are prevalently utilized as fuel due to their capability to sustain nuclear fission reactions. Thorium is also being considered as a potential fuel option owing to its abundance and the reduced production of long-lived radioactive waste it entails. Moreover, actinides are integral to the production of nuclear weapons, attributable to their explosive potential.

?Nonetheless, the handling of actinides necessitates stringent precautions, given their highly radioactive nature and the significant environmental and health hazards they pose.

MSR technology was partly developed, including two demonstration reactors, inthe 1950s and 1960s in the USA (Oak Ridge National Laboratory).

B.6) SCWR- Supercritical Water-cooled Reactors. (potential reduced investment)

Supercritical Water Reactors (SCWRs) represent a class of high-temperature, high-pressure water-cooled reactors. They operate utilizing a direct energy conversion cycle and function above the thermodynamic critical point of water, which is 374°C and 22.1 MPa. Because of the higher thermodynamic efficiency and the simplification opportunities provided by a high-temperature, single-phase coolant, these reactors promise improved economic benefits.

Currently, a range of design options are being explored: both thermal and fast neutron spectra are considered, with either pressure vessel or pressure tube configurations. Consequently, SCWRs can use either light water or heavy water as a moderator. A technology demonstration reactor generating between 30 to 150 MWel is planned to be operational by 2022.

?Unlike the existing water-cooled reactors, the coolant in SCWRs will undergo a significantly higher enthalpy increase in the core. This results in a reduced core mass flow for a given thermal power and allows the core outlet enthalpy to reach superheated conditions. In both pressure vessel and pressure tube designs, a once-through steam cycle is envisioned, thereby eliminating any need for coolant recirculation within the reactor. Similar to a Boiling Water Reactor (BWR), the superheated steam will be sent directly to the high-pressure steam turbine, and the feed water from the steam cycle will be returned to the core.

?Hence, SCWR concepts integrate operational and design insights from hundreds of water-cooled reactors with the extensive experiences from numerous fossil-fired power plants that operate with supercritical water. Unlike some other Generation IV nuclear systems, the development of SCWRs can proceed incrementally, building step-by-step on the foundation of current water-cooled reactors.

?These overall characteristics offer the potential for lower capital costs for a given electrical output and better fuel utilization, presenting a distinct economic advantage over current Light Water Reactors (LWRs).

ADVANTAGES:

• High thermal efficiency. Since SCWRs supply supercritical steam at pressures and temperatures much higher than in conventional LWRs, they will reach higher thermodynamic efficiency of the power plant (~45 % vs. ~33 % for current LWRs).
• Lower coolant mass flow rate. ?The reactor coolant flow rate of SCWR is much lower than that of BWR and PWR because the enthalpy rise in the core is much larger, which results in low capacity components of the primary system. Thus, the coolant pumps, pipes, and other supporting equipment become smaller. Moreover, the only pumps driving the coolant under normal operating conditions are the feedwater pumps and the condensate extraction pumps.
• The higher steam enthalpy allows to decrease the size of the turbine system and thus to lower the capital costs of the conventional island.
• Smaller coolant inventory. The coolant inventory is smaller, which reduces the size of the containment with pressure suppression pools.
• Since supercritical water does not undergo a phase change and exists in a single thermodynamic phase, the boiling crisis (i.e., a departure from nucleate boiling – DNB or dry out) cannot occur.
• With a direct cycle of supercritical coolant, components like steam dryers, separators, and steam generators are omitted entirely, reducing the number of major components and eliminating their associated costs.
• Many components (e.g.,, turbines) are readily developed and available from supercritical fossil-fired power plants.

C) BUDGET ESTIMATION PER KIND OF TECHNOLOGY

for more realistic VPN values, the reader shall apply the VPN techniques and interest rate to budget the CAPEX at one determined time:

NOTES:

• Actinides, are a series of chemical elements that include uranium, thorium, and plutonium, among others. These elements are crucial in nuclear technologies due to their ability to undergo nuclear reactions, releasing significant amounts of energy.?In nuclear reactors, uranium and plutonium are commonly used as fuel because they can sustain nuclear fission reactions. Thorium is also being explored as a potential fuel due to its abundance and lower production of long-lived radioactive waste. Additionally, actinides are used in the production of nuclear weapons due to their explosive potential.?However, handling actinides requires extreme caution because they are highly radioactive and can pose serious environmental and health risks.

• Critical point of water (374.14°C and 22.09 MPa). In nuclear engineering, the supercritical water reactor is considered a promising advancement for nuclear power plants because of its high thermal efficiency (~45 % vs. ~33 % for current LWRs). This concept of reactor operates at supercritical pressure (i.e. greater than 22.1 MPa) and belongs to Generation IV reactor designs.

Above the critical point, there is no constant-temperature vaporization process. At the critical point, the saturated-liquid and saturated-vapor states are identical. The temperature, pressure, and specific volume at the critical point are called the critical temperature, critical pressure, and critical volume. For water, these parameters are the following:

• Pcr = 22.09 MPa
• Tcr = 374.14 °C (or 647.3 K)
• vcr = 0.003155 m3/kg
• uf = ug = 2014 kJ/kg
• hf = hg = 2084 kJ/kg
• sf = sg =4.406 kJ/kg K

Reference:

Reactor Physics and Thermal Hydraulics:

1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading,?MA (1983).
2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering,?Springer; 4th edition, 1994, ISBN:?978-0412985317
5. Todreas Neil E., Kazimi Mujid S. Nuclear Systems Volume I: Thermal Hydraulic Fundamentals, Second Edition. CRC?Press; 2?edition, 2012, ISBN:?978-0415802871
6. Zohuri B., McDaniel P. Thermodynamics in Nuclear Power Plant Systems. Springer; 2015, ISBN:?978-3-319-13419-2
7. Moran Michal J., Shapiro Howard N. Fundamentals of Engineering Thermodynamics, Fifth Edition,?John Wiley & Sons, 2006, ISBN:?978-0-470-03037-0
8. Kleinstreuer C. Modern Fluid Dynamics. Springer, 2010,?ISBN 978-1-4020-8670-0.
9. U.S. Department of Energy, THERMODYNAMICS, HEAT TRANSFER,?AND FLUID FLOW.?DOE Fundamentals Handbook,?Volume 1, 2 and 3. June?1992.
10. US DOe Nuclear Energy Research Advisory committee. Technology Roadmap for Generation IV Nuclear Energy Systems.