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  • 1.
    Bechta, Sevostian
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Miassoedov, Alexei
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Journeau, Christophe
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Okamoto, Koji
    Univ Tokyo, Tokyo, Japan..
    Manara, Dario
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Bottomley, David
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Kurata, Masaki
    JAEA CLADS Lab, Iwaki, Fukushima, Japan..
    Sehgal, Bal Raj
    Stuckert, Jun
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Steinbrueck, Martin
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Fluhrer, Beatrix
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Keim, Torsten
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Fischer, Manfred
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Langrock, Gert
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Piluso, Pascal
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Hozer, Zoltan
    MTA EK, Budapest, Hungary..
    Kiselova, Monika
    UJV REZ As, Hlavni 130, F-25068 Husinec Rez, Czech Republic..
    Belloni, Francesco
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    Schyns, Marc
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    On the EU-Japan roadmap for experimental research on corium behavior2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 124, p. 541-547Article in journal (Refereed)
    Abstract [en]

    A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.

  • 2.
    Chen, Yangli
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sensitivity and uncertainty analysis of trace simulation against FIx-II experiments2016In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper (Refereed)
    Abstract [en]

    In a previous study [1], the US NRC code TRACE was employed to simulate the FIX-II tests which were carried out to investigate the loss of coolant accident (LOCA) of a boiling water reactor (BWR). Results exhibited that the TRACE simulation was sensitive to modelling parameters. In order to further qualify the TRACE code for BWR safety analysis, and to increase our confidence in the simulation results, sensitivity and uncertainty analysis is performed in this paper for the possible uncertain parameters, so as to identify the most influential ones. 12 parameters related to the simulated physical phenomena are selected by resorting to phenomena identification and ranking tables (PIRTs) in relative references. The sensitivity analysis method chosen is based on Finite Mixture Models (FMM) together with Hellinger distance and Kullback-Leibler divergence. Kolmogorov-Smirnov test is first introduced to combine FMM, and it has better performance in screening. Sensitivity analysis results of FMM method show that decay power, choked flow multipliers and break area have the most important influence on calculating peak cladding temperature (PCT). Although previous study failed to predict PCT, uncertainty analysis provides a certain range that successfully covers experiment result.

  • 3.
    Chen, Yangli
    et al.
    KTH.
    Zhang, Huimin
    KTH.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 22-37Article in journal (Refereed)
    Abstract [en]

    This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.

  • 4.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Numerical investigation on quench of an ex-vessel debris bed at prototypical scale2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, p. 47-61Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with the coolability of the heap-like debris beds formed in the cavity of a Nordic-type boiling water reactor (BWR) during a postulated severe accident. A numerical simulation using the MEWA code was performed to investigate the quenching process of the ex-vessel debris bed at post-dryout condition upon its formation. To qualify the simulation tool, the MEWA code was first employed to calculate the quenching tests recently conducted on the PEARL facility. Comparisons of the simulation results with the experimental measurements show a satisfactory agreement. The simulation for the debris bed of the reactor scale shows that the heap-like debris bed flooded from the top is quenched in a multi-dimensional manner. The upper region adjacent to the centerline of the bed is the most difficult for water to reach under the top-flooding condition, and thus is subject to a higher risk of remelting. The oxidation of the residual Zr in the corium has a great impact on the coolability of the debris bed due to (i) large amount of reaction heat and the subsequent positive temperature feedback, (ii) the local accumulation of the produced H2 which may create a “steam starvation” condition and suppresses the oxidation. As possible mitigation measures of oxidation, the effects of bottom-flooding and bypass on quench were also investigated. It is predicted that the debris bed becomes more quenchable with water injected from the bottom, especially for the case with the floor partially flooded in the center. A bypass channel embedded in the center of the debris bed can also promote the quenching process by providing a preferential path for both steam escape and water inflow.

  • 5.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubesManuscript (preprint) (Other academic)
  • 6.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 351, p. 189-202Article in journal (Refereed)
    Abstract [en]

    Since the reactor pressure vessel (RPV) of a typical BWR features a lower head that is penetrated by a forest of control rod guide tubes (CRGTs), coolability of the debris bed formed in the lower head during a severe accident can be realized by coolant injection through the CRGTs (so-called "CRGT cooling"). This paper is concerned with performance assessment of such CRGT cooling system, whose heat removal capacity is determined by two mechanisms: (i) heat-up and boiling of coolant inside the CRGTs; and (ii) evaporation of coolant which reached the top of the debris bed from CRGTs (top flooding). For this purpose, analyses were accomplished by coupling the COCOMO and RELAP5 codes, which simulate the quenching process of the debris bed and the coolant flow inside the CRGTs, respectively. An analysis was first carried out for a unit cell with a single CRGT, whose decay heat removal was limited by heat conduction from debris to the CRGT wall. The simulation indicated that without top flooding, though the temperature of the unit cell was eventually stabilized by the cooling of the CRGT wall, remelting of metallic debris (Zr) in the peripheral region was unavoidable due to low conductivity of corium. Boiling in the CRGT was not only beneficial to heat transfer, but also contributing to a flat axial temperature profile. Given the nominal flowrate of the CRGT cooling, the coolant was not completely boiled off in the CRGT, and therefore the remaining liquid water at the outlet of the CRGT was available for top flooding of the debris bed. The subsequent simulation including the top flooding showed that the debris bed was rapidly quenched without any remelting. However, the top flooding may have a side effect which was Zr oxidation risk at high temperature, leading to production of reaction heat and H-2. Finally analyses were performed for prototypical cases for a reference Nordic BWR, and the results implied that the CRGT cooling could be used as a promising strategy for severe accident mitigation. It is critical that the debris bed is sufficiently cooled down during its formation so that the oxidation risk is eliminated when the CRGT cooling is applied.

  • 7.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Performance evaluation of passive containment cooling system of an advanced PWR using coupled RELAP5/GOTHIC simulation2016In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 83-92Article in journal (Refereed)
    Abstract [en]

    Motivated to investigate the thermal hydraulic characteristics and performance of a passive containment cooling system (PCS) for a Generation III pressurized water reactor (PWR), a coupled RELAP5/GOTHIC model was developed, which was then employed to simultaneously simulate the transient responses of the PCS and the containment during a large break loss of coolant accident of the reactor. The results show that the PCS is capable of lowering the containment pressure to an acceptable level for a long period (up to 3 days). In a separate-effect study, it was found that the height of the PCS loop plays an important role in determining the flow characteristics and heat removal performance of the PCS. Within the range of the considered loop heights, phase change occurs in the riser of the loop after the height exceeds a specific value (between 13 m and 15 m), below which only single-phase flow takes place. With increasing height of the loop, the heat removal capability increases monotonically at first; however, it is no longer sensitive to the height after two-phase flow appears. Finally, a feed-and-bleed operation for the cooling tank of the PCS was proposed as an enhancement measure of the heat removal capacity, and the simulation results show it further mitigates the accident. Moreover, a simplified analytical model is developed to predict the impact of the feed-and-bleed flowrate on the PCS performance, which can be used in engineering design.

  • 8.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 128, p. 330-340Article in journal (Refereed)
    Abstract [en]

    In the aftermath of the Fukushima Daiichi nuclear accident, a great concern has been raised about enhancing the inherent safety of a spent fuel pool (SFP). A passive cooling system using two-phase thermosiphon loops was concerned in this paper. A RELAP5/MELCOR coupling interface was developed, aiming at simultaneously simulating the transient behaviors of the SFP (by MELCOR) and the passive cooling system (by RELAP5). First the RELAP5 model of the thermosiphon loop was qualified against an experiment of a prototypical scale. Comparisons between the experiment and predictions show a good agreement. MELCOR standalone calculations for both station blackout (SBO) and loss of coolant accident (LOCA) without the passive cooling system demonstrate severe degradation of fuel rods. In contrast, for the SBO accident, the coupling simulation shows that the passive cooling system can effectively remove the decay heat, thus keeping fuel rods intact. As for the LOCA scenario, it is more challenging for the passive cooling system due to: (i) the heat transfer power is low during the drainage of water since the natural circulation of steam is blocked by the residual water at the bottom, leading to unavoidable heat-up and oxidation of fuel cladding; (ii) the heat transfer coefficient between steam and the evaporator is very small, which consequently may require a larger heat transfer surface area. Nevertheless, the heat transfer power substantially increases after the pool is emptied and natural circulation is established. The decay heat can be removed by steam convection, thus maintaining the mechanical integrity of fuel rods and stabilizing the fuel temperature eventually. It is also observed that H 2 production is undesirably promoted because the steam supply is enhanced. However such adverse effect can be diminished by increasing the thermosiphon loops number.

  • 9.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation and application of the MEWA code to analysis of debris bed coolability2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 327, p. 22-37Article in journal (Refereed)
    Abstract [en]

    This paper was first aimed at validating the MEWA code against experiments for two-phase flow and dryout in particulate beds, and then investigating the coolability of ex-vessel debris beds with cylindrical, conical and truncated conical shapes assumed to form under severe accident scenarios of a boiling water reactor. The validation was mainly performed against the POMECO-FL and POMECO-HT experiments carried out at KTH for investigating frictional laws and coolability limit (dryout) of particulate beds, respectively. The comparison of the experimental and numerical results shows that the MEWA code is capable of predicting both the pressure drop of two-phase flow through porous media and the dryout condition of various stratified beds. While the coolability of a one-dimensional homogeneous debris bed is bounded by counter-current flow limit (CCFL), the coolability of a heap-like debris bed can be improved due to lateral ingression of coolant in a multi-dimensional geometry. The simulations showed that the dryout power density of a prototypical debris bed was roughly inversely proportional to the bed's height regardless of the bed's shape. The impacts of a debris bed's features on coolability are manifested in three aspects: multidimensionality and contour surface area of the bed, as well as the uniformity of its shape. The contour surface area is defined as the interface between debris bed and water pool, and its effective value depends on the surface orientation that determines the amount of water ingress and vapor escape. The perfect uniformity in bed's shape as cylindrical bed results in even distributions of temperature and void fraction. The dryout power density was also predicted to be strongly correlated to the uniformity of bed's shape. The MEWA simulation also predicted that coolability was improved by an downcomer embedded in the center of debris bed. The efficiency of such enhancement was largely determined by the downcomer's length, whose optimal value was obtained in simulation.

  • 10. Journeau, C.
    et al.
    Bouyer, V.
    Cassiaut-Louis, N.
    Fouquart, P.
    Piluso, P.
    Ducros, G.
    Gossé, S.
    Guéneau, C.
    Quaini, A.
    Fluhrer, B.
    Miassoedov, A.
    Stuckert, J.
    Steinbrück, M.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sehgal, Bal Raj
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Hozer, Z.
    Guba, A.
    Manara, D.
    Bottomley, D.
    Fischer, M.
    Langrock, G.
    Schmidt, H.
    Kiselova, M.
    Ždarek, J.
    Safest roadmap for corium experimental research in Europe2018In: ASCE-ASME Journal of Risk and Uncertainty in Engineering Systems, Part B: Mechanical Engineering, ISSN 2332-9017, Vol. 4, no 3, article id 030901Article in journal (Refereed)
    Abstract [en]

    Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) severe accident (SA) with reactor core melting and formation of hazardous material system known as corium. The main objective of the project is to establish coordinated activities, enabling the development of a common vision and severe accident research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on severe accident experimental research has been developed to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. The roadmap takes into account different SA phenomena and issues identified and prioritized in the analyses of severe accidents at commercial NPPs and in the results of the recent European stress tests carried out after the Fukushima accident. Nineteen relevant issues related to reactor core meltdown accidents have been selected during these efforts. These issues have been compared to a survey of the European SA research experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. The comparison shows certain important lacks in SA research infrastructures in Europe, especially in the domains of core late reflooding impact on source term, reactor pressure vessel failure and molten core release modes, spent fuel pool (SFP) accidents, as well as the need for a large-scale experimental facility operating with up to 500 kg of chemically prototypic corium melt.

  • 11. Li, W.
    et al.
    Qi, Z.
    Ye, Y.
    Yuan, Y.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On improvement of a conditional mornitoring technique for condition-based maintenance2019In: International Conference on Nuclear Engineering, Proceedings, ICONE, ASME Press, 2019Conference paper (Refereed)
    Abstract [en]

    The condition-based maintenance (CMB) is a hot research topic to overcome the drawbacks belonging to the periodic maintenance used in nuclear power plants nowadays. Auto-Associative Kernel Regression (AAKR) is a widely applied condition monitoring technique which is the basis of a CBM. In this paper, the traditional AAKR is improved by using the ensemble learning technique. The modified AAKR is tested by steady-state operational data of a Tennessee-Eastman chemical process and the results show that it can significantly improve the auto- and cross-sensitivity without reducing the accuracy. This indicates a significant improvement in performance of this condition monitoring technique.

  • 12.
    Manickam, Louis
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the fragmentation characteristics of melt jets quenched in water2017In: International Journal of Multiphase Flow, ISSN 0301-9322, E-ISSN 1879-3533, Vol. 91, p. 262-275Article in journal (Refereed)
    Abstract [en]

    Experiments were carried out to investigate the characteristics of jet breakup and debris formation after melt jets fall into a subcooled water pool, which may occur in industrial processes such as the interactions of molten corium with coolant during a severe accident of light water reactors. A high-speed visualization system developed previously at KTH was used to capture the jet fragmentation phenomenon. Molten metal (Woods metal or tin) and mixture of binary oxides (WO3-Bi2O3 or WO3-ZrO2) were employed separately as melt materials to address different breakup mechanisms (e.g., hydrodynamic vs. thermodynamic fragmentation) and material effect. Moreover, the parameters related to melt and water conditions, including superheat, jet diameter and velocity of melt as well as subcooling of water, were scrutinized for their influences on jet fragmentation characteristics. The experimental data were acquired on the melt jet fragmentation patterns, breakup length, droplet size spectrum, droplet breakup and solidification as well as debris morphology, which can be useful for validation of the codes used for the prediction of debris formation phenomena.

  • 13.
    Manickam, Louis
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Guo, Qiang
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Komlev, Andrei A.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Oxidation of molten zirconium droplets in water2019In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 354, article id UNSP 110225Article in journal (Refereed)
    Abstract [en]

    Zirconium, which is used as the cladding material of nuclear fuel rods in LWRs, can react with steam in the case of a core meltdown accident. This results in the release of hydrogen which poses a significant risk of hydrogen explosion. Oxidation of Zr occurs either during the core degradation when the steam flows over the hot fuel rod surfaces or during an FCI when molten corium falls into a water pool (e.g. in the lower head). An experimental study was performed at the MISTEE-OX facility at KTH to observe and quantify the oxidation of molten zirconium droplets in a water pool. During the experimental runs, single droplets of molten zirconium were discharged into a subcooled water pool and the dynamic events were recorded using a high-speed camera. The bubble dynamics indicate a process of cyclic oxidation and hydrogen release from the rear periphery of a droplet while it is quenched in the water. The debris (solidified state of the droplet) after each run was collected for compositional and microstructural analysis via SEM/EDS. The obtained data were employed to estimate the oxidation fractions of the droplets and the results revealed several interesting insights into the oxidation phenomenon of the Zr melt. The water subcooling was observed to have a significant influence on the oxidation: the degree of oxidation decreased with increase in the water subcooling. Furthermore, the degree of oxidation was also influenced by the depth into the debris, forming compounds whose oxygen content decreases from the outer surface towards the core of the debris. Therefore, the qualitative and quantitative results presented in this paper are important in the context of developing a phenomenological understanding of the oxidation behaviour of zirconium melt during the FCI as well as to improve and validate the currently available models implemented in the state-of-art steam explosion codes.

  • 14.
    Manickam, Louis
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Guo, Qiang
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An experimental study on the intense intense heat transfer and phase change during melt and water interactions2019In: Experimental heat transfer, ISSN 0891-6152, E-ISSN 1521-0480, Vol. 32, no 3, p. 251-266Article in journal (Refereed)
    Abstract [en]

    Accidental contact between hot melt and cold water poses fatal hazard in several industries. Vapor explosion during melt-water contact in nuclear power plant accident can result in catastrophic containment failure. The fast transient phenomena as vapor explosion is not comprehensively understood despite several advances in research. It is not clear why certain parameters of melt and water exhibit differences in fragmentation behavior. To examine the influential parameters, we perform a series of experiments. The interactions between melt and water is visualized by high-speed video and X-ray radiograph.

  • 15.
    Manickam, Louis
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the influence of water subcooling and melt jet parameters on debris formation2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 309, p. 265-276Article in journal (Refereed)
    Abstract [en]

    Breakup of melt jet and formation of a porous debris bed at the base-mat of a flooded reactor cavity is expected during the late stages of a severe accident in light water reactors. Debris bed coolability is determined by the bed properties including particle size, morphology, bed height and shape as well as decay heat. Therefore understanding of the debris formation phenomena is important for assessment of debris bed coolability. A series of experiments was conducted in MISTEE-jet facility by discharging binary-oxide mixtures of WO3-Bi2O3 and WO3-ZrO2 into water in order to investigate properties of resulting debris. The effect of water subcooling, nozzle diameter and melt superheat was addressed in the tests. Experimental results reveal significant influence of water subcooling and melt superheat on debris size and morphology. Significant differences in size and morphology of the debris at different melt release conditions is attributed to the competition between hydrodynamic fragmentation of liquid melt and thermal fracture of the solidifying melt droplets. The particle fracture rate increases with increased sub cooling. Further the results are compared with the data from larger scale experiments to discern the effects of spatial scales. The present work provides data that can be useful for validation of the codes used for the prediction of debris formation phenomena.

  • 16. Mei, Y.
    et al.
    Gong, Shengjie
    KTH. School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, China.
    Gu, H.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    A study on steam-water two phase flow distribution in a rectangular channel with different channel orientations2018In: Experimental Thermal and Fluid Science, ISSN 0894-1777, E-ISSN 1879-2286, Vol. 99, p. 219-232Article in journal (Refereed)
    Abstract [en]

    Experimental study on steam-water two phase vertical and inclined upward flow (15–90°) was performed in a rectangular channel with cross section of 17 mm × 10 mm under atmospheric pressure to investigate the phase distribution and the average void fraction in the cross section which were obtained from the local void fraction measurement by a conductivity probe. The inlet superficial velocities of the steam and water varied from 0.72 to 3.85 m/s and from 0.11 to 0.3 m/s respectively. A high speed camera was used to identify the flow patterns. Experimental results show that the phase distribution curves are significantly affected by channel orientation and the average void fraction first decreases and then increases with the increase of orientation. Based on the drift-flux model, two parameters, namely, the distribution parameter (C0) and the drift velocity (Ugm) have been studied in detail. Both the distribution parameter and the drift velocity are found to be functions of orientation. The distribution parameter decreases with the increase of orientation while the drift velocity first increases and then decreases with the increase of orientation., Based on the experimental data, an improved drift-flux model is proposed especially for the slug and churn flow, which predicts the void fraction in an inclined channel with good accuracy.

  • 17. Qi, Z.
    et al.
    Hong, J.
    Li, W.
    Yuan, Y.
    Zhang, Y.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Application of nonlinear principal component analysis technique to nuclear power plants2019In: International Conference on Nuclear Engineering, Proceedings, ICONE, ASME Press, 2019Conference paper (Refereed)
    Abstract [en]

    Traditionally, manual calibration of sensors is required and performed during each refueling outage. If the traditional time-directed calibration is replaced by an online monitoring technique, the maintenance cost will be significantly reduced since only the abnormal sensors identified in on-line monitoring need to be re-calibrated or replaced off-line. The Nonlinear Principal Component Analysis (NLPCA), such as Auto-Associative Neural Network (AANN) and Auto-Associative Kernel Principal Component Analysis (AAKPCA), can describe the nonlinear correlation between sensors such as power, temperature, pressure and flowrate. In this paper, AANN and AAKPCA model are tested by simulated redundant data and Tennessee-Eastman process data. The results show that both of them have a high ability of prediction and a low sensitivity. Therefore, they are can be used in on-line monitoring.

  • 18.
    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, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    On the Relation between Nucleation Site Density and Critical Heat Flux of Pool Boiling2018In: Heat Transfer Engineering, ISSN 0145-7632, E-ISSN 1521-0537, Vol. 39, no 17-18, p. 1498-1506Article in journal (Refereed)
    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.

  • 19.
    Yu, Peng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. KTH.
    Komlev, Andrei A.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Li, Yichuan
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Pre-Test Simulations of SIMECO-2 Experiments on Stratified Melt Pool Heat Transfer2016Conference paper (Refereed)
    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.

  • 20.
    Yu, Peng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Royal Inst Technol KTH, Div Nucl Power Safety, Roslagstullsbacken 21, S-10691 Stockholm, Sweden..
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 133, p. 637-648Article in journal (Refereed)
    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.

  • 21.
    Yu, Peng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR2018Conference paper (Refereed)
    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.

1 - 21 of 21
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