SCWRs are high temperature, high-pressure, light-water-cooled reactors that operate above the thermodynamic critical point of water (374°C, 22.1 MPa).
The reactor core may have a thermal or a fast-neutron spectrum, depending on the core design. The concept may be based on current pressure vessel or on pressure tube reactors, and thus use light water or heavy water as moderator. Unlike current water-cooled reactors, the coolant will experience a significantly higher enthalpy rise in the core, which reduces the core mass flow for a given thermal power and increases the core outlet enthalpy to superheated conditions. For both pressure vessel and pressure-tube designs, a once through steam cycle has been envisaged, omitting any coolant recirculation inside the reactor. As in a boiling water reactor, the superheated steam will be supplied directly to the high pressure steam turbine and the feed water from the steam cycle will be supplied back to the core. Thus, the SCWR concepts combine the design and operation experiences gained from hundreds of water-cooled reactors with those experiences from hundreds of fossil-fired power plants operated with supercritical water (SCW). In contrast to some of the other Generation IV nuclear systems, the SCWR can be developed incrementally step-by-step from current water-cooled reactors.
Advantages and Challenges
Such SCWR designs have unique features that offer many advantages compared to state-of the-art water-cooled reactors:
SCWRs offer increases in thermal efficiency relative to current-generation water-cooled reactors. The efficiency of a SCWR can approach 44% or more, compared to 34-36% for current reactors.
Reactor coolant pumps are not required. The only pumps driving the coolant under normal operating conditions are the feed water pumps and the condensate extraction pumps.
The steam generators used in pressurized water reactors and the steam separators and dryers used in boiling water reactors can be omitted since the coolant is superheated in the core.
Containment, designed with pressure suppression pools and with emergency cooling and residual heat removal systems, can be significantly smaller than those of current water-cooled reactors.
The higher steam enthalpy allows to decrease the size of the turbine system and thus to lower the capital costs of the conventional island.
These general features offer the potential of lower capital costs for a given electric power of the plant and of better fuel utilization, and thus a clear economic advantage compared with current light water reactors.
However, there are several technological challenges associated with the development of the SCWR, and particularly the need to validate transient heat transfer models (for describing the depressurization from supercritical to sub-critical conditions), qualification of materials (namely advanced steels for cladding), and demonstration of the passive safety systems.
GIF progress up to 2012
Pre-conceptual core design studies for a core outlet temperature of more than 500°C have been performed in Japan, assuming either a thermal neutron spectrum or a fast neutron spectrum. Both options are based on a coolant heat-up in two steps with intermediate mixing underneath the core. Additional moderator for a thermal neutron spectrum is provided by feed water inside water rods. The fast-spectrum option uses zirconium-hydride (ZrH2) layers to minimize hardening of the neutron spectrum in case of core voiding. A pre-conceptual design of safety systems for both options has been studied with transient analyses.
A pre-conceptual plant design with 1700 MW net electric power based on a pressure-vessel-type reactor has been studied by Yamada et al. and has been assessed with respect to efficiency, safety and cost. The study confirms the target net efficiency of 44% and estimates a cost reduction potential of 30% compared with current pressurized water reactors. Safety features are expected to be similar to advanced boiling water reactors.
A pre-conceptual design of a pressure-vessel-type reactor with a 500°C core outlet temperature and 1000 MW electric power has been developed in Europe, as summarized by Schulenberg and Starflinger. The core design is based on coolant heat-up in 3 steps. Additional moderator for the thermal neutron spectrum is provided in water rods and in gaps between assembly boxes. The design of the nuclear island and of the balance of the plant confirms results obtained in Japan, namely an efficiency improvement up to 43.5% and a cost reduction potential of 20 to 30% compared with latest boiling water reactors. Safety features as defined by the stringent European Utility Requirements are expected to be met.
Canada is developing a pressure-tube-type SCWR concept with a 625°C core outlet temperature at the pressure of 25 MPa. The concept is designed to generate 1200 MW electric power (a 300 MW concept is also being considered). It has a modular fuel channel configuration with separate coolant and moderator. A high-efficiency fuel channel is incorporated to house the fuel assembly. The heavy-water moderator directly contacts the pressure tube and is contained inside a low-pressure calandria vessel. In addition to providing moderation during normal operation, it is designed to remove decay heat from the high-efficiency fuel channel during long-term cooling using a passive moderator cooling system. A mixture of thorium oxide and plutonium is introduced as the reference fuel, which aligns with the GIF position paper on thorium fuel. The safety system design of the Canadian SCWR is similar to that of the ESBWR. However, the introduction of the passive moderator cooling system coupled with the high-efficiency channel could reduce significantly the core damage frequency during postulated severe accidents such as large-break loss-of-coolant or station black-out events.
Pre-conceptual designs of three variants of pressure vessel supercritical reactors with thermal, mixed and fast neutron spectrum have been developed in Russia, which joined the SCWR System Arrangement in 2011.
Outside of the GIF framework, two conceptual SCWR designs with thermal and mixed neutron spectrum cores have been established by some research institutes in China under framework of the Chinese national R&D projects from 2007-2012, covering some basic research projects on materials and thermo hydraulics, the core/fuel design, the main system design (including the conventional part), safety systems design, reactor structure design and fuel assembly structure design. The related feasibility studies have also been completed, and show that the design concept has promising prospects in terms of the overall performance, integration of design, component structure feasibility and manufacturability.
Prediction of heat transfer in SCW can be based on data from fossil fired power plants as discussed by Pioro et al. Computational tools for more complex geometries like fuel assemblies are available but still need to be validated with bundle experiments. System codes for transient safety analyses have been upgraded to include SCW, including depressurization transients to subcritical conditions. Flow stability in the core has been studied numerically. As in boiling water reactors, flow stability can be ensured using suitable inlet orifices in fuel assemblies.
A number of candidate cladding materials have been tested in capsules, autoclaves and recirculating loops up to 700°C at a pressure of 25 MPa. Stainless steels with more than 20% chromium (Cr) are expected to have the required corrosion resistance up to a peak cladding temperature of 650°C. More work is needed to develop alloys suitable for use at the design peak cladding temperatures of 850°C for the Canadian SCWR concept. Further work is also needed to better identify the coolant conditions that lead to stress corrosion cracking. It has been shown that the creep resistance of existing alloys can be improved by adding small amounts of elements, such as zirconium (Zr), as reported by Kaneda et al. In the longer term, the steel experimental oxide dispersion strengthened (ODS) alloys offer an even higher potential, whereas nickel-base alloys are being considered for use in ultrasupercritical fossil fired plants are less favourable for use in SCWRs due to their high neutron absorption and associated swelling and embrittlement.
Key water chemistry issues have been identified by Guzonas et al.; predicting and controlling water radiolysis and corrosion product transport (including fission products) remain the major R&D areas. In this regard, the operating experience using nuclear steam reheat at the Beloyarsk nuclear power plant in Russia is extremely valuable.
Recent SCWR research papers and links
Y. Oka, S. Koshizuka, Y. Ishiwatari, A. Yamaji, Super light water reactors and super fast reactors, Springer 2010.
K. Yamada, S. Sakurai, Y. Asanuma, R. Hamazaki, Y. Ishiwatari, K. Kitoh, Overview of the Japanese SCWR concept developed under the GIF collaboration, Proc. ISSCWR-5, Vancouver, Canada, March 13-16, 2011.
M. Yetisir, W. Diamond, L.K.H. Leung, D. Martin and R. Duffey, Conceptual Mechanical Design for A Pressure-Tube Type Supercritical Water-Cooled Reactor, Proc. 5th International Symposium on Supercritical Water-cooled Reactors, Vancouver, Canada, March 13-17, 2011.
S.B. Ryzhov, V.A.Mokhov, M.P.Nikitenko, A.K.Podshibyakin, I.G. Schekin, A.N. Churkin, Advanced designs of VVER reactor plant, The 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8), October 10-14, 2010, Shanghai, China, Paper N8P0184.
S. B. Ryzhov, P. L. Kirillov, et al., Concept of a Single-Circuit RP with Vessel Type Supercritical Water-Cooled Reactor, Proc. ISSCWR-5, Vancouver, Canada, March 13-16, 2011.
I.L. Pioro, R.B. Duffey, Heat transfer and hydraulic resistance at supercritical pressures in power engineering applications, ASME Press 2007.
J. Kaneda, S. Kasahara, F. Kano, N. Saito, T. Shikama, H. Matsui, Material development for supercritical water-cooled reactor, Proc. ISSCWR-5, Vancouver, Canada, March 13-16, 2011.
D. Guzonas, F. Brosseau, P. Tremaine, J. Meesungnoen, J.-P. Jay-Gerin, Water chemistry in a supercritical watercooled pressure tube reactor, Nuclear Technology Vol. 179, 2012.
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