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Today, Small Modular Light Water Reactors based on the use of natural circulation for removing core power in steady operational and transient condition, are one of the key design options for the deployment of nuclear reactor technology considering its advantage in term of inherent safety, design simplicity, simplified parallel construction, reduction of construction time, and reduction in finance and operation cost. In order to develop deterministic safety analyses, best estimate thermal-hydraulic system codes need to be validated against natural circulation phenomena typical of an integral design during steady-state and transient conditions. The goal of this paper is to summarize the results of the validation activity of the TRACE code against the experimental database developed in the OSU-MASLWR facility, designed to thermal hydraulically characterize the Multi-Application Small Light Water Reactor (MASLWR) basis for the NuScale design. The TRACE code qualitative and quantitative accuracy, by using the Fast Fourier Transform based Methods (FFTBM), will be presented and discussed. Particular attention will be focused on the capability of the code to predict the natural circulation in the integral design, the primary-to-secondary heat transfer in helical coil steam generator, and the passive primary-to-containment coupling. The tests selected for the analyses are the OSU-MASLWR-OOl, aiming at characterizing the thermal-hydraulic coupling of the passive primary-to-containment in design basis accident conditions, and the OSU-MASLWR-002 test aiming at characterizing the single-phase natural circulation and the primary-to-secondary heat transfer in a helical coil steam generator. © 2019 American Nuclear Society.

Recent developments of special models and correlations extended the system TH codes capabilities to simulate 3D phenomena. A code assessment process is always needed whenever the new code features are intended for nuclear reactor design and/or licensing applications. The Korean Thermal-Hydraulic code MARS (developed by Korea Atomic Energy Research Institute) experiences such an improvement, extending the 1D flow field formulation. In this respect the present paper describes the activity conducted to assess the 3D features of the MARS code by independent users. The adopted experimental data are gathered in a test conducted at the ROCOM (Rossendorf Coolant Mixing Model) experimental facility, which reproduced a pump start-up scenario. In addition, to support the interpretation of experimental data and system code results, a CFD analysis has been also performed. The assessment activity includes a comparison with RELAP5-3D© code, a set of sensitivity calculations and the use of the FFTBM package.

Pressurized Thermal Shock (PTS) and Direct Contact Condensation (DCC) were identified by the European project EUROFASTNET as two of the most important industrial needs related to nuclear reactor safety where CFD may bring a real benefit. One typical PTS scenario limiting the Reactor Pressure Vessel (RPV) lifetime is cold water Emergency Core Cooling (ECC) injection into the cold leg during a hypothetical SB-LOCA. The injected water mixes with the hot fluid present in the cold leg and the mixture flows towards the downcomer where further mixing with the ambient fluid takes place. Such a scenario may lead to high thermal gradients in the structural components and consequently to thermal stresses. Therefore, the loads upon the RPV must be reliably assessed. The NURESIM sub-project 2 (Thermohydraulics) Work Package 2.1 focuses on a two-phase flow configuration resulting from a partially or fully uncovered cold leg. In the case of a partially uncovered cold leg, a stratification of cold water on the bottom of the cold leg with counter-current flow of hot water and steam on top of this cold-water layer may occur. There is mixing between hot and cold water. Condensation takes place at the free surfaces between steam and water, e.g. at the cooling water jet. Mixing and condensation are strongly dependent on the turbulence in the fluids. Reliable numerical simulations are required. Two-phase PTS constitutes one of the most challenging exercises for a Computational Fluid Dynamics (CFD) simulation. Presently available CFD tools are not yet able to reproduce all the separate phenomena taking place in the cold leg and the downcomer during the ECC injection, let alone an accurate simulation of the whole process. Improvements of the two-phase modelling capabilities have to be undertaken to qualify the codes for the simulation of such flows. A really accurate simulation of all the phenomena that occur in the scenario will only be possible in the far future and a step-by-step improvement of the quality of the forecasts is necessary. However, a reasonable prediction of the most important phenomena may be reached in a short or medium term and the use of CFD in industrial studies related to PTS is already possible in the frame of some limitations.

Extensive analytical work: - 65 pre-test calculations - 45 post-test calculations Comparison of all code results against exp data Main findings: - 3rd Group (steady pump operation + tracer injection) - Perturbation morphology correctly described - Quantitative discrepancies (degree of mixing) - Can be handled by proper accuracy and uncertainty evaluation 2nd Group (tracer slug + onset of NC) - Crucial role played by density effects

This report summarizes the results of the NURESIM project for the work package 2.1 “Pressurized Thermal Shock (PTS)”. It includes summaries of the single tasks done by the partners involved in this work package. In the Introduction chapter some more general information on the PTS issue is given, which should help to clarify the integration of the single activities. Since the PTS scenario involves different flow situations, for which also different modelling approaches are necessary, the tasks are sorted according to these flow situations. The relation of the work done to the general aim of the NURESIM project, which is to establish a new code platform, is indicated by assigning the activities to 6 different types. The results achieved in the PTS work package are in agreement with the expectations to the NURESIM project. The conclusion drawn from the single investigations and recommendations for future work are discussed in a separate chapter. It was shown, that for further improvement of the CFD-code capabilities for the two-phase PTS case new well-instrumented experimental data are needed especially for condensation at the surface of a sub-cooled liquid jet in a steam environment as well as on free surfaces, turbulence production and bubble entrainment below the jet and mixing in a stratified flow. Integral experiments, which reflect the PTS flow situations, are important to test the interplay between all the sub-models. Some of the local flow situations can be already captured quite well by presently available CFD codes, for other still many open questions exist. In general more flexible models are required which allow switching between different approaches within one flow domain but for the different local flow situations. Examples for such model approaches are the Large Scale Simulation (LSS) which should allow the application of a two-fluid model for dispersed flows and Interface Tracking Methods for large surfaces and the Scale Adaptive Simulations (SAS) which allow the simulation of large eddies while modelling the turbulence at the unresolved scales. The work done leads to a clear improvement of the simulation capabilities regarding a two-phase PTS situation, but caused by the complexity of the issue it will still be a long way to enable predictive simulations for all the different phenomena that occur in this application. In the near term, one may envisage a simplified treatment of two-phase PTS transients by neglecting some effects which are not yet controlled.

A commercial CFD code was applied, for validation purposes, to the simulation of a slug mixing experiment carried out at OKB “Gidropress” scaled facility in the framework of a TACIS project: “Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. Such experimental model reproduces a VVER-1000 nuclear reactor and is aimed at investigating the in-vessel mixing phenomena. The addressed experiment involves the start-up of one of the four reactor coolant pumps (the other three remaining idle), and the presence of a tracer slug on the starting loop, which is thus transported to the reactor pressure vessel where it mixes with the clear water. Such conditions may occur in a boron dilution scenario, hence the relevance of the addressed phenomena for the nuclear reactor safety. Both a pre-test and a post-test CFD simulation of the mentioned experiment were performed, which differ in the definition of the boundary conditions (based either on nominal quantities or on measured quantities, respectively). The numerical results are qualitatively and quantitatively analyzed and compared against the measured data in terms of space and time tracer distribution at the core inlet. The improvement of the results due to the optimization of the boundary conditions is evidenced, and a quantification of the simulation accuracy is proposed.

The main objective for the quantification of the fluid mixing in the downcomer and the lower plenum is the demonstration of the safety of the nuclear plant during non-symmetrical transients. This concerns two main topics: The risk of fragile brittle of the Reactor Pressure Vessel (RPV) during Pressurized Thermal Shock (PTS) transients and the risk of core reactivity excursion during non-symmetrical transient such as Main Steam Line Breaks (MSLB) or Boron Dilution Transients (BDT). The corresponding fluid mixing scenarios are studied in the 1:5 scaled VVER-1000 reactor model at OKB “Gidropress” in the framework of a TACIS project: “Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. An extensive experimental program was running, studying different flow conditions in the reactor mock up, like the start of the 1st coolant pump or natural circulation conditions with density differences of the primary coolant. Pre and post test CFD simulations were carried out for code validation and for a deeper understanding of the flow and mixing behavior in the VVER-1000 reactor also in the future of the project. The 3-D computational fluid dynamics (CFD) codes provide an effective tool for mixing calculations. The CFD-Code used was ANSYS CFX. The geometric details of the construction internals inside the RPV have a strong influence on the flow field and on the mixing. Therefore, an exact representation of the inlet region, the spacer in the downcomer, the elliptical perforated plate and the complicated structures in the lower plenum was necessary. All parts of the lower plenum structures were modeled in detail. The final computational grid contained 6.5 Million nodes. Results of selected experiments and corresponding CFD calculations will be described and discussed in the paper and conclusions will be drawn.

This paper reports some activity about the Pressurized Thermal Shock (PTS) performed within the European Integrated Project NURESIM. The PTS phenomenon is expected to take place in some water cooled nuclear reactors equipped with pressure vessels during selected accident scenarios. The PTS implies the formation of temperature gradients in the thick vessel walls with consequent localized stresses and the potential for propagation of possible flaws present in the material. Current generation Pressurized Water Reactors, PWR (including the Russian VVER types), are primarily affected by the phenomenon which is investigated within three broad areas: material damage originated by irradiation, thermal-hydraulics (including single and two-phase flow conditions in the region of the ‘shock’) and structural mechanics with main reference to fracture mechanics. The present paper, in the area of thermal-hydraulics, focuses on the study of two-phase conditions that are potentially at the origin of PTS. Within the above context, the paper summarizes recent advances in the understanding of the two-phase phenomena occurring within the geometric region of the nuclear reactor, i.e. the cold leg and the downcomer, where the ‘PTS fluid-dynamics’ is relevant. Available experimental data for validation of two-phase CFD simulation tools are reviewed and the capabilities of such tools to capture each basic phenomenon are discussed. Key conclusions show that several two phase mechanisms (or subphenomena) are involved and can individually be simulated at least at a qualitative level, but the capability to simulate their interaction and the overall system performance (case of two phase flow) is presently not available.

The main objective for the quantification of the fluid mixing in the downcomer and the lower plenum is the demonstration of the safety of the nuclear plant during non-symmetrical transients. This concerns two main topics: The risk of fragile brittle of the Reactor Pressure Vessel (RPV) during Pressurized Thermal Shock (PTS) transients and the risk of core reactivity excursion during non-symmetrical transient such as Main Steam Line Breaks (MSLB) or Boron Dilution Transients (BDT). These scenarios are studied in the 1:5 scaled VVER-1000 reactor model at OKB “Gidropress” in the framework of a TACIS project: ““Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. The 3-D computational fluid dynamics (CFD) codes provide an effective tool for mixing calculations. In recent years, the rapid development of both the software and the computers has made it feasible to study the coolant mixing in sufficient detail and to perform the calculations for transient conditions. The CFD-Code used was ANSYS CFX. The geometric details of the construction internals inside the RPV have a strong influence on the flow field and on the mixing. Therefore, a detailed representation of the inlet region, the spacer in the downcomer, the elliptical perforated plate and the complicated structures in the lower plenum was necessary. All parts of the lower plenum structures were modeled in detail. The computational grid contained 4.3 Million nodes. In the VVER-1000 reactor, similar characteristic flow and mixing pattern are observed in the case of nominal flow conditions like for Western type PWR. Sensitivity analyses were performed following recommendations included in the ECORA Best Practice Guidelines. Regarding the flow field and mixing in the downcomer during four loop operation at nominal flow rates, it has been shown that a sharp sector formation like in western 4-loop reactors appears, the flow field is inhomogeneous; in fact high velocity values occur beside the loop positions, and not below the inlet nozzles, which indicates the presence of recirculation areas or stagnant zones. Regarding the flow field and mixing at the core inlet, it has been shown that the mass flow rate distribution is more or less homogenous over the core diameter due to the lower plenum internals, the perturbed sector covers more or less one fourth of the core; a sharp sector formation like in western 4-loop reactors appears, weak mixing zones appear (around 97% of the unperturbed concentration). In most cases, the sensitivity analyses performed did not show any appreciable dependence of the results with respect to the addressed parameters. A three loop operation was chosen to show the differences of the flow and mixing behavior compared to the four loop operation. An extensive experimental program is now running, studying different flow conditions in the reactor mock up, like the start of the 1st coolant pump or natural circulation conditions with density differences of the primary coolant. Pre and post test CFD simulations are carried out for code validation and for a deeper understanding of the flow and mixing behavior in the VVER-1000 reactor also in the future of the project.

The modeling of complex transients in Nuclear Power Plants (NPP) remains a challenging topic for Best Estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulating and more precise consideration of multidimensional effects under complex transients in NPPs. Therefore, large international activities are in progress aiming to assess the capabilities of coupled codes and the new frontiers for the nuclear technology that could be opened by this technique. In the current paper a contribution to the assessment and validation of coupled code technique through the Kolzoduy VVER100 pump trip test is performed. For this purpose, the coupled REALP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data and the overall data comparison shows good agreements between the calculations and the global kinetic and thermal-hydraulic aspects observed experimentally. Further investigations through the use of two “antagonist” GRS and the CIAU methods, in order to evaluate and quantify the origin of the observed discrepancies. It was revealed on one hand that relative error quantification discrepancies exist between the two approaches, and on the other hand deviations between code predictions and measurements are mainly due to the used models for evaluating and modeling of the Doppler feedback effect.

This report identifies PTS-scenarios for the French 900 MW CPY PWR, the German 1300 MW Kon-voi reactor, the Loviisa 400 MW VVER, the Russian VVER-1000 and the Czech VVER-100. Accord-ing to the resulting basic flow conditions relevant physical phenomena for the simulation of the scenes during Emergency Core Cooling (ECC) injection into the cold leg are identified. The main focus is on two-phase flow phenomena. The state of the art of modelling these phenomena and needs for models improvement are discussed. Thus the report is a suitable basis for the specification of the main topics to be provided in Task T2.1.4 of the NURESIM project.

The Code with – the capability of – Internal Assessment of Uncertainty (CIAU) is a tool proposed by the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione (DIMNP) of the University of Pisa. Other Institutions including the nuclear regulatory body from Brazil, “Comissäo Nacional de Energia Nuclear”, contributed to the development of the tool. The CIAU aims at providing the currently available Relap5/Mod3.2 system code with the integrated capability of performing not only relevant transient calculations but also the related estimates of uncertainty bands. The Uncertainty Methodology based on Accuracy Extrapolation (UMAE) is used to characterize the uncertainty in the prediction of system code calculations for light water reactors and is ‘internally’ coupled with the above system code. Following an overview of the CIAU development, the present paper deals with the independent qualification of the tool. The qualification test is performed by estimating the uncertainty bands that should envelope the prediction of the LOBI test Al-93 (small break loss of coolant accident). The current limitation of the ‘error’ database, implemented into the CIAU prevented a final demonstration of the qualification. However, all the steps for the qualification process are demonstrated.

The mixing phenomenon is relevant for the normal and off normal operation of the Nuclear Power Plant (NPP), because it influences the safety, mitigating the reactivity and structural consequences. The study of these issues was investigated performing experimental campaigns in large-scale test facilities and in real NPPs. In this framework ROCOM (Rossendorf Coolant Mixing Model) test facility was built with the purpose of investigating the coolant mixing phenomena occurring in the Reactor Pressure Vessel (RPV) of a Pressurizer Water Reactor (PWR). The experiments executed in this facility provide experimental data for code validation (CFD and TH-SYS codes). The purpose of this work is to address the capability of the RELAP5-3D© to reproduce ROCOM facility dynamic in simulating the mixing for a large range of operational and accident conditions. In particular, the attention is focused on the effects in the vessel downcomer, lower plenum and core inlet. Three experiments were selected for the analyses: two steady states and one transient. The ROCOM steady states are slug mixing experiments that analyze the mixing scalar trend inside the facility at different mass flow rates. The ROCOM transient represents the injection of a mixing scalar slug from one cold leg with an increasing mass flow rate in the same loop. A systematic comparison between ROCOM experimental data and the results of the simulations with the RELAP5-3D© is presented including a complete set of sensitivity analyses to find out the most relevant parameters which influence the results from nodalization and user effects points of view.