Computer calculations of air and steam blowdown suppression

We describe a computer code that combines an Eulerian incompressible-fluid algorithm (SOLA) with a Lagrangian finite-element shell algorithm. The former models the fluid and the latter models the containing structure in an analysis of pressure suppression in boiling-water reactors. The code (PELE-IC) calculates loads and structural response from air blowdown and from the oscillatory condensation of steam bubbles in a water pool. The fluid, structure, and coupling algorithms are tested by recalculating problems that have known analytical solutions, including tank drainage, spherical bubble growth, coupling for circular plates, and submerged cylinder vibration. Code calculations are also compared with the results of small-scale blowdown experiments.     

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.     

Design of the HDR experimental program on blowdown loading and dynamic response of PWR-vessel internals

In reactor safety, a postulated breach of the primary coolant circuit, i.e. a reactor vessel blowdown, must not result in an uncontrolled failure propagation. Therefore deformations of the pressure vessel internals caused by the sudden pressure release within the first milliseconds of the blowdown must be within certain limits. To guarantee this, up to now conservative assessments are used based on theoretical models and a number of small scale experiments.In the next two years full scale blowdown experiments will be performed at the former HDR-reactor. They will be used to verify state-of-the-art and newly developed computer codes. In order to obtain reliable information on the safety margins in case of a blowdown, conservative approaches are replaced by the more detailed models of best estimate codes. Especially the coupling between fluid- and structural dynamics, i.e. the feedback of the structural deformations on the blowdown loading will be taken into account.In this paper the design of the HDR-experiments and the accompanying program of computer code development is discussed.     

RELAP5与CFX程序耦合研究

以RELAP5与CFX程序为基础,利用并行虚拟机技术和CFX用户函数进行编程,开发了RELAP5/CFX耦合程序。在单相范围内,首先利用水平圆管喷放问题验证了程序间耦合的正确性。然后,针对双T型接管混合实验进行了模拟,相对于单独的RELAP5程序,耦合程序能更好地揭示真实的物理现象。通过后续的开发完善,耦合程序可用于反应堆安全分析中存在着显著三维混合现象的问题。     

Analysis of blowdown experiment of scale steam generator model

A subcooled blowdown experiment in a scale steam generator (SG) model is analyzed by the use of a fluid-structure computer code (MULTIFLEX). The experimental model simulates the secondary side of a SG with a preheater. The MULTIFLEX code that solves simultaneously a coupled set of one-dimensional hydraulic conservation equations and structural dynamic equations is used to analyze the experiment, taking into account the fluid structure interaction between the secondary coolant and the SG structure, the baffle and tube support plates and the divider plate. The computed values of pressure and wall displacement histories agree well with the experimental data. The success of the analysis supports the use of the one-dimensional MULTIFLEX code to analyses of thermal hydraulic transients in the SG secondary side and the validity of the method for modeling the complicated system of the fluid-structure interactions.     

Pre-and post-test analyses with K-FIX for an HDR blowdown test

The coupled fluid-structure dynamics of a pressurized water reactor core support barrel can be calculated with the K-FIX(3D, FLX) code for blowdown and seismic induced transients. The K-FIX solution method has been used to perform pre- and post-test analyses of a full scale blowdown test at the HDR facility in Frankfurt, Germany. The results verified the accuracy of the fully three-dimensional method, which solves the nonequilibrium, two-fluid equations for the fluid dynamics and the Timoshenko elastic shell equations for the core barrel motion.     

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.     

Analysis of blowdown experiment of 110 scale steam generator model

A subcooled blowdown experiment in a $\text{1}\text{10}$ scale steam generator (SG) model is analyzed by the use of a fluid-structure computer code (MULTIFLEX). The experimental model simulates the secondary side of a SG with a preheater. The MULTIFLEX code that solves simultaneously a coupled set of one-dimensional hydraulic conservation equations and structural dynamic equations is used to analyze the experiment, taking into account the fluid structure interaction between the secondary coolant and the SG structure, the baffle and tube support plates and the divider plate. The computed values of pressure and wall displacement histories agree well with the experimental data. The success of the analysis supports the use of the one-dimensional MULTIFLEX code to analyses of thermal hydraulic transients in the SG secondary side and the validity of the method for modeling the complicated system of the fluid-structure interactions.     

Homogenized model for fluid-structure interaction of the pressurized water reactor core internals during blowdown

A method for the numerical simulation of the pressurized water reactor core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. In order to investigate the global core motion during the blowdown accident, the core model describes the coupled fluid-rod motion with Homogenization methods. The heterogeneous fluid-rod mixture thus is treated as a special continuum with anisotropic material properties. Furthermore, the core model considers elastical rod forces against bending and axial straining and the direct interaction of neighbouring fuel elements, which is a highly nonlinear process due to the finite gaps. Because this effect is very important, two simulation models have been developed and are compared. All these models have been implemented into the blowdown code FLUX-4. With the new code version FLUX-5 the PWR-blowdown is parametrically investigated.     

Blowdown force analysis of piping system under LOCA conditions using BLOWDOWN code

The BLOWDOWN code was developed for blowdown force analysis of piping system under LOCA conditions. This is a post-processor of the thermal-hydraulic analysis code RELAP4/MOD6. The results obtained from the RELAP4/MOD6 code are converted into blowdown forces by the BLOWDOWN code.In the paper, the physics and algorithms of the BLOWDOWN code are outlined. Some numerical examples are also presented to show the effectiveness of the code.     

A Validation of ATR LOCA Thermal-hydraulic Code with a Statistical Approach

When cladding temperatures are measured for a blowdown experiment, cladding temperatures at the same elevation in the fuel bundle have usually some differences due to eccentricity of the fuel bundle and other reasons such as biased two-phase flow. In the present paper, manufacturing tolerances and uncertainties of thermal-hydraulics are incorporated into a LOCA code that is applied with the statistical method. The present method was validated with the results of different blowdown experiments conducted using the 6 MW blowdown facility simulating the Advanced Thermal Reactor (ATR). In the present statistical method, the code was modified to run fast in order to calculate the blowdown thermal-hydraulics a lot of times with the code using different sets of input data. These input data for sizes and empirical correlations are prepared by the effective Monte-Carlo method based on the distribution functions deduced by the measured manufacturing errors and the uncertainties of thermal hydraulics. The calculated curves express uncertainties due to the different input deck. The uncertainty band and tendency of the cladding temperature were dependent on the beak sizes in the experiment. The measured results were traced by the present method.     

Fluid-structure analysis for the HDR blowdown and snapback experiments with FLUX

FLUX is a special purpose code to analyse three-dimensional fluid-structure interactions during blowdown of a pressurized water reactor. Such a blowdown has been simulated in the HDR experiments. For the first series of blowdown experiments and for snapback experiments in the same facility the results of precomputations are reported and compared with the experimental results. Refinements are desirable with respect to two-phase-damping of pressure waves in the blowdown pipe and vessel wall flexibility. The general quantitative agreement between measurement and computation is satisfactory.     

Pressure suppression pool mixing in passive advanced BWR plants

In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data.     

多用途流固耦合结构动力学计算程序研究

结合Eulerian流体弹塑性计算方法和Lagrangian结构动力学计算方法的特点,自主提出了一种比较通用的Euler-Lagrange流固耦合数值模拟计算技术,并编制形成了一套包含多种材料模型、状态方程及加载方式的多用途流固耦合计算分析程序。测试算例表明,自主开发的流固耦合算法使得流固间的相互作用可以更好地被监控、评价,从而可以更好地分析流场传递到结构中的载荷作用及流固相互作用过程。同时可为核工程中的爆炸、冲击载荷下结构安全性评价提供了一个较好的分析平台,也可为相关多物理场耦合计算技术的开发提供基础平台。      

LOCA hydroloads calculations with multidimensional nonlinear fluid-structure interaction

A methodology for realistically analyzing three-dimensional fluid-structure interaction effects and the resulting hydrodynamic loads during the subcooled portion of a hypothetical loss of coolant accident (LOCA) in a pressurized water reactor (PWR) is discussed. The methodology uses a hydrodynamic computer program, STEALTH 3D, coupled with a structural response program, WHAMSE 3D, to calculate the dynamic interaction of fluid and structure during a reactor vessel blowdown. This coupled program is user oriented and highly versatile in modeling the various components in complex reactor systems. Assessment of the methodology is provided by STEALTH/WHAMSE 3D calculations of blowdown tests in the German Battelle-Frankfurt RS-16B facility (Test DWR5) and the HDR facility (Test V31.1). The calculations are described and the results compared with experimental data. Agreement between the calculational results and experimental data is extremely good.     

Application of the COBRA/TRAC best-estimate methodology to the AP600 large-break LOCA analysis

The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safety systems that perform the same function as the active emergency core cooling systems (ECCSs) on the current reactors. In order to verify the effectiveness of the AP600 design features for mitigation of a postulated large-break loss-of-coolant accident (LOCA), the recently United States Nuclear Regulatory Commission (USNRC)-approved best-estimate LOCA methodology (BELOCA) was applied to perform the AP600 standard safety analysis report large-break LOCA analysis. The applicability of the COBRA/TRAC code to model the AP600 unique features was validated against cylindrical core test facility (CCTF) and upper plenum test facility (UPTF) downcomer injection tests, the blowdown and reflood cooling heat transfer uncertainties were re-assessed for the AP600 large-break LOCA conditions and a conservative minimum film boiling temperature was applied as a bounded parameter for blowdown cooling. The BELOCA methodology was simplified to quantify the code uncertainties due to local and global models, as well as the statistical approximation methods, with the other uncertainties being bounded by limiting assumptions on the initial and boundary conditions. The final 95th percentile peak cladding temperature (PCT^95%) was 1186 K, which meets the 10CFR50.46 criteria with a considerable margin. It is therefore concluded that the AP600 design is effective in mitigation of a postulated large-break LOCA.     

Vibration of nuclear fuel bundles

Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors (LMFBRs) is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code (AMASS) is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods.     

Experimental and computed results for fluid-structure interactions with impacts in the HDR blowdown experiment

The HDR experimental facility has been used for several blowdown experiments in order to study fluid-structure interactions and loadings on the pressure vessel internal structures of a pressurized water reactor. We have developed the code FLUX to analyse the motions in the initial blowdown period.This paper describes a new type of HDR experiments (V34) and compares the experimental results with the FLUX-code results. As novel feature, the core barrel is not rigidly clamped to the vessel as in earlier experiments but supported with gaps such that the core barrel can move freely upwards for about 2 mm and horizontally for 0.3 mm at the upper flange. At the lower core-barrel edge, snubbers restrict the horizontal motion to about + 1.4 mm and −2.8 mm.The experimental results show that the core barrel is deflected sidewards until it hits the snubber at the lower edge and then swings back to hit the opposite snubber. By this some kinetic energy is lost due to plastic snubber deformations. At the same time, the measurements show that the core barrel lifts rather uniformly from its support upwards until it hits the upper constraint. Several bounces up and down are observed until the core barrel becomes fixed probably due to friction from the side.This situation has been pre- and post-computed with the new FLUX-version which contains a very effective algorithm to treat supports with gaps and resultant impacts. For treatment of plastic supports, a simple model is added. Pre-computations were not meaningful because of large deviations in the pre-estimated initials gaps. However the computed pressure-field is not influenced very much by these parameters and predicted very well. This was favoured by the isothermal fluid initial conditions. Post-computations show sufficient agreement with respect to computed core barrel motion. The axial motion is described very well. Some problems remain which are due to the model for the upper flange support.Impacts do not results in greatly enlarged loadings, strains or accelerations for this situation.     

Vapor generation model for flashing in the initial blowdown phase

A vapor generation model for flashing in the initial blowdown phase is proposed based on a wall nucleation theory and a bubble transport model. Comparisons are made between the proposed model and the TRAC-PF1 model by using the MINCS code through analyses of three blowdown experiments with different scales. The present model well predicts the pressure undershoot in the vessel, while the TRAC model can not predict this typical thermodynamic nonequilibrium phenomenon.     

Numerical Approach to the Safety Evaluation of Sodium—Water Reaction

A numerical simulation method of multi-dimensional and multi-phase reacting flow (SERAPHIM code) has been developed to evaluate the sodium-water reaction (SWR) phenomena in a steam generator of liquid metal fast reactor (LMFR). A compressible multi-fluid and one-pressure model is adopted and pressure and velocity fields are updated simultaneously by the HSMAC method. Two types of reaction models are considered; one is a surface reaction and the other is a gas-phase reaction. The surface reaction model assumes that water vapor reacts with the liquid sodium at the gas-liquid interface. If chemical reaction heating is large enough, liquid sodium is vaporized resulting in a gas-phase reaction. In the surface reaction, the reaction rate is assumed to be infinitely large. Several overall reaction equations are taken into account in the gas-phase reaction and the reaction rates are described in the form of the Arrhenius law. In the present study, adequacy of the analytical procedures for compressible multi-phase flow is validated by a benchmark calculation of the Edwards pipe blowdown problem. As a numerical example, two- and three-dimension analyses of the single-tube geometry and the two-dimension analyses of the 43-tubes geometry are carried out. It is concluded that the numerical quantification of the SWR accident by the SERAPHIM code is practicable and further use of the SERAPHIM code is useful to resolve safety issues immanent in the SWR.