Two-Phase Flow Phenomena in Broken Recirculation Line of BWR

When a primary recirculation line of BWR is ruptured, a primary recirculation pump may be subjected to very high velocity two-phase flow and its speed may be accelerated by this flow. It is important for safety evaluation to estimate the pump behavior during blowdown. There are two problems involved in analyzing this behavior. One problem concerns the pump characteristics under two-phase flow. The other involves the two-phase conditions at the pump inlet. If the rupture occurs at a suction side of the pump, choking is considered to occur at a broken jet pump nozzle. Then, a void fraction becomes larger downstream from the jet pump nozzle and volumetric flow through the pump will be very high. However, there is little experimental data available on two-phase flow downstream from a choking plane. Blowdown tests were performed using a simulated broken recirculation line and measured data were analyzed by TRAC-PIA. Analytical results agreed with measured data.     

Analysis and simulation of the flow signal in the recirculation loop of a nuclear power station during a bistable flow event

This study represents a major advance in terms of research into bistable flow and the vibrations of jet pumps inside a vessel of boiling water. Firstly, an analysis is made of a flow signal in a recirculation loop using the Hilbert transform. This study of the temporal flow series concludes that this is a particular case of noise-induced transitions, which, in this case, are induced by turbulence. From this we conclude that the variations in the flow in the recirculation loops are due to turbulence and vortices. Therefore, the cause of the vibrations in the jet pumps is in the flow.Secondly, the identified phenomenon is simulated and the model validated. The model shows a very precise degree of prediction and perfectly reproduces the observed behaviour.Finally, a recommendation to eliminate this problem is set out. The central idea is that if we eliminate turbulence in the central header of recirculation loops, we eliminate or reduce the factors that produce the vortices and turbulence and therefore the bistable flow.     

RELAP4 Modeling for Reverse Flow Property of BWR Jet Pump under Large Break LOCA Condition

It is very important to identify the reverse loss coefficient of BWR jet pump in the evaluation of core inlet flow at the beginning phase of BWR LBLOCA (Large Break Loss-of-Coolant Accident) analyses. Hence, the reverse flow property of jet pump was investigated in relation between the momentum equation, pressure loss coefficient and RELAP4 noding, and a new modeling has been proposed. In the proposed modeling, an equivalent pressure loss coefficient is used to take into account of the effect of accellerating pressure loss by the continuous flow area reduction from the tale pipe to the throat. The effectiveness of this model was studied by analyses for the LOFT 1/6 scale jet pump experiment and typical BWR LBLOCA. It has been, consequently, shown that this proposed model gives better jet pump property than a previous model which is used in the WREM sample problem and which gives very conservative result in core inlet flow and in the peak cladding temperature through whole transient.     

Assessment of SAFER03 Code Using ROSA-III Break Area Spectrum Tests on Boiling Water Reactor Loss-of-Coolant Accidents

The SAFER03 computer code has a newly developed evaluation model for the analysis of various boiling water reactor (BWR) loss-of-coolant accidents (LOCAs). Analyses of the ROSA-III break area spectrum tests in a recirculation line were performed using the SAFER03 to assess the predictive capability of the code for a BWR LOCA. The ROSA-III test facility at the Japan Atomic Energy Research Institute (JAERI) was constructed to simulate a LOCA in a BWR/6-251 plant with 848 fuel bundles and 24 jet pumps. This paper summarizes the assessment results of SAFER03 which predicted the system responses and key phenomena well and the conservative peak cladding temperature (PCT) for recirculation line break tests with different break areas.     

Similarity study of ROSA-III and fist large break cuonterpart tests to BWR large break LOCA

A large break test in a recirculation pump suction line with the assumption of LPCI-diesel generator failure was conducted at the ROSA-III test facility of Japan Atomic Energy Research Institute. A counterpart test was also performed at the FIST test facility of General Electric Company. The objective of the tests was to develop common understanding and interpretation of the controlling thermal-hydraulic phenomena during a large break LOCA of a BWR. The fundamental thermal-hydraulic phenomena in the ROSA-III and FIST tests such as the system pressure, mixture level and fuel rod surface temperatures agreed well. The FIST test had more bundle uncovery than that in ROSA-III since lower plenum steam in the FIST test flowed out of the jet pumps when they uncovered allowing more liquid to drain from the bundle. The ROSA-III and FIST tests and a BWR counterpart were analyzed with the RELAP5/MODI (cycle 018) code. The similarity of the ROSA-III and FIST large break tests to a BWR large break LOCA has been confirmed through comparison of calculated results though they are slightly different in details. It is perhaps desirable to reexamine the DNB and interphase drag correlations and the jet pump models usedin the code.     

An analytical model for the analysis of BWR/4 long-term cooling with either intact or broken jet pump seals

An analytical model for the long-term emergency core cooling (ECC) of a boiling water nuclear reactor (BWR) has been developed. This one dimensional drift-flux model, is an extension of a previous study by Lahey and Kamath [1]. It considers both subcooled and bulk boiling in the core, allows the drift-flux parameters, C₀ and V_gj, to be functions of void fraction (α), and can accommodate both broken and intact jet pump seals. The results of this analytical model compare well with data from simulated full scale BWR fuel rod bundles, and experiments in the PCE facility at RPI.It has been found that the unlikely failure of jet pump seals can have a detrimental effect on the long term cooling capabilities of a BWR/4.     

Experimental investigation of jet fragment diameter in BWR complicated structures using image processing techniques

ABSTRACT

To estimate the status of the Fukushima Daiichi nuclear power plant’s reactor pressure vessel, it is important to understand the breakup and fragmentation behavior of a molten jet of core fuel in the lower plenum of a boiling water reactor (BWR). To clarify the effects of complicated structures on jet breakup and fragmentation, we conducted experiments to visualize jet falling behavior, simulating conditions of a severe BWR accident using a multi-channel experimental apparatus. In this study, we developed a new image-processing method that could recognize fragments in a liquid-liquid flow and estimate fragment diameter of a molten jet in the BWR lower plenum under several inlet conditions. We clarified that complicated structures, such as control rod guide tubes, have little effect on fragment diameter, and that inlet velocity is the dominant factor affecting fragment diameter. This indicates that shear stress would occur at the crest of interfacial waves at the side of the jet, leading to its fragmentation. Finally, the fragment diameters measured in this study show good agreement with the correlation based on shear stress, and a new correlation to predict the jet fragment diameter was developed by fitting the constant correlation value from the experimental result using a simulant fluid.     

Study and application of boiling water reactor jet pump characteristic

RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model.In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible.In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K_d, the suction flow junction form loss coefficients, K_S, the diffuser form loss coefficient, K_e, and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve.     

文丘里管空化限流现象数值模拟和实验研究

压水堆核电厂中发生超流量工况时,要求补水泵下游的文丘里流量计形成空化限流,以保护管道流量不超过限值。采用FLUENT数值模拟和高速摄像实验结合的方法,使用3种不同空化模型,对文丘里管的空化限流现象、空化发展规律和流动特性进行了研究。结果表明:采用Zwart-Gerber-Belamri（ZGB）空化模型和剪切应力输运（SST）k-ω湍流模型可对文丘里管空化限流现象进行较为准确的模拟;空化限流时文丘里管内部将发生周期性空化现象,同时将在壁面回射流的作用下发生小气泡脱落、尾部气泡脱落和空化云整体断裂式脱落等微观流动行为。       

Investigation of BWR LOCA at ROSA-III — Effect of break configuration on system transients

The ROSA-III test facility is a volumetrically scaled ( ) BWR/6 system with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA).Six loss-of-coolant experiments with a break area of 15%, 50% or 200% at the main recirculation pump inlet line were conducted at the ROSA-III test facility with a high pressure core spray failure. A sharp-edged orifice or a long throat nozzle was used as a break plane. It was found in the experiments that the break flow differences between the orifice and the nozzle break configurations with the same flow area were observed only in the subcooled break flow region. Subcooled break flow rate through the orifice was much larger than that through the nozzle. The break configuration difference had little influence on the other system responses, especially on the peak cladding temperature. The applicability of the test results to a BWR/6 has been confirmed through analyses of the 15% break ROSA-III LOCA experiments and BWR/6 LOCAs by using RELAP4/MOD6/U4/J3 code. The experimental results of the ROSA-III LOCA experiments were calculated well by the code, and the same trends were calculated in the BWR analyses.     

Investigation of Break Location Effects on Thermal-Hydraulics during Intermediate Break Loss-of-Coolant Accident Experiments at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25% main recirculation pump suction line break (MRPS-B) experiments, the 21% single-ended jet pump drive line break (JPD-B) experiment and the 15% main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests.

In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop.     

Fuzzy Logic Control Application for BWR Recirculation Flow Control System

There has been increasing necessity for load following and/or AFC (Automatic Frequency Control) operation along with the growth in the share of nuclear power generation in the electric power network. Fuzzy logic control was investigated for application to a BWR recirculation flow control system, in order to obtain a rapid generator power response within an allowable neutron flux overshoot. The proposed controller has two control loops, generator power and neutron flux loop. The fuzzy logic is utilized for weighing these control loops and for controlling the neutron flux. By evaluating the controller performance by numerical simulations on the step response for generator power demand with the model BWR recirculation flow system, more rapid response was obtained than that for conventional proportional plus integral controllers with no neutron flux overshoot beyond alarm activation level.     

Breakup and fragmentation behavior of molten material jet in multi-channel of BWR lower plenum

To estimate the current status of reactor pressure vessel of Fukushima Daiichi nuclear power plant, it is important to clarify the breakup and the fragmentation behavior of molten material jet in boiling water reactor (BWR) lower plenum by a numerical simulation. To clarify the effects of complicated structures on jet breakup and fragmentation behavior, we conducted visualized experiments simulating the severe accident in the BWR by using the multi-channel experimental apparatus. In this study, the jet falling behavior, the jet breakup length, the fragmentation behavior and internal/external velocity profiles of the jet are observed by the backlight method and the particle image velocimetry. It is clarified that the complicated structures prolong the jet breakup length or make the fragments fall together to the lower plenum similar to the bulk state. In addition, it is clarified that strong shearing stress occurs at the crest of interfacial waves at the side of the jet. Finally, the fragment diameters measured in the present study well agree with the theory based on the shering stress by changing the coefficient term. Thus, it is suggested that the fragmentation mechanism is controlled by shearing stress and the fragment diameter can be estimated by adjusting the characteristic value.     

核电厂主给水系统再循环阀设计布置试验研究

对某核电厂主给水系统再循环阀的设计布置进行试验研究,分析引起再循环管道在启泵瞬间突然跳动并伴随爆破声的根源,以及泵组基础错位及振动超标与再循环阀异常情况之间的关系。结果表明,多级笼式调节阀不能布置于有空气残留的高压给水管道中,否则在启动阶段将诱发破坏性水锤。通过优化再循环阀的设计布置,最终解决了主给水系统的非正常启动问题。      

Experimental study on breakup and fragmentation behavior of molten material jet in complicated structure of BWR lower plenum

To estimate the state of reactor pressure vessel of Fukushima Daiichi nuclear power plant, it is important to clarify the breakup and fragmentation of molten material jet in the lower plenum of boiling water reactor (BWR) by a numerical simulation. To clarify the effects of complicated structures on the jet behavior experimentally and validate the simulation code, we conduct the visualized experiments simulating the severe accident in the BWR lower plenum. In this study, jet breakup, fragmentation and surrounding velocity profiles of the jet were observed by the backlight method and the particle image velocimetry (PIV) method. From experimental results using the backlight method, it was clarified that jet tip velocity depends on the conditions whether complicated structures exist or not and also clarified that the structures prevent the core of the jet from expanding. From measurements by the PIV method, the surrounding velocity profiles of the jet in the complicated structures were relatively larger than the condition without structure. Finally, fragment diameters measured in the present study well agree with the theory suggested by Kataoka and Ishii by changing the coefficient term. Thus, it was suggested that the fragmentation mechanism was mainly controlled by shearing stress.     

Undesirable flow behavior in a proposed validation data set

The next generation nuclear plant (NGNP), whose development is supported by the U.S. Department of Energy, will be a very high temperature reactor (VHTR). The VHTR is a single-phase helium-cooled reactor that will provide helium at up to 1000 °C. The prospect of a coolant at these temperatures circulating in the reactor vessel demands that careful analysis be performed to ensure that excessively hot spots are not created and that sufficient mixing of the coolant is obtained. Computational fluid dynamics (CFD) coupled with heat transfer will be used to perform the desired analyses. However, primarily because of the imperfect nature of modeling turbulent flow, any CFD calculations used to perform nuclear reactor safety analysis must be validated against experimental data. Experimental data have been taken in a scaled section of the lower plenum of a prismatic VHTR at the matched index of refraction (MIR) facility at the Idaho National Laboratory. These data were taken with the intent that they be examined for use as validation data. A series of investigations have been conducted to assess the MIR data. Issues that have already been examined include the extent of the required computational domain, the outlet boundary condition, the inlet data and the effect of the turbulence model. One of the jets that flow into the model impacts on a wedge, which represents a portion of a hexagonal graphite block that lines the inner wall of the lower plenum. The nature of the flow below this particular jet is such that a randomly varying recirculation zone is created. This recirculation zone is seen to change in size, causing a relatively long-time scale of motion or disturbance of the flow in the model. It is concluded that such a feature is undesirable in a validation data set, firstly because of its apparent random nature and, secondly, because to obtain an appropriate long-time average would be impractical because of the compute time required. It is predicted computationally that by eliminating the first of the four inlet jets into the scaled model, the resulting recirculation zone is rendered stable.     

BWR online monitoring system based on noise analysis

A monitoring system for during operation early detection of an anomaly and/or faulty behavior of equipment and systems related to the dynamics of a boiling water reactor (BWR) has been developed. The monitoring system is based on the analysis of the “noise” or fluctuations of a signal from a sensor or measurement device. An efficient prime factor algorithm to compute the fast Fourier transform allows the continuous, real-time comparison of the normalized power spectrum density function of the signal against previously stored reference patterns in a continuously evolving matrix.The monitoring system has been successfully tested offline. Four examples of the application of the monitoring system to the detection and diagnostic of faulty equipment behavior are presented in this work: the detection of two different events of partial blockage at the jet pump inlet nozzle, miss-calibration of a recirculation mass flow sensor, and detection of a faulty data acquisition card. The events occurred at the two BWR Units of the Laguna Verde Nuclear Power Plant.The monitoring system and its possible coupling to the data and processing information system of the Laguna Verde Nuclear Power Plant are described. The signal processing methodology is presented along with the introduction of the application of the evolutionary matrix concept for determining the base signature of reactor equipment or component and the detection of off normal operation conditions.     

Simulation model of a nuclear reactor turbine

In applying optimal control theory to a boiling water nuclear reactor (BWR) system which includes the primary recirculation loop, the turbine and their associated auxiliaries, it is necessary to have a linearized mathematical model. Nonlinear and linearized models of a turbine coupled to a BWR, and open-loop responses for specific disturbances are presented.     

BWR small break LOCA counterpart tests at ROSA-III and FIST test facilities

The Japan Atomic Energy Research Institute performed a 2.8% recirculation pump suction line break BWR LOCA test at the ROSA-III test facility. The test was a counterpart test to the 2.8% break test performed at the FIST test facility by the General Electric Company. The objective of the test was to develop a common understanding and interpretation of the controlling phenomena for a small break LOCA of a BWR. Similar phenomena were observed in the two tests in a similar time sequence and with magnitudes. These two test results and a 2.8% break reference BWR LOCA were analyzed using the THYDE-B1 computer code. It was confirmed from the analysis that the THYDE-B1 code has enough capability to analyze a BWR small break LOCA. The applicability of the tests performed at the two facilities to a BWR was also confirmed through the analyses.     

核反应堆冷却剂循环泵全流道三维数值模拟及性能预估

为实现核反应堆冷却剂循环泵（核主泵）的设计自主化及制造国产化,通过CFD数值模拟软件FLUENT,应用RNGk-ε湍流模型及SIMPLE算法对某核主泵进行全流道三维数值模拟,获得了在不同工况下的叶轮内部流动情况,分析了压力和速度分布规律,并进行了性能预估。结果表明,稳态工况下叶片的工作面与背面的压力分布与速度分布合理;泵段压力总体上由进口端至出水端呈递增趋势且在叶轮段出现最大值;在设计工况点得到了较为理想的泵效率与扬程值;随着流量的增加,核主泵的轴功率也逐步增加。模拟结果有助于认识核主泵在运转状态下的内部流场变化情况,为核主泵的国产化前期探索和理论研究提供支持。