| 1. |
- Anglart, Henryk, et al.
(författare)
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Measurement of Wall Temperature Fluctuations during Thermal Mixing of Non-isothermal Water Streams
- 2015
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Ingår i: Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). - : American Nuclear Society.
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Konferensbidrag (refereegranskat)abstract
- This paper is dealing with measurement of temperature fluctuations during mixing of two water streams in an annular test section at BWR operational conditions. The experiments are simulating conditions existing in a guide tube of BWR control rods, where relatively cold water at about 333 K is mixing with hot water at ~550 K. It is shown that the mixing is causing high amplitude temperature fluctuations in the solid walls of the control rod extender. Using new movable multi-sensors it became possible to obtain a large experimental database, containing wall temperature measurements at 8 azimuthal and 5 axial positions, with 13 thermocouples at each position. In total 520 temperature readings were performed, each lasting about 2 minutes and recording transient temperature with frequency of at least 100 samples per second and with estimated non-calibrated uncertainty less than 3.9 K. The present experimental results can be used to analyze the governing phenomena during thermal mixing and also to validate CFD conjugate heat transfer models of thermal mixing applied to actual reactor geometries.
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| 2. |
- Bednarski, M, et al.
(författare)
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Identification of sensitivities in Bayesian networks
- 2004
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Ingår i: Engineering applications of artificial intelligence. - : Elsevier BV. - 0952-1976 .- 1873-6769. ; 17:4, s. 327-335
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Tidskriftsartikel (refereegranskat)abstract
- This paper presents a methodology for sensitivity analysis that can be applied to Bayesian belief networks, i.e. analysis of the influence of the quality of network parameters (such as conditional and a priori probabilities) on the values of the hypothesis variable(s). The presented methodology makes use of one-way sensitivity analysis and makes it possible to apply a particular mathematical model for relations between the considered parameter and distribution of values in the node of interest (hypothesis node). The sensitivity analysis has been applied to a network describing a Nuclear Power Plant during fault conditions.
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| 3. |
- Caraghiaur, Diana, 1978-, et al.
(författare)
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Detailed pressure drop measurements in single- and two-phase adiabatic air-water turbulent flows in realistic BWR fuel assembly geometry with spacer grids
- 2004
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Ingår i: The 6<sup>th</sup> International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6) Nara, Japan, October 4-8, 2004.
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Konferensbidrag (refereegranskat)abstract
- In recent years, advanced numerical simulation tools based on CFD methods have been increasingly used in various multi-phase flow applications. One of these is two-phase flow in fuel assemblies of Boiling Water Reactors. The important and often missing aspect of this development is the validation of CFD codes against proper experimental data. The purpose of the current paper is to present detailed pressure measurements over a spacer grid in adiabatic single- and two-phase flow, which will be used to validate and further develop a CFD code for BWR fuel bundle analysis. The experiments have been carried out in an asymmetric 24-rod sub-bundle, representing ¼ of Westinghouse SVEA-96 nuclear reactor fuel assembly. Single-phase measurements have been performed at superficial velocities comprised between jliq: 0.90 – 4.50 m/s and in the two-phase, which was simulated by air-water mixture, measurements have been performed at void fractions ranging from 4 to 12% and liquid superficial velocity jliq : 4.50 m/s. In order to increase the number of the measured points, five pressure taps were drilled in one of the rods, which was easily moved vertically by a traverse system, covering most of the points in axial direction. The possibility to substitute any of the rods in the fuel bundle by the pressure sensing rod and the possibility to change the pressure taps facing-angle provides more measuring points inside the subchannels. A detailed pressure distribution comparison between single- and two-phase flows for different subchannel positions and different flow conditions was performed over one of the spacers. In addition, single-phase pressure drop measurements on the upper part of the test section comprising two spacer grids has been carried out.
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| 4. |
- Caraghiaur, Diana, 1978-, et al.
(författare)
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Experimental investigation of turbulent flow through spacer grids in fuel rod bundles
- 2009
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 239:10, s. 2013-2021
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Tidskriftsartikel (refereegranskat)abstract
- This paper contains experimental data of pressure, velocity and turbulence intensity in a 24‐rod fuel bundle withspacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressuresensingrod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuatingcomponent upstream and downstream of the spacer grid in subchannels with different blockage ratios. Themeasurements show a changing pattern in function of radial position in the cross‐section of the fuel bundle. Forsubchannels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer andthen gradually decays. In inner subchannels, however, the turbulence intensity downstream of the spacer decreasesbelow its upstream value and then gradually increases until it reaches the maximum value at approximately twospacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulenceintensity enhancement, do not depend exclusively on the local geometry details, but also on the location in thecross‐section of the rod bundle.
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| 5. |
- Cholewa, W, et al.
(författare)
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Identification of loss-of-coolant accidents in LWRs by inverse models
- 2004
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 147:2, s. 216-226
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Tidskriftsartikel (refereegranskat)abstract
- This paper describes a novel diagnostic method based on inverse models that could be applied to identification of transients and accidents in nuclear power plants. In particular, it is shown that such models could be successfully applied to identification of loss-of-coolant accidents (LOCAs). This is demonstrated for LOCA scenarios for a boiling water reactor. Two classes of inverse models are discussed: local models valid only in a selected neighborhood of an unknown element in the data set, representing a state of a considered object, and global models, in the form of partially unilateral models, valid over the whole learning data set. An interesting and useful property of local inverse models is that they can be considered as example based models, i.e., models that are spanned on particular sets of pattern data. It is concluded that the optimal diagnostic method should combine the advantages of both models, i.e., the high quality of results obtained from a local inverse model and the information about the confidence interval for the expected output provided by a partially unilateral model.
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| 6. |
- Frid, Wiktor, et al.
(författare)
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Development of a Bayesian belief network for a boiling water reactor during fault conditions
- 2005
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Ingår i: Computer Assisted Mechanics and Engineering Sciences. - 1232-308X. ; 12:2-3, s. 133-145
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Tidskriftsartikel (refereegranskat)abstract
- This paper describes briefly the development and verification of a probabilistic system for the rapid diagnosis of plant status and radioactive releases during postulated severe accidents in a Boiling Water Reactor nuclear power plant. The probabilistic approach uses Bayesian belief network methodology, and was developed in the STERPS project in the European Union 5-th Euroatom Framework program.
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| 7. |
- Knochenhauer, M., et al.
(författare)
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Using bayesian belief networks to predict radioactive releases after a severe accident in a BWR
- 2006
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Ingår i: Proceedings of the 8th International Conference on Probabilistic Safety Assessment and Management, PSAM 2006. - 0791802442 - 9780791802441 ; , s. 1-8
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Konferensbidrag (refereegranskat)abstract
- The paper deals with an application within the STERPS project (Source Term Indicator Based on Plant Status) which is part of the 5th Euratom Framework Program. The project aims at developing a computer based tool for the rapid and early diagnosis of plant status and for the prediction of environmental releases. The approach is based on a probabilistic plant model using the Bayesian Belief Network (BBN) methodology. A BBN models relations between variables which are relevant to some problem. In this way, meaningful results can be obtained in spite of incomplete or uncertain information. In the STERPS application, the outcome is typically a number of possible plant states ranked according to probability, each with an associated environmental source term. The paper describes the development of a BBN for the Swedish boiling water reactor (BWR) Oskarshamn 3. The analysis used the generic BBN software Netica (developed by Norsys Inc.), with the user interface SPRINT, which was developed within the project for handling of the BBN. The user interface includes a set of questions and background information, which are used in order to gain information about crucial plant parameters during the course of a severe accident. SPRINT also includes graphical presentation of analysis results, both in terms of node probabilities and as characteristics for radioactive releases (amount, composition, and timing). The customization to the Oskarshamn 3 nuclear power plant included identification of key plant parameters for inclusion in the BBN, through a systematic consideration of fission product transport and retention phenomena in plant. Plant systems for mitigation of accidents as well as implemented severe accident management strategies at the plant were also considered. The functionality and practicability of the SPRINT software is being demonstrated in the recently started EU project EURANOS (European approach to nuclear and radiological emergency management and rehabilitation strategies). For the Oskarshamn 3 plant, SPRINT has been tested in connection with an emergency exercise at Oskarshamn 3. Preliminary results and conclusions show that the project has successfully demonstrated the suitability of the BBN technique for modeling the complex conditions after a severe accident in a nuclear power plant. The user interface is simple, and after some adaptation SPRINT will be suitable for use in plant technical support centers and at national emergency centers. The plant BBN models will also be a very useful tool for training and education. The prediction capabilities of the resulting models can be efficiently verified using results from plant PSA quantification and from accident analysis codes. In conclusion, the described technique has proved to be a very promising prediction tool for plant status diagnosis and estimation of the source term in severe accident situations.
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| 8. |
- Tinoco, H., et al.
(författare)
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Numerical simulation of boron injection in a BWR
- 2007
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Ingår i: Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12. - 0894480588 - 9780894480584
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Konferensbidrag (refereegranskat)abstract
- The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion.The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause fuel damage due to local reactivity transients. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several minutes.
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| 9. |
- Tinoco, Hernan, et al.
(författare)
-
Numerical simulation of boron injection in a BWR
- 2010
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 240:2, s. 221-234
-
Tidskriftsartikel (refereegranskat)abstract
- The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several minutes.
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