Microfluidic Quantification of Blood Pressure and Compliance Properties Using Velocity Fields under Periodic On–Off Blood Flows

To monitor variations of blood samples effectively, it is required to quantify static and dynamic properties simultaneously. With previous approaches, the viscosity and elasticity of blood samples are obtained for static and transient flows with two syringe pumps. In this study, simultaneous measurement of pressure and equivalent compliance is suggested by analyzing the velocity fields of blood flows, where a blood sample is delivered in a periodic on-off fashion with a single syringe pump. The microfluidic device is composed of a main channel (mc) for quantifying the equivalent compliance and a pressure channel (pc) for measuring the blood pressure. Based on the mathematical relation, blood pressure at junction (Px) is expressed as Px = kβ. Here, β is calculated by integrating the averaged velocity in the pressure channel (<Upc>). The equivalent compliance (Ceq) is then quantified as Ceq = λoff · Q0/Px with a discrete fluidic model. The time constant (λoff ) is obtained from the transient behavior of the averaged blood velocity in the main channel (<Umc>). According to results, Px and Ceq varied considerably with respect to the hematocrit and flow rate. The present method (i.e., blood pressure, compliance) shows a strong correlation with the previous method (i.e., blood viscosity, elasticity). In conclusion, the present method can be considered as a potential tool for monitoring the mechanical properties of blood samples supplied periodically from a single syringe pump.

Analysis and calculation of the process of low-pressure reverse osmosis during recycling of hygienic water

Analysis and calculation have been conducted of the process of low-pressure reverse osmosis in the membrane apparatus of the system for recycling hygiene water for the space station. The paper describes the physics of the reverse osmosis treatment and determines the motive force of the process, which is the difference of effective pressures (operating pressure minus osmotic pressure) in the solution near the surface of the membrane and in the purified water. It is demonstrated that the membrane scrubbing action is accompanied by diffusion outflow of the cleaning agent components away from the membrane. The mass transfer coefficient and the difference of concentrations (and, accordingly, the difference of osmotic pressures) in the boundary layer of the pressure channel can be determined using an extended analogy between mass transfer and heat transfer. A procedure has been proposed and proven in an experiment for calculating the throughput of a reverse osmosis apparatus purifying the hygiene water obtained through the use of a cleaning agent used in sanitation and housekeeping procedures on Earth. Key words: life support system, hygiene water, water processing, low-pressure reverse osmosis, space station.

Underwater Acoustic Communication by a Single-Vector Sensor: Performance Comparison Using Three Different Algorithms

In November, 2014, the underwater acoustic (UWA) communication experiment by a single-vector sensor was conducted in shallow water environment. In this paper, three different algorithms are used to process the experimental data and their performance are compared in terms of equalized output signal to noise ratio (OSNR) and bit error rate (BER). The three algorithms are P-DFE, B-DFE, and T-DFE, respectively. P-DFE uses only the pressure channel of the vector sensor to realize the decision feedback equalizer (DFE). B-DFE linearly combines the pressure channel and velocity channel first and then uses DFE to equalize the combined signal. T-DFE adopts time reversal to combine all the channels of the vector sensor and then is followed by a single-channel DFE to remove residual intersymbol interference (ISI). According to the data processing results, both B-DFE and T-DFE can achieve better performance compared with P-DFE. This paper also finds that the performance of B-DFE depends on the beam pattern of the combined signal while the performance of T-DFE depends on the q function of the combined signal. Which algorithm should be used to process real data, B-DFE or T-DFE, depends on the degree of coherence between different channels of the vector sensor.

Investigation of Thermal Efficiency of Generic Pressure-Channel Reactors With Steam Reheat

Currently, i.e., in 2016, 4361 nuclear-power reactors operate in the world. 96.6% of these reactors are water-cooled (373 reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 reactors — PHWRs are cooled with heavy water. 15% of all water-cooled reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plants (up to 62%) and supercritical-pressure coal-fired plants (up to 55%). Therefore, it is very important to propose ways of improvement of thermal efficiency for this largest group of nuclear-power reactors. It should be noted that among six Generation-IV nuclear-reactor concepts one concept is a SCWR, which might reach thermal efficiencies within the range of 45–50% and even beyond. However, this concept has been never tested, and the most difficult problem on the way of implementation of this type of reactor is the reliability of materials at supercritical pressures and temperatures, very aggressive reactor coolant – supercritical water, and high neutron flux. Up till now, no experiments on behavior of various core materials at these conditions have been reported so far in the open literature. As an interim way of thermal-efficiency improvement for water-cooled NPPs nuclear steam reheat can be considered. However, this way is more appropriate only for pressure-channel reactors, for example, CANDU-type or PHWRs. Moreover, in the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat in subcritical-pressure experimental boiling reactors. Therefore, an objective of the current paper is to summarize this experience and to estimate effect of a number of parameters on thermal efficiencies of a generic pressure-channel reactors with nuclear steam reheat. For this purpose the DE-TOP program has been used.

Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts, one of which is supercritical water-cooled reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of water-cooled reactors (pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), and light-water, graphite-moderated reactors (LGRs)), which comprise 96% of the current fleet of operating nuclear power reactors and are categorized under Generation II, III, and III+ nuclear reactors. The latter water-cooled reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressure-vessel (PV) SCWRs and pressure-channel (PCh) SCWRs. A generic pressure-channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350°C and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel-channel design. Second, a nuclear fuel and fuel cycle should be selected. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and reentrant channel concepts. The objective of this paper is to study thermal-hydraulic and neutronic aspects of a reentrant fuel-channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB, which calculates fuel-centerline-temperature, sheath-temperature, coolant-temperature, and heat-transfer-coefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850°C for fuel.