UB Paderborn / Katalog / Suche / in CDI

Sortiert nach

Einfacher Volltextzugriff von (fast) überall

Installieren Sie die LibKey Nomad-Browsererweiterung und erhalten Sie von vielen Webseiten einfachen Zugriff auf lizenzierte Volltexte oder Open Access Publikationen. Mehr erfahren...

Validation of interfacial area concentration model for simulating bubbly and cap/slug flow behaviors in large diameter pipes using LSTF, ATLAS, and horizontal pipe experiment data

Progress in nuclear energy (New series), 2020-10, Vol.128, p.103499, Article 103499

2020

Volltextzugriff (PDF)

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel bundle for SFR

Nuclear Engineering and Technology, 2015, 47(5), , pp.523-533

2015

Volltextzugriff (PDF)

Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

Nuclear engineering and technology, 2016-08, Vol.48 (4), p.952-964

2016

Volltextzugriff (PDF)

Development of singular value decomposition based solution scheme for a numerical simulation of multiphase flow

Nuclear engineering and design, 2021-12, Vol.385, p.111499, Article 111499

2021

Volltextzugriff (PDF)

A numerical derivation of internal energy equation under a staggered mesh system for 1-D thermal hydraulic system analysis codes

Nuclear engineering and design, 2021-01, Vol.371, p.110935, Article 110935

2021

Volltextzugriff (PDF)

Fuel fragmentation and relocation (FFR) model in SPACE code PART 1: Development and verification

Annals of nuclear energy, 2024-05, Vol.199, p.110323, Article 110323

2024

Volltextzugriff (PDF)

Analyses of LOFT LP-FW-1 using SPACE code

Annals of nuclear energy, 2020-01, Vol.135, p.107001, Article 107001

2020

Volltextzugriff (PDF)

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

Nuclear engineering and technology, 2012-06, Vol.44 (5), p.535-542

2012

Volltextzugriff (PDF)

Sensitivity and uncertainty analysis for ULOF of PGSFR using PAPIRUS

Annals of nuclear energy, 2017-12, Vol.110, p.1232-1241

2017

Volltextzugriff (PDF)

An experimental study on natural convection heat transfer of liquid gallium in a rectangular loop

International journal of heat and mass transfer, 2013-11, Vol.66, p.192-199

2013

Volltextzugriff (PDF)

Assessment study about the risk of Wilks' formula for uncertainty quantification of design extension condition scenarios in prototype Gen-IV sodium fast reactor

Journal of nuclear science and technology, 2018-07, Vol.55 (7), p.746-755

2018

Volltextzugriff (PDF)

Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code

Annals of nuclear energy, 2018-05, Vol.115, p.138-153

2018

Volltextzugriff (PDF)

Validation and parametric study of FFRD model in DRACCAR code

Annals of nuclear energy, 2024-12, Vol.209, p.110832, Article 110832

2024

Volltextzugriff (PDF)

A dominant geometrical parameter affecting the turbulent mixing rate in rod bundles

International journal of heat and mass transfer, 2007-03, Vol.50 (5), p.908-918

2007

Volltextzugriff (PDF)

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

Nuclear engineering and technology, 2009-08, Vol.41 (6), p.797-806

2009

Volltextzugriff (PDF)

A ‘must-go path’ scenario for sustainable development and the role of nuclear energy in the 21st century

Energy policy, 2010-04, Vol.38 (4), p.1962-1968

2010

Volltextzugriff (PDF)

Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

Nuclear technology, 2010-02, Vol.169 (2), p.134-142

2010

Volltextzugriff (PDF)

Fuel fragmentation and relocation (FFR) model in SPACE code PART 2: Validation and sensitivity analyses

Annals of nuclear energy, 2024-11, Vol.207, p.110682, Article 110682

2024

Volltextzugriff (PDF)

Consideration of H2 generation in fluid governing equations of SPACE code

Annals of nuclear energy, 2019-09, Vol.131, p.291-296

2019

Volltextzugriff (PDF)

ADVANCED SFR DESIGN CONCEPTS AND R&D ACTIVITIES

Nuclear engineering and technology, 2009-05, Vol.41 (4), p.427-446

2009

Volltextzugriff (PDF)

Evaluation of the conduction shape factor with a CFD code for a liquid–metal heat transfer in heated triangular rod bundles

Nuclear engineering and design, 2007-03, Vol.237 (6), p.648-654

2007

Volltextzugriff (PDF)

Development of a phenomena identification and ranking table (PIRT) for intermediate break loss-of-coolant accident in PWRs

Nuclear engineering and technology, 2024-08

2024

Volltextzugriff (PDF)

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

Nuclear technology, 2005-01, Vol.149 (1), p.71-87

2005

Volltextzugriff (PDF)

An Enhanced Code for the Safety Analysis of Pool-Type Sodium-Cooled Fast Reactors

Nuclear technology, 2008-11, Vol.164 (2), p.221-231

2008

Volltextzugriff (PDF)

DEVELOPMENT OF A TWO-DIMENSIONAL THERMOHYDRAULIC HOT POOL MODEL AND ITS EFFECTS ON REACTIVITY FEEDBACK DURING A UTOP IN LIQUID METAL REACTORS

Nuclear engineering and technology, 2009-10, Vol.41 (8), p.1053-1064

2009

Volltextzugriff (PDF)

1 – 25 von 38

Suche als RSS-Feed

Suchergebnisse filtern

Filter anzeigen

Nur zeigen

Geprüft (Peer-Review)
(36)
Mit Volltextzugriff
(31)

Publikationsform

Artikel
(38)

Sprache

Erscheinungsjahr

n.n

Thema

Science & Technology
(26)
Technology
(26)
Nuclear Science & Technology
(22)
Energy
(10)
Applied Sciences
(9)
Energy. Thermal Use Of Fuels
(9)
Exact Sciences And Technology
(9)
Fission Nuclear Power Plants
(9)
Installations For Energy Generation And Conversion: Thermal And Electrical Energy
(9)
Nuclear Fuels
(5)
원자력공학
(5)
Mars-Lmr
(4)
Engineering
(3)
Ffrd
(3)
Fuel Fragmentation
(3)
Fuels
(3)
Mechanics
(3)
Papirus
(3)
Physical Sciences
(3)
Pulverization
(3)

15 weitere anzeigen...

Weniger anzeigen...

Neu hinzugefügt

... seit 90 Tagen
(2)

Menü

Weitere Dienste

Einstellungen

Validation of interfacial area concentration model for simulating bubbly and cap/slug flow behaviors in large diameter pipes using LSTF, ATLAS, and horizontal pipe experiment data

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel bundle for SFR

Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

Development of singular value decomposition based solution scheme for a numerical simulation of multiphase flow

A numerical derivation of internal energy equation under a staggered mesh system for 1-D thermal hydraulic system analysis codes

Fuel fragmentation and relocation (FFR) model in SPACE code PART 1: Development and verification

Analyses of LOFT LP-FW-1 using SPACE code

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

Sensitivity and uncertainty analysis for ULOF of PGSFR using PAPIRUS

An experimental study on natural convection heat transfer of liquid gallium in a rectangular loop

Assessment study about the risk of Wilks' formula for uncertainty quantification of design extension condition scenarios in prototype Gen-IV sodium fast reactor

Best estimate calculation and uncertainty quantification of sodium-cooled fast reactor using MARS-LMR code

Validation and parametric study of FFRD model in DRACCAR code

A dominant geometrical parameter affecting the turbulent mixing rate in rod bundles

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

A ‘must-go path’ scenario for sustainable development and the role of nuclear energy in the 21st century

Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

Fuel fragmentation and relocation (FFR) model in SPACE code PART 2: Validation and sensitivity analyses

Consideration of H2 generation in fluid governing equations of SPACE code

ADVANCED SFR DESIGN CONCEPTS AND R&D ACTIVITIES

Evaluation of the conduction shape factor with a CFD code for a liquid–metal heat transfer in heated triangular rod bundles

Development of a phenomena identification and ranking table (PIRT) for intermediate break loss-of-coolant accident in PWRs

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

An Enhanced Code for the Safety Analysis of Pool-Type Sodium-Cooled Fast Reactors

DEVELOPMENT OF A TWO-DIMENSIONAL THERMOHYDRAULIC HOT POOL MODEL AND ITS EFFECTS ON REACTIVITY FEEDBACK DURING A UTOP IN LIQUID METAL REACTORS

Suchergebnisse filtern