Optimization of a Fuel Assembly for Supercritical Water-Cooled Reactor CSR1000

As one of the Generation IV nuclear reactors, the SCWR (supercritical water-cooled reactor) has high economy and safety margin, good mechanical properties for its high thermal efficiency, and simplified structure design. As the key component of nuclear reactor, the fuel assembly has always been the main issue for the design of the SCWR. The design of the fuel assembly for CSR1000 proposed by the Nuclear Power Institute of China (NPIC) has been optimized and presented in this study, which is composed of four subassemblies welded by four filler strips and guide thimbles arranged close together in the cross-shaped passage. Aiming at improving the hydraulic buffer performance of the cruciform control rod, the scram time and terminal velocity of control rod assembly were calculated to assess the scram performance based on the computational fluid dynamics and dynamic mesh method, and the mechanical property and neutronic performance of assemblies were also investigated. It has been demonstrated that the optimized fuel assembly had good feasibility and performance, which was a promising design for CSR1000.

Study on Cooling Heat Transfer of Supercritical Carbon Dioxide Applied to Transcritical Carbon Dioxide Heat Pump

Understanding the heat transfer characteristics of supercritical fluids is of fundamental importance in many industrial processes such as transcritical heat pump system, supercritical water-cooled reactor, supercritical separation, and supercritical extraction processes. This chapter addresses recent experimental, theoretical, and numerical studies on cooling heat transfer of supercritical CO2. A systematic study on heat transfer coefficient and pressure drop of supercritical CO2 was carried out at wide ranges of tube diameter, mass flux, heat flux, temperature, and pressure. Based on the understanding of temperature and velocity distributions at cross-sectional direction provided by the numerical simulation, a new prediction model was proposed, which agreed well with the experimental results. In addition, the effect of lubricating oil was also discussed with the focus on the change in flow pattern and heat transfer performance of oil and supercritical CO2.

PRELIMINARY POWER TRANSIENT ANALYSIS OF THE SUPER FR WITH AXIALLY HETEROGENEOUS CORE

The Super FR is one of the SuperCritical Water cooled Reactor (SCWR) concepts with once-through direct cycle plant system. Recently, new design concept of axially heterogeneous core has been proposed, which consists of multiple layers of MOX and blanket fuels. To clarify the safety performance during power transient, safety analyses have been conducted for uncontrolled control rod (CR) withdrawal and CR ejection at full power. RELAP/SCDAPSIM code was used for the safety analysis. The results show that the peak cladding surface temperature (PCST) is high in the upper MOX fuel layer. It is also shown that axial temperature gradient of cladding greatly increases in a short period. Suppressing such large temperature gradient may be a design issue for the axially heterogeneous core from the viewpoint of ensuring fuel integrity.

Energy Conversion Using the Supercritical Steam Cycle

A supercritical steam (or Rankine) cycle is used today for more most of the new coal-fired power plants. More recently, it has been proposed as well for future water-cooled nuclear reactors to enhance their efficiency and to reduce their costs. This chapter provides the technical background explaining this technology. Some criteria for boiler design and operation, like drum or once-through boiler design, fixed or sliding pressure operation, and coolant mixing, are discussed in general to explain the particular challenges of supercritical steam cycles. Examples of technical solutions are given for two large-scale applications: a coal-fired power plant and a supercritical water-cooled reactor, both producing around 1000 MW electric power.

European Research Program on Supercritical Water-Cooled Reactor

The article summarizes the major achievements and scope of the projects supported by the European Commission with the main focus on the supercritical water-cooled reactor, one of the concepts of the Generation IV reactors. The presented projects are focused mainly on the design of the future reactor, the study of the very specific knowledge gaps related to the technology cooled by supercritical water in large-scale installation as well as the small modular reactor type. Major research topics cover the fields of the constructional materials, their corrosion and mechanical stability, validation of the thermal-hydraulic computing codes and design of the future reactor core including the neutronic calculations. Main goals and future direction of the supercritical water-cooled reactor research program are stated to gain as much as possible from the already performed research not only in the nuclear field—but also in the area of fossil-fueled supercritical-water cooled power plants.

A Numerical Study of Turbulent Upward Flow of Super Critical Water in a 2 × 2 Rod Bundle With Nonuniform Heating

This work is part of a benchmarking exercise organized by an IAEA in supercritical water-cooled reactor (SCWR) thermal-hydraulics aimed at improving the understanding and prediction accuracy of the thermal-hydraulic phenomena relevant to SCWRs. An experiment carried out using a 2 × 2 SCWR bundle at University of Wisconsin-Madison was modeled using an open-source computational fluid dynamics (CFD) code—Code_Saturne. The k–ω shear stress transport (SST) model was used to account for the buoyancy-aided turbulent flow in the fuel channel. Significant heat transfer deterioration (HTD) was observed in the boundary layer, which is commonly expected to occur in buoyancy-aided flows. For comparison, simulations were also conducted using ansysfluent with similar model setups.

Numerical Simulation of a Steam Injector for Passive Residual Heat Removal

In the case of depressurization of a supercritical water-cooled reactor (SCWR), the steam pressure can be used to pump coolant back into the reactor, either by using a turbine driven pump or a steam injector. The latter option is cheaper, more robust, and requires less periodic inspections. The numerical simulation of a steam injector, however, has still been a challenge. In principle, a steam injector consists of a steam nozzle, a mixing chamber, and a diffuser. This paper describes a one-dimensional two-phase flow model simulating the flow and condensation phenomena occurring in the steam nozzle, the structure of shock fronts at the nozzle outlet, the direct contact condensation of steam with cold water in the mixing chamber, and the sudden pressure increase in the final condensation shock front. The model is based only on first principles and on the existing correlations. Correction factors adapting the model to experimental results have not been used yet. For validation, the model has been applied both to separate effect tests and to steam injector tests.

Evaluation of Friction-Factor Correlations at Supercritical Water Conditions in Support of the Canadian SCWR

Canada is participating in the Generation IV International Forum with the main focus on the pressure-tube-type supercritical water-cooled reactor (SCWR) concept. The Canadian SCWR concept is a heavy-water moderated and light-water-cooled reactor. The R&D framework for the development of the Canadian SCWR fuel-assembly concept includes experiments and analyses, including subchannel code development and applications. This paper focuses on the modeling of the hydraulic resistance under supercritical conditions, with or without the wire-wrap spacers. More specifically, it presents an assessment of three friction factor correlations developed for supercritical conditions. A literature survey of wire-wrap hydraulic models is presented. The assessment of the supercritical friction factor correlations and wire-wrap hydraulic models is carried out using three pressure drop experimental datasets, using the subchannel code ASSERT-PV V3.2m2.