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Boiling heat transfer system keeps a nuclear power plant safe without getting over-heated. Crisis will occur if the dissipated heat flux exceeds the critical heat flux value. This study assumes the flow boiling system at high heat flux is characterized by the existence of a very thin liquid layer, known as the &#34sublayer&#34, which is trapped between the heated surface and the vapor blankets. In the present study, it is hypothesized that the heat transfer through the liquid sublayer is dominated by the heat conduction and the sublayer is dried out due to occurrence of Helmholtz instability as the relative velocity of the vapor blanket to the local liquid in the sublayer reaches a critical value. By recognizing this hypothesis, a theoretical model for low-quality flow is developed to predict boiling heat transfer and Critical Heat Flux (CHF). To verify the validity of the present model, the predictions are compared with the experimental data of flow boiling heat transfer in the simulation of Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) conditions. For the PWR low-quality flow, the comparison demonstrates that the Helmholtz instability is the trigger condition for the onset of CHF.

The purpose of this study is to evaluate the performance of a passive core cooling system (PCCS) with passive injection during the cold-leg small break loss-of-coolant accidents (SBLOCAs) experiments conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. Four tests were performed simulating break sizes of 0.2–2% (approximately corresponding to 1.25–4″ breaks for a referenced nuclear power plant) at cold-leg for assessing the PCCS capability in accident management. The key thermal–hydraulic phenomena to core heat removal for PCCS are observed and discussed. The experimental results show that the PCCS has successfully provided a continuous removal of core heat and a long term core cooling can be reached for all cases of SBLOCA.

A complete scheme of scaling methods to design the reduced-height, reduced-pressure (RHRP) Institute of Nuclear Energy Research Integral System Test (IIST) facility and to specify test conditions for incident simulation was developed. In order to preserve core decay power history and coolant mass inventory during a transient, a unique power-to-mass scaling method is proposed and utilized for RHRP and full-height, full-pressure (FHFP) systems. To validate the current scaling method, three counterpart tests done at the IIST facility are compared with the FHFP tests in small-break loss-of-coolant, station blackout, and loss-of-feedwater accidents performed at the Large-Scale Test Facility (LSTF) and the BETHSY test facility. Although differences appeared in design, scaling, and operation conditions among the IIST, LSTF, and BETHSY test facilities, the important physical phenomena shown in the facilities are almost the same. The physics involved in incident transient phenomena are well measured and modeled by showing the common thermal-hydraulic behavior of key parameters and the general consistency of chronological events. The results also confirm the adequacy of power-to-mass scaling methodology.

A passive core cooling system (PCCS) has been installed at the Institute of Nuclear Energy Research Integral System Test facility. It includes three core makeup tanks (CMTs), three accumulators (ACCs), a four-stage automatic depressurization system (ADS), a passive residual heat removal (PRHR) heat exchanger submerged into an in-containment refueling water storage tank (IRWST). The purpose of this research is to study the performance of the PCCS with passive injection during either a pressure balance line (PBL) break or a direct vessel injection (DVI) line break. Five experiments were performed simulating break area ratios of 0.5 to 2.0% (1.88 to 3.77 mm) at either a PBL or a DVI line. The general system response and the interactions of CMT, ACC, PRHR, and IRWST to the effect of core heat removal are observed and discussed. The experimental results show long-term core cooling can be reached for the cases of the PBL break and the DVI-line break following the PCCS actuation procedures.

The thermal-hydraulic behavior of a postulated 1% cold-leg break loss-of-coolant accident (LOCA) in a pressurized water reactor system was investigated experimentally by the three-loop Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility with the passive core cooling system (PCCS) and numerically by the RELAP5/MOD3.2 computer code. The PCCS of the IIST facility includes three core makeup tanks (CMTs), three accumulators, and a four-stage automatic depressurization system. The aim of this research is to study the performance of the CMTs with the actuation of the ADS during a small-break LOCA. The experimental results show that the IIST PCCS has the capability to maintain long-term cooling under a postulated 1% cold-leg break LOCA. The comparison of the RELAP5/MOD3.2 simulation against the experimental data shows good agreement in major thermal-hydraulic phenomena in the reactor coolant system, but the prediction of the asymmetric behavior for the three CMTs during a gravity drain period is inadequate.

The current emergency operating procedures (EOPs) for a three-loop pressurized water reactor on steam generator tube rupture (SGTR) incident are experimentally evaluated at the Institute of Nuclear Energy Research Integral System Test Facility. The focus is on the adequacy of EOPs to limit primary-to-secondary leakage with particular emphasis on the number of ruptured U-tubes on the severity of an incident and the response time available for operator actions. By comparing test data with plant records of a Mihama-2 SGTR event, the key thermal-hydraulic phenomena during SGTR transients can be successfully simulated. The test results indicated that the current EOPs can function well in terminating the break flow and maintaining adequate core cooling. However, the effectiveness in minimizing the radioactive release demands an early and substantial operator involvement. To mitigate the consequences of such an event, the timing to isolate the faulted steam generator (SG) and to terminate safety injection flow will be critical. Furthermore, to avoid overfilling, the water level in the faulted SG needs to be drained prior to the implementation of the cooldown process.

Four scaled small break loss-of-coolant accident (LOCA) tests simulating the pressurizer power-operated relief valves (PORVs) stuck-open accidents and the recovery actions in a pressurized water reactor (PWR) were performed at the Institute of Nuclear Energy Research (INER) integral system test (IIST) facility. The objectives of this study are to verify the effectiveness of emergency operating procedure (EOP) and emergency core cooling system (ECCS) on reactor safety. The break sizes were volumetrically scaled down based on one and all three fully-opened PORVs which is equivalent to 0.23% and 0.69% hot leg flow area of the reference plant. The experimental results indicate that in case of high pressure injection (HPI) system failure, the rapid depressurization of the steam generators is proved to be an effective way in the depressurization of the reactor coolant system and the core cooling. In contrast, if only one HPI charging pump operates normally, which injected half (or minimum) flow rate of normal cooling water, the core cooling can be adequately provided without operating the secondary bleeding during PORV stuck-open transient. This paper also presents the scaling methods for the reduced-height, reduced-pressure (RHRP) IIST facility and the test conditions. The validity of the present scaling methodology is confirmed by the results from previous IIST counterpart tests and comparison of the present results with those of the tests performed at the full-height, full-pressure(FHFP) stuck-open tests.

A reduced-height, reduced-pressure (RHRP) integral system test facility at the Institute of Nuclear Energy Research (INER) has been established for simulating the thermal-hydraulics of a Westinghouse three-loop pressurized water reactor (PWR). To understand whether or not the physical phenomena observed in this RHRP integral system test facility during a station blackout (SB) transient can be reliably extrapolated to those for an actual plant, a counterpart test based on the same scenarios as those of the full-height, full-pressure (FHFP) large-scale test facility (LSTF) test was performed. To see the result of differences in the design, scaling approach and facility operational conditions in the systems, the present study examines their effects on the SB transient, particularly for the tests performed at full and reduced pressures. We also identify the occurrence of key thermal-hydraulic phenomena, as well as their possible distortions. Results of the INER integral system test (ISST) facility and LSTF tests showed the common thermal-hydraulic phenomena, such as the secondary coolant boil-off and the subsequent primary coolant saturation, pressurization, coolant inventory depletion and redistribution, and core uncovery caused by coolant boil-off. The sequence and timing of the significant events during the SB transient studied in the RHRP IIST facility are also consistent (in most cases) with those for the SB transient studied in the FHFP LSTF.

Thermal-hydraulic responses in the station blackout experiment conducted at the IIST facility are simulated through the use of the advanced system code RELAP5/MOD3. Typical behaviors occurring in the IIST station blackout transient are characterized by secondary boiloff, primary saturation and pressurization, and subsequent core uncovery and heatup. As the coolant inventory within the steam generator secondary system boils dry, the primary system pressure increases as a result of degradation of the heat removal ability of the steam generator secondary side. This pressurization phenomenon causes the pressurizer safety valve to open and the primary coolant to deplete through the valve, causing the core to eventually become uncovered. The same response can be exactly simulated by the current model. The current calculated results show fairly good agreement with the experimental data, but the timing of the events occurring in the station blackout transient is calculated earlier than the measured value. The overall comparison of key parameters between the calculated results and IIST test data, however, reveals that the current RELAP5/MOD3 model can provide rear sonable station blackout modeling for simulating longterm system behavior.

The simulation capabilities of RELAP5/MOD3 are analyzed and assessed in comparison with the IIST experiments conducted to investigate the system response to the loss of the residual heat removal (RHR) system during midloop operation. Two IIST experiments are simulated; a one-loop test under closed system conditions and a three-loop test with a vent at the top of the pressurizer. Once the RHR cooling system is lost and if alternate heat sinks are not established in time, the primary system will be heated up by the decay power, causing core boiling, system pressurization, and potential core uncovery and fuel heatup. The predicted responses of system parameters by the current model show reasonable agreement with the experimental data. These key parameters consist of the system pressure transient, temperature histories, and variation in the active heat transfer length within the steam generator. The liquid flooding in the pressurizer and the steam generator can also be captured in the current simulation. A periodic fill-and-down cycle developed in the steam generator U-tubes has been observed in the IIST measured data of oscillatory differential pressure across the steam generator. This phenomenon is not simulated in the calculation. However, the calculated differential pressure will follow the experimentalmore » trend and agree qualitatively with the measured data averaged over one fill-and-down cycle. As shown in the comparison of the calculated and experimental data, the overall system responses to the loss-of-RHR system event during midloop operation can be appropriately simulated by the current RELAP5/MOD3 model.« less

A series of experiments dealing with variable secondary-side cooling conditions have been conducted at the IIST facility, including the natural circulation experiments under the secondary-side conditions of normal feedwater, loss of feedwater, and full of air. Different cooling conditions at the secondary side directly affect the primary-to-secondary heat transfer and then may influence the heat removal capability of natural circulation in the primary system. The corresponding analytical work is performed using the RELAP5/MOD3 code. Good agreement is reached both qualitatively and quantitatively between the experimental data and calculated results, demonstrating the satisfactory assessment of RELAP5/MOD3 code compared with the IIST natural circulation experiments. The cooling conditions at the secondary side have no significant effect on the heat removal capability of natural circulation as long as sufficient coolant exists on the steam generator secondary side, based on current IIST data and analytical results. Continuous increase of the core temperature and system pressure is also demonstrated experimentally and analytically in the test with the secondary side dry for the sake of deficient heat transfer capability at the steam generator secondary system.

With high efficiency and very low emissions, fuel cells have been one of the choices of research in current energy development. The Solid Oxide Fuel Cell (SOFC) is a high temperature type fuel cell. It has the characteristic of very high operating temperature 1,027°C (1,300K). The SOFC has the main advantage of very high performance efficiency (over 50%), but also has very high exhaust temperature. Current studies point out that the combination of SOFC and Gas Turbine (GT) can produce efficiency more than 60%. The exhaust temperature of this hybrid power system can be as high as 227–327°C (500–600K). With this waste heat utilized, we can further improve the overall efficiency of the system. A simulation program of SOFC/GT system and the introduction of the concept of Combined Cooling, Heating, and Power System (CCHP) have been used in this study. The waste heat of SOFC/GT hybrid power generation system was used as the heat source to drive an Absorption Refrigeration System (ARS) for cooling. This waste heat enables the SOFC/GT to generate electricity in the system while providing additional cooling and heating capacity. Therefore, we have a combined CCHP system developed using three major modules which are SOFC, GT, and ARS modules. The SOFC module was verified by our test data. The GT and SOFC/GT modules were compared to a commercial code and literature data. Both the single- and double-effect ARS modules were verified with available literature results. Finally, the CCHP analysis simulation system, which combines SOFC, GT, and ARS, has been completed. With this CCHP configuration system, the fuel usability of the system by our definition could be above 100%, especially for the double effect ARS. This analysis system has demonstrated to be a useful tool for future CCHP designs with SOFC/GT systems.Copyright © 2011 by ASME

An experimental investigation of flow transient critical heat flux (CHF) for water at pressures 6.9 and 15.5 MPa is conducted under the simulation of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) conditions. The influence of flow decay rates on local mass velocity and quality at CHF is explored to ensure the adequacy of local-condition approach in the CHF predictions. The assesment of transient upstream correction factor to modify the predictions of steady-state CHF correlations during the BWR flow transients are presented. A mechanistic CHF model based on the dryout of a sublayer and CHF correlations are evaluated in the predictions of PWR flow transient CHF.

In this paper a new approach to bundle critical power predictions is presented. In addition to a very accurate critical heat flux (CHF) model, correction factors that account for the effects of grid spacers, heat flux non-uniformities, and cold walls, which are needed for critical power predictions for practical fuel bundles, are developed. By using the subchannel analysis code COBRA IIIC/MIT-1, local flow conditions needed as input to CHF correlations are obtained. Critical power is therefore obtained iteratively to ensure that the bundle power value from the subchannel analysis will cause CHF at only one point in the bundle. Good agreement with the experimental data is obtained. The accuracy is higher than that of the W-3 and EPRI-1 correlations for the limited data base used in this study. The effects of three types of fuel abnormalities, namely, local heat flux spikes, local flow blockages, and rod bowing, on bundle critical power are also analyzed. The local heat flux spikes and flow blockages have no significant influence on critical power. However, rod bowing phenomena have some effect, the severity of which depends on system pressure, the gap closure between adjacent rods, and the presence or absence of thimble tubes (cold walls).more » A correlation for the influence of various rod bowing phenomena on bundle critical power is developed. Good agreement with experimental data is shown.« less

Conventional fully differential HF CMOS operational amplifiers (op amps) use an extra circuit to perform common-mode feedback (CMFB), which results in signal distortion and/or asymmetry as the signal swing increases. To solve this problem, two fully differential CMOS op amps that are based on merging two single-ended-output op amps are proposed and analyzed. Through the inherent differential-to-single-ended conversion circuitry, CMFB can be efficiently performed without extra circuits. The output signals are not distorted, and the HF performance is not degraded. Simulation and experimental results have confirmed the performance of the proposed new op amps. >

The Universal Serial Bus (USB) technology is now becoming an integral part of the personal computer platform. In this paper, the transceiver architecture and circuits are proposed and implemented for USB2 high-speed mode with 480 Mb/s bandwidth. This physical layer of USB2 consists of transmitter, receiver, two envelope detectors and an all-digital clock recovery/data synchronization blocks. It has been implemented with UMC 0.35 /spl mu/m 1P4M 3.3 V CMOS technology and consumes only 156 mW.

The purpose of this study is to conduct the experiments at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility for evaluation of the performance of the passive core cooling system (PCCS) during the cold-leg small break loss-of-coolant accidents (SBLOCAs). Five experiments were performed with (1) three different break sizes, 2%, 0.5%, and 0.2% (approximately corresponding to 1 1/4”, 2”, and 4” breaks for Maanshan nuclear power plant), and (2) 0.2% and 0.5% without actuation of the first-stage and third-stage automatic depressurization valve (ADS-1 and ADS-3) to initiate PCCS for assessing its capacity in accident management. The detailed descriptions of general system response and the interactions of core makeup tanks (CMTs), accumulators (ACCs), automatic depressurization system (ADS), passive residual heat Removal (PRHR), and in-containment refueling water storage tank (IRWST) on the core heat removal are included. The results show: (1) core long term cooling can be maintained for all cases following the PCCS procedures, (2) the core can be covered for the cases of the 0.2% and 0.5% breaks without actuation of ADS-1 and ADS-3.Copyright © 2002 by ASME

A theoretical critical heat flux (CHF) model based on microlayer dryout and Helmholtz instability for subcooled tube flow under pressurized water reactor operation conditions is first extended to t...