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Publications (10 of 16) Show all publications
Galushin, S. & Kudinov, P. (2019). Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code. Science and Technology of Nuclear Installations, Article ID 5310808.
Open this publication in new window or tab >>Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code
2019 (English)In: Science and Technology of Nuclear Installations, ISSN 1687-6075, E-ISSN 1687-6083, article id 5310808Article in journal (Refereed) Published
Abstract [en]

Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

Place, publisher, year, edition, pages
Hindawi Publishing Corporation, 2019
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-251723 (URN)10.1155/2019/5310808 (DOI)000466590400001 ()2-s2.0-85065230309 (Scopus ID)
Note

QC 20190520

Available from: 2019-05-20 Created: 2019-05-20 Last updated: 2019-05-29Bibliographically approved
Grishchenko, D., Galushin, S. & Kudinov, P. (2019). Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a Nordic type BWR. Nuclear Engineering and Design, 343, 63-75
Open this publication in new window or tab >>Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a Nordic type BWR
2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 63-75Article in journal (Refereed) Published
Abstract [en]

Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel to fragment and quench core melt and provide long term cooling of the debris. One of the risks associated with this strategy is early containment failure due to ex-vessel steam explosion. Assessment of the risk of steam explosion is subject to significant (i) epistemic uncertainties in modelling and (ii) aleatory uncertainties in scenarios of melt release. For quantification of the uncertainties and the risk a full model (FM) based on TEXAS-V code and a computationally efficient surrogate model (SM) have been previously developed. FM is used to provide a database of solutions that is used for development of a SM, while SM is used in extensive sensitivity and uncertainty analysis. In this work, we compare the risk of containment failure with non-reinforced and reinforced hatch door for metallic and oxidic melt release scenarios. We quantify the error of SM in the approximation of the FM and assess the effect of the approximation uncertainty on risk assessment. We analyze the results and suggest a simplified approach for decision making considering predicted failure probabilities, expected costs, and scenario frequencies.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Severe accident, Artificial neural networks, Aleatory and epistemic uncertainties, Surrogate model uncertainty
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-244081 (URN)10.1016/j.nucengdes.2018.12.013 (DOI)000456923500007 ()2-s2.0-85059446676 (Scopus ID)
Note

QC 20190219

Available from: 2019-02-19 Created: 2019-02-19 Last updated: 2019-02-19Bibliographically approved
Galushin, S. & Kudinov, P. (2018). Analysis of the effect of severe accident scenario on the vessel lower head failure in Nordic BWR using MELCOR code. In: PSAM 2018 - Probabilistic Safety Assessment and Management: . Paper presented at 14th Probabilistic Safety Assessment and Management, PSAM 2018, 16 September 2018 through 21 September 2018. International Association for Probablistic Safety Assessment and Management (IAPSAM)
Open this publication in new window or tab >>Analysis of the effect of severe accident scenario on the vessel lower head failure in Nordic BWR using MELCOR code
2018 (English)In: PSAM 2018 - Probabilistic Safety Assessment and Management, International Association for Probablistic Safety Assessment and Management (IAPSAM) , 2018Conference paper, Published paper (Refereed)
Abstract [en]

Severe accident management (SAM) in Nordic boiling water reactors (BWR) relies on ex-vessel core debris coolability. In case of core melt and vessel failure, melt is poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by natural circulation of water. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) properties and thus coolability of the debris bed, and (ii) potential for energetic steam explosion. Both non-coolable debris bed and steam explosion are credible threats to containment integrity. Melt release conditions are the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs using ROAAM+ Framework. The melt release conditions, including in-vessel\ex-vessel pressure, lower drywell pool depth and temperature, are affected by aleatory (severe accident scenario) and epistemic (modeling) uncertainties. In this work we use MELCOR code to perform the analysis of the effects of Severe accident scenarios and modelling options in MELCOR on the properties of debris relocated to the lower head, the time and the mode of vessel lower head failure. We identify the most influential uncertain factors and discuss the needs for improvements in the modeling approaches. 

Place, publisher, year, edition, pages
International Association for Probablistic Safety Assessment and Management (IAPSAM), 2018
Keywords
MELCOR, Nordic BWR, ROAAM, Severe accident, Accidents, Boiling water reactors, Debris, Explosions, Uncertainty analysis, Boiling water reactor (BWR), Containment integrity, Natural circulation, Severe accident management, Failure (mechanical)
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-252279 (URN)2-s2.0-85063129205 (Scopus ID)
Conference
14th Probabilistic Safety Assessment and Management, PSAM 2018, 16 September 2018 through 21 September 2018
Note

QC20190607

Available from: 2019-06-07 Created: 2019-06-07 Last updated: 2019-09-24Bibliographically approved
Yu, P., Villanueva, W., Galushin, S., Ma, W. & Bechta, S. (2018). Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR. In: : . Paper presented at 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12).
Open this publication in new window or tab >>Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR
Show others...
2018 (English)Conference paper, Published 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.

Keywords
Boiling Water Reactor, Severe Accident, Vessel Failure, Thermo-Mechanical Analysis
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-259590 (URN)
Conference
12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12)
Note

QCR 20191015

Available from: 2019-09-18 Created: 2019-09-18 Last updated: 2019-10-15Bibliographically approved
Galushin, S., Ranlöf, L., Bäckström, O., Adolfsson, Y., Grishchenko, D., Kudinov, P. & Marklund, A. R. (2018). Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR. In: PSAM 2018 - Probabilistic Safety Assessment and Management: . Paper presented at 14th Probabilistic Safety Assessment and Management, PSAM 2018, 16 September 2018 through 21 September 2018. International Association for Probablistic Safety Assessment and Management (IAPSAM)
Open this publication in new window or tab >>Joint application of risk oriented accident analysis methodology and PSA level 2 to severe accident issues in Nordic BWR
Show others...
2018 (English)In: PSAM 2018 - Probabilistic Safety Assessment and Management, International Association for Probablistic Safety Assessment and Management (IAPSAM) , 2018Conference paper, Published paper (Refereed)
Abstract [en]

A comprehensive and robust assessment of severe accident management effectiveness in preventing unacceptable releases is a challenge for a today’s real life PSA. This is mainly due to the fact that major uncertainty is determined by the physical phenomena and timing of the events. The static PSA is built on choosing scenario parameters to describe the accident progression sequence and typically uses a limited number of simulations in the underlying deterministic analysis. Risk Oriented Accident Analysis Methodology framework (ROAAM+) is being developed in order to enable consistent and comprehensive treatment of both epistemic and aleatory uncertainties. The framework is based on a set of deterministic models that describe different stages of the accident progression. The results are presented in terms of distributions of conditional containment failure probabilities for given combinations of the scenario parameters. This information is used for enhanced modeling in the PSA-L2. Specifically, it includes improved definitions of the sequences determined by the physical phenomena rather than stochastic failures of the equipment, improved knowledge of timing in sequences and estimation of probabilities determined by the uncertainties in the phenomena. In this work we present an example of application of the dynamic approach in a large scale PSA model and show that the integration of the ROAAM+ results and the PSA model can potentially lead to a considerable change in PSA Level 2 analysis results. 

Place, publisher, year, edition, pages
International Association for Probablistic Safety Assessment and Management (IAPSAM), 2018
Keywords
Nordic BWR, PSA L2, ROAAM, Severe accident, Accident prevention, Accidents, Probability distributions, Risk analysis, Risk assessment, Stochastic systems, Uncertainty analysis, Aleatory uncertainty, Deterministic analysis, Deterministic models, Severe accident management, Boiling water reactors
National Category
Probability Theory and Statistics
Identifiers
urn:nbn:se:kth:diva-252277 (URN)2-s2.0-85063145364 (Scopus ID)
Conference
14th Probabilistic Safety Assessment and Management, PSAM 2018, 16 September 2018 through 21 September 2018
Note

QC20190610

Available from: 2019-06-10 Created: 2019-06-10 Last updated: 2019-06-10Bibliographically approved
Phung, V.-A., Grishchenko, D., Galushin, S. & Kudinov, P. (2018). Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks. Annals of Nuclear Energy, 461-476
Open this publication in new window or tab >>Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks
2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, p. 461-476Article in journal (Refereed) Published
Abstract [en]

Severe accident management strategy in Nordic boiling water reactors (BWRs) employs ex-vessel corium debris coolability. In-vessel core degradation and relocation provide initial conditions for further accident progression. Characteristics of debris bed in the reactor vessel lower plenum affect corium interactions with the vessel structures, vessel failure and melt ejection mode. Melt release from the vessel determines conditions for ex-vessel accident progression and potential threats to containment integrity, i.e. steam explosion, debris bed formation and coolability. Outcomes of core relocation depend on the interplay between (i) accident scenarios, e.g. timing and characteristics of failure and recovery of safety systems and (ii) accident phenomena. Uncertainty analysis is necessary for comprehensive risk assessment. However, computational efficiency of system analysis codes such as MELCOR is one of the big obstacles. Another problem is that code predictions can be quite sensitive to small variations of the input parameters and numerical discretization, due to the highly non-linear feedbacks and discrete thresholds employed in the code models.

 

The goal of this work is to develop a computationally efficient surrogate model (SM) for prediction of main characteristics of corium debris in the vessel lower plenum of a Nordic BWR. Station black out scenario with a delayed power recovery is considered. The SM has been developed using artificial neural networks (ANNs). The networks were trained with a database of MELCOR solutions. The effect of the noisy data in the full model (FM) database was addressed by introducing scenario classification (grouping) according to the ranges of the output parameters. SMs using different number of scenario groups with/without weighting between predictions of different ANNs were compared. The obtained SM can be used for failure domain and failure probability analysis in the risk assessment framework for Nordic BWRs.

Keywords
Core relocation; boiling water reactor; MELCOR; surrogate model; artificial neural network.
National Category
Other Engineering and Technologies not elsewhere specified
Identifiers
urn:nbn:se:kth:diva-202955 (URN)10.1016/j.anucene.2018.06.007 (DOI)000441485700040 ()2-s2.0-85048449676 (Scopus ID)
Note

QC 20170309

Available from: 2017-03-08 Created: 2017-03-08 Last updated: 2019-04-26Bibliographically approved
Galushin, S. & Kudinov, P. (2018). Sensitivity analysis of debris properties in lower plenum of a Nordic BWR. Nuclear Engineering and Design, 332, 374-382
Open this publication in new window or tab >>Sensitivity analysis of debris properties in lower plenum of a Nordic BWR
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 374-382Article in journal (Refereed) Published
Abstract [en]

Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. The melt released from the vessel is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the melt release mode from the vessel, which determine conditions for (i) the debris bed coolability, (ii) steam explosion that present credible threats to containment integrity. Melt release conditions are recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena, subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine conditions for the corium interactions with the vessel structures (such as instrumentation guide tubes IGTs, control rod guide tubes CRGTs), vessel failure and melt release. This work is focused on the evaluation of uncertainty in core degradation progression and its effect on the resultant properties of relocated debris in lower plenum of Nordic BWR. We use MELCOR code for prediction of the accident progression. The main goal of this paper is to characterize the range of possible debris properties in lower plenum and its sensitivity towards different modelling parameters, which is of paramount importance for the analysis of in-vessel debris coolability and vessel failure mode in the risk assessment framework.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Severe accident, Nordic BWR, ROAAM, MELCOR
National Category
Other Chemistry Topics
Identifiers
urn:nbn:se:kth:diva-227209 (URN)10.1016/j.nucengdes.2018.03.029 (DOI)000430395700033 ()2-s2.0-85056238508 (Scopus ID)
Note

QC 20180529

Available from: 2018-05-29 Created: 2018-05-29 Last updated: 2019-03-18Bibliographically approved
Grishchenko, D., Galushin, S., Basso, S. & Kudinov, P. (2017). Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR. In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017: . Paper presented at 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, 3 September 2017 through 8 September 2017. Association for Computing Machinery, Inc
Open this publication in new window or tab >>Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a nordic type BWR
2017 (English)In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery, Inc , 2017Conference paper, Published paper (Refereed)
Abstract [en]

Sever accident mitigation strategy adopted in Nordic Boiling Water Reactors (BWRs) employs a deep water pool below the reactor vessel in order to fragment and quench core melt and provide long term cooling of the debris. One of the risk factors associated with this accident management strategy is early failure of the containment due to steam explosion. Assessment of the risk is subject to significant epistemic and aleatory uncertainties in (i) modelling of steam explosion and (ii) scenarios of melt release from the vessel and water pool conditions. High computational efficiency of the models is required for such assessment. A surrogate model (SM) approach has been previously developed using artificial neural network and the database of Texas-V code solutions for steam explosion loads in the Nordic type BWRs. In this paper we extend our surrogate model to allow analysis of steam explosion in relatively shallow water pools (>2 m), address effects of melt emissivity and resolve more accurately variation of pressure in the drywell. We provide detailed comparison of metallic vs oxidic melt release scenarios, incorporate uncertainty of the SM into modelling and analyze the sensitivity of our results to SM uncertainty. We estimate risks of containment failure with non-reinforced and reinforced hatch door and demonstrate the effect of the surrogate model uncertainty on the results. We analyze the results and develop a simplified approach for decision making considering predicted failure probabilities, expected costs and scenario frequencies. 

Place, publisher, year, edition, pages
Association for Computing Machinery, Inc, 2017
Keywords
Aleatory and epistemic uncertainties, Artificial Neural Networks, Severe accident, Surrogate model uncertainty, Boiling water reactors, Computational efficiency, Explosions, Failure (mechanical), Fuel additives, Hydraulics, Lakes, Neural networks, Nuclear reactor accidents, Reinforcement, Risk assessment, Risk perception, Steam, Aleatory uncertainty, Failure Probability, Mitigation strategy, Shallow water pools, Surrogate model, Uncertainty quantifications, Uncertainty analysis
National Category
Other Civil Engineering
Identifiers
urn:nbn:se:kth:diva-236832 (URN)2-s2.0-85052599683 (Scopus ID)
Conference
17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, 3 September 2017 through 8 September 2017
Funder
Swedish Radiation Safety Authority
Note

QC 20181221

Available from: 2018-12-21 Created: 2018-12-21 Last updated: 2018-12-21Bibliographically approved
Galushin, S. & Kudinov, P. (2016). Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR. NUCLEAR ENGINEERING AND DESIGN, 310, 125-141
Open this publication in new window or tab >>Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR
2016 (English)In: NUCLEAR ENGINEERING AND DESIGN, ISSN 0029-5493, Vol. 310, p. 125-141Article in journal (Refereed) Published
Abstract [en]

Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario parameters. Pattern analysis is employed in order to characterize typical behavior of core relocation transients. Clustering analysis is employed for grouping of different accident scenarios, which result in similar core relocation behavior and properties of the debris.

Place, publisher, year, edition, pages
Elsevier, 2016
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-200215 (URN)10.1016/j.nucengdes.2016.09.029 (DOI)000390736400011 ()2-s2.0-84993993448 (Scopus ID)
Note

QC 20170202

Available from: 2017-02-02 Created: 2017-01-23 Last updated: 2019-01-30Bibliographically approved
Galushin, S. & Kudinov, P. (2016). Comparison of melcor code versions predictions of the properties of relocated debris in lower plenum of nordic BWR. In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017: . Paper presented at 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Qujiang Int'l Conference CenterXi'an, Shaanxi, China, 3 September 2017 through 8 September 2017. Association for Computing Machinery (ACM)
Open this publication in new window or tab >>Comparison of melcor code versions predictions of the properties of relocated debris in lower plenum of nordic BWR
2016 (English)In: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Association for Computing Machinery (ACM), 2016Conference paper, Published paper (Refereed)
Abstract [en]

Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the vessel failure and melt release mode from the vessel, which determine conditions for (i) the formation of debris bed and its coolability, and (ii) steam explosion. Non-coolable debris and strong explosions present credible threats to containment integrity. A risk oriented accident analysis framework (ROAAM+) is under development for assessment of the effectiveness of the severe accident management strategy. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine conditions for the corium interactions with the vessel structures, vessel failure and melt release. In this work we perform comparison of predictions of different MELCOR code versions used for the analysis of the effect severe accident scenario and uncertainties on the processes of core degradation and relocation, and resulting properties of relocated debris in Nordic BWR lower plenum. Properties of relocated debris are obtained as functions of the accident scenario parameters, such as timing of activation of different safety systems. We perform the analysis of the codes predictions and discuss possible reasons for the discrepancies in observations. The main goal of this work is to provide insights regarding the effect of code uncertainty, sensitivity coefficients and user effect on the code predictions, which is of importance for the analysis of in-vessel debris coolability and vessel failure mode in the ROAAM+ framework.

Place, publisher, year, edition, pages
Association for Computing Machinery (ACM), 2016
Keywords
MELCOR, Nordic BWR, Severe Accident
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-234524 (URN)2-s2.0-85052503219 (Scopus ID)
Conference
17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017, Qujiang Int'l Conference CenterXi'an, Shaanxi, China, 3 September 2017 through 8 September 2017
Note

QC 20180907

Available from: 2018-09-07 Created: 2018-09-07 Last updated: 2018-09-07Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0001-8216-9376

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