小型压水堆下腔室交混特性实验研究

为研究小型压水堆下腔室的交混特性,本文基于比例模化方法,开展小型压水堆1∶3比例模型水力学实验,通过测量溶液浓度变化,获得在冷管流量均衡和非均衡工况下堆芯入口的交混因子矩阵。研究结果表明,均衡流量工况下,冷管流量的变化对堆芯入口交混因子矩阵未产生明显影响;非均衡流量工况下,靠近出口管的燃料组件交混因子受流量不均衡的影响较大,而中心区域的交混因子变化幅度较小。由此可见,小型压水堆在均衡流量下具有较稳定的下腔室交混特性,而在非均衡工况下需要重点关注出口附近燃料组件交混特性的变化。      

CAP1400反应堆堆芯入口流量分配试验研究

反应堆堆芯入口流量分配是反应堆水力性能研究的重要内容之一,其与堆芯热裕量和燃料组件燃料棒的流致振动密切相关,从而影响反应堆的运行。CAP1400反应堆堆芯入口流量分配试验是验证CAP1400反应堆结构设计与分析的一个重要环节,旨在验证CAP1400反应堆堆芯入口流量分配的均匀程度。本文通过1/6比例模型试验,获得无均流板结构工况和带均流板结构3种工况（均匀流量工况、非均匀流量工况、偏回路流量工况）下CAP1400反应堆堆芯入口流量分配结果,并进行了各工况下流量分配均匀程度的分析。试验结果表明,CAP1400反应堆堆芯入口具有较好的流量分配效果。     

反应堆压力容器直接安注热分布特性研究

我国非能动系列压水堆将应急冷却系统冷却水的注入管道直接连接于压力容器上,与传统的冷管段安注不同,这种安注方式被称之为反应堆压力容器直接安注。本文以安注条件下的反应堆压力容器为研究对象,采用物理实验与数值分析结合的方法,对安注流体在压力容器表面形成的热分布形态进行研究。研究发现,不同于传统的主管道冷段斜接管安注方式,直接安注条件下安注流体在下降环腔中的分布形态接近于等腰三角形。以实验结果为基础,结合数值计算验证,发现了压力容器热分布角与流速比成正比关系,并进一步提出了安注流体分布计算模型,从而为反应堆安全设计提供参考。      

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

反应堆压力容器下降段水-蒸汽CCFL实验与模型研究

为探究反应堆压力容器下降段在喷放末期冷段安注过程中的水-蒸汽逆流特性,建立下降段逆向流动限制(CCFL)模型,开展了基于压力容器模化本体的下降段CCFL实验研究以及建模分析。通过实验研究获得了不同入口安注水流量、安注水过冷度、堆芯蒸汽流量等条件下的下降段环腔内的安注特性数据,并基于实验数据进行了CCFL建模分析。结果表明,开始发生CCFL的蒸汽无量纲流速与入口安注水无量纲流速呈现正相关,基于无量纲流速建立的模型斜率与入口安注水无量纲流速呈现高度指数关联。本文建立了适用于从不发生CCFL至不完全CCFL,再到完全CCFL的下降段水-蒸汽气液逆流全过程预测模型。     

一维自然循环比例分析的理论模型

整体性能试验研究是验证先进非能动压水堆核电站堆芯冷却系统设计有效性的核心技术,一回路系统两相自然循环热工水力特性比例分析是确定整体性能试验装置尺度的主要理论依据。以一维漂移流模型为基础,对整个一回路两相自然循环系统控制方程积分,并求得稳态解,由此获得了系统的流动条件。应用初始流动条件与边界条件,对两相自然循环系统控制方程直接无量纲化,最终得到了整体性能试验装置与实际非能动电站热工水力特性的相似准则。     

Two-phase flow in a full-scale loop seal facility

An experimental program has been carried out to study two-phase behaviour of a PWR cold leg loop seal during loss-of-coolant accidents. The experimental facility comprises a full-scale cold leg with a reactor coolant pump simulator. Three separate air/water test series were performed to determine the onset of slugging in the horizontal pipe, the residual water mass and the total two-phase pressure drop in the loop seal.The results of flow regime transition experiments have been compared with smaller-scale experiments and with theoretical predictions to evaluate scaling criteria. The strong hysteresis of transitions found between the stratified and slug flow regimes depends on the loop seal geometry and U-tube oscillations.     

Algorithm for Cause Analysis Using Cause Consequence Model

In an integral system test on the reflood phenomena of a PWR LOCA with the Cylindrical Core Test Facility (CCTF), a high pressure drop was observed through the broken cold leg of the pressure vessel side. In order to understand the pressure drop and to assess the applicability of the CCTF result to the LOCA analyses for PWRs, the pressure drop characteristics through the broken cold leg is analyzed with CCTF and FLECHT SET data. The high pressure drop is explained quantitatively with the homogeneous flow model of the two-phase flow. The difference of the pressure drop between the FLECHT SET and the CCTF is attributed to the differences of the flow area scaling of the broken cold leg and the ECC water injection method. It is confirmed analytically that the high pressure drop as in the CCTF tests is expectable in a PWR system with a cold leg break due to the pressure losses at the broken cold leg nozzle and the break.     

Analysis of an RPV upper head SBLOCA at the ROSA facility using TRACE

Inspections of existing nuclear power plants have pointed out the possibility that small break loss-of-coolant accidents (SBLOCAs) could be initiated by a small break located in the upper head (UH) of the reactor pressure vessel (RPV). Such type of breaks has been the subject of investigation in some of the tests carried out in the framework of the OECD/NEA ROSA test program for safety research and safety assessment of light water reactors. The ROSA/LSTF test facility simulates a Westinghouse design PWR with a four-loop configuration and 3423 MWth. Areas, volumes and power are scaled down by a factor of 1:48 while the elevations are kept at full height. Only two loops, sized to conserve the volume scaling (2:48), are simulated. The present paper is focused on test 6-1 that simulated a RPV upper head SBLOCA with a break size equivalent to 1.9% cold leg break. The experiment assumes a total failure of the high pressure injection system (HPIS) and a loss of off-site power concurrent with the scram. The main purpose of the present study is the assessment of the capabilities of the best estimate system code, TRACE, to reproduce and understand the physical phenomena involved in this type of SBLOCA scenarios. Special attention was dedicated to the modelling of the leakage flows, necessary to correctly simulate the distribution of the water inventory in the primary side. In addition, the particular location of the break in test 6-1 allows the verification of the chocked flow model in the same way as for a separate-effect test.     

Design and analysis of a thermal hydraulic integral test facility for Bushehr nuclear power plant

In this paper, design and analysis of a thermal hydraulic integral test facility for Bushehr Nuclear Power Plant (NPP) is presented. The Bushehr Integral Test Facility (BITF) is a test facility designed to model the thermal-hydraulic behaviours of the Bushehr NPP (VVER-1000) pressurized water reactors currently in use in IRAN. These reactors have unique features that differ from other PWR designs. The BITF simulates the major components and systems of the reference NPP, making it possible to examine postulated small and medium break a loss of coolant accidents (LOCAs) and operational transients. The BITF is a volume-scaled model (1:1375). To ensure that gravitational forces remain equal to those in the reference reactor, the major components and systems in the BITF preserve 1:1 elevation equivalence to the reference reactor. The facility has four loops (each one consists of a hot leg, a steam generator, a loop seal, a main circulation pump and a cold leg), a pressurizer connected via two surge line to the hot leg of the loops 2, 4, the emergency-core-cooling system (ECCS) which is provided by an active pump simulating high and low pressure injection systems, and four hydro-accumulators. The report also contains a comparison between experimental data of PSB test facility and RELAP5 calculations of BITF facility under steady state condition of the reactor power 15% from the nominal.     

Study on post-blowdown of hot leg large break loss-of-coolant-accident in the view of mass and energy release analysis

This study conducted mass and energy release experiment for the hot leg large break loss-of-coolant-accident (LBLOCA) during post-blowdown with an integral test facility, Seoul National University Facility (SNUF), and its RELAP5 simulation. This facility simulated the Young Kwang Nuclear Power Plant Units 3 and 4 (YGN3&4) with volume ratio of 1:1140 based on Ishii's three level scaling. The experiments showed that safety injection (SI) water refilled the cold leg first and later the core. The SI water was vaporized in the core, which resulted in the repressurization of the reactor. This increase in pressure drove the water in the cold leg to flow up to half the height of the U tubes. However, since the water was drained back not long after, the release through the SG side broken section by evaporation was negligible. The SNUF experiment was assessed by RELAP5 simulations. Overall, the analysis of the post-blowdown phase showed that the transient of the primary pressure can be properly simulated by RELAP5 when a sufficient heat source is modeled. Consequently, the releases from reactor side broken section and steam generator side broken section were properly predicted. The pressure rise by steam generation in the core was partially well predicted. The release from the steam generator side broken section was predicted to be small except when there exists a large pressure difference between the primary system and the break boundary.     

Advanced DVI for ECC direct bypass mitigation

An ECC direct bypass fraction during a late reflood phase of a LBLOCA is strongly dependent on the characteristics of the cross flow and the geometrical configuration of a DVI in the downcomer of a pressurized light water reactor. The important design parameters of a DVI are the elevation, the azimuthal angle, and the separator to prevent a steam-water interaction. An ECC sub-channel to separate or to isolate an ECC water from a high-speed cross flow is one of the important design features to mitigate the ECC bypass phenomena. A dual core barrel cylinder as an ECC flow separator is located between a reactor vessel and a core barrel outer wall in the downcomer annulus. A new narrow gap between the core barrel and the additional dual core barrel plays the role of a downward ECC flow channel or an ECC flow separator in a high-speed cross flow field of the downcomer annulus. The flow zone around a broken cold leg in the downcomer annulus has the role of a high ECC direct bypass due to a strong suction force while the wake zone of a hot leg has the role of an ECC penetration. Thus, the relative azimuthal angle of the DVI nozzle from the broken cold leg is an important design parameter. A large azimuthal angle from a cold leg to a hot leg needs to avoid a high suction flow zone when an ECC water is being injected. The other enhancing mechanism of an ECC penetration is a grooved core barrel which has small rectangular-shaped grooves vertically arranged on the core barrel wall of the reactor vessel downcomer annulus. These grooves have the role for a generation of a vortex induced by a high-speed cross flow. Since the stagnant flow in a lateral direction and rotational vortex provides the pulling force of an ECC drop or film to flow down into the lower downcomer annulus by gravity, the ECC direct bypass fraction is reduced when compared to the current design of a smoothed wall. An open channel of grooves generates a stagnant vortex, while a closed channel of grooves creates an isolated ECC downward flow channel from a high-speed lateral flow. In this study, new design concepts for a dual core barrel cylinder, grooved core barrel, and a reallocation of the DVI azimuthal angle are proposed and tested by using an air-water 1/5 scaled air-water test facility. The ECC direct bypass reduction performances of the new design concepts have been compared with that of the standard type of a DVI injection. The azimuthal angle of the DVI nozzle from a broken cold leg varies from −15° to +52° toward a hot leg. The test results show that the azimuthal injection angle is an effective parameter to reduce the ECC direct bypass fraction. The elevation of the DVI nozzle is also an important parameter to reduce the ECC direct bypass fraction. The most effective design for reducing the ECC direct bypass fraction is a dual core barrel. The reduction fraction when compared to the standard DVI is about −30% for the dual core barrel while it is −15% for the grooved core barrel.     

Three-dimensional analysis of flow characteristics on the reactor vessel downcomer during the late reflood phase of a postulated LBLOCA

A three-dimensional CFD analysis has been performed on the flow characteristics in the reactor vessel downcomer during the late reflood phase of a postulated large-break loss-of-coolant accident (LBLOCA), in order to validate the modified linear scaling methodology that was applied in the MIDAS test facility of Korea Atomic Energy Research Institute. The vertical and circumferential velocity similarities are numerically tested for the 1/1 and 1/5 linear scale models for the APR1400 reactor vessel downcomer. The effects of scale on flow patterns, pressure and velocity distributions, and the impinging jet behavior are analyzed with the FLUENT code. In addition, a simplified half cylinder model with a single emergency core cooling (ECC) nozzle is numerically tested to investigate the scale effect on the spreading width and break-up of ECC water film. The qualitative and quantitative results indicate that the 1/5 modified linear scale model of the reactor vessel downcomer would reasonably preserve the hydrodynamic similarity with APR1400.     

Simulation of mixing effects in a VVER-1000 reactor

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found.Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created.The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.     

安全棒系统试验样机冷、热态性能试验研究

安全棒系统是空间核反应堆的关键设备之一,它具有结构紧凑、传动精度高、与反应堆容器连接接口多、工作温度高等特点。通过采用全尺寸的安全棒系统试验样机,确定了冷、热态性能试验方案,设计了专用的试验装置开展冷、热态性能试验。试验结果表明,安全棒系统试验样机运行正常,性能达到设计要求,为试验样机的抗震试验提供了条件,也为安全棒系统后续设计及试验装置的改进提供了参考依据。     

Scaling of passive condenser system for separate effect testing

The passive containment cooling system (PCCS) of the simplified boiling water reactor (SBWR) is a passive condenser system designed to remove energy from the containment for long term cooling period after a postulated reactor accident. Depending on pressure condition and noncondensable (NC) gas fraction in drywell (DW) and suppression pool (SP), three different modes are possible in the PCCS operation namely the forced flow, cyclic venting and complete condensation modes. The prototype SBWR has total of six condenser units with each unit consisting of hundreds of condenser tubes. Simulation of such prototype system is very expensive and complex. Hence a scaling analysis is used in designing an experimental model for the prototype PCCS condenser system. The motive for scaling is to achieve a homologous relationship between an experiment and the prototype which it represents. A scaling method for separate effect test facility is first presented. The design of the scaled test facility for PCCS condenser is then given. Data from the test facility are presented and scaling approach to relate the scaled test facility data to prototype is discussed.     

Experimental investigation of natural circulation instability in a BWR-type small modular reactor

The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV). Specifically, the NMR is one third the height and area of a conventional BWR RPV with an electrical output of 50 MWe. Experiments are performed in a well-scaled test facility to investigate the thermal hydraulic flow instabilities during the startup transients for the NMR. The scaling analysis for the design of natural circulation test facility uses a three-level scaling methodology. Scaling criteria are derived from non-dimensional field and constitutive equations. Important thermal hydraulic parameters, e.g. system pressure, inlet coolant flow velocity and local void fraction, are analyzed for slow and fast normal startup transients. Flashing instability and density wave oscillation are the main flow instabilities observed when system pressure is below 0.5 MPa. And the flashing instability and density wave oscillation show different type of oscillations in void fraction profile. Finally, the pressurized startup procedure is recommended and tested in current research to effectively eliminate the flow instabilities during the NMR startup transients.     

PTS瞬态流动与换热数值模拟研究

为了研究压水堆因“直接安注”冷水注入压力容器下降环腔而导致的承压热冲击（PTS）热工水力问题,基于1:10比例模型,应用计算流体力学软件FLUENT5.4进行了紊流流动换热的数值模拟分析,同时进行了常压瞬态传热实验研究。针对下降环腔折算流速0.5 m/s,安注流速10m/s的典型工况,研究了安注水开启后下降环腔内的瞬态流动换热特性,数值模拟与实验结果吻合良好。考察了压力容器安注接管出口区环形焊缝区及堆芯段筒体中子强辐照区所承受的热冲击状况,基于稳态流动研究了下降环腔内流体混合特性及流动机理,为热冲击分析提供参考。     

Scaling for the ECC bypass phenomena during the LBLOCA reflood phase

As one of the advanced design features of the APR1400 (Advanced Power Reactor), a direct vessel injection (DVI) system is adopted instead of the conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is underway. In this paper, a new scaling method, using the time and velocity reduced “modified linear scaling law”, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in the PWR downcomer.     

铅基快堆自然循环实验台架比例分析方法研究

铅基快堆具有良好的自然循环能力，研究其自然循环特性对提高反应堆固有安全性具有重要价值，而比例分析方法是建立合理可行铅基快堆自然循环实验台架的理论基础。本文通过无量纲化典型自然循环铅基快堆一回路系统的流体控制方程，确定主要的无量纲相似准则群；基于所构建的无量纲相似准则数对小型自然循环铅基快堆SNCLFR-10开展比例分析，获得双环路单相自然循环实验台架的几何和热工水力设计参数；对比分析额定工况下SNCLFR-10和缩比实验台架的关键热工水力参数，开展铅基快堆自然循环实验台架比例分析方法验证。研究结果表明，SNCLFR-10和缩比台架的关键热工参数模拟结果比值与理论推导比例关系吻合良好，建立的铅基快堆自然循环实验台架比例分析方法合理可行。