| 1. |
- Banati, Jozsef, 1960, et al.
(författare)
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Validation of relap5/mod3.3 against a load step transient at Ringhals 4 u
- 2015
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Ingår i: 16th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) Chicago, IL, August 30 - September 4 2015. - 9781510811843 ; 6, s. 4548-4561
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Konferensbidrag (refereegranskat)abstract
- ABSTRACTThe Unit 4 of the Ringhals Nuclear Power Plant has recently undergone a large component replacement project with installation of new steam generators and a pressurizer, targeting a power uprate. A series of startup and maneuverability tests has been performed, mainly focusing on evaluation of the system responses for various perturbations. The subject of the present numerical analysis is a test with a ± 10 % steps applied in the load. The data collected during test provided a good opportunity for validation of the full plant model, which was prepared recently for the RELAP5/Mod3.3 Patch04 computer code, with incorporation of the new component models.The paper introduces the test procedure and shows an overview on the key parameters that are utilized as initial and boundary conditions. Strategies applied for achievement of steady-state conditions are addressed in the document. Furthermore, the paper summarizes the results of the validation study using the transient data to simulate the startup test. It has been proven that the stand-alone RELAP5 thermal-hydraulic model is capable of reproduction of the key features and events of the test. Sufficiently good agreement has been achieved between the measured and simulated thermal hydraulic parameters, already in its current stage of model development. On the basis of successful verification at the original power, it is expectable that the new Ringhals 4 model will be able to predict the fluid conditions in other types of transients, even at uprated conditions.
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| 2. |
- Bergagio, Mattia, et al.
(författare)
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Instrumentation for Temperature and Heat Flux Measurement on a Solid Surface under BWR Operating Conditions
- 2015
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Ingår i: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). - : American Nuclear Society. - 9781510811843 ; , s. 5962-5975
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Konferensbidrag (refereegranskat)abstract
- A new experimental facility has been developed at KTH Royal Institute of Technology to measure temperature and heat flux propagations in solid walls due to mixing of non-isothermal water streams in their vicinity. The main purpose of the measurements has been to obtain a high-precision experimental database suitable for validation of Computational Fluid Dynamics (CFD) codes. Consequently, a set of experiments have been performed in a test section simulating the annular region in the BWR control-rod guide tubes. Since preliminary CFD results implied that 0.1-1 Hz temperature oscillations were to be expected, this experimental research intends to assess the magnitude of temperature fluctuations within the abovementioned frequency range. To this end, water and wall temperatures have been measured in the innermost part of the test-section annulus, with a variety of boundary conditions. As thermocouples would otherwise be available at few axial and azimuthal coordinates only, the tube they are installed on has been lifted, lowered and rotated by a software-controlled motor to record temperature fluctuations in the whole mixing region. At each measurement point, data have been collected over a time long enough to detect the existence of the aforesaid fluctuations. Moreover, an uncertainty analysis has been carried out concerning water temperatures. Thermocouples meant to monitor these temperatures have been modelled with a finite-element method for this very purpose. The wall heat flux has also been estimated using experimental data, thanks to a corrected finite-difference Crank-Nicolson scheme.
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| 3. |
- Corre, J. -ML., et al.
(författare)
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Detailed measurements of local parameters in annular Two-Phase FLOW IN FUEL BUNDLE under BWR operating conditions
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - : American Nuclear Society. - 9781510811843 ; , s. 4337-4351
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Konferensbidrag (refereegranskat)abstract
- The Westinghouse FRIGG facility, in Vasteras/Sweden, is dedicated to the measurement of critical power, stability and pressure drop in fuel rod bundles under BWR operating conditions (steady-state and transient). The facility is particularly relevant to test modern BWR fuel designs which typically have complex features, such as part-length rods and mixing vanes that make the flow heterogeneous and challenging to accurately simulate (e.g. using sub-channel analysis codes or CFD tools). In order to support the validation of advanced thermal-hydraulics codes for detailed BWR fuel assembly simulation, new local instrumentation techniques have been tested at the FRIGG facility for the measurement of two-phase dynamic pressure (Pitot tubes) and high time resolution phase detection (optical sensor). The optical sensors were custom-made by RBI Instrumentation for the FRIGG facility and optimized for annular two-phase flow (drop/steam) under BWR operating conditions. This new instrumentation was successfully tested and allows the first-time measurement, under BWR operating conditions, of relevant two-phase flow parameters such as the local void fraction in the steam core, the local drop/steam velocity, the volumetric interfacial area, the drop collision frequency and the assessment of drop size distribution during BWR steady-state and transient operations.
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| 4. |
- Fischer, M., et al.
(författare)
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Core melt stabilization concepts for existing and future LWRs and associated R&D needs
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 7578-7592
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Konferensbidrag (refereegranskat)abstract
- In the event of a severe accident with core melting in a NPP the stabilization of the molten corium is an important mitigation issue, as it can avoid late containment failure caused by basemat penetration, overpressure, or severe damage of internal structures. The related failure modes may result in significant long-term radiological consequences and high related costs. Because of this, the licensing framework of several countries now includes the request to implement mitigative core melt stabilization measures. This does not only apply to new builds but also to existing LWR plants. The paper gives an overview of the ex-vessel core melt stabilization strategies developed during the last decades. These strategies are based on a variety of physical principles like: melt fragmentation in a deep water pool or during molten core concrete interaction with top-flooding, water injection from the bottom (COMET concept), and retention in an outside-cooled crucible structure. The provided overview covers the physical background and functional principles of these concepts, as well as their status of validation and, if applicable, the remaining open issues and R&D needs. For concepts based on melt retention inside a cooled crucible that reached sufficient maturity to be implemented in current Gen-III+ designs, like the VVER-1000/1200 and the EPR™, more detailed descriptions are provided, which include key aspects of the related technical realization. The paper is compiled using contributions from the main developers of the individual concepts.
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| 5. |
- Ghione, Alberto, 1989, et al.
(författare)
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Wall superheat prediction in narrow rectangular channels under fully developed boiling of water at low pressures
- 2015
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Ingår i: Proc. 16th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-16). - 9781510811843 ; 10, s. 8360-8373
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Konferensbidrag (refereegranskat)abstract
- The modeling of two-phase heat transfer is a crucial issue in the safety analysis of nuclear reactors. Thethermal-hydraulic correlations employed in this kind of simulations are usually derived from experimentsthat were carried out over limited ranges of parameters and for specific geometries. Therefore theirapplicability to systems with different characteristics has to be carefully scrutinized.In this paper, an assessment study of wall superheat correlations under fully developed boiling ispresented. This is a contribution to the validation and improvement of the thermal-hydraulic modeling ofthe Jules Horowitz Reactor, which is a research reactor under construction at CEA-Cadarache (France).The SULTAN-JHR experiments are used. These tests were performed at CEA-Grenoble with upwardwater flow in a vertical uniformly heated narrow rectangular channel with gap of 2.16 mm. Theexperimental conditions ranged between 0.2 and 0.9 MPa for the pressure and between 0.5 and 4.4MW/m2 for the heat flux.It is shown that the correlations of Thom and Jens-Lottes significantly overestimate the wall superheat.The correlation of Belhadj and Qiu, which were developed for narrow channels at low heat fluxes, cannotaccurately predict the experimental data. On the other hand, satisfactory results can be obtained withGorenflo (standard deviation of 11.9%) and a simplified version of the Forster-Greif (standard deviationof 10.1%) correlations. In conclusion, considering the validity range of the above correlations along withthe outcomes of the current assessment, the simplified Forster-Greif correlation is thus recommended forthe analysis of the JHR.
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| 6. |
- Grishchenko, Dmitry, et al.
(författare)
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Development of Texas-V code surrogate model for assessment of steam explosion impact in Nordic BWR
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015. - : American Nuclear Society. - 9781510811843 ; , s. 7222-7235
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Konferensbidrag (refereegranskat)abstract
- Severe accident mitigation strategies in Nordic boiling water reactors (BWRs) employ core melt cooling in a deep pool of water under the reactor pressure vessel. Corium melt released from the vessel is expected to fragment, solidify and form a porous debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. Significant aleatory and epistemic uncertainties exist in accident scenarios, melt release conditions, and modeling of steam explosion phenomena. Assessment of the risk of ex-vessel steam explosion requires application of the Integrated Deterministic Probabilistic Safety Analysis (IDPSA). IDPSA is a computationally demanding task which makes unfeasible direct application of Fuel-Coolant Interaction codes. The goal of the current work is to develop a Surrogate Model (SM) of the Texas-V code and demonstrate its application to the analysis of explosion impact in the Nordic BWR. The SM should be computationally affordable for IDPSA analysis. We focus on prediction of the steam explosion loads in a reference Nordic BWR design assuming a scenario of coherent corium jet release into a deep water pool. We start with the review of the Texas-V sub-models in order to identify a list of parameters to be considered in implementation of the SM. We demonstrate that Texas-V exhibits chaotic response in terms of the explosion impulse as a function of the triggering time and introduce a statistical representation of the explosion impulse for given melt release conditions and arbitrary triggering time. We demonstrate that characteristics of the distribution are well-posed. We then separate out the essential portion of modelling uncertainty by identification of the most influential uncertain parameters using sensitivity analysis. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in FCI modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a Nordic BWR. A database of Texas-V solutions is generated and used for the development of the SM. Performance, predictive capability and application of the SM to risk analysis are discussed in detail.
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| 7. |
- Konovalenko, Alexander, et al.
(författare)
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Experiments and modeling of particulate debris spreading in a pool
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 8055-8068
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Konferensbidrag (refereegranskat)abstract
- Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. Success of the strategy is contingent upon effectiveness of natural circulation in removing the decay heat generated by the porous debris bed. Geometrical configuration of the bed is one of the factors which affect coolability of the bed. Boiling and two-phase turbulent flows in the pool serve as a source of mechanical energy which can affect the initial geometry as well as dynamically change the shape of already formed debris bed. The main goal of this work is to provide experimental data on spreading of solid particles in the pool by large scale two-phase flow structures induced by gas injection from the bottom. These data are necessary for development and validation of predictive capabilities of computer codes allowing numerical modeling of the debris bed formation at prototypic severe accident conditions. In PDS-P experiments air injection at the bottom of the test section is employed in order to create large scale flow in the pool. The test section is constructed as a rectangular tank. It has close to 2D geometry with fixed width (72 mm), variable length (up to 1.6 m) and allows water filling depth of up to 1 m. The variable pool length and depth allows formation of the different in size and pattern two-phase circulating flows. Experimental conditions such as gas-phase flow rate and particle properties (density and size) are scaled to maintain relevancy to the prototypic accident conditions. The average void fraction in the pool is determined by video recording and image processing. Particles are supplied from the top of the facility above the water surface. In the separate-effect studies of the influence of two-phase currents on particle trajectories and bed formation, a low particle flow rate is required in order to minimize or completely exclude particle-particle interaction. Results of several series of PDS-P (Particulate Debris Spreading in the Pool) reported in this paper are analyzed analytically. The preliminary scaling approach is proposed and has good agreement with experimental findings.
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| 8. |
- Kudinov, Pavel, et al.
(författare)
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Experimental investigation of debris bed agglomeration and particle size distribution using W03-ZR02 melt
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 8046-8054
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Konferensbidrag (refereegranskat)abstract
- Nordic BWR severe accident management strategy employs reactor cavity flooding to terminate ex-vessel accident progression. Corium melt released from the reactor pressure vessel is expected to fragment and form a porous debris bed. Success of the SAM strategy is contingent upon possibility to remove the decay heat generated in the debris bed by natural circulation of the coolant. Properties of the debris bed such as particle size, porosity and shape of the bed determine resistance for the coolant flow and thus dryout heat flux. Agglomeration of incompletely solidified debris can create additional obstacles for coolant circulation and thus reduce debris coolability margin. The goal of DEFOR (debris bed formation) experimental work is to provide data necessary for the development of analytical models and approaches for prediction of debris bed formation and agglomeration phenomena. Different corium simulant materials are used in the experiments. Liquid melt jet fragmentation and debris bed formation are considered at different conditions such as melt release (jet diameter, free fall height, etc.), melt superheat, water subcooling and water pool depth. A series of confirmatory DEFOR-A experiments has been carried out with ZrO2-WO3 simulant material. The data on particle size distribution, debris bed porosity and agglomeration is in good agreement with the previous DEFOR-S, DEFOR-A and FARO tests. On average, larger particles were obtained with ZrO2-WO3 melt than with previously used Bi2O3-WO3, size distributions for both melt simulant materials are within the ranges of size distributions observed in FARO tests. The difference between particle sizes in the tests with free falling jets was found to be insignificant. There is a tendency to form slightly larger particles only in the tests with submerged nozzles where melt is released under water with initially small jet velocity. Initial jet velocity also seems to have no visible effect on the fraction of agglomerated debris.
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| 9. |
- Manickam, Louis, et al.
(författare)
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An experimental study on void generation around hot metal partcle quenched into water pool
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015. - : American Nuclear Society. - 9781510811843 ; , s. 6569-6578
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Konferensbidrag (refereegranskat)abstract
- Motivated to understand the void formation in fuel coolant interactions during a postulated severe accident of light water reactors, an experimental study was carried out to characterize void generation and evolution from stainless steel spheres quenched into a subcooled water pool. The major experimental parameters ranged from variations in water subcooling as well as the initial temperature and the diameter of the spheres which were kept still or falling in the water pool. The experimental results show that under the stagnant condition, axisymmetric vapor waves develop at the stagnation point which travels through the droplet’s rear periphery and subsequently detaches into almost equal size bubbles at regular intervals in the stable film boiling regime, but a sharp decrease in the bubble detachment frequency was observed due to a transition in boiling regime. The effect of the water subcooling on the detached vapor volume is significant. The number of bubbles detached from the sphere periphery drastically reduces at high subcooling conditions. Under the falling condition, the temperature and diameter of the sphere as well as water subcooling significantly affect the dynamics of the bubbles entrained to the sphere’s periphery.
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| 10. |
- Mickus, Ignas, et al.
(författare)
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Development of tall-3d test matrix for APROS code validation
- 2015
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Ingår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015. - 9781510811843 ; , s. 4562-4575
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Konferensbidrag (refereegranskat)abstract
- APROS code is a multifunctional process simulator which combines System Thermal-Hydraulic (STH) capabilities with ID,'3D reactor core neutronics and full automation system modeling. It is applied for various tasks throughout the complete power plant life cycle including R&D, process and control engineering, and operator training. Currently APROS is being developed for evaluation of Generation IV conceptual designs using Lead-Bismuth Eutectic (LBt) alloy coolant. TALL-3D facility has been built at KTH in order to provide validation data for standalone and coupled STH and Computational Fluid Dynamics (CFD) codes. The facility consists of sections with measured inlet and outlet conditions for separate effect and integral effect tests (SETs and lETs). The design is aimed at reducing experimental uncertainties and allowing fall separation of code validation from model input calibration. In this paper we present the development of experimental TALL-3D lest matrix for comprehensive validation of APROS code. First, the representative separate effect and integral system response quantities (SRQs) arc defined. Second, sources of uncertainties are identified and code sensitivity analysis is carried out to quantify the effects of code input uncertainties on the code prediction. Based on these results the test matrixes for calibration and validation experiments arc determined in order to minimize the code input uncertainties. The applied methodology and the results arc discussed in detail.
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