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Grishchenko, Dmitry
Publications (10 of 26) Show all publications
Kudinov, P., Galushin, S., Grishchenko, D. & Yakush, S. E. (2019). Development of risk oriented accident analysis methodology (ROAAM+) for assessment of ex-vessel severe accident management effectiveness. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019: . Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019; Marriott Portland Downtown WaterfrontPortland; United States; 18 August 2019 through 23 August 2019 (pp. 2519-2535).
Open this publication in new window or tab >>Development of risk oriented accident analysis methodology (ROAAM+) for assessment of ex-vessel severe accident management effectiveness
2019 (English)In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 2019, p. 2519-2535Conference paper, Published paper (Refereed)
Abstract [en]

In this work we present results of development and application of Risk Oriented Accident Analysis framework (ROAAM+) to assessment of effectiveness of ex-vessel severe accident management strategy. In case of a core melt accident in Nordic type boiling water reactor (BWR) corium is released into a deep pool of water below reactor vessel to form a porous bed of debris. Energetic steam explosion or formation of non-coolable debris can threaten containment integrity. Both stochastic (aleatory) accident scenario and modeling (epistemic) uncertainties contribute to uncertainty. ROAAM+ framework is developed to simulate the whole accident progression The analysis starts from plant damage states determined in PSA Level-1 and continues with analysis of core degradation, vessel failure, melt release, steam explosion and debris bed formation and coolability. In order to achieve computational efficiency sufficient for extensive sensitivity, uncertainty, and risk analysis the surrogate modeling approach is used. In the paper we present results of the analysis aimed at quantification of uncertainty in the conditional containment failure probability. Specifically, we carry out sensitivity analysis using standalone and coupled models in order to identify the most influential scenario and modeling parameters for each sub-model. We assess the impact of the parameters on the prediction of the “load”, “capacity” and also failure probability. Then we quantify the effect of the most influential parameters on the failure probability. The results are presented using the failure domain approach and second order probability analysis, considering the uncertainty in distributions of the input parameters.

National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-268324 (URN)2-s2.0-85073714269 (Scopus ID)
Conference
18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019; Marriott Portland Downtown WaterfrontPortland; United States; 18 August 2019 through 23 August 2019
Note

QC 20200310

Available from: 2020-03-10 Created: 2020-03-10 Last updated: 2020-03-10Bibliographically approved
Moreau, V., Profir, M., Alemberti, A., Frignani, M., Merli, F., Belka, M., . . . Martelli, D. (2019). Pool CFD modelling: lessons from the SESAME project. Nuclear Engineering and Design, 355, Article ID UNSP 110343.
Open this publication in new window or tab >>Pool CFD modelling: lessons from the SESAME project
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2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 355, article id UNSP 110343Article in journal (Refereed) Published
Abstract [en]

The current paper describes the Computational Fluid-Dynamics (CFD) modelling of Heavy Liquid Metal (HLM) flows in a pool configuration and in particular how this is approached within the Horizon 2020 SESAME project. SESAME's work package structure, based on a systematic approach of redundancy and diversification, is explained along with its motivation. The main achievements obtained and the main lessons learned during the project are illustrated. The paper focuses on the strong coupling between the experimental activities and CFD simulations performed within the SESAME project. Two different HLM fluids are investigated: pure lead and Lead-Bismuth Eutectic. The objective is to make CFD a valid instrument used during the design of safe and innovative Gen-IV nuclear plants. Some effort has also been devoted to Proper Orthogonal Decomposition with Galerkin projection modelling (POD-Galerkin), a reduced order model suited for Uncertainty Quantification that operates by post-processing CFD results. Assessment of Uncertainty highly improves the reliability of CFD simulations. Dedicated experimental campaigns on heavily instrumented facilities have been conceived with the specific objective to build a series of datasets suited for the calibration and validation of the CFD modelling. In pool configuration, the attention is focused on the balance between conductive and convective heat transfer phenomena, on transient test-cases representative of incidental scenarios and on the possible occurrence of solidification phenomena. Four test sections have been selected to generate the datasets: (i) the CIRCE facility from ENEA, (ii) the TALL-3D pool test section from KTH, (iii) the TALL-3D Solidification Test Section (STS) from KTH and (iv) the SESAME Stand facility from CVR. While CIRCE and TALL-3D were existing facilities, the STS and SESAME Stand facility have been conceived, built and operated within the project, heavily relying on the use of CFD support. Care has been taken to ensure that almost all tasks were performed by at least two partners. Specific examples are given on how this strategy has allowed to uncover flaws and overcome pitfalls. Furthermore, an overview of the performed work and the achieved results is presented, as well as remaining or new uncovered issues. Finally, the paper is concluded with a description of one of the main goals of the SESAME project: the construction of the Gen-IV ALFRED CFD model and an investigation of its general circulation.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
CFD, Numerical simulation, Pool thermal-hydraulics, Lead solidification, Gen-IV reactors
National Category
Mechanical Engineering
Identifiers
urn:nbn:se:kth:diva-264342 (URN)10.1016/j.nucengdes.2019.110343 (DOI)000493898800029 ()2-s2.0-85072246527 (Scopus ID)
Note

QC 20191126

Available from: 2019-11-26 Created: 2019-11-26 Last updated: 2019-11-26Bibliographically approved
Wang, X., Gallego-Marcos, I., Grishchenko, D. & Kudinov, P. (2019). Post-test calibration of the Effective Momentum Source (EMS) model for steam injection through multi-hole spargers.. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019: . Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 18 August 2019 through 23 August 2019 (pp. 6176-6189). American Nuclear Society
Open this publication in new window or tab >>Post-test calibration of the Effective Momentum Source (EMS) model for steam injection through multi-hole spargers.
2019 (English)In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 6176-6189Conference paper, Published paper (Refereed)
Abstract [en]

Steam condensation in a large pool is often used in light water reactors to prevent containment overpressure. In boiling water reactors, steam from the primary system can be released into a pressure suppression pool (PSP) in normal operation and during accidents through multi-hole spargers to control the pressure in the reactor vessel. Steam injection into the pool can lead to the development of thermal stratification that affects (i) pressure suppression capacity of the pool, (ii) operation of the safety systems that use PSP as a source of water (e.g. emergency core cooling system and containment spray). Modeling of direct contact condensation of steam presents a challenge for contemporary codes. Therefore, Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed to enable prediction of thermal stratification and mixing induced by steam condensation in a large pool. EMS defines the time-averaged effect of steam injection into the pool in terms of a momentum source. For multi-hole spargers, the momentum source requires models for (i) momentum induced by multi-holes steam injection, (ii) direction (vertical angle) of the induced momentum, and profile of velocity in (iii) vertical and (iv) azimuthal directions. Previous works on EMS model validation and sensitivity study against PPOOLEX and HYMERES PANDA pool tests suggest the importance of all these factors for accurate prediction of the pool mixing behaviour. All these parameters, except the velocity profile in the azimuthal direction, were measured in PANDA facility and in Separate Effect Facility (SEF) at Lappeenranta Institute of Technology. The goal of this work is to develop a model for the azimuthal profile of radial velocity (APV) of water induced by steam injection through multi-hole spargers in a pressure suppression pool of a Nordic Boiling Water Reactor (BWR). In previous work, it was assumed that the APV is the same as the radial velocity profile in vertical cross section (which was measured in PANDA experiments using PIV) and can be described by axisymmetric jet expansion model. In this paper, APV is defined as a separate model with own closure for the jet diffusion rate. The effect of the steam mass flow rate is taken into account in the APV and respective jet expansion factor according to the experimental observations. Finally, we compare the pool temperature evolution in the experiment and simulations with the EMS model.

Place, publisher, year, edition, pages
American Nuclear Society, 2019
Keywords
Azimuthal profile of velocity, EHS/EMS model, Pressure suppression pool, Sparger, Thermal stratification and mixing, Condensation, Cooling systems, Electron injection, Expansion, Hydraulics, Lakes, Light water reactors, Mixing, Momentum, Nuclear reactor accidents, Plant shutdowns, Steam, Steam condensers, Thermal stratification, Velocity, Accurate prediction, Azimuthal direction, Boiling water reactor (BWR), Direct contact condensation, Emergency Core Cooling System, Sensitivity studies, Temperature evolution, Boiling water reactors
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-268501 (URN)2-s2.0-85073716090 (Scopus ID)
Conference
18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 18 August 2019 through 23 August 2019
Note

QC 20200402

Available from: 2020-04-02 Created: 2020-04-02 Last updated: 2020-04-02Bibliographically approved
Wang, X., Gallego-Marcos, I., Grishchenko, D. & Kudinov, P. (2019). Pre-test analysis for HYMERES-2 PANDA tests series for steam injection into pool through spargers. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019: . Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 18 August 2019 through 23 August 2019 (pp. 6190-6203). American Nuclear Society
Open this publication in new window or tab >>Pre-test analysis for HYMERES-2 PANDA tests series for steam injection into pool through spargers
2019 (English)In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 6190-6203Conference paper, Published paper (Refereed)
Abstract [en]

Steam condensation in a pool of water is often used in light water reactors. Steam injection provides sources of heat and momentum, which can lead to the development of thermal stratification or mixing of the pool. Modelling of the direct steam condensation is computationally challenging, especially considering the complex geometry of the pool and spargers and duration of transients. Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed and implemented to enable prediction of thermal stratification and mixing induced by steam condensation in a large pool. These models are utilized in this work and are subject for further development. The goal of this work is to provide pre-test analysis to support the design and selection of the test conditions for the Pressure Suppression Pool (PSP) test series (OECD/HYMERES-2). This test series is aiming to provide data for the development and validation of the EMS model predictive capabilities for the PSP phenomena. Specifically, it was proposed to extend the experimental database with regards to the sparger design, pool depth and depth of sparger submergence. In order to maximize the value of obtained data for the model development and validation, the pre-test analysis of HYMERES-2 is carried out. Compared to the HYMERES-1 test series, the sparger elevation above pool bottom will be increased in order to study the effect of the distance between the sparger head and thermocline interface on the rate of erosion of the stratified layer. The Particle Image Velocimetry (PIV) will be used for the measurement of the azimuthal profile of flow radial velocity around the sparger. A large number of thermocouples are provided in the vertical direction to capture the transient location of the thermocline. Selection of specific values for the sparger elevation, location of the PIV window, number of thermocouples, and steam injection conditions are based on the analysis provided in this work. In the pre-test analysis we use the previous EHS/EMS models which show a good agreement with the HYMERES-1 test series.

Place, publisher, year, edition, pages
American Nuclear Society, 2019
Keywords
HYMERES-2, PIV, Pre-test analysis, Pressure suppression pool, Sparger, Condensation, Hydraulics, Lakes, Light water reactors, Mixing, Steam, Steam condensers, Stream flow, Thermal stratification, Thermocouples, Velocity measurement, Complex geometries, Experimental database, Particle image velocimetries, Steam condensation, Vertical direction, Testing
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-268495 (URN)2-s2.0-85073755425 (Scopus ID)
Conference
18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 18 August 2019 through 23 August 2019
Note

QC 20200402

Available from: 2020-04-02 Created: 2020-04-02 Last updated: 2020-04-02Bibliographically approved
Grishchenko, D., Galushin, S. & Kudinov, P. (2019). Risk of containment failure due to ex-vessel steam explosion for Nordic BWRs. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019: . Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 18 August 2019 through 23 August 2019 (pp. 4032-4038). American Nuclear Society
Open this publication in new window or tab >>Risk of containment failure due to ex-vessel steam explosion for Nordic BWRs
2019 (English)In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, American Nuclear Society , 2019, p. 4032-4038Conference paper, Published paper (Refereed)
Abstract [en]

In case of a severe accident in a Light Water Reactor (LWR) degraded core relocates into the lower head of the reactor pressure vessel. Under thermal and mechanical loads from the core debris the vessel can fail releasing hot debris into the containment. In some designs of LWRs the severe accident mitigation strategy aims to prevent early containment failure by providing a pool of water below the reactor vessel. The melt is expected to form a coolable debris configuration preventing or delaying release of radioactive materials to the environment. One of the risk factors associated with melt-water interaction is containment failure due to ex-vessel steam explosion. Energetics of the steam explosion is contingent upon characteristics of melt release, pool and containment geometry. A general purpose full and surrogate models for estimation of the steam explosion loads in various conditions prototypic to boiling and pressurized water reactors have been proposed. In this paper, we rely on our recent results in model validation to develop a new surrogate model for the estimation of the steam explosion loads in LWRs using less conservative assumptions. We sample model output using Risk Oriented Accident Analysis Methodology code (ROAAM+) and provide estimates for the risk of containment failure for Nordic BWR given different accident scenarios. We plot Failure Domain maps and discuss implication of the steam explosion for different designs (fragility levels) and severe accident management strategies (pool depths). Importantly, we analyze the effect of the reduced model conservatism on the results of the risk analysis and discuss its implications to the decision making.

Place, publisher, year, edition, pages
American Nuclear Society, 2019
Keywords
Failure Domain maps, ROAAM+, Severe accident, Uncertainty analysis, Boiling water reactors, Debris, Explosions, Failure (mechanical), Hydraulics, Lakes, Nuclear reactor accidents, Pressure vessels, Pressurized water reactors, Radioactive materials, Risk analysis, Risk assessment, Risk perception, Steam, Ex-vessel steam explosions, Failure domains, Light water reactor (LWR), Reactor Pressure Vessel, ROAAM, Severe accident management, Thermal and mechanical loads, Light water reactors
National Category
Other Engineering and Technologies not elsewhere specified
Identifiers
urn:nbn:se:kth:diva-268498 (URN)2-s2.0-85073733781 (Scopus ID)
Conference
18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 18 August 2019 through 23 August 2019
Note

QC 20200402

Available from: 2020-04-02 Created: 2020-04-02 Last updated: 2020-04-02Bibliographically approved
Galushin, S., Grishchenko, D. & Kudinov, P. (2019). Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code. In: : . Paper presented at ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019.
Open this publication in new window or tab >>Surrogate Model Development for Prediction of Vessel Failure Mode and Melt Release Conditions in Nordic BWR based on MELCOR code
2019 (English)Conference paper, Published paper (Refereed)
National Category
Other Engineering and Technologies
Identifiers
urn:nbn:se:kth:diva-242348 (URN)
Conference
ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
Galushin, S., Grishchenko, D. & Kudinov, P. (2019). The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR. In: : . Paper presented at ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019.
Open this publication in new window or tab >>The Effect of the Uncertainty in Prediction of Vessel Failure Mode and Melt Release Conditions on Risk of Containment Failure due to Ex-Vessel Steam Explosion in Nordic BWR
2019 (English)Conference paper, Published paper (Refereed)
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-242346 (URN)
Conference
ICONE-27, 27th International Conference on Nuclear Engineering, Tsukuba, Ibaraki, Japan, May 19-24, 2019
Note

QC 20190130

Available from: 2019-01-30 Created: 2019-01-30 Last updated: 2019-01-30Bibliographically approved
Gallego-Marcos, I., Grishchenko, D. & Kudinov, P. (2019). Thermal stratification and mixing in a Nordic BWR pressure suppression pool. Annals of Nuclear Energy, 132, 442-450
Open this publication in new window or tab >>Thermal stratification and mixing in a Nordic BWR pressure suppression pool
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 132, p. 442-450Article in journal (Refereed) Published
Abstract [en]

The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Sparger, Relief vales, Steam injection, Condensation, CFD, Effective momentum
National Category
Atom and Molecular Physics and Optics
Identifiers
urn:nbn:se:kth:diva-259408 (URN)10.1016/j.anucene.2019.04.054 (DOI)000482247600042 ()2-s2.0-85065229097 (Scopus ID)
Note

QC 20190925

Available from: 2019-09-25 Created: 2019-09-25 Last updated: 2019-09-25Bibliographically approved
Grishchenko, D. & Kudinov, P. (2019). Validation of a full model for the analysis of ex-vessel steam explosion in LWRs. In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019: . Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019; Marriott Portland Downtown WaterfrontPortland; United States; 18 August 2019 through 23 August 2019 (pp. 4568-4574).
Open this publication in new window or tab >>Validation of a full model for the analysis of ex-vessel steam explosion in LWRs
2019 (English)In: 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 2019, p. 4568-4574Conference paper, Published paper (Refereed)
Abstract [en]

In a Light Water Reactor (LWR) severe accident, the reactor core can be melted and released from the reactor vessel at ~3000K. In most of reactor designs ex-vessel severe accident mitigation strategy employs a pool of water underneath the reactor vessel. If water pool is deep enough, the melt is expected to be fragmented and quenched and form a coolable debris bed preventing further accident progression. However, there is a possibility that upon contact with volatile coolant thermal energy stored in the hot melt will be converted into mechanical energy of rapidly expanding steam in the process of so called “steam explosion”. Energetics of the steam explosion is contingent upon conditions of melt release, pool characteristics and containment geometry. Containment failure due to the ex-vessel steam explosion can be a factor of risk for the “wet cavity” strategy if fragility limits are close to the expected loads. In order to assess the risk, we develop so called full model (based on TEXAS-V code) for the estimation of the steam explosion loads. To ensure model applicability to a wide range of LWR designs, a number of modifications have been introduced in comparison to previous works. A large database of Full Model solutions is used then for the development of a Surrogate Model based on the Artificial Neural Networks (ANN) to enable extensive sensitivity analysis and uncertainty quantification. The uncertainty in the SM approximation of the FM is considered explicitly in the assessment of failure probability. In this work, we demonstrate an approach to the validation of the Full Model against previous steam explosion experiments using a statistical approach in which a joint distribution of the experimental data is compared to a database of explosion distributions obtained using the full model.

National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-268322 (URN)2-s2.0-85073726445 (Scopus ID)
Conference
18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019; Marriott Portland Downtown WaterfrontPortland; United States; 18 August 2019 through 23 August 2019
Note

QC 20200310

Available from: 2020-03-10 Created: 2020-03-10 Last updated: 2020-03-10Bibliographically approved
Jeltsov, M., Grishchenko, D. & Kudinov, P. (2019). Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment. Nuclear Engineering and Design, 341, 306-325
Open this publication in new window or tab >>Validation of Star-CCM plus for liquid metal thermal-hydraulics using TALL-3D experiment
2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 341, p. 306-325Article in journal (Refereed) Published
Abstract [en]

Computational Fluid Dynamics (CFD) provides means for high-fidelity 3D thermal-hydraulics analysis of Generation IV pool-type nuclear reactors. However, to be used in the decision making process a proof of code adequacy for intended application is required. This paper describes the Verification, Validation and Uncertainty Quantification (VVUQ) of a commercial CFD code Star-CCM + for forced, natural and mixed convection regimes in lead-bismuth eutectic (LBE) coolant pool flows. Code qualification is carried out according to an iterative VVUQ process aiming to reduce user effects. Validation data is produced in TALL-3D experimental facility - a 7 m high LBE loop featuring a 3D pool-type test section. Accurate prediction of mutual interaction between thermal stratification and mixing in the pool and the loop dynamics requires 3D analysis, especially during natural circulation. Solution verification is used to reduce the numerical uncertainty during code validation activities. Sensitivity Analysis (SA) is used to identify the effect of the most influential uncertain input parameters (UIPs) on numerical results. Two new visualization methods are proposed to enhance interpretation of the SA results. Dedicated experiments are performed according to the SA results to reduce the uncertainties in the most important UIPs. Automated calibration method for large CFD models is tested and demonstrated in combination with manual calibration using detailed temperature profile measurements in the pool. Calibration reveals the deficiencies in the modeling of heat losses owing to the presence of thermal bridges and other local effects in thermal insulation that are not explicitly modeled. It is demonstrated that Star-CCM + is able to predict thermal stratification and mixing phenomena in the pool type geometries. The results are supported by an Uncertainty Analysis (UA).

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
LFR, CFD, VVUQ, Pool thermal-hydraulics, Star-CCM
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-240701 (URN)10.1016/j.nucengdes.2018.11.015 (DOI)000453016700028 ()2-s2.0-85056899609 (Scopus ID)
Funder
EU, FP7, Seventh Framework Programme, FP7-249337
Note

QC 20190111

Available from: 2019-01-11 Created: 2019-01-11 Last updated: 2020-03-09Bibliographically approved
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