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- Cadinu, Francesco, et al.
(author)
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Relating system-to-CFD coupled code analyses to theoretical framework of a multiscale method
- 2008
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In: Societe Francaise d'Energie Nucleaire - International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work". - 9781604238716 ; , s. 2959-2967
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Conference paper (peer-reviewed)abstract
- Over past decades, analyses of transient processes and accidents in a nuclear power plan t have been performed, to a significant extent and with an admirable success, by means of so called system codes, e.g. RELAP5, CATHARE, ATHLET codes. These computer codes, based on a multi-fluid model of two-phase flow, provide an effective, one-dimensional description of the coolant thermal-hydraulics in the reactor system. For some components in the system, wherever needed, the effect of multi-dimensional flow is accounted for through approximate models. The later are derived from scaled experiments conducted for selected accident scenarios. Increasingly, however, we have to deal with newer and ever more complex accident scenarios. In some such cases the system codes fail to serve as simulation vehicle, largely due to its deficient treatment of multi-dimensional flow (in e.g. downcomer, lower plenum). Enter Computational Fluid Dynamics (CFD). Based on solving Navier-Stokes equations, CFD codes have been developed and used, broadly, to perform analysis of multi-dimensional flow, dominantly in non-nuclear industry and for single-phase flow applications. Although not always straightforward, CFD codes can be, and have been, used to analyze thermo-fluid processes in a certain component of the reactor system at a well-defined point during the accident progression. It is natural to think that CFD codes provide the much-needed complementary capability to the system codes. Furthermore, due to the CFD excessive demand on computational resources, ideas were proposed, and attempts were reported in the literature, to use a coupled system-to-CFD code to maximize the benefit of both tools. Easy as it might sound, progress in this area has been sluggish. In this paper, we take a close look at the progress in coupled system-to-CFD code analyses, including coupling algorithms, their implementation and performance. Tackling thermo-fluid dynamics at largely different scales, system codes and CFD codes employ different models and governing equations. This notion led us to the idea to examine the system-to-CFD coupling in the language of multiscale simulations. As a theoretical framework, we bring to bear the heterogeneous multiscale method pioneered by E and Engquist and problem classification offered by E et al.[16]. Viewing system-to-CFD coupling as a multiscale problem, the ultimate objective of the present effort is to define requirements for models and numerical methods, and develop suggestions on a coupling strategy that ensures robust and effective generation and transfer of information between scale-specific simulations (system and CFD).
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| 4. |
- Cadinu, Francesco, et al.
(author)
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Study of algorithmic requirements for a system-to-CFD coupling strategy
- 2008
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In: Experiments and CFD Code Application to Nuclear Reactor Safety (XCFD4NRS).
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Conference paper (peer-reviewed)abstract
- Over the last decades, the analysis of transients and accidents in nuclear power plants has beenperformed by system codes. Though they will remain the analyst’s tool of choice for the foreseeablefuture, their limitations are also well known. It has been suggested that an improvement in thesimulation technology can be obtained by “coupling” system codes with Computational FluidDynamics (CFD) calculations. This is usually attempted in a domain decomposition fashion: the CFDsimulation is only performed in a selected subdomain and its solution is “matched” with the systemcode solution at the interface. However, another coupling strategy can be envisioned. Namely, CFDsimulations can be used to provide closures to a system code.This strategy is based on the following two assumptions. The first assumption is that there aretransients which cannot be simulated by system codes because of the lack of adequate closures. Thesecond assumption is that appropriate closures can be provided by CFD simulations. In this paper,such a coupling strategy, inspired by the Heterogeneous Multiscale Method (HMM), is presented. Thephilosophy underlying this strategy is discussed with the help of a computational example.
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- Kozlowski, Tomasz, et al.
(author)
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Evaluation of coupled codes RELAP5/PARCS capability for BWR global stability prediction
- 2007
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In: Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12. - 9780894480584 - 0894480588 ; , s. 1619-1640
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Conference paper (peer-reviewed)abstract
- The present study is concerned with capability of a coupled neutron-kinetic/thermal-hydraulic code system RELAP5/PARCS for the numerical prediction of global core stability condition and instability transients. The work is motivated by the need to assess safety significance of a number of stability transients which trigger core instability and challenge reactor protection systems. The technical approach adopted is both to learn from real stability events and to perform analysis of idealized well-defined transients in a real plant and core configuration. In this paper, we show that the code system can serve as a unique and powerful tool to provide a consistent and reasonably reliable prediction of stability boundary even in complex plant transients. However, the prediction quality of the instability transients, i.e. core behavior without scram, namely parameters of the limit cycle remains questionable. We identify two main factors for future studies (two-phase flow regimes in oscillatory flow and algorithm for effective grouping of thermal-hydraulic channels) as key to enhancing the predictive capability of the existing coupled code system for BWR stability.
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- Kubarev, Andrej, et al.
(author)
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Design of BWR instability suppression system
- 2008
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In: International Conference on the Physics of Reactors 2008, PHYSOR 08. - 9781617821219 ; , s. 2062-2068
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Conference paper (peer-reviewed)abstract
- This paper explores the concept of instability suppression system and its performance in a BWR plant. The key idea adopted from the work of Aleksakov et al. (1980) is to utilize information provided by the in-core power monitoring detectors to guide movement of control rods in a way that suppress the global, regional and local instability. In the paper, effectiveness of a simplified suppression algorithm is characterized by implementing it on a real BWR model, using the RELAP5/PARCS coupled thermal-hydraulics and neutron kinetics code. Both forced power oscillations and realistic reactor transients (feedwater temperature transients, control rod drop) were analyzed. The results suggest that, without requiring any modifications for the in-reactor diagnostics and equipment, the proposed suppression system is capable of significantly mitigating the impact of core instability events on plant performance by maintaining the core parameters within the safe operational range.
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| 12. |
- Peltonen, Joanna, et al.
(author)
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Spatial coupling for coupled code safety analysis of BWR design-basis accidents
- 2008
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Conference paper (peer-reviewed)abstract
- Analyses of nuclear reactor safety have increasingly requiredcoupling of full three dimensional neutron kinetics(NK) core models with system transient thermal-hydraulics(TH) codes. To produce results within a reasonable computingtime, the coupled codes use different spatial descriptionof the reactor core. The TH code uses few, typically 5 to20 TH channels which represent the core. The NK code usesexplicit node for each fuel assembly. Therefore, a spatialmapping of coarse grid TH and fine grid NK domain is necessary.However, improper mappings may result in loss ofvaluable information, thus causing inaccurate prediction ofsafety parameters. The purpose of this paper is to study thesensitivity of spatial coupling (channel refinement and spatialmapping) and develop recommendations for NK-THmapping in simulation of safety transients.The research methodology consists of spatial couplingconvergence study, as increasing number of TH channelsand different mapping approach the reference case. Thereference case consists of 700 TH channels, which givesone TH channel per one fuel assembly. The comparison ofresults has been done under steady-state and transientconditions. Obtained results and conclusions are presentedin this paper.
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- Roshan Ghias, Sean, et al.
(author)
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A study of reactor systems during a loss of offsite electric power in Forsmark-1 Plant
- 2008
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In: Proceedings of 2007 International congress on advances in nuclear power plants (ICAPP 2007). - 9781604238716 ; , s. 2642-2651
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Conference paper (peer-reviewed)abstract
- On Tuesday the 25. of July 2006 at around 13:15, Forsmark-1 nuclear power plant experienced a loss of external power event, initiated by a short circuit in the offsite 400 kV switchyard. Due to voltage and frequency fluctuations that followed, together with additional component failures, two of the four auxiliary diesel generators did not start, causing loss of power in 2 of four redundant trains existing in the power plant. The loss of power in trains A,B resulted in reactor shutdown and abnormal intervention of safety systems. After 20 minutes, the water level inside the Reactor Pressure Vessel (RPV) decreased to 1,9 m above the reactor core, and the pressure inside the RPV decreased to 1,5 MPa. The aim of the present study is to evaluate the capabilities of U.S. NRC codes RELAP5 and MELCOR to simulate the Forsmark-1 event, and then to reconstruct the sequence of the event based on the known behavior of the plant systems, such as activation of depressurization valves. To examine the safety margin, it is of interest to address 'what if' questions related to this event, such as i) what if the operator would delay the recovery of the two failing diesel generators, and ii) what if all 4 diesel generators would fail. The results show that both RELAP5 and MELCOR codes are able to reproduce the system thermal-hydraulic behavior during such an event. The intervention of emergency cooling systems and effort of operators to start the remaining two auxiliary generators have prevented the core from becoming uncovered. The analysis also shows that even in case of failure of all 4 auxiliary generators, the timely action of the plan operator, as demonstrated in the action during the event, would prevent a core damage from occurring.
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| 15. |
- Sánchez, V. H., et al.
(author)
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Qualification of the 3D Thermal Hydraulics Model of the Code System TRACE Based on plant data
- 2008
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In: Societe Francaise d'Energie Nucleaire - International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work". - 9781604238716 ; , s. 1084-1092
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Conference paper (peer-reviewed)abstract
- The Institute of Reactor Safety is involved in the qualification of best-estimate coupled code systems for reactor safety evaluations since it is a key step toward improving their prediction capability and acceptability. In the frame of the WER-1000 Coolant Transient Benchmark PhaseI the coupled code RELAP5/PARCS has been extensively assessed. The Phase 2 of this benchmark- currently underway- is focused on both multidimensional thermal hydraulics phenomena within the reactor pressure vessel (RPV) such as coolant mixing and core physics. Hence it is an excellent opportunity to qualify the prediction capability of the new coupled code system TRACE/PARCS taking into account plant data obtained in the Kozloduy nuclear power plant unit 6. In addition a lose coupling of CFX with RELAP5 is applied for the posttest calculation of the coolant mixing experiment. The developed multidimensional models of the WER-1000 reactor pressure vessel as well as the performed calculations using these models are described in some detail. The predicted results are in good agreement with the data. It was demonstrated that the chosen 3D-nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a WER-1000 reactor. In addition selected results of the code TRACE/PARCS for a postulated main steam line transient (MSLB) are given. The investigations have shown that the multidimensional neutronics and thermal hydraulic model developed for the RPV of the WER-1000 reactor are well qualified and consequently they are ready for their integration into a overall plant model so that the exercise 3 of the Phase 2 can be investigated as next.
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| 16. |
- Suchoszek, Joanna, et al.
(author)
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RELAP5 and TRACE codes comparison and validation under steady-state and transient conditions on the basis of NUPEC data
- 2007
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In: 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12. - 0894480588 - 9780894480584
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Conference paper (peer-reviewed)abstract
- One-dimensional system codes are one of the main tools for the safety analysis of nuclear power plants. For BWR applications, it is of particular interest to know the performance of system codes in predicting void fraction, pressure drops, and, possibly, critical power in a wide range of conditions. The BFBT (BWR Full-Scale Fine-Mesh Bundle Tests) Benchmark [1], based on the NUPEC experiments, allows an accurate evaluation of the capabilities of the thermal-hydraulic codes in predicting the earlier mentioned quantities. The goal of this work is to evaluate the performance of RELAP5 and TRACE against the NUPEC experimental data. The fuel bundle employed in the benchmark has been modeled in the RELAP5 and TRACE input by a simple PIPE component using the fluid temperature and the mass flow rate as the inlet boundary condition and the pressure as the outlet boundary condition. We assessed the capability of RELAP5 and TRACE in predicting the void fraction, pressure drops and critical power in both steady-state and transient conditions. The obtained results were compared with the NUPEC measured data.
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