Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was aimed 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.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.     

Evaluation of computational fluid dynamic methods for reactor safety analysis (ECORA)

The objective of the ECORA project is the evaluation of computational fluid dynamics (CFD) software for reactor safety applications, resulting in best practice guidelines (BPG) for an efficient use of CFD for reactor safety problems. The project schedule is as follows: (i) establishment of BPGs for use of CFD codes, for judgement of CFD calculations and for assessment of experimental data; (ii) assessment of CFD simulations for three-dimensional flows in LWR primary systems and containments; (iii) quality-controlled CFD simulations for selected UPTF and SETH PANDA test cases; and (iv) demonstration of CFD code customisation for PTS analysis by implementation and validation of improved turbulence and two-phase flow models.The project started in October 2001 and is for a period of 36 months. The project consortium consists of 12 partners combining thermal-hydraulic experts, code developers, safety experts and engineers from nuclear industry and research organizations. At mid-term, the following results were achieved: (i) BPGs are available for simulations of reactor safety relevant flows. These BPGs have found interest in the European projects FLOMIX-R, ASTAR and ITEM; (ii) important flow phenomena for PTS and containment flows have been identified; (iii) experimental data featuring these phenomena have been selected and described in a standardised manner suitable for simulation with CFD methods; (iii) surveys of existing CFD calculations and experimental data for containment and primary loop flows have been performed and documented; (iv) first results for simulations of PTS-relevant single-phase and two-phase flow cases are available.Documentation is available via the internet at http://domino.grs.de/ecora/ecora.nsf. The models developed within the project are implemented in industrial and commercial CFD software packages and are therefore accessible by industry and research institutions.     

How to make a fuel rod modelling code meet the requirements of design applications

The tasks of the fuel rod designer and the resulting requirements on fuel rod modelling codes are described in the first part. These requirements have increased during recent years in connection with the goal to increase the burnup. Cutting of overconservatism can contribute to this goal, but this needs good and accurately calibrated models. The second part of the paper discusses the special rules which control the use of a fuel rod modelling code in design applications. It is demonstrated how an uncontrolled piling-up of scatter bands and parameter bounds will very rapidly end in hypothetic results. Only a reasonable coordination of unfavourable input parameters leads to “realistic” conservatism from an engineer's point of view. A sound data base is the prerequisite for the respective methods. Further efforts will be necessary to qualify codes and procedures for future probabilistic methodologies.     

The prediction of 2D thermal detonations and resulting damage potential

The main purpose of this paper is to introduce a new concept for the processes responsible for the escalation and propagation of steam explosions. The concept recognizes that initially only a small quantity of coolant around each coarsely premixed melt mass “sees” the fragmenting debris coming off it, hence it is called the concept of “microinteractions”. We also derive the analytical basis for it, define the nature of the requisite constitutive laws and related experimental data, and demonstrate that this concept is essential for the prediction of steam explosion energetics in large-scale premixtures in 2D geometries. We also provide the first numerical illustrations of this concept, implemented in the computer code .m. Further, we provide the first numerical results of steam explosions in large water pools, i.e. ex-vessel explosions. These results reveal two important mechanisms for explosion “venting” and thus for reducing the dynamic loads on adjacent structures. We conclude that, taken together, the “microinteractions” and “venting” make realistic predictions of steam explosion loads feasible and within reach in the near future.     

Novel experimental measuring techniques required to provide data for CFD validation

CFD code validation requires experimental data that characterize the distributions of parameters within large flow domains. On the other hand, the development of geometry-independent closure relations for CFD codes have to rely on instrumentation and experimental techniques appropriate for the phenomena that are to be modelled, which usually requires high spatial and time resolution. The paper reports about the use of wire-mesh sensors to study turbulent mixing processes in single-phase flow as well as to characterize the dynamics of the gas–liquid interface in a vertical pipe flow. Experiments at a pipe of a nominal diameter of 200 mm are taken as the basis for the development and test of closure relations describing bubble coalescence and break-up, interfacial momentum transfer and turbulence modulation for a multi-bubble-class model. This is done by measuring the evolution of the flow structure along the pipe. The transferability of the extended CFD code to more complicated 3D flow situations is assessed against measured data from tests involving two-phase flow around an asymmetric obstacle placed in a vertical pipe. The obstacle, a half-moon-shaped diaphragm, is movable in the direction of the pipe axis; this allows the 3D gas fraction field to be recorded without changing the sensor position. In the outlook, the pressure chamber of TOPFLOW is presented, which will be used as the containment for a test facility, in which experiments can be conducted in pressure equilibrium with the inner atmosphere of the tank. In this way, flow structures can be observed by optical means through large-scale windows even at pressures of up to 5 MPa. The so-called “Diving Chamber” technology will be used for Pressurized Thermal Shock (PTS) tests. Finally, some important trends in instrumentation for multi-phase flows will be given. This includes the state-of-art of X-ray and gamma tomography, new multi-component wire-mesh sensors, and a discussion of the potential of other non-intrusive techniques, such as neutron radiography and magnetic resonance imaging (MRI).     

Two-phase flow models in unbounded two-phase critical flows

Referring to a Loss-of-Coolant Accident situation in LWRs, an analysis of the two-phase region just downstream from the broken pipe, in which a two-phase critical flow takes place, has been performed. A characterization of the flow pattern inside the unbounded two-phase jet is given considering:

• - jet's external shape, obtained by means of photographic pictures;

• - pressure profiles inside the jet, obtained by means of a movable “Pitot” gauge;

• - jet phase's distribution information, obtained by means of X-ray pictures.

Jet's X-ray pictures show the existence of a central high-density “core” gradually evaporating all around, which gives place to a characteristic “dart flow” the length of which depends on stagnation thermodynamic conditions.     

The frequency-dependent elements in the code SASSI: A bridge between civil engineers and the soil–structure interaction specialists

After four decades of the intensive studies of the soil-structure interaction (SSI) effects in the field of the NPP seismic analysis there is a certain gap between the SSI specialists and civil engineers. The results obtained using the advanced SSI codes like SASSI are often rather far from the results obtained using general codes (though match the experimental and field data). The reasons for the discrepancies are not clear because none of the parties can recall the results of the “other party” and investigate the influence of various factors causing the difference step by step. As a result, civil engineers neither feel the SSI effects, nor control them. The author believes that the SSI specialists should do the first step forward (a) recalling “viscous” damping in the structures versus the “material” one and (b) convoluting all the SSI wave effects into the format of “soil springs and dashpots”, more or less clear for civil engineers. The tool for both tasks could be a special finite element with frequency-dependent stiffness developed by the author for the code SASSI. This element can represent both soil and structure in the SSI model and help to split various factors influencing seismic response. In the paper the theory and some practical issues concerning the new element are presented.     

Numeric̀al modelling of ultrasonic scattering by cracks

Two basic aspects of numerical modelling of ultrasonic wave scattering by cracks in solids are discussed in the present paper: how do several standard mathematical procedures to obtain numerical values for scattered fields compare with respect to validity and accuracy, and what conclusions can be drawn from the spatial and temporal structure of these fields for identification and imaging purposes. For simplicity, the model of a plane twodimensional strip-like scatterer with stress-free boundary condition embedded in a linear, homogeneous and isotropic solid has been chosen, and scattering amplitudes for P → P, P → SV, SV → SV, SV → P wave fields are compared with respect to the methods of eigenfunction expansions (EIFU), numerical solution of integral equations (INT), elastodynamic physical optics (EPO) and elastodynamic geometric theory of diffraction (GTD). The first two methods are supposed to yield “exact” numerical results, whereas EPO and GTD rely on physical assumptions to yield simplified approximate but analytical expressions whose range of validity could then be investigated by comparison with the “exact” results giving rise to an estimation of their applicability.The integral equation method is then utilized to yield scattered fields in the time domain for pulsed excitation. Comparison with easily interpretative GTD-results exhibits specific features of these transients in terms of wavefronts and resonances, which form the basis of appropriate imaging and identification algorithms. Preliminary fundamental experiments support the underlying theory.     

Advances in two-phase flow modeling for LMFBR applications

This paper summarizes the development of numerical models for analysis of sodium boiling phenomena in LMFBR which has been carried out at M.I.T. over the last five years.With regard to the degree of spatial averaging, our models use the porous body approach, in two and three-dimensional configurations. One important advantage of this model is the ability to accommodate homogenization of arbitrary-sized regions of interest.From a numerical point of view our basic approach is a semi-implicit method in which pressure pulse propagation and local effects characterized by short time constraints are treated implicitly, while convective transport and diffusion heat transfer phenomena, associated with longer time constants, are handled explicitly. This method remains tractable and efficient in multi-dimensional applications.Both a six-equation (“two-fluid”) model and a four-equation (“mixture”) model have been pursued. A considerable effort has been devoted to the development of constitutive relations. Our current package provides an adequate simulation capability for a wide range of applications.This paper will present the general physical formulation of the codes, the constitutive relations, the general numerical approach, applications, and finally some concluding remarks based on our experience with these codes.     

PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions.RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.The objective of the experiment “loss of feed water”, which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as “integral system effects” and ”natural circulation“. For assessment of the RELAP5 capability to predict the “Integral system effect” phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the “Natural circulation” phenomenon the hot and cold leg temperatures behavior have been investigated.This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Progress for on-line acoustic emission monitoring of cracks in reactor systems

This paper reviews accomplishments and planned tasks for the NRC-sponsored research program concerned with “Acoustic Emission/Flaw Relationships for Inservice Monitoring of Nuclear Reactor Pressure Boundaries”. The objective of the acoustic emission (AE) monitoring program is to develop and validate the use of AE methods for continuous surveillance of reactor pressure boundaries to detect flaw growth. Topics discussed include testing AE monitoring on reactors, refinement of an AE signal identification relationship, study of slow crack growth rate effects on AE generation, and activity to produce an ASTM standard for AE monitoring and to gain ASME code acceptance of AE monitoring.     

蒸汽发生器一级汽水分离器两相流动数值模拟

采用计算流体力学方法并采用非结构化网格和多块网格技术对流动区域进行了网格划分,用两相流模型对蒸汽发生器一级汽水分离器两相流动进行模拟,得到了汽-液两相流动细节,将出口蒸汽干度与蒸汽发生器热工水力专用计算程序计算结果进行比较,吻合良好.     

Development of the evaluation method for crack propagation due to thermal striping

The evaluation code “THERST” was developed to estimate the fatigue crack propagation behavior under thermal stresses due to high-frequency temperature fluctuations, called “thermal striping”. This paper presents fundamental formulations of the evaluation method and verifications of the evaluation method by FEM analyses. Experimental data were obtained in high cycle thermal fatigue tests and the effect of a multiple crack which is characteristic for a crack under thermal stress is discussed in addition to the results of the FEM analyses. A modification of the evaluation method was performed to take multiple crack effects into account.     

Flooding and flow reversal of two-phase annular flow

The flooding and flow reversal conditions of two-phase annular flow are mathematically defined in terms of a characteristic function representing a force balance. Sufficiently below the flooding point in counter-current flow, the interface is smooth and the characteristic equation reduces to the Nusselt relationship. Just below the flooding point and above the flow reversal point in cocurrent flow, the interface is “wavy”, so that the interfacial shear effect plays an important role. The theoretical analysis is compared with experimental results by others. It is suggested that the various length effects which have been experimentally observed may be accounted for by the spatial variation of the droplet entrainment.     

A two-dimensional model for fluid-structure interaction in curved pipes

A “channel” model was developed for the purpose of simulating the interactive fluid-structural response of curved pipes to pressure pulses. Simulation is shown to have been achieved analytically in both the axisymmetric (“breathing”) and transverse (“bending”) modes of interactive behavior.An experimental program which was aimed at the validation of the model is also described. Tests were run in both straight and curved pipe configurations. Comparisons between measurements and model calculations demonstrate the validity of the model within the range of parameters under consideration.The model was implemented into the DISCO code for nonlinear fluid-shell interaction.     

Reliability evaluation of a natural circulation system

This paper discusses a reliability study performed with reference to a passive thermohydraulic natural circulation (NC) system, named TTL-1. A methodology based on probabilistic techniques has been applied with the main purpose to optimize the system design. The obtained results have been adopted to estimate the thermal-hydraulic reliability (TH-R) of the same system.A total of 29 relevant parameters (including nominal values and plausible ranges of variations) affecting the design and the NC performance of the TTL-1 loop are identified and a probability of occurrence is assigned for each value based on expert judgment. Following procedures established for the uncertainty evaluation of thermal-hydraulic system codes results, 137 system configurations have been selected and each configuration has been analyzed via the Relap5 best-estimate code. The reference system configuration and the failure criteria derived from the “mission” of the passive system are adopted for the evaluation of the system TH-R.Four different definitions of a less-than-unity “reliability-values” (where unity represents the maximum achievable reliability) are proposed for the performance of the selected passive system. This is normally considered fully reliable, i.e. reliability-value equal one, in typical Probabilistic Safety Assessment (PSA) applications in nuclear reactor safety. The two ‘point’ TH-R values for the considered NC system were found equal to 0.70 and 0.85, i.e. values comparable with the reliability of a pump installed in an “equivalent” forced circulation (active) system having the same “mission.” The design optimization study was completed by a regression analysis addressing the output of the 137 calculations: heat losses, undetected leakage, loop length, riser diameter, and equivalent diameter of the test section have been found as the most important parameters bringing to the optimal system design and affecting the TH-R.As added values for this work, the comparison has been made between results from this study and results from a previous analysis where the same methodology was adopted for the evaluation of the TH-R of a different passive system named Isolation Condenser (IC). The comparison shows that the current single-phase NC system is ‘more reliable’ than the two-phase IC system. This constitutes a proof of qualification and of consistency for the adopted methodology.     

Local thermal–hydraulic behaviour in tight 7-rod bundles

Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal–hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes.In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.     

Investigation of the two-phase flow in a boiling water reactor using neutron-noise technique

In this paper, the fluctuations of the neutron flux (“neutron noise”) of the Mühleberg BWR are investigated. Above 2 Hz, the noise measured by the in-core neutron detectors is driven exclusively by local fluctuations of the void fraction. Characteristic changes of the neutron-noise signature along the axis can be attributed to changes of flow pattern. By measuring the phase lag between pairs of axially placed neutron detectors, the transit time of the steam between the detectors can be evaluated. The measured transit times are applied to the study of two-phase flow in the core. The neutron-noise method has the advantage of providing in-core information under operational conditions.