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Ma, Weimin
Publications (10 of 21) Show all publications
Chen, Y., Zhang, H., Villanueva, W., Ma, W. & Bechta, S. (2019). A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor. Nuclear Engineering and Design, 343, 22-37
Open this publication in new window or tab >>A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor
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2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 22-37Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Boiling water reactor, Reactor safety, Severe accident, MELCOR simulation, Mesh sensitivity analysis
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-244082 (URN)10.1016/j.nucengdes.2018.12.011 (DOI)000456923500003 ()2-s2.0-85059233155 (Scopus ID)
Note

QC 20190219

Available from: 2019-02-19 Created: 2019-02-19 Last updated: 2019-04-29Bibliographically approved
Manickam, L., Guo, Q., Ma, W. & Bechta, S. (2019). An experimental study on the intense intense heat transfer and phase change during melt and water interactions. Experimental heat transfer, 32(3), 251-266
Open this publication in new window or tab >>An experimental study on the intense intense heat transfer and phase change during melt and water interactions
2019 (English)In: Experimental heat transfer, ISSN 0891-6152, E-ISSN 1521-0480, Vol. 32, no 3, p. 251-266Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Taylor & Francis Group, 2019
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-239897 (URN)10.1080/08916152.2018.1505786 (DOI)000462344000004 ()2-s2.0-85052292697 (Scopus ID)
Note

QC 20181214

Available from: 2018-12-05 Created: 2018-12-05 Last updated: 2019-05-03Bibliographically approved
Qi, Z., Hong, J., Li, W., Yuan, Y., Zhang, Y. & Ma, W. (2019). Application of nonlinear principal component analysis technique to nuclear power plants. In: International Conference on Nuclear Engineering, Proceedings, ICONE: . Paper presented at 27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019; Tsukuba International Congress Center,Tsukuba, Ibaraki; Japan; 19 May 2019 through 24 May 2019. ASME Press
Open this publication in new window or tab >>Application of nonlinear principal component analysis technique to nuclear power plants
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2019 (English)In: International Conference on Nuclear Engineering, Proceedings, ICONE, ASME Press, 2019Conference paper, Published 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.

Place, publisher, year, edition, pages
ASME Press, 2019
Keywords
Auto-associative kernel principal component analysis, Auto-associative neural network, Feature extraction, Nonlinear correlation, Online monitoring
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-258166 (URN)2-s2.0-85071385804 (Scopus ID)9784888982566 (ISBN)
Conference
27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019; Tsukuba International Congress Center,Tsukuba, Ibaraki; Japan; 19 May 2019 through 24 May 2019
Note

QC 20191002

Available from: 2019-10-02 Created: 2019-10-02 Last updated: 2019-10-02Bibliographically approved
Li, W., Qi, Z., Ye, Y., Yuan, Y. & Ma, W. (2019). On improvement of a conditional mornitoring technique for condition-based maintenance. In: International Conference on Nuclear Engineering, Proceedings, ICONE: . Paper presented at 27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019, 19 May 2019 through 24 May 2019. ASME Press
Open this publication in new window or tab >>On improvement of a conditional mornitoring technique for condition-based maintenance
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2019 (English)In: International Conference on Nuclear Engineering, Proceedings, ICONE, ASME Press, 2019Conference paper, Published 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.

Place, publisher, year, edition, pages
ASME Press, 2019
Keywords
Auto-associative kernel regression, Condition-based maintenance, Ensemble learning
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-258163 (URN)2-s2.0-85071394966 (Scopus ID)9784888982566 (ISBN)
Conference
27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019, 19 May 2019 through 24 May 2019
Note

QC 20191007

Available from: 2019-10-07 Created: 2019-10-07 Last updated: 2019-10-07Bibliographically approved
Bechta, S., Ma, W., Miassoedov, A., Journeau, C., Okamoto, K., Manara, D., . . . Schyns, M. (2019). On the EU-Japan roadmap for experimental research on corium behavior. Annals of Nuclear Energy, 124, 541-547
Open this publication in new window or tab >>On the EU-Japan roadmap for experimental research on corium behavior
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2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 124, p. 541-547Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Light water reactor, Severe accident, Corium, Accident phenomena, Research priority, Experimental facility
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-240340 (URN)10.1016/j.anucene.2018.10.019 (DOI)000451498100046 ()2-s2.0-85055353709 (Scopus ID)
Note

QC 20181218

Available from: 2018-12-18 Created: 2018-12-18 Last updated: 2018-12-18Bibliographically approved
Huang, Z. & Ma, W. (2019). On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes. Nuclear Engineering and Design, 351, 189-202
Open this publication in new window or tab >>On the quench of a debris bed in the lower head of a Nordic BWR by coolant injection through control rod guide tubes
2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 351, p. 189-202Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Severe accident, Debris bed coolability, CRGT cooling, Quench, Coupled simulation
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-255547 (URN)10.1016/j.nucengdes.2019.06.001 (DOI)000475396200017 ()2-s2.0-85067063084 (Scopus ID)
Note

QC 20190806

Available from: 2019-08-06 Created: 2019-08-06 Last updated: 2019-08-06Bibliographically approved
Manickam, L., Guo, Q., Komlev, A. A., Ma, W. & Bechta, S. (2019). Oxidation of molten zirconium droplets in water. Nuclear Engineering and Design, 354, Article ID UNSP 110225.
Open this publication in new window or tab >>Oxidation of molten zirconium droplets in water
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2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 354, article id UNSP 110225Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Reactor safety, Severe accident, Fuel coolant interaction, Zirconium oxidation, Hydrogen production
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-257790 (URN)10.1016/j.nucengdes.2019.110225 (DOI)000481647400019 ()2-s2.0-85069953862 (Scopus ID)
Note

QC 20190913

Available from: 2019-09-13 Created: 2019-09-13 Last updated: 2019-09-13Bibliographically approved
Huang, Z. & Ma, W. (2019). Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis. Annals of Nuclear Energy, 128, 330-340
Open this publication in new window or tab >>Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 128, p. 330-340Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Coupling analysis, Severe accident, Spent fuel pool, Thermosiphon
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-246416 (URN)10.1016/j.anucene.2019.01.024 (DOI)000465054700035 ()2-s2.0-85060329248 (Scopus ID)
Note

QC 20190329

Available from: 2019-03-29 Created: 2019-03-29 Last updated: 2019-05-14Bibliographically approved
Yu, P., Ma, W., Villanueva, W., Karbojian, A. & Bechta, S. (2019). Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment. Annals of Nuclear Energy, 133, 637-648
Open this publication in new window or tab >>Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment
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2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 133, p. 637-648Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Reactor pressure vessel, Creep failure, Thermal fluid-structure interaction, Computational fluid dynamics, Computational structural mechanics, Coupled analysis
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-260983 (URN)10.1016/j.anucene.2019.06.067 (DOI)000484649800061 ()2-s2.0-85068784394 (Scopus ID)
Note

QC 20191010

Available from: 2019-10-10 Created: 2019-10-10 Last updated: 2019-10-16Bibliographically approved
Mei, Y., Gong, S., Gu, H. & Ma, W. (2018). A study on steam-water two phase flow distribution in a rectangular channel with different channel orientations. Experimental Thermal and Fluid Science, 99, 219-232
Open this publication in new window or tab >>A study on steam-water two phase flow distribution in a rectangular channel with different channel orientations
2018 (English)In: Experimental Thermal and Fluid Science, ISSN 0894-1777, E-ISSN 1879-2286, Vol. 99, p. 219-232Article in journal (Refereed) Published
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.

Place, publisher, year, edition, pages
Elsevier Inc., 2018
Keywords
Distribution parameter, Drift velocity, Orientation, Rectangular channel, Void fraction, Atmospheric pressure, Crystal orientation, High speed cameras, Steam, Velocity, Average void fraction, Channel orientations, Distribution parameters, Drift velocities, Local void fraction, Steam water two phase flow, Superficial velocity, Two phase flow
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-236570 (URN)10.1016/j.expthermflusci.2018.07.022 (DOI)000446146600019 ()2-s2.0-85051040996 (Scopus ID)
Note

Export Date: 22 October 2018; Article; CODEN: ETFSE; Correspondence Address: Gong, S.; School of Nuclear Science and Engineering, Shanghai Jiao Tong UniversityChina; email: gsj@sjtu.edu.cn; Funding details: 51306112, NSFC, National Natural Science Foundation of China; Funding text: This work was supported by the National Natural Science Foundation of China (Grant No. 51306112 ). QC 20181127

Available from: 2018-11-27 Created: 2018-11-27 Last updated: 2018-11-27Bibliographically approved
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