James J. Sienicki

Ex-vessel melt-coolant interactions in deep water pool: studies and accident management for Swedish BWRs

Abstract In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The Thirmal-1 code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the Thirmal-1 analyses. A description of the modeling incorporated in Thirmal-1 is also provided.

Stability analysis on single-phase natural circulation in Argonne lead loop facility

One-dimensional linear stability analysis was performed for single-phase lead–bismuth eutectic (LBE) natural circulation. The Nyquist criterion and a root search method were employed to find the linear stability boundary of both forward and backward circulations. It was found that the natural circulations could be linearly unstable in a high Reynolds number region. Increasing loop friction makes a forward circulation more stable, but destabilizes the corresponding backward circulation under the same heating/cooling conditions. The characteristic wavelength of an unstable disturbance is roughly equal to the entire loop length.

Natural Circulation and Linear Stability Analysis for Liquid-Metal Reactors with the Effect of Fluid Axial Conduction

The effect of fluid axial thermal conduction on one-dimensional liquid metal natural circulation and its linear stability was performed through nondimensional analysis, steady-state assessment, and linear perturbation evaluation. The Nyquist criterion and a root-search method were employed to find the linear stability boundary of both forward and backward circulations. The study provided a relatively complete analysis method for one-dimensional natural circulation problems with the consideration of fluid axial heat conduction. The results suggest that fluid axial heat conduction in a natural circulation loop should be considered only when the modified Peclet number is [approximately]1 or less, which is significantly smaller than the practical value of a lead liquid metal-cooled reactor.

Power optimization in the STAR-LM modular natural convection reactor system.

STAR-LM is a modular, factory-fabricated, overland transportable, proliferation resistant, autonomous load following, and passively safe liquid metal-cooled fast reactor system that utilizes inert lead-bismuth eutectic coolant combined with 100+% natural circulation heat transport to achieve radial design simplification, enhanced reliability, and cost savings. STAR-LM has the potential to meet all of the U.S.DOE Generation IV goals of sustainable energy development, safety and reliability, and economics. Previous development of STAR-LM resulted in a 300 MWt modular pool type reactor in which modular steam generators are immersed directly inside the primary coolant and the core is part of a reactor module/flow-thru fuel cartridge that provides no access to fuel during the core lifetime. Recent concept development has focused on raising the power achievable in a small module size based on preserving key criteria for: i) full spectrum of modes of module transport from factory to site (including rail transport); ii) ultralong core cartridge lifetime; iii) 100% natural circulation heat transport; and iv) coolant and cladding peak temperatures well within the existing (mainly Russian) database for lead-bismuth eutectic coolant and ferritic steel core materials. For example, natural circulation is found to be capable of transporting 400 MWt in a fully transportable module size with 15 year core life (at 100% capacity factor;100,000 MWd/tonne average burnup; 1.5 peaking factor) with a core outlet temperature of 489 C, peak cladding inner surface temperature of 580 C, and steam outlet superheat of 82 C at 10 MPa.Copyright

Experimental Investigations on Sodium Plugging in Narrow Flow Channels

Abstract A series of experiments was performed to investigate the potential for plugging of narrow flow channels of sodium by impurities (e.g., oxides). In the first phase of the experiments, clean sodium was circulated through the test sections simulating flow channels in a compact diffusion-bonded heat exchanger such as a printed circuit heat exchanger. The primary objective was to see if small channels whose cross sections are semicircles of 2, 4, and 6 mm in diameter are usable in liquid sodium applications where sodium purity is carefully controlled. It was concluded that the 2-mm channels, the smallest of the three, could be used in clean sodium systems at temperatures even as low as 100 to 110°C without plugging. In the second phase, sodium oxide was added to the loop, and the oxygen concentration in the liquid sodium was controlled by means of varying the cold-trap temperature. Intentional plugging was induced by creating a cold spot in the test sections, and the subsequent plugging behavior was observed. It was found that plugging in the 2-mm test section was initiated by lowering the cold spot temperature below the cold-trap temperature by 10 to 30°C. Unplugging of the plugged channels was accomplished by heating the affected test section.

Lead-Cooled Fast Reactors

This chapter describes fast neutron reactors utilizing either of two so-called Heavy Liquid Metal Coolants; namely, lead (Pb) and lead-bismuth eutectic (LBE), both having significantly higher densities and boiling temperatures than sodium. At the current time, the main driver for interest in such reactors is the potential for reductions in the nuclear power plant capital cost per unit electrical power, realized by taking advantage of the particular properties of the Heavy Liquid Metal Coolant. A lead-cooled fast reactor (LFR) is not a sodium-cooled fast reactor (SFR) with a different coolant. Effective LFR designs differ in significant ways from SFR designs and the coolant technologies for the Heavy Liquid Metal Coolants are significantly different from the coolant technology for sodium. The present chapter discusses the distinctive features of the Heavy Liquid Metal Coolants and their consideration in LFR design. Three examples of LFR concepts are described.

Lessons Learned and Improvements in ANL Plant Dynamics Code Simulation of Experimental S-CO2 Loops

Validation of the ANL Plant Dynamics Code with the experimental data from integral S-CO2 cycle facilities has been continued. Several code modifications as well as modeling approaches and assumptions were introduced to improve both the code’s capabilities in modeling the experimental loops and the agreement of the code prediction with the experimental data. The lessons learned from the code improvement and modeling experience important for the validation of the codes with the experimental data from small-scale integral loops are presented.Copyright

A Supercritical CO2 Brayton Cycle Power Converter for a Sodium-Cooled Fast Reactor Small Modular Reactor

Although a number of power conversion applications have been identified or have even been developed (e.g., waste heat recovery) for supercritical carbon dioxide (S-CO2) cycles including fossil fuel combustors, concentrated solar power (i.e., solar power towers), and marine propulsion, the benefits of S-CO2 Brayton cycle power conversion are especially prominent for applications to nuclear power reactors. In particular, the S-CO2 Brayton cycle is well matched to the Sodium-Cooled Fast Reactor (SFR) nuclear power reactor system and offers significant benefits for SFRs. The recompression closed Brayton cycle is highly recuperated and wants to operate with an approximate optimal S-CO2 temperature rise in the sodium-to-CO2 heat exchangers of about 150 °C which is well matched to the sodium temperature rise through the core that is also about 150 °C. Use of the S-CO2 Brayton cycle eliminates sodium-water reactions and can reduce the nuclear power plant cost per unit electrical power. A conceptual design of an optimized S-CO2 Brayton cycle power converter and supporting systems has been developed for the Advanced Fast Reactor – 100 (AFR-100) 100 MWe-class (250 MWt) SFR Small Modular Reactor (SMR). The AFR-100 is under ongoing development at Argonne National Laboratory (ANL) to target emerging markets where a clean, secure, and stable source of electricity is required but a large-scale power plant cannot be accommodated. The S-CO2 Brayton cycle components and cycle conditions were optimized to minimize the power plant cost per unit electrical power (i.e.,

Validation of the ANL Plant Dynamics Code With the SNL S-CO2 Loop Transient Data

/kWe). For a core outlet temperature of 550 °C and turbine inlet temperature of 517 °C, a cycle efficiency of 42.3 % is calculated that exceeds that obtained with a traditional superheated steam cycle by one percentage point or more. A normal shutdown heat removal system incorporating a pressurized pumped S-CO2 loop slightly above the critical pressure on each of the two intermediate sodium loops has been developed to remove heat from the reactor when the power converter is shut down. Three-dimensional layouts of S-CO2 Brayton cycle power converter and shutdown heat removal components and piping have been determined and three-dimensional CAD drawings prepared. The S-CO2 Brayton cycle power converter is found to have a small footprint reducing the space requirements for components and systems inside of both the turbine generator building and reactor building. The results continue to validate earlier notions about the benefits of S-CO2 Brayton cycle power conversion for SFRs including higher efficiency, improved economics, elimination of sodium-water reactions, load following, and smaller footprint.Copyright

Transient Accident Analysis of a Supercritical Carbon Dioxide Brayton Cycle Energy Converter Coupled to an Autonomous Lead-Cooled Fast Reactor

The ANL Plant Dynamics Code (PDC) is the current state-of-the-art capability for one-dimensional system level transient analysis of supercritical carbon dioxide (S-CO2) Brayton cycle power converters. Earlier validation of code models was carried out with data from testing of individual S-CO2 components such as a small-scale compact diffusion-bonded heat exchanger and compressor tests. The steady-state part of the PDC has been compared with experimental data from the Sandia National Laboratories (SNL) small-scale S-CO2 Brayton cycle demonstration. In this work, predictions of the PDC code are assessed through comparison with SNL S-CO2 loop transient data. Code modifications were needed to properly simulate the actual experimental runs due to the unique features of the small-scale SNL loop. Overall, good agreement with the measured data is predicted by the PDC, although the code predictions could be improved in some cases. Future code improvements for comparisons with future SNL loop data are identified based upon the results.Copyright