A mixed core for supercritical water-cooled reactors

Corrosion of engineering materials in a supercritical water cooled reactor: Characterization of oxide scales on Alloy 800H and stainless steel 316

The Supercritical water-cooled reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification compared to current light water reactors (LWRs). Reactor operating conditions call for a core coolant temperature between 280 and 620 °C at a pressure of 25 MPa and maximum expected neutron damage levels to any replaceable or permanent core component of 15 dpa (thermal reactor design) and 100 dpa (fast reactor design). Throughout the core, corrosion, stress corrosion cracking, and the effect of irradiation on these degradation modes are critical issues. This chapter reviews the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-offs associated with certain alloy systems.

Material Performance in Supercritical Water

Due to the pressing needs to develop and improve compressors to be used in supercritical carbon dioxide Brayton (S-CO2) cycle, a 3D numerical study has been carried out for the full S-CO2 compressor geometry including diffuser and volute. The predictions were compared with measurements obtained from the recently constructed S-CO2 compressor test facility called SCO2PE (Supercritical CO2 Pressurizing Experiment), based on existing liquid water technology. The objective of the experimental and numerical work is to obtain fundamental data for the design optimization of compressor and to measure the overall performance near the critical point and in the supercritical state.

To simulate nonlinear behavior near the critical point of CO2, the fluid properties were implemented, via property table, in the computational analysis code. Before embarking in the CFD approach evaluation, a number of parametric runs were conducted to examine the order of errors induced by the property table resolution and to achieve grid convergence.

The steady-state numerical predictions using the k–ω SST model were found to return satisfactory results for liquid water and in the case of S-CO2 operating condition, quite far from the critical point. However, as the compressor operating condition approaches more toward the critical point; deviation from the reference data start to become more apparent. In the more challenging case, the disagreement with experimental data might be partially due to the modeling limitations but is also attributed to the subcritical region observed in the contour plot of the static pressure.

CFD investigation of a centrifugal compressor derived from pump technology for supercritical carbon dioxide as a working fluid

A study has been completed to develop a new baseline core design for the liquid-salt-cooled very high-temperature reactor (LS-VHTR) that is better optimized for liquid coolant and that satisfies the top-level operational and safety targets, including strong passive safety performance, acceptable fuel cycle parameters, and favorable core reactivity response to coolant voiding. Three organizations participated in the study: Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Although the intent was to generate a new reference LS-VHTR core design, the emphasis was on performing parametric studies of the many variables that constitute a design. The results of the parametric studies not only provide the basis for choosing the optimum balance of design options, they also provide a valuable understanding of the fundamental behavior of the core, which will be the basis of future design trade-off studies. A new 2400-MW(t) baseline design was established that consists of a cylindrical, nonannular core cooled by liquid {sup 7}Li{sub 2}BeF{sub 4} (Flibe) salt. The inlet and outlet coolant temperatures were decreased by 50 C, and the coolant channel diameter was increased to help lower the maximum fuel and vessel temperatures. An 18-month fuel cycle length with 156 GWD/t burnup was achieved with a two-batch shuffling scheme, while maintaining a core power density of 10 MW/m{sup 3} using graphite-coated uranium oxicarbide particle fuel enriched to 15% {sup 235}U and assuming a 25 vol-% packing of the coated particles in the fuel compacts. The revised design appears to have excellent steady-state and transient performance. The previous concern regarding the core's response to coolant voiding has been resolved for the case of Flibe coolant by increasing the coolant channel diameter and the fuel loading. Also, the LSVHTR has a strong decay heat removal performance and appears capable of surviving a loss of forced circulation (LOFC) even with failure to scram. Significant natural convection of the coolant salt occurs, resulting in fuel temperatures below steady-state values and nearly uniform temperature distributions during the transient.

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Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)

The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor

/pdf/feasibility-study-of-supercritical-light-water-cooled-51p5mhlxve.pdf

Robert Witt

Papers

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production