| 1. |
- Anglart, Henryk, et al.
(författare)
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Prediction and Analysis of Onset of Turbulent Convective Heat Transfer Deterioration in Supercritical Water Flows
- 2007
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Ingår i: Proceedings of the International Congress on Advances in Nuclear Power Plants. - 9781604238716 ; , s. 1552-1558
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Konferensbidrag (refereegranskat)abstract
- Supercritical water is considered as a coolant in one of the six systems defined as Generation IV reactors. Such reactor will operate at pressures higher than the thermodynamic critical point of water (374 °C and 22.1 MPa), allowing for a significant increase of the system thermal efficiency. During normal operation no boiling crisis will occur, thereby sudden temperature excursions will be avoided. However, since the physical properties of supercritical fluids change rapidly with temperature in the pseudocritical region, the local heat transfer coefficient may still show unusual behaviour depending upon the heat flux. It can be either enhanced or deteriorated, depending on flow conditions and heat flux. This phenomenon has to be properly modelled in order to correctly predict the cladding temperatures in the core of the nuclear reactor.
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| 2. |
- Cadinu, Francesco, et al.
(författare)
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Relating system-to-CFD coupled code analyses to theoretical framework of a multiscale method
- 2008
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Ingår i: 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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Konferensbidrag (refereegranskat)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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| 3. |
- Dinh, Truc-Nam
(författare)
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Multiphase flow phenomena of steam generator tube rupture in a lead-cooled reactor system : A scoping analysis
- 2008
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Ingår i: Societe Francaise d'Energie Nucleaire - International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work". - 9781604238716 ; , s. 2765-2775
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Konferensbidrag (refereegranskat)abstract
- The paper is concerned with understanding and quantification of intense multiphase interactions in a Steam Generator Tube Rupture (SGTR) scenario in advanced lead-cooled reactor systems. The scoping approach taken in this paper is to focus on key flow physics that complements other ongoing detailed computational and experimental efforts on SGTG analysis. The present study suggests that (i) the initial pressure shock wave poses no credible threat to invessel structures, except for limited pressure loading on very few adjacent heat-exchange tubes; (ii) the sloshing-relatedfluid motion is well bounded in a domain beyond the heat exchanger; (iii) the pre-mixture is not pre-conditioned for triggering and a postulated steam explosion would have limited energetics; and (iv) an initial discharge of steam/water mass is amenable for entrapment in the primary coolant flow. Implications for further research are discussed in the paper.
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| 4. |
- Kudinov, Pavel, et al.
(författare)
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A study of ex-vessel debris formation in a LWR severe accident
- 2007
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Ingår i: Proceedings of the International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work". - : Curran Associates, Inc.. - 9781604238716 ; , s. 2848-2859
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Konferensbidrag (refereegranskat)abstract
- In the paper we analyze phenomena that governdebrisformationand introduceacomprehensive framework to exhibit their interrelationship duringahypotheticalsevereaccidentinaboiling water reactor (BWR). We focus on phenomena feedbacks and identify key parameters which are believed to have significant effect ondebrispacking, including boiling regimes on fragments, their settling against steam flow stemming fromabottom bed. Based on scoping calculations for reactor scenarios, the prototypic range of the key parameters is delineated. Taking into account the practical and technical constraints of laboratory experiments with simulant fluids and results from calculations for experimental conditions, we establish feasibility and parameter ranges, under whichanew series of DEFOR-S "snap-shot" experiments shall be conducted to provide reactor relevant data and insights. Requirements on DEFOR-S experimental measurements and data analysis are also discussed in the paper.
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| 5. |
- Roshan Ghias, Sean, et al.
(författare)
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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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Ingår i: Proceedings of 2007 International congress on advances in nuclear power plants (ICAPP 2007). - 9781604238716 ; , s. 2642-2651
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Konferensbidrag (refereegranskat)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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| 6. |
- Sánchez, V. H., et al.
(författare)
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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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Ingår i: 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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Konferensbidrag (refereegranskat)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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| 7. |
- Starflinger, J., et al.
(författare)
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European research activities within the project : "High Performance Light Water Reactor Phase 2" (HPLWR Phase 2)
- 2008
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Ingår i: Int. Congr. Adv. Nucl. Power Plants - ICAPP, "Nucl. Renaiss. Work". - 9781604238716 ; , s. 867-876
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Konferensbidrag (refereegranskat)abstract
- The High Performance Light Water Reactor (HPLWR) is a Light Water Reactor (LWR) operating at supercritical pressure (25MPa). It belongs to the six reactors currently being investigated under the framework of the Generation IV International Forum. The most visible advantage of the HPLWR shall be the low construction costs in the order of1000€/kWe, because of size reduction of components and buildings compared to current LWR, and the low electricity production costs which are targeted at 3-4cents/kWh. In Europe, investigations on the HPLWR have been integrated into a joint research project, called High Performance Light Water Reactor Phase 2 (HPLWR Phase 2), which is co-funded by the European Commission. Within 42 months, ten partners from eight European countries working on critical scientific issues shall show the feasibility of the HPLWR concept. The final goal is to assess the future potential of this reactor in the electricity market.
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| 8. |
- Tran, Chi Thanh, et al.
(författare)
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An effective convectivity model for simulation of in-vessel core melt progression in a boiling water reactor
- 2008
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Ingår i: Societe Francaise d'Energie Nucleaire - International Congress on Advances in Nuclear Power Plants - ICAPP 2007, "The Nuclear Renaissance at Work". - 9781604238716 ; , s. 925-935
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Konferensbidrag (refereegranskat)abstract
- The present paper is concerned with development and application of a so-called Effective Convectivity Model (ECM), which aims to provide a detailed, mechanistic description of heat transfer processes in a BWR lower plenum. The ECM is a Computational Fluid Dynamics (CFD)-like tool which employs a simpler and more effective approach to compute heat transfer by solving only energy conservation equation instead of solving the full set of Navier-Stokes and energy equations by a CFD code. We implement the ECM in a CFD code (Fluent), with detailed description of the ECM development, implementation and validation. A dual approach is used to validate the ECM, namely validation against experimental data and against heat transfer results obtained by CFD predictions in the same geometries and conditions. Insights gained from CFD simulations are also used to improve ECM. The ECM capability as an effective tool to simulate heat transfer of an internally heated volume in 3D complex geometry is demonstrated through examples of heat transfer analysis in a BWR lower plenum being cooled by coolant flow in Control Rod Guide Tubes. Simulation results and key findings of this case are reported and discussed.
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| 9. |
- Von Lensa, W., et al.
(författare)
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Red-impact : A European research programme to assess the impact of partitioning and transmutation on final nuclear waste disposal
- 2008
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Ingår i: Int. Congr. Adv. Nucl. Power Plants - ICAPP, "Nucl. Renaiss. Work". - 9781604238716 ; , s. 2564-2573
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Konferensbidrag (refereegranskat)abstract
- It is the objective of the EU-funded 'Red-Impact' project to analyse the impact of partitioning, transmutation and waste reduction technologies on the final nuclear waste disposal. The partnership of 25 organisations is originating from European nuclear industry, waste agencies, research centres and universities. The system studies focus on a realistic evolution of P&T technologies and advanced fuel cycles which can be deployed incrementally on an industrial scale as well as on future developments such as reactors of the third and fourth generation (Gen III & Gen IV) and Accelerator Driven Systems (ADS). A comprehensive inventory of all existing and foreseen nuclear fuel cycle facilities in Europe has been performed including a review on worldwide ongoing R&D programs on P&T. Thus, it was possible to select a set of three so-called "industrial scenarios", taking into account industrial feasibility of alternate strategies leading to increased actiniae burning and reduced actinide generation based on direct disposal (reference case) or MOXfuel for LWR and plutonium recycle in Sodium Fast Reactors (SFR). R&D needs for the development of processes and technologies have also been addressed. In addition, three 'innovative scenarios ' have been identified allowing multi-recycling of plutonium and minor actinides in SFR and Accelerator-Driven Systems (ADS) as well as GANEX or COEXprocess and PYRO reprocessing technologies. Waste streams have been calculated for all of these scenarios including the transition from the present situation towards new fuel cycle options. These data provide the input to specific analyses on the impact on geological disposal in different host formations such as granite, clay and salt. The results show that advanced fuel cycles influence the required size of the geological repository in case of disposal in clay, salt or hard rock formations. Recycling of all the actinides results in a reduction of the necessary gallery length (depending on geology and design) at least by a factor 3. If additionally cesium and strontium are extracted from the high-level waste for separate decay, the reduction factor will become 10 or more. In the frame of the project, the feasibility and the impact of the Cs or Sr separated management were not assessed or evaluated. Transmutation of the actinides fast neutron spectrum reactors (FR or ADS) results in a limited reduction of the maximum dose because the dose is essentially due to long-lived fission and activation products. On the other hand, reprocessing the spent fuel decreases the maximum dose at the storage with a factor 5 because a considerable fraction of the iodine is separated from the high level waste during reprocessing. The radiotoxicity in the high level waste or spent fuel as well as human intrusion doses after 500 years are drastically reduced by the transmutation of the actinides. Evaluating actinide minimization systems and industrialised P&T in general requires an assessment of relevant nuclear fuel cycles especially with regard to the economic, environmental and societal advantages/disadvantages (i.e. the sustainability of the fuel cycles). Thus, a set of indicators has been derived for each of these areas. The results are analysed using the multi-criterion analysis approach which allows the importance of each of the indicators to be specified.
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