As discussed in Section 9.2, evolutionary reactor designs have concentrated on improving the economics and the performance of existing nuclear reac­tors. Beyond 2030, it is anticipated that new reactor designs will address key issues such as the closure of the fuel cycle or proliferation concerns while possibly ensuring competitive economics, safety and performance.

In order for nuclear energy to remain a long-term option in the world’s energy mix, nuclear power technology development must meet sustainabil­ity goals with regard to natural resource utilization and radioactive waste management. Interest in fast neutron systems and the related fuel cycles has reappeared with the realization of their requisite role in meeting these goals.

Also in alignment with the above sustainability goals, there has been an increased interest in expanding the range of energy products provided by nuclear fission beyond electricity production to include industrial heat, hydrogen and energy for transportation. In this sense, interest in high- temperature gas-cooled reactors that may be able to realize these applica­tions in the most effective manner has also intensified.

Both fast reactors and high-temperature gas-cooled reactors, among others, are innovative reactor designs and the achievement of their full potential is conditional on continued advances in research and technology development.

A few of the topics in which significant work is presently taking place are briefly presented below.

• Core physics: neutronics and thermal-hydraulics. Improvements to safety and economics are met through stringently accurate design require­ments, which must be demonstrated with well-validated calculation tools. Presently, in the area of neutronics, the uncertainties on the nuclear data relevant to several of these innovative reactor designs in many cases are such that they negate the benefits offered by advanced mod­eling and simulation techniques. In the area of thermal-hydraulics, the extreme thermal performance expected from these systems imposes the need for the accurate determination of thermal-hydraulic parameters to relatively high resolution in order to ensure that the relevant safety criteria are met in both normal and off-normal operation.

• Fuels and structural materials. Oxide-, metallic-, carbide — and nitride — based fuels which incorporate depleted, natural and recycled uranium as well as possibly recycled plutonium, and even thorium, are considered for use in innovative reactors. Advanced structural materials with increased strength and creep resistance are sought as a means by which to improve reactor performance by allowing for greater design and safety margins, longer lifetimes and higher operating temperatures; in
parallel, this leads to improved economics through reduction of plant materials and permitted flexibility for design simplifications.

• Coolants and coolant technologies. The coolants under consideration for innovative reactors include alkali metals (e. g. sodium), heavy liquid metals (e. g. lead and lead-bismuth eutectic) and gases (e. g. helium and supercritical water). Each demonstrates distinct advantages and disad­vantages with regard to performance and safety, so much so that serious studies are devoted to all.

Several suppliers and designer organizations are working on the develop­ment of innovative nuclear reactor concepts that address the above con­cerns, although most of them are still in an early stage of development. Some of these are as follows.

• 4S. Toshiba’s 4S (super-safe, small and simple; see Fig. 9.4) is a small sodium-cooled reactor without on-site refueling in which the core has a lifetime of approximately 30 years. Being developed as a distributed energy source for multipurpose applications, the 4S offers two outputs of 30 MWt and 135 MWt, respectively selected from demand analyses. Although 4S has a fast neutron spectrum it is not a breeder reactor since blanket fuel, usually consisting of depleted uranium located around the core to absorb leakage neutrons from the core to achieve breeding of fissile materials, is not provided in its basic design. The reactor power can be controlled by the water/steam system without affecting the core operation directly. The capability of power self-adjustment makes the reactor applicable for a load-follow operation mode. The reactor is a pool type, integral type as all primary components are installed inside the reactor vessel.

Nuclear

• PRISM. The PRISM reactor is a 2200 MWe modular sodium-cooled fast reactor whose development began as a joint project between General Electric (GE) and the US Department of Energy as part of the advanced liquid-metal reactor (ALMR) program. Development has since contin­ued at GE-Hitachi, and the design today incorporates Generation IV objectives. PRISM employs passive safety and uses a modular fabrica­tion technique to expedite plant construction. PRISM uses metallic fuel for better compatibility with the coolant, inherent safety, and ease of fabrication in a hot cell.

Beginning in 2000, 10 countries joined together in the Generation IV International Forum (GIF) to perform the necessary research and develop­ment to support the development and deployment of innovative nuclear energy systems that can be licensed, constructed, and operated in a manner that will provide competitively priced and reliable energy products while satisfactorily addressing nuclear safety, waste, proliferation, and public per­ception concerns (GIF, 2002). GIF experts assessed a large number of candidate reactor concepts by using the common GIF evaluation methodol­ogy resulting in the following six reactor systems selected.

The Gas-Cooled Fast Reactor (GFR, Fig. 9.5) system features a fast — neutron spectrum and closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The reference reactor is a 600-MWth/288-MWe, helium-cooled system operating with an outlet tem­perature of 850°C using a direct Brayton cycle gas turbine for high thermal efficiency. Several fuel forms are being considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products: composite ceramic fuel, advanced fuel particles, or ceramic clad elements of actinide compounds. Core configurations are being con­sidered based on pin — or plate-based fuel assemblies or prismatic blocks. The GFR is estimated to be deployable by 2025.

The Lead-Cooled Fast Reactor (LFR, Fig. 9.6) system features a fast — neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The system uses a lead or lead/ bismuth eutectic liquid-metal cooled reactor. Several options with different plant sizes have been proposed including a battery of 50-150 MWe that features a very long refueling interval, a modular system rated at 300-400 MWe, and a large monolithic plant option at 1200 MWe. The fuel is metal or nitride-based, containing fertile uranium and transuranics. The LFR system is estimated to be deployable by 2025.

The Molten Salt Reactor (MSR, Fig. 9.7) system features an epithermal to thermal neutron spectrum and a closed fuel cycle tailored to the efficient utilization of plutonium and minor actinides. In the MSR system, the fuel is a circulating liquid mixture of sodium, zirconium, and uranium fluorides.

The molten salt fuel flows through graphite core channels, producing a thermal spectrum. The heat generated in the molten salt is transferred to a secondary coolant system through an intermediate heat exchanger, and then through another heat exchanger to the power conversion system. There is no need for fuel fabrication. The reference plant has a power level of 1000 MWe. The system operates at low pressure (<0.5 MPa) and has a coolant outlet temperature above 700°C, affording improved thermal effi­ciency. The MSR is estimated to be deployable by 2025.

The Sodium-Cooled Fast Reactor (SFR, Fig. 9.8) system features a fast — neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides. Two options have been envisioned. The first is an intermediate-size (150 to 500 MWe) sodium-cooled reactor with a uranium-plutonium-minor-actinide-zirconium metal alloy fuel,

supported by a fuel cycle based on pyrometallurgical processing in colo­cated facilities. The second option is a medium to large (500 to 1500 MWe) sodium-cooled fast reactor with mixed uranium-plutonium oxide fuel, sup­ported by a fuel cycle based upon advanced aqueous processing at a central location serving a number of reactors. The outlet temperature is approxi­mately 550°C for both. The SFR is estimated to be deployable by 2015.

The Supercritical-Water-Cooled Reactor (SCWR, Fig. 9.9) system features two fuel cycle options: the first is an open cycle with a thermal neutron spectrum reactor; the second is a closed cycle with a fast-neutron spectrum reactor and full actinide recycle. Both options use a high-temperature, high — pressure, water-cooled reactor that operates above the thermodynamic critical point of water (22.1 MPa, 374°C) to achieve a thermal efficiency approaching 44%. In either option, the reference plant has a 1700-MWe power level, an operating pressure of 25 MPa, and a reactor outlet tempera­ture of 550°C. Passive safety features similar to those of the simplified boiling water reactor are incorporated. The balance-of-plant is considerably

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Generator

Purified salt

Pre I cooler

Heat sink

Intercooler

Compressor

Emergency dump tanks

9.7
GIF Molten Salt Reactor (MSR) generic concept (illustration courtesy of Idaho National Laboratory).

9.8 GIF Sodium-cooled Fast Reactor (SFR) generic concept (illustration courtesy of Idaho National Laboratory).

Electrical power

;o Pump

9.9 GIF Supercritical Water Cooled Reactor (SCWR) generic concept (illustration courtesy of Idaho National Laboratory).

simplified because the coolant does not change phase in the reactor. The SCWR system is estimated to be deployable by 2025.

The Very-High-Temperature Reactor (VHTR, Fig. 9.10) system uses a thermal neutron spectrum and a once-through uranium cycle. The VHTR system is primarily aimed at relatively faster deployment of a system for high-temperature process heat applications with superior efficiency (see Section 9.5). The reference reactor concept has a 600-MWth helium-cooled core based on either the prismatic block fuel of the Gas Turbine-Modular Helium Reactor (GT-MHR) or the pebble fuel of the Pebble Bed Modular Reactor (PBMR). The primary circuit is connected to a steam reformer/ steam generator to deliver process heat. The VHTR system has coolant outlet temperatures above 1000°C. The system may incorporate electricity generation equipment to meet cogeneration needs. The system also has the flexibility to adopt U/Pu fuel cycles and to offer enhanced waste minimiza­tion. The VHTR system is the nearest-term hydrogen production system, estimated to be deployable by 2020.

. Heat Reactor He|ium exchanger coolant