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  • 401.
    Villanueva, Walter
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head2012Inngår i: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 249, s. 146-153Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a 'ballooning' of the vessel bottom and (ii) a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel cooling was able to suppress the creep and subsequently prevent vessel failure for the considered pool depths.

  • 402.
    Villanueva, Walter
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Tran, Chi-Thanh
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Effect Of CRGT Cooling On Modes Of Global Vessel Failure Of A BWR Lower Head2012Inngår i: Proceedings Of The 20th International Conference On Nuclear Engineering And The ASME 2012 Power Conference - 2012, Vol2, 2012, s. 467-477Konferansepaper (Fagfellevurdert)
    Abstract [en]

    An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head. A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis. It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) 'ballooning' of the vessel bottom for smaller pools, and (b) 'localized creep' concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.

  • 403. Vorobyev, Y.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Development and Application of a Genetic Algorithm Based Dynamic PRA Methodology to Plant Vulnerability Search2011Inngår i: International Topical Meeting on Probabilistic Safety Assessment and Analysis 2011, PSA 2011, 2011, s. 559-573Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The paper describes recent achievements in development and application of the Dynamic Probabilistic Risk Analysis (DPRA) methodology based on the Genetic Algorithm (GA). The aim of the GA-DPRA approach is to enable identification of safety vulnerabilities and quantification of accident risks related to operation of nuclear power plants (NPP). The approach combines a system code as a deterministic model of the plant and a GA search engine for the exploration of the plant scenarios space. A point in this space represents a scenario (transient) which is defined by unique combination of initial plant state and time dependent sequence of changes in the plant state parameters implemented in the system code input. The GA-DPRA is used to address two main types of safety analysis problems: (i) identification of a "worst case" scenario with most severe violation of safety limits (failure of safety barriers); (ii) identification of "failure domains" (subdomains in the space of plant scenarios where at least one of the safety limits (barriers) is violated). Safety critical parameters (safety limits) are used by GA as fitness functions to guide selection of the system code input parameters in process of the global optimum search. The GA controls selection of system code input parameters within predefined diapasons and time windows. Unlike "brute force" approaches or Monte Carlo type methods the GA-DPRA is much less demanding to computational resources due to intelligent and adaptive resolution in the exploration of the plant scenarios space. Stochastic properties of GA and Importance Sampling technique are applied to estimate probabilistic characteristics of the identified vulnerabilities. Solutions of benchmark problems and comparison with other methods are discussed in the paper.

  • 404. Vorobyev, Y.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Development of methodology for identification of failure domains with GA-DPSA2012Inngår i: 11th International Probabilistic Safety Assessment and Management Conference and the Annual European Safety and Reliability Conference 2012, PSAM11 ESREL 2012: Volume 3, 2012, s. 2480-2489Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Methods of PSA/PRA play important role in understanding of threats to Nuclear Power Plants (NPPs) safety. However, static logic of PSA has difficulties in considering the dynamic nature of physical processes and their interaction with stochastic events. Different Dynamic PSA (DPSA) methods have been proposed to resolve the influence of timing and order of events in safety analysis of NPP. In this work we discuss a DPSA approach which employs global optimum search methods (particularly genetic algorithm (GA)) for the exploration of the uncertainty space (the space of plant accident scenarios and uncertain parameters) and a system code as a deterministic model of the plant. The GA is used to generate the system code input for probing the uncertainty space. Safety important parameters (e.g. peak cladding temperature etc.) are used by GA as objective functions (which are also often called fitness functions in GA) to guide selection of the system code input to find conditions at which safety limits are exceeded (failure domains (FD) in the uncertainty space). The biggest challenges in the problem of FD identification are (i) difficulties caused by enormous dimensionality of the space, (ii) large variations in sensitivity of the fitness function to different input parameters, (iii) significant cross correlations between input parameters, and (iv) non-monotonic behavior of the fitness function in the whole uncertainty space. In this paper we report most recent developments of the GA-DPSA methodology. Specifically we investigate the influence of the selection of GA internal parameters on the efficiency of failure domain identification. A method for probability estimation based on the neural networks is discussed. A test case for DPSA methods is proposed based on the LOCA scenario for a PWR model distributed along with the RELAP5 code. Presented test case reveals intricate dynamic interactions in different accident scenarios despite relative simplicity of the model.

  • 405. Vorobyev, Yu.B.,
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Jeltsov, Marti
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kööp, Kaspar
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Nhat, T.V.K.
    Application of information technologies (genetic algorithms, neural networks, parallel calculations) in safety analysis of Nuclear Power Plants2014Inngår i: Proceedings of the Institute for System Programming, ISSN 2220-6426, Vol. 26, nr 2, s. 137-158Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This paper investigates important issues in three types of safety assessment methodologies commonly applied for Nuclear Power Plants (NPP). These methodologies are i) dynamic probabilistic safety assessment (DPSA) where application of genetic algorithm (GA) is shown to improve the efficiency of the analysis, ii) deterministic safety assessment (DSA) with meta model representation of the system using pre-performed computational fluid dynamics (CFD) code and iii) vulnerability search (e.g. identification of accident scenarios in an NPP) with application of neural network (NN). The use of advanced computational tools and methods such as genetic algorithms, neural networks and parallel computations improve the efficiency of safety analysis. To achieve the best effect, these advanced technologies are to be integrated with existing classical methods of safety analysis of the NPP.

  • 406.
    Wallenius, Janne
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Maximum efficiency nuclear waste transmutation2019Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 125, s. 74-79Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Efficient recycle of high level waste in spent nuclear fuel decreases the time required to isolate residual waste by a factor of 100 and reduces the volume of the waste repository by a factor of 4-6. Technical approaches to accomplish this feat include fast neutron Generation IV reactors and accelerator driven systems. Here, I present a novel design of a very small, passively safe lead-cooled reactor with (Np,Am)N fuel, which is shown to achieve maximum possible efficiency in transmutation of long-lived high level waste, while producing a nuclear fuel that is difficult to use for weapons production. Using this reactor for waste management minimises costs for demonstrating closure of the nuclear fuel cycle.

  • 407.
    Wallenius, Janne
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. LeadCold Reactors, Valhallavägen 79, S-11428 Stockholm, Sweden..
    Bortot, Sara
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnenergiteknik. LeadCold Reactors, Valhallavägen 79, S-11428 Stockholm, Sweden..
    A small lead-cooled reactor with improved Am-burning and non-proliferation characteristics2018Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, s. 193-200Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In this paper, a novel approach for transmutation of americium in fast reactors is presented. Using enriched uranium as fissile support, rather than plutonium, it is shown that a minor actinide burning rate of 25 kg/TWh(th) is possible to achieve in a passively safe, critical lead-cooled reactor. Moreover, the plutonium produced by transmutation of Am-241 features up to 38% (PU)-P-238, making it difficult to use for weapons production.

  • 408.
    Wallenius, Janne
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Henriksson, Krister O. E.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Sandberg, Nils
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Olsson, P.
    The role of magnetism in modelling of fuels and structural materials2008Inngår i: Transactions of the American Nuclear Society, 2008Konferansepaper (Fagfellevurdert)
  • 409. Wang, K.
    et al.
    Bai, B.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A model for droplet entrainment in churn flow2013Inngår i: Chemical Engineering Science, ISSN 0009-2509, E-ISSN 1873-4405, Vol. 104, s. 1045-1055Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Understanding the mechanism of droplet entrainment is of great importance for the churn flow. So far, the droplet entrainment mechanism has been experimentally studied; however, no detailed model is available for this particular flow pattern. To address this, the author established an analytical model to better understand the drop entrainment in churn flow. In this model, only the entrainment mechanism named shear-off in equilibrium state is considered and detailed analysis performed for the interface stability based on the Kelvin-Helmholtz instability and force balance acting on the wave crest. The model has been verified using experimental data and different parameters (e.g., pipe diameter, gas and liquid flowrate and pressure) influencing the entrainment is presented. The author proposed a more accurate and reasonable formula for the entrained rate in churn flow based on the existing formula for annular flow. The model developed in this paper predicts the entrainment mechanism under churn flow condition to an accuracy of 30% which is essential for the development of mechanistic models to predict the dryout condition in the future.

  • 410. Wang, K.
    et al.
    Bai, B.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    An improved liquid film model to predict the CHF based on the influence of churn flow2014Inngår i: Applied Thermal Engineering, ISSN 1359-4311, E-ISSN 1873-5606, Vol. 64, nr 1-2, s. 422-429Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The critical heat flux (CHF) for boiling crisis is one of the most important parameters in thermal management and safe operation of many engineering systems. Traditionally, the liquid film flow model for "dryout" mechanism shows a good prediction in heated annular two-phase flow. However, a general assumption that the initial entrained fraction at the onset of annular flow shows a lack of reasonable physical interpretation. Since the droplets have great momentum and the length of churn flow is short, the droplets in churn flow show an inevitable effect on the downstream annular flow. To address this, we considered the effect of churn flow and developed the original liquid film flow model in vertical upward flow by suggesting that calculation starts from the onset of churn flow rather than annular flow. The results indicated satisfactory predictions with the experimental data and the developed model provided a better understanding about the effect of flow pattern on the CHF prediction.

  • 411. Wang, K.
    et al.
    Bai, B.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Huge wave and drop entrainment mechanism in gas-liquid churn flow2013Inngår i: Chemical Engineering Science, ISSN 0009-2509, E-ISSN 1873-4405, Vol. 104, s. 638-646Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    A profound knowledge of huge wave and droplet entrainment mechanism is crucial for the thorough study on the gas-liquid churn flow. Although studies have shown that the entrained fraction is high in churn flow and reaches the minimum around the churn-annular transition, the underlying mechanism of the drop entrainment in churn flow is still not well explored. To address this, we investigated the properties of the huge waves and the droplet entrainment in two vertical pipes with the inner diameter of 19. mm and 34. mm under churn flow conditions. We found that the flooding of the film was a characteristic of the churn flow throughout the regime. In addition, increasing the gas or liquid flow rate could lead to the transition from churn flow to annular flow or reverse to slug flow, providing the insight into the differences among slug, churn and annular flow. We also discussed the film instability under different flow conditions and tried to reveal the physical mechanism based on the instability analysis. In our study, the bag breakup and the ligament breakup were observed to coexist. The analysis of the liquid distribution in the cross-section of the pipes not only revealed the variations of the entrained fraction of churn flow from that of annular flow, but also indirectly illustrated the differences between their breakup mechanisms. Moreover, the wave properties (amplitude and frequency) were also analyzed in detail.

  • 412. Wang, Ke
    et al.
    Bai, Bofeng
    Cui, Jiahuan
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A physical model for huge wave movement in gas-liquid churn flow2012Inngår i: Chemical Engineering Science, ISSN 0009-2509, E-ISSN 1873-4405, Vol. 79, s. 19-28Artikkel, forskningsoversikt (Fagfellevurdert)
    Abstract [en]

    A complete knowledge of the huge wave in churn flow is of great importance for the characterization of its entrainment. Huge wave in churn flow is experimentally identified as a highly disturbed wave; however, no specific model is available for this particular wave. Based on the force balance over the wave, we established an analytical model to study its growth and levitation and analyzed the effects of the parameters (including gravity, pressure force of gas and liquid, wall shear stress and interfacial shear stress) on the wave and the gas and liquid flow field. We proposed that the boundary layer in liquid film is more likely to be turbulent rather than laminar and the gas pressure force is the most influential factor. The proposed model was verified qualitatively and quantitatively. We hence theoretically concluded that the churn flow is characterized by the flooding of the film, the flow reversal is attributed to the transition to the annular flow and the pressure gradient decreases with the increase of the gas flow rate. These findings provided insight into the distinction between the churn flow and the annular flow. The wave properties (amplitude and velocity) were analyzed in detail and the churn/annular transition occurred at U* = 1.12. The model helps understand the droplet entrainment in churn flow which is essential for the development of mechanistic models to predict the dryout condition.

  • 413.
    Wang, Ke
    et al.
    China Univ Petr, Beijing Key Lab Proc Fluid Filtrat & Separat, Beijing, Peoples R China.;KTH, Dept Phys, Stockholm, Sweden..
    Gong, Shengjie
    Shanghai Jiao Tong Univ, Sch Nucl Sci & Engn, Shanghai, Peoples R China..
    Bai, Bofeng
    Xi An Jiao Tong Univ, State Key Lab Multiphase Flow Power Engn, Xian, Shaanxi, Peoples R China..
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    On the Relation between Nucleation Site Density and Critical Heat Flux of Pool Boiling2018Inngår i: Heat Transfer Engineering, ISSN 0145-7632, E-ISSN 1521-0537, Vol. 39, nr 17-18, s. 1498-1506Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    It is traditionally accepted that the critical heat flux (CHF) decreases with increasing nucleation site density (NSD). However, such a CHF-NSD relation was no longer observed in the BETA-B experiment performed on nano-film heaters; instead the increase of NSD resulted in a gain in CHF. To address this seeming contradiction in the relation between critical heat flux and nucleation site density, the present work employed probabilistic analysis to reveal the different tendencies. A concept of effective NSD was proposed, which concerns the active nucleation sites appear within a bubble lifetime, and the resulting bubbles have the chance of direct interaction. We assumed that the boiling crisis on a heater surface is mainly induced by two mechanisms: dry spot expanding in isolated bubble regime for low-NSD surface, coalescence of dry spots under multiple bubbles in fully developed nucleate boiling regime for high-NSD surface, or a combination of the two in the transition regime for medium-NSD surface. Accordingly, we estimated the critical heat flux of each boiling regime at which the boiling crisis occurs. The result indicated that there is a threshold of nucleation site density below which the increase of NSD is contributing to CHF enhancement, while the trend is inverted beyond the threshold.

  • 414.
    Wang, Ke
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Zhang, Youjia
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    An experimental study on the dynamics of a liquid film under shearing force and thermal influence2015Inngår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, 2015, Vol. 5, s. 3668-3677Konferansepaper (Fagfellevurdert)
    Abstract [en]

    A profound knowledge of dynamics and instability of a liquid film is crucial for the thorough study on the physical mechanisms of boiling crisis (i.e., departure from nucleation boiling - DNB, and dryout), which is important to the performance and safety of light water reactors. To address this, a series of tests were carried out at various water and air flow rates under atmospheric pressure and different heat fluxes. A confocal optical sensor system was employed to investigate the dynamics of a liquid film in a horizontal aluminum channel. The liquid film was sheared by co-flowing air from above and heated from below. We obtained the experimental data about the variation of thickness of the liquid film under different flow conditions and tried to analyze how it is affected by liquid/gas flow rates, shearing force and heat flux. We found that the shear force, evaporation and the generation of the bubbles enhanced the instability of the liquid film. We also found that the occurrence of the rupture was random and the critical thickness at the rupture increased with increasing heat flux. The spectrum analysis indicated that the effect of the shear force on the liquid film instability became weak when the liquid film is very thin, but the heat flux always enhanced the instability of the liquid film.

  • 415. Willschütz, H. G.
    et al.
    Altstadt, E.
    Sehgal, Balraj R.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Weiss, F. P.
    Recursively coupled thermal and mechanical FEM-analysis of lower plenum creep failure experiments2006Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 33, nr 2, s. 126-148Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible failure mode of the reactor pressure vessel (RPV) and its failure time have to be investigated for a determination of the load conditions for subsequent containment analyses. Worldwide several experiments have been performed in this field accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model (FEM) has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post-test calculations for the FOREVER test series representing the lower head RPV of a pressurised water reactor (PWR) in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. In this paper the differences between the results of a simple coupled and a recursive coupled FE-simulation are highlighted. Due to the thermal expansion at the beginning and the accumulating creep strain later on the shape of the melt pool and of the vessel wall are changing. Despite of the fact that these relative small geometrical changes take place relatively slowly over time, the effect on the temperature field is rather significant concerning the mechanical material behaviour and the resulting failure time. Assuming the same loading conditions, the change in the predicted failure time between the simple and the recursive coupled model is in the order of magnitude of the total failure time of the simple model. The comparison with results from the FOREVER-experiments shows that the recursive coupled model is closer to reality than the one-way coupled model.

  • 416. Xing, M.
    et al.
    Hu, X.
    Chen, Y.
    Li, L.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of trace code against ROSA/LSTF test for SBLOCA of pressure vessel upper-head small break2014Inngår i: 2013 21st International Conference on Nuclear Engineering, ICONE 2013: Volume 4: Thermal Hydraulics, ASME Press, 2014, Vol. 4Konferansepaper (Fagfellevurdert)
    Abstract [en]

    OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, fullpressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. ROSA-V Test 6-1 simulates a pressure vessel (PV) upper-head small break loss-of-coolant accident (SBLOCA) with a break size equivalent to 1.9% of the volumetrically scaled cross-sectional area of the reference PWR cold leg. By building a TRACE calculation model of LSTF and PV upper-head, the paper dedicated to assess the effect of different modeling options and parameters on simulating thermal hydraulic behaviors of TRACE code. The results show that TRACE code well reproduce the physical phenomena involved in this type of SBLOCA scenarios. Almost all the events in the experiment are well predicted by the model based on TRACE code. In addition, the sensitivity of different models and parameters are investigated. For example, the code slightly overestimated the break mass flow from upper head which could affect the accuracy of the results significantly. The rising of core exit temperature (CET) is significantly influenced by the bypass flow area between downcomer and hot leg.

  • 417.
    Xing, Mian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Validation of TRACE Code against ROSA/LSTF Test for SBLOCA of Pressure Vessel Upper-Head Small Break2012Independent thesis Advanced level (degree of Master (Two Years)), 20 poäng / 30 hpOppgave
    Abstract [en]

    OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, full-pressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. ROSA-V Test 6-1 simulates a pressure vessel (PV) upper-head small break loss-of-coolant accident (SBLOCA) with a break size equivalent to 1.9% of the volumetrically scaled cross-sectional area of the reference PWR cold leg.The main objective of present thesis is to build a TRACE calculation model for simulating thermal hydraulic behaviors in LSTF and PV upper-head SBLOCA, so as to assess different modeling options and parameters of TRACE code. The results show that TRACE code well reproduce the complex physical phenomena involved in this type of SBLOCA scenarios. Almost all the events in the experiment are well predicted by the model based on TRACE code. In addition, the sensitivity of different models and parameters are investigated. For example, the code slightly overestimates the break mass flow from upper head which affects the accuracy of the results significantly. The rising of core exit temperature (CET) is significantly influenced by the flow area of leakage between downcomer and hot leg. Besides, the effect of the break location, low pressure injection system (LPIS) and accumulator setup are also studied.

  • 418. Xu, Y.
    et al.
    Downar, T.
    Ward, A.
    Kozlowski, Tomasz
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ivanov, K.
    Multi-physics coupled code reactor analysis with the U.S. NRC code system TRACE/PARCS2006Inngår i: PHYSOR-2006: American Nuclear Society's Topical Meeting on Reactor Physics, 2006Konferansepaper (Fagfellevurdert)
  • 419. Yakush, S. E.
    et al.
    Konovalenko, Alexander
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Effect of particle spreading on coolability of Ex-Vessel debris BED2015Inngår i: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, American Nuclear Society, 2015, s. 1210-1222Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Debris bed formation and coolability are studied by DECOSIM code. Main physical mechanisms affecting dispersed particle spreading in the course of melt-water interaction are considered, and their relevance to the formation of porous debris bed in various melt ejection modes is discussed. Numerical simulations of gradually growing and instantly formed debris beds are performed by DECOSIM code. Also, coupled simulations are carried out in which all mechanisms are taken into account simultaneously. It is shown that particle spreading limits the height of debris bed. Also, it is obtained that in some parameter ranges even if local dryout occurs, further particle spreading can render the debris bed coolable, resulting in its reflooding and quenching of the material.

  • 420. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Melt agglomeration influence on ex-vessel debris bed coolability2016Inngår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Konferansepaper (Fagfellevurdert)
    Abstract [en]

    A deep pool of water below reactor vessel is employed in some light water reactors in order to arrest the core melt progression in case of severe accident. The melt is expected to fragment and quench in the pool. Coolability of a porous debris bed in a water pool is considered in this work, with emphasis on the effect of agglomerates formed due to incomplete fragmentation of a melt jet, or incomplete freezing of melt particles reaching the debris bed. Agglomerates block the escape paths for vapor generated in the debris bed, increasing resistance to coolant flow and facilitating occurrence of dry zones. Numerical simulations are carried out by DECOSIM multiphase code. Two main types of debris bed inhomogeneity due to agglomeration are considered: (i) solid impermeable “cake” on the debris bed top, and (ii) distributed low-permeability low-porosity zone with properties determined according to prediction of an agglomeration model. In the presence of either type of flow obstruction, dryout occurs at significantly lower decay heat power than in the case of homogeneous non- agglomerated debris. There is a critical agglomeration fraction above which a large dry zone develops in the debris bed. Debris temperature escalates in the dry zone leading to subsequent remelting of the material. On the contrary, below the critical agglomeration fraction, the dry zone temperature is stabilized by vapor cooling. Implications of the obtained results for assessment of severe accident risks are discussed.

  • 421. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    On the evaluation of dryout conditions for a heat-releasing porous bed in a water pool2019Inngår i: International Journal of Heat and Mass Transfer, ISSN 0017-9310, E-ISSN 1879-2189, s. 895-905Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    This work is motivated by the problem of restraining temperature escalation inside a porous heat-releasing media submerged in a pool of liquid coolant. When coolant temperature reaches saturation, boiling begins in the bulk of the porous bed, with void generation rate determined by the heating power. Amount of void determines hydrostatic pressure difference that drives natural circulation of two-phase flow through the porous material. At a certain critical value of the heat release rate, the driving head cannot overcome drag of the two phase porous media flow, which results in complete evaporation of coolant in some zone. Temperature of material in the dry zone increases significantly due to deterioration of heat exchange with single phase vapor flow in comparison with boiling flow heat transfer. The paper considers the problem of determining the critical conditions for onset of dryout in a heat-releasing porous bed of an arbitrary shape. The well-known one-dimensional problem for a flat top-flooded bed is revisited, and the functional form of the dryout boundary (expressed as the dryout heat flux, DHF) is derived using non-dimensional parameters. Asymptotic behavior of the solution is analyzed, and, by the method of asymptotic interpolation, a surrogate model is proposed consisting of three single-argument, non-dimensional functions. It is shown that such a model provides acceptable accuracy even in the cases where complete similarity of solutions is not achieved. The results obtained provide important insights into the physics of the problem, reduce the number of free parameters, and enable fast evaluation of dryout conditions without the need of numerical solution of algebraic equations involved in the exact formulation. The ultimate goal of the surrogate model development, i.e. its application to multidimensional configurations, is discussed.

  • 422. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Transient phenomena of ex-vessel debris bed formation in an LWR severe accident2009Inngår i: Transactions of the American Nuclear Society, 2009, s. 546-547Konferansepaper (Fagfellevurdert)
  • 423. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Lubchenko, N. T.
    Coolability of heat-releasing debris bed. Part 1: Sensitivity analysis and model calibration2013Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 52, nr SI, s. 59-71Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Coolability of heat-releasing debris bed is an important issue in the severe accident analysis and management. Traditionally, theoretical studies of top or bottom-fed debris bed coolability have been focused on obtaining a "best estimate" value for the Dryout Heat Flux (DHF) as a function of debris bed parameters (mean particle diameter and porosity). However, an important question for safety analysis is the quantification of uncertainties inherent in the problem. In this paper, a one-dimensional coolability problem is considered, with the aim of analyzing the influence of aleatory uncertainties in input physical parameters and modeling (epistemic) uncertainties on the prediction of DHF. Global sensitivity analysis is applied to rank the aleatory and epistemic parameters according to their effects on DHF and average pressure drop. The most influential model parameters are then calibrated to achieve the best fit to experimental data available. On the one hand, we demonstrate that model calibration is instrumental in achieving considerable improvement of quantitative agreement between the experimental and simulation data. On the other hand, experience of model calibration also suggested that (i) optimization of model parameters with respect to available experimental data on DHF is an ill-posed problem, and (ii) model calibration with respect to one-dimensional pressure drop experiments does not automatically improve the prediction of DHF and in some cases can even worsen it. Based on these insights, one can speculate that further analytical and experimental efforts are necessary to establish a better consistency between model form and experimental data on pressure drop and DHF.

  • 424. Yakush, S. E.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Lubchenko, N. T.
    Coolability of heat-releasing debris bed. Part 2: Uncertainty of dryout heat flux2013Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 52, nr SI, s. 72-79Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    One-dimensional coolability problem for a flat homogeneous heat-releasing debris bed is considered, with the critical conditions for steady-state cooling characterized by the Dryout Heat Flux (DHF). DHF is determined for top-fed and bottom-fed debris beds from several two-phase models differing by the treatment of porous and interphase drag. Aleatory uncertainties due to randomness of the debris bed formation scenario and respective physical parameters (particle diameter, porosity) are quantified. The influence of ranges and distributions of input parameters on the uncertainty in the DHF are studied. Major contributors to the DHF uncertainty are identified. Influence of model uncertainty on the prediction of the lower boundary for DHF is discussed.

  • 425. Yakush, S. E.
    et al.
    Kudinov, Pavel Yu
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    A model for prediction of maximum post-Dryout temperature in decay-heated debris BED2014Inngår i: International Conference on Nuclear Engineering, Proceedings, ICONE, 2014, Vol. 2BKonferansepaper (Fagfellevurdert)
    Abstract [en]

    Several designs of light water reactors consider melt fragmentation and cooling of corium debris bed in a deep pool as important part of their severe accident management strategies. Traditional approach to assessment of debris coolability is based on the bed dryout criterion. However, this is the most conservative criterion which doesn’t take into account possibility of debris temperature stabilization in steam cooling regime. In this work numerical simulations of cooling of a decay heat-releasing porous debris bed in a water pool are carried out for the conditions where local dryout of porous material occurs. It is shown that the temperature of solid material in the dry zone can be stabilized if sufficient vapor flow is generated in the wetted part of the debris bed beneath the dry zone. A simple one-dimensional model which connects the maximum temperature and the relative size of the dry zone is proposed and verified against the numerical simulations with DECOSIM code for different shapes of the debris beds relevant to severe accident conditions in a Nordic type boiling water reactor (BWR). On the basis of this model, a criterion is obtained which defines the critical relative height of the dry zone corresponding to specific temperature of debris material which can be considered as a safety limit (e.g. start of zirconium oxidation, remelting of metallic debris or oxidic corium, etc.). The criterion allows one to evaluate the safety margins and degree of conservatism introduced by the dryoutbased approach to assessment of debris coolability.

  • 426. Yakush, S. E.
    et al.
    Lubchenko, N. T.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Risk-informed approach to debris bed coolability issue2012Inngår i: Proceedings Of The 20th International Conference On Nuclear Engineering And The ASME 2012 Power Conference - 2012, Vol 2, ASME Press, 2012, nr 1, s. 531-543Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Coolability of an ex-vessel debris bed in severe accident conditions is considered from the risk perspective. The concept of "load versus capacity" is employed to quantify the probability of failure (local dryout). Possible choices of "load" and "capacity" in terms of heat fluxes, thermal power or melt mass are discussed. Results of Monte Carlo simulations of distribution functions for the local heat flux and the dryout heat flux at the debris bed top point (defined as the extensions of one-dimensional counterparts) are presented. A surrogate model for the dryout heat flux is developed by the generalization of two-dimensional simulation results. Dryout probabilities are obtained under the conservative assumptions (neglecting the coolability improvement due to side ingress of water into a non-flat debris bed), and from the surrogate model. Outlook is given for the prospective development of the risk-informed approach to debris bed coolability in the context of comprehensive severe accident risk analysis.

  • 427.
    Yakush, S.
    et al.
    Institute for Problems in Mechanics, Russian Academy of Sciences, Moscow, Russia.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Simulation of Ex-Vessel Debris Bed Formation and Coolability in a LWR Severe Accident2009Inngår i: Proceedings of ISAMM 2009: Implementation of severe accident management measures, Paul Scherrer Institute , 2009Konferansepaper (Fagfellevurdert)
  • 428.
    Yakush, S.
    et al.
    Institute for Problems in Mechanics, Russian Academy of Sciences, Moscow, Russia.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Transient Phenomena of Ex-vessel Debris Bed Formation in a LWR Severe Accident2009Inngår i: American Nuclear Society Transactions 2009, American Nuclear Society, 2009, s. 546-547Konferansepaper (Fagfellevurdert)
  • 429. Yakush, S.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Modeling of Two-Phase Natural Convection Flows in a Water Pool with a Decay-Heated Debris Bed2008Inngår i: Proceedings of International Congress on Advances in Nuclear Power Plants (ICAPP 2008), 2008Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Coolability of a debris bed in a water pool is studied numerically with the emphasis on the effects of natural convection flows driven by vapor release in the porous bed. A multifluid computer code DECOSIM is developed and applied to the simulation of liquid-vapor flows in the saturated conditions relevant to the long transient of the debris bed formation and coolability problems. Numerical calculations carried out reveal the influence of the debris bed shape on the structure of natural circulation flows in the water pool, as well as on the void fraction distributions inside the debris bed. The effect of large scale flow structures in the pool on transport of debris particles is demonstrated. It is shown that the vortical flow can capture the smaller particle which may result in considerable spreading of the sedimented debris over the pool floor.

  • 430. Yakush, S.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Multiscale Simulations of Self-organization Phenomena in the Formation and Coolability of Corium Debris Bed2009Inngår i: Proc. The 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13), 2009Konferansepaper (Fagfellevurdert)
  • 431. Yakush, S.
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Lubchenko, N.T.
    Sensitivity and Uncertainty Analysis of Debris Bed Coolability2011Inngår i: The 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics (NURETH-14), 2011Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Theoretical studies of top-fed debris bed coolability available so far have been focused on obtaining the Dryout Heat Flux (DHF) as a function of debris bed parameters (mean particle diameter and porosity). In this paper, uncertainty analysis is carried out to quantify the influence of different factors on DHF. Global sensitivity analysis is applied to rank the drag model parameters according to their effects on DHF and average pressure drop (epistemic uncertainty). The most influential model parameters are then optimized to achieve the best fit to experimental data available. Finally, aleatory uncertainties due to randomness of the debris bed format

  • 432.
    Yakush, Sergey
    et al.
    Institute for Problems in Mechanics, Russian Academy of Sciences.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Effects of Water Pool Subcooling on the Debris Bed Spreading by Coolant Flow2011Inngår i: ICAPP2011: Proceedings of the 2011 international congress on advances in nuclear power plants, American Nuclear Society, 2011Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Deep water pool in the reactor pit is considered an effective means of arrest and long term cooling thecorium released from the reactor pressure vessel in the case of a hypothetical severe accident with lightwater reactors. Corium melt interaction with water is expected to result in the material fragmentation andformation of a porous debris bed on the containment basemat. From the safety point of view, it isessential to quantify the conditions for debris bed coolability without corium remelting due to the internalheat release. Previous research results have clearly demonstrated that, given the total mass of the material,the conditions for the occurrence of local dryout are contingent on the particle diameters, porosity, as wellas on the debris bed shape. In particular, a tall heap-shaped debris bed would be more prone to dryoutthan that uniformly spread over the pool basemat.For the BWR designs adopted in Sweden and Finland, one of the severe accident scenarios to beconsidered is the gradual melt release, for which the molten corium is released over a rather long time(hours) rather than as a concentrated jet. This case was studied recently by using the dedicated computercode DECOSIM (DEbris COolability SIMulator) developed at KTH. The code implements the multifluidmodel for liquid-vapor flows in the heat-releasing debris bed and in the volume of pool, where effects ofturbulence are taken into account by the k-epsilon turbulence model. The model used in DECOSIMdescribes the following phenomena: i) filtration of water and vapor in the porous debris bed with heatrelease; ii) turbulent natural convection flows in the water pool; iii) sedimentation of melt particles andtheir interaction with circulatory flow in the pool due to drag and turbulent dispersion; iv) fallout ofparticles, their packing and growth of debris bed. To address the multiscale nature of the problem, acomputationally efficient “gap-tooth” algorithm was developed to speed up considerably the simulationsof long transients typical of gradual melt release mode.Previous DECOSIM simulations of debris bed growth from falling melt particles were performedassuming saturated conditions in the pool at the system pressure. It was shown that the large scale naturalcirculation flows developing in the pool due to vapor production in the porous debris bed, affectsignificantly the debris bed shape because they capture the particles and cause their spreading over thebase-mat of the containment, making the debris bed more flattened and increasing its potentialcoolability. However, an important question remained on how these results would be affected if the poolwater is subcooled. In this case, condensation of vapor in the pool reduces the mixture buoyancy, whichcan decrease the effectiveness of particle spreading.In this work, numerical simulations by DECOSIM code are presented, in which the effects of poolsubcooling on the formation of debris bed are studied. It is shown that, in the subcooled case, thereduction of void due to vapor condensation is offset by the localized heating of water above the debrisbed. The density non-uniformities in the water pool caused by water heating are shown to be sufficient forthe development of large scale natural circulation flows, strong enough for the particle spreading to beefficient. The effect of water subcooling on the ultimate shape of debris bed is demonstrated bycomparing the results obtained for high and low subcooling. This work is performed within the MSWI project supported by APRI/ENSI/NKS

  • 433. Yakush, Sergey
    et al.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Basso, Simone
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    In-vessel Debris Bed Coolability and its Influence on the Vessel Failure2013Inngår i: Proc. 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15), 2013Konferansepaper (Fagfellevurdert)
  • 434.
    Yakush, Sergey
    et al.
    A. Yu. Ishlinskii Institute for Problems in Mechanics of the Russian Academy of Sciences.
    Lubchenko, Nazar
    Department of Nuclear Science and Engineering, Massachusetts Institute of Technology.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Risk And Uncertainty Quantification In Debris Bed Coolability2013Inngår i: 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Pisa, Italy, 2013Konferansepaper (Fagfellevurdert)
    Abstract [en]

    An approach to the quantification of uncertainties inherent in the debris bed coolability problem and associated risks analysis is presented. The approach relies on the “Load vs Capacity” concept which compares the thermal loads due to decay heat release and the maximum heat which can be safely removed by water evaporation, both quantities considered as uncertain values due to the uncertainties in the porous material properties, accident scenarios, debris bed shape etc. At the current level of knowledge, not only the distribution functions, but also the ranges of the input parameters cannot be fully quantified. To study the influence of these factors, distribution functions for the Load (decay heat power) and Capacity (maximum power which can be safety removed) are obtained using different ranges and distribution functions for the input parameters. In each case, the probabilities of dryout are evaluated for flat and cone-shaped debris beds. The results obtained demonstrate the advantages of risk-oriented approach and importance of its embedding into the technical decision-making process.

  • 435.
    Yakush, Sergey
    et al.
    A. Yu. Ishlinskii Institute for Problems in Mechanics of the Russian Academy of Sciences.
    Lubchenko, Nazar
    Department of Nuclear Science and Engineering, Massachusetts Institute of Technology.
    Kudinov, Pavel
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Surrogate Models for Debris Bed Dryout2013Inngår i: The 15th International Topical Meeting on Nuclear Reactor Thermal - Hydraulics, NURETH-15, 2013Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The problem of debris bed coolability is important for nuclear power plant severe accident management strategies which employ corium melt fragmentation in a deep pool of water as a means to terminate the accident progression. This work is concerned with development of computationally efficient methods for analysis of corium debris bed coolability. To evaluate the likelihood of severe accident progression due to reheating and remelting of the debris it is important to determine conditions for onset of the dryout in the bed. In the case of a flat one-dimensional debris bed, such conditions are called Dryout Heat Flux (DHF). The DHF is determined by the properties of the bed such as mean particle diameter, porosity, etc., by decay heat power and system pressure and can be obtained analytically. For non-flat configurations of the bed, numerical solution of the multidimensional problem of heat and mass transfer in a porous heat generating media is required in order to find the coolability boundary. In this paper we develop the general functional form of coolability conditions for arbitrary-shape debris beds. It is shown that the DHF concept for a flat debris bed can be extended naturally to multidimensional cases. On the basis of this analysis, a surrogate model is proposed which provides approximation of the coolability boundary, enabling fast calculation of the dryout conditions. The surrogate models enables application of the sensitivity, uncertainty and risk analysis of debris bed coolability.

  • 436. Yang, Zhi Lin
    et al.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Direct numerical simulation of dynamic three-fluid flow2007Inngår i: Progress in Computational Fluid Dynamics, An International Journal, ISSN 1468-4349, E-ISSN 1741-5233, Vol. 7, nr 2-4, s. 176-182Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    Three-fluid flow is characterised by three different interfaces between fluids, which are among the most important issues in a three-fluid flow. The objective of this paper is to track the dynamic interfaces between fluids implicitly and then to computerise the flow hydrodynamics. We use the level-set method for the interface tracking and a high-order Navier-Stokes solver, called Cubic-Interpolated Pseudo-Particle (CIP) algorithm to solve the equation system. Firstly, a two-fluid situation is simulated to demonstrate the capability of the level-set method. Then, as an example, we simulate a drop which is covered by a layer of gas/vapour and is moving in a water pool. The dynamic processes, such as deformation of vapour layer and its detachment from the drop, are simulated.

  • 437.
    Yu, Peng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. KTH.
    Komlev, Andrei A.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Li, Yichuan
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Pre-Test Simulations of SIMECO-2 Experiments on Stratified Melt Pool Heat Transfer2016Konferansepaper (Fagfellevurdert)
    Abstract [en]

    Severe accident progression in light water reactors can lead to the formation of a melt pool in the lower head that can impose thermo-mechanical loads on the pressure vessel, and subsequently can lead to vessel failure. For quantification of the thermal load which is important to in-vessel corium coolability and retention, various experiments have been carried out to investigate the heat transfer characteristics of melt pools, including the SIMECO experiment accomplished at KTH (Sehgal et al., 1998), which used low melting-point materials as the simulant of corium. In order to reduce the gaps in temperature and scale between experimental and prototypical conditions, a new test facility named SIMECO-2 is being designed at KTH (supported by the EU project IVMR), which features higher temperature (up to 900 ℃℃) and larger scale (1 meter in diameter), aiming to investigate the natural convection heat transfer of a stratified melt pool and the effects of different parameters/factors such as temperature of melt, thickness of boundary crust, thickness of top layer, top layer cooling. The present study is to provide pre-test calculations using the PECM method (Tran and Dinh, 2009), with the objectives to provide insights and analytical support to the design of the SIMECO-2 facility, including determination of required input power, as well as estimate of the temperature and heat flux distributions in the layers and time to reach steady state mode. A calculation was first performed for a reference base case with one-layer pool for which a CFD simulation was also conducted as benchmark. The calculations were then carried on to investigate the influences of different boundary conditions and internal heat sources on heat transfer. Finally the thermal behavior of a two-layer melt pool configuration was addressed in detail, and suggestions for the experimental conditions were provided.

  • 438.
    Yu, Peng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet. Royal Inst Technol KTH, Div Nucl Power Safety, Roslagstullsbacken 21, S-10691 Stockholm, Sweden..
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Karbojian, Aram
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment2019Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 133, s. 637-648Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The failure of reactor pressure vessel (RPV) during a severe accident of light water reactors is a thermal fluid-structure interaction (FSI) problem which involves melt pool heat transfer and creep deformation of the RPV. The present study is intended to explore a reliable coupling approach of thermo-fluid-structure analyses which will not only be able to reflect the transient thermal FSI feature, but also apply the advanced models and computational platforms to melt pool convection and structural mechanics, so as to improve simulation fidelity. For this purpose, the multi-physics platform of ANSYS encompassing Fluent and Structural capabilities was employed to simulate the fluid dynamics and structural mechanics in a coupled manner. In particular, the FOREVER-EC2 experiment was chosen to validate the coupling approach. The natural convection in melt pool was modeled with the SST turbulence model with a well-resolved boundary layer, while the creep deformation for the vessel made of 16MND5 steel was analyzed with a new three-stage creep model (modified theta projection model). A utility tool was introduced to transfer the transient thermal loads from Fluent to Structural which minimizes the user effort in performing the coupled analysis. The validation work demonstrated the well-posed capability of the coupling approach for prediction of the key parameters of interest, including temperature profile, total displacement of vessel bottom point and the evolution of wall thickness profile in the experiment. Ltd. All rights reserved.

  • 439.
    Yu, Peng
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Villanueva, Walter
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Galushin, Sergey
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR2018Konferansepaper (Fagfellevurdert)
    Abstract [en]

    We present a coupled thermo-mechanical creep analysis for a Nordic BWR lower head with a non-homogeneous debris bed configuration generated with MELCOR code. A one-way coupling approach was adopted which uses the Phase-Change Effective Convectivity Model implemented in Fluent to simulate the convective heat transfer in the melt pool and the ANSYS Mechanical to simulate the vessel wall deformation induced by the thermal and mechanical load from the debris. An initial non-homogeneity of debris bed was estimated using MELCOR core relocation simulation results specifying the mass of each component (UO2/Zr/ZrO2/SS/SSOX) and temperature in each MELCOR cell of the lower head. A mapping scheme was designed to transfer this non-homogeneities debris bed to Fluent through User Defined Functions. All components were locally treated in Fluent as one ideal phase by averaging the weights of element-specific mass fractions inside each cell. Material properties (density, heat capacity, etc.) and volumetric heat in the debris were both spatial- and temperature-dependent. Meanwhile, additional simulations using homogeneous debris bed configuration but with the same amount of mass compositions were run for comparison. Results including temperature escalation, vessel failure timing and location were analyzed and compared.

  • 440. Zambaux, J. A.
    et al.
    Manickam, Louis
    Meignen, R.
    Ma, W. M.
    KTH.
    Bechta, Sevostian
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Picchi, S.
    Study on thermal fragmentation characteristics of a superheated alumina droplet2018Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 119, s. 352-361Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In the frame of the European Commission FP7 SAFEST project, IRSN proposed to experimentally investigate the steam explosion triggering mechanisms of a superheated alumina droplet falling into water, through a set of experiments in the Micro Interactions in Steam Explosion Energetics facility (MISTEE) at KTH. Since thermal fragmentation is considered to be a likely process for the triggering of Steam Explosions in the KROTOS tests (performed at CEA) with alumina, the ability of a single droplet of such material to undergo thermally induced fine fragmentation is studied on the MISTEE facility with a close-up visualization. A series of experiments were conducted, where droplets of molten alumina were discharged into a water pool and potentially exposed to a small pressure wave. The intense interactions were recorded with a high-speed camera along with the pressure in the droplet vicinity. The ability of alumina to undergo thermal fragmentation is expected to be firstly contingent on the stability of the vapour film enshrouding the melt droplet. The water and melt temperatures may then play a crucial role on the vapour film stability, and therefore on the observation of a steam explosion. Indeed, under high to moderate water sub-cooling conditions, experimental observations indicate that fine fragmentation of the melt can occur when the droplet is exposed to even a weak pressure wave, in the range of 0.15 MPa. In contrast, melt fine fragmentation is suppressed at low water sub-cooling conditions (less than 30 °C), where the formation of a thick vapour film (and large wake) is observed, and which is probably too stable to be destabilized by the weak pressure wave. The effect of the melt temperature on thermal fragmentation is also assessed. This parameter influences the solidification of the droplet and the strength of the explosion as it determines the available heat energy. In the present conditions, fine fragmentation of melt occurred even at quite low melt superheat (≈60 °C). For a high melt superheat (above 200 °C) a very energetic spontaneous steam explosion was observed. A physical analysis on the debris particles acquired indicates a mass median diameter of ≈100 µm, comparable to the one observed in the KROTOS alumina experiments. The MISTEE experimental results are finally used to assess the heat and mass transfer modelling of the coolant during the fragmentation process in the FCI code MC3D.

  • 441.
    Zerkak, Omar
    et al.
    Paul Scherrer Institut.
    Gajev, Ivan
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Manera, Annalisa
    Paul Scherrer Institut.
    Kozlowski, Tomasz
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Gommlich, Andre
    Helmholtz-Zentrum Dresden-Rossendorf.
    Zimmer, Stéphanie
    Commissariat à l’Energie Atomique.
    Kliem, Sören
    Helmholtz-Zentrum Dresden-Rossendorf.
    Crouzet, Nicolas
    Commissariat à l’Energie Atomique.
    Zimmermann, Martin A.
    Paul Scherrer Institut.
    Revisiting Temporal Accuracy in Neutronics/THCode Coupling Using the NURESIM LWRSimulation Platform2011Inngår i: The 14th Internation Topical Meetingon Nuclear Reactor Thermalhydraulics (NURETH-14), 2011Konferansepaper (Fagfellevurdert)
    Abstract [en]

    The first part of this paper reviews the different temporal coupling methodologies that are currentlyemployed for the transient simulation of LWR cores. The second part shows preliminary resultsfrom the implementation of some suggested coupling improvements, including high-ordercorrections to the exchanged coupling fields and a dynamic time step control technique, for thesimulation of an exemplary reactivity insertion transient analysed using the European NURESIMLWR simulation platform.

  • 442. Zerkak, Omar
    et al.
    Kozlowski, Tomasz
    Gajev, Ivan
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Review of multi-physics temporal coupling methods for analysis of nuclear reactors2015Inngår i: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 84, s. 225-233Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    The advanced numerical simulation of a realistic physical system typically involves multi-physics problem. For example, analysis of a LWR core involves the intricate simulation of neutron production and transport, heat transfer throughout the structures of the system and the flowing, possibly two-phase, coolant. Such analysis involves the dynamic coupling of multiple simulation codes, each one devoted to the solving of one of the coupled physics. Multiple temporal coupling methods exist, yet the accuracy of such coupling is generally driven by the least accurate numerical scheme. The goal of this paper is to review in detail the approaches and numerical methods that can be used for the multi-physics temporal coupling, including a comprehensive discussion of the issues associated with the temporal coupling, and define approaches that can be used to perform multi-physics analysis. The paper is not limited to any particular multi-physics process or situation, but is intended to provide a generic description of multi-physics temporal coupling schemes for any development stage of the individual (single-physics) tools and methods. This includes a wide spectrum of situation, where the individual (single-physics) solvers are based on pre-existing computation codes embedded as individual components, or a new development where the temporal coupling can be developed and implemented as a part of code development. The discussed coupling methods are demonstrated in the framework of LWR core analysis.

  • 443.
    Zhang, Youjia
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftssäkerhet.
    Gong, S.
    An experimental study on liquid film dynamics and interfacial wave of air-water two-phase flow in a horizontal channel2013Inngår i: 2013 21st International Conference on Nuclear Engineering, ICONE 2013: Volume 4: Thermal Hydraulics, ASME Press, 2013Konferansepaper (Fagfellevurdert)
    Abstract [en]

    This study is concerned with liquid film dynamics and stability of annular flow, which plays an important role in understanding film rupture and dryout in boiling heat transfer. The research work starts from designing and making a test facility which enables the visualization and measurement of liquid film dynamics. A confocal optical sensor is applied to track the evolution of film thickness. A horizontal rectangular channel made of glass is used as the test section. Deionized water and air are supplied into that channel in such a way that an initial stratified flow forms, with the liquid film on the bottom wall. The present study is focused on characterization of liquid film profile and dynamics in term of interfacial wave and shear force induced film rupture under adiabatic condition. Based on the experimental data and analysis, it is found that given a constant water flowrate, the average thickness of water film decreases with increasing air flowrate, while the interfacial wave of the two-phase flow is intensified. As the air flowrate reaches a critical value, a localized rupture of the water film occurs.

  • 444.
    Zhong, Huaqiang
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    A Study on the Coolability of Ex-vessel Corium by Late Top Water Flooding2011Independent thesis Advanced level (degree of Master (Two Years)), 20 poäng / 30 hpOppgave
    Abstract [en]

    The molten core-concrete interaction (MCCI) is treated as one of the important phenomena that may lead to the late containment failure by basemat penetration in a hypothetical severe accident of light water reactors (LWRs). The earlier research has showed that heat transfer limitation exists for the coolability of ex-vessel corium by atop water flooding due to crust formation on the melt/water interface that will isolate melt from water. However, several cooling mechanisms were identified in a series of intense investigations. A code (CORQUENCH) was developed and updated to incorporate the newly identified cooling mechanisms for the better predictions of cavity erosion and corium cooling behaviors. A description about such cooling mechanisms (i.e., bulking cooling, water ingression, eruption and crust breach) and the concrete ablation models implemented in the code is presented in this thesis.

    The technical work in the thesis includes two parts: first, the verification and validation of the code were performed against the CCI tests from the OECD/MCCI projects; and then a reactor-scale simulation was carried out for MCCI and ex-vessel corium coolability of a reference PWR with LCS concrete. The calculations of CCI tests have a plausible agreement with the experimental data.

    The calculation predicts an optimistic result for the reactor case, and a fast quenching achieved at about 145 minutes. In addition, a sensitivity study was also conducted on several important parameters, i.e., concrete type, corium composition, water flooding time, atmosphere pressure, concrete ablation temperature, initial temperature, decay power, cavity geometry, concrete decomposition model and melt upper heat transfer model. An attempt to explain the physics of the different predicted phenomena is presented as well.

    Finally, comparative calculations were performed by the other codes (ASTEC and FinCCI) for the same reactor-scale configuration. Discrepancies are found in the results. Some suggestions are proposed to improve the CORQUENCH code.

  • 445.
    Ålander, Alexandra
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Dufek, Jan
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Gudowski, Waclaw
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    From once-through nuclear fuel cycle to accelerator-driven transmutation2006Inngår i: Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, ISSN 0168-9002, E-ISSN 1872-9576, Vol. 562, nr 2, s. 630-633Artikkel i tidsskrift (Fagfellevurdert)
    Abstract [en]

    In this study simulation of different nuclear fuel cycle scenarios are performed. The reference scenario corresponds to a medium size nuclear power country, with 10 light water reactors (LWRs). The study addresses long-term, equilibrium fuel cycle scenarios, with and without plutonium recycling (MOX) in LWRs and transuranics (TRU) burning in accelerator-driven systems (ADS). However, also short-term phase-out scenarios, including TRU burning in ADS, are performed. The equilibrium simulation showed that four ADS units, each of 800 MWt, are sufficient to burn an amount of plutonium and americium corresponding to the build-up of those elements. The phase-out simulation of a country adopting an approach to reduce the spent nuclear fuel inventory, showed that complementary burning of TRU in three to four ADS units appear suitable. The fuel cycle simulations have been performed using the Nuclear Fuel Cycle Simulation (NFCSim) code [C.G. Bathke, E.A. Schneider, NFCSim User's Manual, Los Alamos National Laboratory Report LA-UR-04-8369, 2004.] and the Monteburns code [D.I. Poston, H.R. Trellue, User's Manual, Version 2.0 for Monteburns, Version 1.0, LA-UR-99-4999, 1999.].

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