The Ninth International Symposium on Supercritical Water-Cooled Reactors (IS-SCWRs) was successfully held on Mar. 11–13, 2019 at the Vancouver Marriott Hotel, Vancouver, BC, Canada. It was hosted by the Canadian Nuclear Society in cooperation with the Generation-IV International Forum (GIF) and the International Atomic Energy Agency (IAEA). As in previous symposia (Tokyo (Japan) in 2000 and 2003, Shanghai (China) in 2007, Heidelberg (Germany) in 2009, Vancouver (Canada) in 2011, Shenzhen (China) in 2013, Helsinki (Finland) in 2015, and Chengdu (China) in 2017), the Ninth IS-SCWRs provided a forum for discussion of advances and issues, sharing of information, transfer of technology, and establishment of future collaborations on research and development (R&D) in key technology areas for SCWRs.
The Ninth IS-SCWRs was attended by scientists and engineers from 12 countries (Austria, Canada, China, Czech Republic, Finland, France, Germany, Hungary, Japan, Spain, the Netherlands, and United Kingdom). Six plenary speeches were presented highlighting advances in SCWR design and development by the key signatories of the GIF SCWR System. Fifty-four technical presentations were delivered covering SCWR core and fuel designs, neutronics, thermal-hydraulics, materials, chemistry, and safety analyses. Selected full manuscripts are compiled in this Special Section of the ASME Journal of Nuclear Engineering and Radiation Science.
The SCWR is the only water-cooled reactor among the six concepts selected for cooperative R&D within the GIF. It is considered as one of the most promising concepts due to its simplicity in design, high thermal efficiency, and over 50 years of industrial experience from thermal-power stations using a supercritical water cycle. Organizations from four countries (Canada, China, Japan, and the Russian Federation) and the European Union are pursuing the SCWR design, and other countries have expressed interest in the technology. Many technological advances being developed for SCWR concepts are synergetic to those of the currently operating water-cooled reactor fleets. In addition, high-temperature materials development and heat transfer to supercritical carbon dioxide (relevant to the Brayton cycle) are of interest to other GIF reactor concepts, such as the very high temperature reactor and the metal-cooled fast reactor.
The SCWR is a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water (374 °C, 22.1 MPa). Two SCWR design concepts are being developed: (a) the pressure-vessel type with a reactor core (fuelled) heat source, analogous to conventional pressurized water reactors and boiling water reactors and (b) the pressure-tube type consisting of fuel bundles in distributed pressure tubes or channels, analogous to conventional CANDU® (Canada Deuterium Uranium, a registered trademark of Atomic Energy of Canada Limited) and RBMK (Reactor Bolshoy Moshchnosti Kanalnyy) nuclear reactors. These core designs are based on thermal neutron, fast neutron, or mixed (thermal and fast) spectra. The majority of SCWR plants are developed for power generation higher than 1000 MWe at operating pressures of about 25 MPa and reactor outlet temperatures up to 625 °C. Under these conditions, the coolant does not change phase (boil) in the reactor and the SCWR adopts a direct cycle eliminating moisture separator reheaters and recirculation pumps as in boiling-water reactors, or steam generators as in pressurized light-water and heavy-water reactors. This has significantly reduced the sizes and footprints of the containment, building, and plant of SCWR concepts. The balance-of-plant configuration is the same as that of a modern fossil-fuel power plant, which is based on over 50 years of design and operation experience.
The main mission of the SCWR is to generate electricity efficiently, economically, and safely. All SCWR concepts currently under development could generate electricity with thermal efficiencies ranging from 43% to 48%, which is better than the typical 35% achieved by the current fleet of water-cooled nuclear reactors. The high core-outlet temperature of SCWR concepts facilitates cogeneration, such as hydrogen production, space heating, and steam production.
Two fast-spectrum SCWR concepts (one with thorium fuel and the other with uranium fuel) were introduced at the symposium. Both concepts are still in the preliminary development phase and details have not been provided. Neutronics analyses were presented for the SCWR concepts being developed in Canada and China. A new fuel concept was also introduced for the Chinese SCWR. An economic analysis was performed for a Canadian small SCWR concept being developed to generate 350 MWe power.
Studies on materials focused on corrosion and cracking characteristics at high pressure and high temperature conditions. Materials examined in these studies included stainless steels, nickel-based alloys, titanium and zirconium alloys, and oxide dispersion-strengthened austenitic steel. Results of the second round-robin corrosion testing of Alloy 800H and stainless-steel 310S were also presented. Water chemistry has been shown to have a significant impact on material corrosion characteristics. Radiolysis processes in water under conditions expected in an SCWR core have been modeled and appear to increase the formation of hydrogen and could lead to localized transient increases in the coolant acidity.
Thermal-hydraulics studies presented at the symposium mainly covered heat transfer, flow instability, and critical flow. Fundamental investigations of heat transfer in tubes and annuli were presented using computational fluid dynamics tools and direct numerical simulations. Correlations were examined for normal and deteriorated heat-transfer phenomena in tubes at supercritical pressures. A number of experimental and analytical studies were shown for heat transfer in three- or four-rod bundles. Results of flow instability and natural circulation studies were presented for a single tube and parallel channels. Experimental investigations and modeling of critical flow through orifices and nozzles at supercritical pressures were described. Implementation of the latest critical-flow model into the CATHENA safety analysis code was introduced.
Experimental heat transfer data were presented for depressurization from supercritical to subcritical pressures. Transient and system analyses of SCWR concepts were demonstrated using safety analysis tools. Coupled thermal-hydraulics and neutronic analyses were described for a thorium-fueled SCWR concept.
The symposium was completed with a technical tour of Canada's Particle Accelerator Center (TRIUMF) in Vancouver, BC, Canada, on March 14. This center is a key partner for SCWR material R&D through collaborations with several academic institutes.
Overall, the Ninth IS-SCWRs facilitated interaction between stakeholders, SCWR designers, engineers, researchers, and academia. State-of-the-art information was presented on various topics of importance to SCWR concept development. Ample time was allocated for fruitful discussions and the exchange of ideas as well as experience. All participants enjoyed the symposium and emphasized the importance to maintain this type of interaction every 2 years.
