Application of BWR Simulator to Neutronics Calculation Tool in Fast Reactors

Scaling considerations have been made for liquid line small break simulation tests of a BWR affected by the stored heat released from loop structures. Necessary scaling parameters were given from lumped energy and mass balances equations. It was possible to compensate for excess stored heat released in the reduced-scale facility by increasing the coolant discharge flow area A_D . Although this compensation accelerated the blowdown process, actual pressure and coolant mass changes of the reactor system could be obtained by altering the time scale of the test results by the ratio of A_D/V (V: vessel volume) between the test loop and the reactor system. A semi-experimental evaluation of the stored heat release rate in the Two Bundle Loop (TBL) during blowdown after the operation of Automatic Depressurization System (ADS) was also presented on the basis of comparison of experimental diathermic and analytical adiabatic blowdown pressure changes. The obtained results were used in the determination of simulated ADS nozzle diameter for small break tests at the TBL.     

Application of signature analysis for determining the operational readiness of motor-operated valves under blowdown test conditions

In support of the NRC-funded Nuclear Plant Aging Research (NPAR) program, Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of motor-operated valves (MOVs).As part of this work, ORNL participated in the gate valve flow interruption blowdown (GVFJB) tests carried out in Huntsville, Alabama, The tests provided an excellent opportunity to evaluate signature analysis methods for determining the operability of MOVs under accident conditions.ORNL acquired motor current and torque switch shaft angular position signauresnon two test MOVs during several GVFIB tests. The reduction in operating “margin” of both MOVs due to the presence of additional value running loads imposed by high flow was clearly observed in motor current and troque switch angular signatures. In addition, the effects of differential pressure, fluid temperature, and line voltage on MOV operations were observed and more clearly understood as a result of utilizing the signature analysis techniques.     

PRTHRUST-J1 code for calculation of blowdown thrust force and its experimental verification

This paper presents an outline of the PRTHRUST-J1 code for calculating blowdown thrust force and gives two numerical examples to show the effectiveness of this code. One numerical example is the problem of saturated steam blowdown. The blowdown thrust forces obtained from the PRTHRUST-J1 code were compared with those of the simplified method of Moody. Fairly good agreement was found between these two results. The other numerical example is the problem of jet discharging tests with stop valve performed in Japan Atomic Energy Research Institute. Analysis was carried out by varying the discharge coefficient. The analytical blowdown thrust force and pressure in the discharging nozzle were compared with experimental results. Qualitative agreement was found between the analytical and experimental results of the blowdown thrust force. Generally speaking, the blowdown thrust forces obtained from the experiment were between the analytical results for discharge coefficients of 1.0 and 0.6.     

An analytical study of the dynamics and stability of a spring loaded safety valve

Spring loaded self-actuating safety valves are employed as part of the overpressure protection systems in various industrial applications. In order to design and predict their performance it is necessary to study the dynamic behavior of the valve over a range of fluid and system conditions. A one-dimensional model has been developed to study the effects of different valve parameters such as the spring-mass characteristics, geometry of internal parts, adjustment ring settings, bellows etc. which influence the dynamic behavior and stability of the valve. Analytical results for steam flow conditions are presented to demonstrate the relative effects of these parameters on the valve opening time, maximum lift, blowdown (upstream pressure differential between the valve opening and closing) and any oscillations of the valve stem. If the valve is not properly backpressure compensated, it may become unstable as the stagnation pressure at the valve inlet decreases. Lowering of the guide adjustment ring position or raising the nozzle adjustment ring generally results in improved stability, shorter valve opening time, higher lift and longer blowdown. The effect of damping on the valve stability is also demonstrated. The model can be used to evaluate the design of safety valves and damping devices to eliminate unstable valve dynamic behavior.     

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.     

LOCA下喷放参数对冷却剂喷放特性影响数值研究

冷却剂喷放过程是失水事故（LOCA）的重要过程之一,研究冷却剂喷放过程的热工水力特性对认识LOCA以及预测事故后放射性源项迁移过程有着重要意义。本文利用FLUNET软件建立冷却剂喷放数值计算模型,并对其进行验证。利用模型研究喷口直径、喷放距离和喷放压力等喷放参数对计算域内流场温度、液滴速度和蒸汽流速等特性的影响。研究结果表明：喷口直径的提高使得喷放参数均有提高;随喷放距离的增大,流场温度和液滴速度先上升后下降,而蒸汽流速先上升后趋于平稳;喷放压力越大,喷放参数的最大值离喷放出口越远,液滴速度和蒸汽流速的最大值随喷放压力的增大逐渐上升,而流场温度最大值没有变化。     

Numerical Study on Influence of Blowdown Parameter on Coolant Blowdown Characteristic in LOCA

The coolant blowdown process is one of the important processes of the loss of coolant accident (LOCA). It is of great significance to study the thermal hydraulic characteristics of coolant blowdown process for understanding LOCA and predicting the migration process of radioactive source term after accident. The numerical simulation model of coolant blowdown was established by FLUENT software and verified. The model was used to study the effects of blowdown parameters such as diameter of nozzle, blowdown distance and blowdown pressure on flow field temperature, droplet velocity and vapor velocity. The results show that the increase of diameter of nozzle increases the blowdown parameters. As the blowdown distance increases, the flow field temperature and the droplet velocity increase first and then decrease, while the vapor velocity first rises and then stabilizes. The greater the blowdown pressure is, the farther the blowdown parameter is from the blowdown outlet. The maximum values of droplet velocity and vapor velocity increase gradually with the blowdown pressure, while the maximum value of the flow field temperature does not change.     

Safety Relief Valve Performance for Two-Phase Flow

The performance of main steam safety relief valve has been evaluated with respect only to the steam. In the present study, two-phase flow and subcooled water blow-out tests with model valves were performed in order to evaluate the valve's characteristics and performance. From the test results, it was made clear that not only for the steam but also for the two-phase flow the measurement data were hardly affected by scaling and also that the reaction force of the fluid to the valve stem was hardly dependent upon the void fraction. Analytical study was performed using the two-phase flow model in the valve. The results of the analysis showed good agreement with the test data. It was shown from the test and analysis results that the reaction force of the two-phase flow and subcooled water to the valve stem was almost as much as that of the steam flow, and the integrity of the safety relief valve could be maintained.     

A coupled RELAP5-3D/CFD methodology with a proof-of-principle calculation

The RELAP5-3D computer code was modified to make the explicit coupling capability in the code fully functional. As a test of the modified code, a coupled RELAP5/RELAP5 analysis of the Edwards–O'Brien blowdown problem was performed which showed no significant deviations from the standard RELAP5-3D predictions. In addition, a multiphase Computational Fluid Dynamics (CFD) code was modified to permit explicit coupling to RELAP5-3D. Several calculations were performed with this code. The first analysis used the experimental pressure history from a point just upstream of the break as a boundary condition. This analysis showed that a multiphase CFD code could calculate the thermodynamic and hydrodynamic conditions during a rapid blowdown transient. Finally, a coupled RELAP5/CFD analysis was performed. The results are presented in this paper.     

Counterpart Tests of SBLOCA on the Effect of Break Location with the ATLAS: DVI Nozzle versus Cold-Leg Breaks

Results from two integral effect tests were compared to discuss the effect of break location between the direct vessel injection (DVI) nozzle and cold leg during the small-break loss of coolant accident (SBLOCA) scenario. One is the SB-DVI-09 test for a 6-inch (50% break area of a DVI nozzle) DVI line break and the other is the SB-CL-06 test for an equivalent break size of cold leg. Both counterpart tests were performed with the same control logic and initial/boundary conditions except for different break locations of the DVI line and cold leg. Experimental results showed that the maximum heater surface temperature increased more with the broken DVI nozzle (SB-DVI-09) than with the broken cold leg (SB-CL-06) due to the delayed and simultaneous occurrence of the loop seal clearing and the momentary decrease in the collapsed water level in the core region.     

PRHRS隔离阀前后破口事故对非能动堆芯冷却系统的影响分析

对于AP型核电站小破口失水事故(SBLOCA)试验进程,国内外有较为一致的认识,但对于相同尺寸破口在不同破口位置对试验进程、非能动堆芯冷却系统的影响仍需进一步研究。本文利用大型非能动堆芯冷却整体试验台架ACME开展了非能动余热排出系统（PRHRS）隔离阀前后破口事故试验工况研究,并以堆芯补水箱（CMT）侧冷管底部破口事故工况作为对比工况。试验结果表明：ACME开展的PRHRS隔离阀前后破口事故模拟工况事故进程符合典型SBLOCA进程,堆芯始终处在良好的冷却状态,非能动堆芯冷却系统的安全性得到有效验证;相同破口尺寸工况下,不同破口位置对事故进程有一定的影响,其中破口位置对CMT液位、安注流量的影响较为关键。对比工况中,PRHRS设备换热量也有较大不同,冷管破口和隔离阀后破口工况较隔离阀前破口工况换热量更大,但PRHRS换热管内部流动换热机理需进一步研究。     

Analysis of Effect of SBLOCA before and behind Isolation Valve of PRHRS on Passive Reactor Core Cooling System

For the test process of small break loss of coolant accident (SBLOCA) of AP type nuclear power plant, there is a more consistent understanding at home and abroad. However, the influence of the same size of the break on the test process and passive core cooling system in different locations still needs further study. In this paper, a large passive core cooling integrated test facility ACME was used to study the break accident test conditions of passive residual heat removal system (PRHRS) before and behind the isolation valve, and the bottom break test of the cold pipe of core makeup tank (CMT) was used as the contrast condition. The test results show that the accident process of PRHRS before and behind the isolation valve is in accordance with the process of SBLOCA, the core is always in a good cooling statement and the safety of passive core cooling system is effectively verified. There is a certain impact on the accident process for the same break size and different break locations, and the location of the break has a key impact on the CMT level and safety injection flow. In contrast, the heat transfer of PRHRS equipment is also quite different. The heat transfer of cold pipe break and break behind the isolation valve is greater than break before the isolation valve, however, the flow and heat transfer mechanism of PRHRS heat exchange tube needs further study.     

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.     

Blowdown thrust force under pipe rupture accident: II. Analytical evaluations of partial break and ramped opening break

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.     

Blowdown thrust force under pipe rupture accident: I. Experimental evaluations of blowdown thrust force and decompression characteristics

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.     

The LOCA air-injection loads in BWR Mark II pressure suppression containment systems

Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.     

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.     

Acoustics impact on steam dryer induced by main steam line break in a boiling water reactor

In boiling water reactor (BWR) design, significant acoustic pressure loads impact the steam dryer hood as a result of the main steam line break outside containment (MSLB) event. When a main steam line breaks, it is assumed that the pipe instantaneously breaks completely open to the ambient environment (double-ended guillotine break). Due to the huge pressure difference between the inside of the reactor pressure vessel (RPV) and surrounding ambient environment, a shock wave will form at the break point and burst into the surrounding environment. At the same time, an expansion wave will travel upstream through the main steam line to the RPV, which results in a pressure reduction on the outside of the steam dryer hood. This expansion wave will create a substantial pressure difference between the two sides of the steam dryer hood with a resultant high stress on the hood. This differential pressure load is the acoustic load used in the structure design evaluations for this event. A key design basis requirement for the steam dryer is to maintain structural integrity during transient, and accident conditions. Demonstration that the steam dryers meet this design basis requires a calculation of the magnitude of the acoustic load on the steam dryer during a MSLB. In this study, computational fluid dynamics (CFD) is used as an alternate calculation method to investigate the phenomenon of MSLB. Transient simulations with fine time steps were carried out. The results show that CFD is a useful tool to provide additional information on the acoustic load as compared to the traditional methods. From the CFD results, the minimum pressure value and its distribution area at different flow times was identified. Through the modeling, an understanding of the detailed transient flow field, particularly the acoustic pressure field near the dryer hood during the MSLB was achieved.     

Fluid mixing and flow distribution in the reactor circuit, measurement data base

Experimental investigations and computational fluid dynamics (CFD) calculations on coolant mixing in pressurised water reactors (PWR) have been performed within the EC project FLOMIX-R. The project aims at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. Measurement data from a set of mixing experiments have been gained by using advanced measurement techniques with enhanced resolution in time and space. Slug mixing tests simulating the start-up of the first main circulation pump are performed with two 1:5 scaled facilities: the Rossendorf Coolant Mixing model ROCOM and the Vattenfall test facility. Additional data on slug mixing in a VVER-1000 type reactor have been gained at a 1:5 scaled metal mock-up at EDO Gidropress. Experimental results on buoyancy driven mixing of fluids with density differences have been obtained at ROCOM and the Fortum PTS test facility.Concerning mixing phenomena of interest for operational issues and thermal fatigue, flow distribution data available from commissioning tests at PWRs and VVER are used together with the data from the ROCOM facility as a basis for the flow distribution studies.In the paper, the experiments performed are described, results of the mixing experiments are shown and discussed. Efforts on computational fluid dynamics codes validation on selected mixing tests applying Best Practice Guidelines in code validation will be reported about in a separate paper.