Helmholtz-Zentrum Dresden-Rossendorf

Reactor Safety

Dr. Sören Kliem

Head

Phone:+49 351 260 2318
Email: s.kliemAthzdr.de
Link:ORCID: 0000-0001-6654-6434
AddressBautzner Landstraße 400
01328 Dresden
Building/Office:250/108

Scientific career

  • Seit 2010:
    Department head "Reactor Safety" at the Helmholtz-Zentrum Dresden- Rossendorf, Institute of Resource Ecology
  • 2010:
    PhD at Technical University Dresden, Faculty of Machinery Building
  • 1994-2010:
    Scientific co-worker at the Forschungszentrum Rossendorf, Institute of Safety Research
  • 1992:
    Graduated from Moscow Power Engineering Institute (MEI - Technical University)

Research work

  • Reactor dynamics
  • Nuclear reactor accident analysis
  • Coupling of neutron kinetic and thermal hydraulic codes
  • Experimental and numerical investigation of coolant mixing phenomena in pressurized water reactors

Projects

  • EU FP 7 Nuclear Reactor Safety Simulation Platform (NURESAFE) project (2013-2015)
  • OECD NEA Primary Coolant Loop Test Facility (PKL-3) Project (2012-2016)
  • Horizon 2020 McSAFE: High-Performance Monte Carlo Methods for SAFEty Demonstration, 2017-2020
  • Horizon 2020 R2CA: Reduction of radiological consequences of design basis and design extension accidents

Publications

2021

Advanced modelling of complex boron dilution transients in PWRs—Validation of CFD simulation with ANSYS CFX against the ROCOM E2.3 experiment

Grahn, A.; Diaz Pescador, E.; Kliem, S.; Schäfer, F.; Höhne, T.


2020

Detailed Simulation of the Nominal Flow and Temperature Conditions in a Pre-Konvoi PWR Using Coupled CFD and Neutron Kinetics

Höhne, T.; Kliem, S.

  • Contribution to proceedings
    CFD4NRS-8 : Computational Fluid Dynamics for Nuclear Reactor Safety - OECD/NEA Workshop, 25.-27.11.2020, Palaiseau, Frankreich
  • Lecture (Conference)
    CFD4NRS-8 : Computational Fluid Dynamics for Nuclear Reactor Safety - OECD/NEA Workshop, 25.-27.11.2020, Palaiseau, Frankreich
  • Open Access Logo Fluids 5(2020)3, 161
    DOI: 10.3390/fluids5030161

Advanced modelling of complex boron dilution transients in PWRs – Validation of ATHLET 3D-Module against the experiment ROCOM E2.3

Diaz Pescador, E.; Grahn, A.; Kliem, S.; Schäfer, F.; Höhne, T.


A SERPENT2-SUBCHANFLOW-TRANSURANUS coupling for pin-by-pin depletion calculations in Light Water Reactors

Garcia, M.; Tuominen, R.; Gommlich, A.; Ferraro, D.; Valtavirta, V.; Imke, U.; van Uffelen, P.; Mercatali, L.; Sanchez, V.; Leppänen, J.; Kliem, S.


The efficiency of sequential accident management measures for a German PWR under prolonged SBO conditions

Kozmenkov, Y.; Jobst, M.; Kliem, S.; Kosowski, K.; Schäfer, F.; Wilhelm, P.


2019

Thermal-hydraulic insights during a main steam line break in a generic PWR KONVOI reactor with ATHLET 3.1A

Diaz Pescador, E.; Schäfer, F.; Kliem, S.


A realistic approach for the assessment of the consequences of heterogeneous boron dilution events in pressurized water reactors

Kliem, S.; Grahn, A.; Bilodid, Y.; Höhne, T.


2018

Severe Accident Management Measures for a Generic German PWR. Part II: Small-break loss-of-coolant accident

Jobst, M.; Wilhelm, P.; Kozmenkov, Y.; Kliem, S.


Severe accident management measures for a generic German PWR. Part I: Station blackout

Wilhelm, P.; Jobst, M.; Kozmenkov, Y.; Schäfer, F.; Kliem, S.


The HEXNEM3 nodal flux expansion method for the hexagonal geometry in the code DYN3D

Bilodid, Y.; Grundmann, U.; Kliem, S.


Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part II: DYN3D calculations.

Rachamin, R.; Kliem, S.


IAEA CRP benchmark of ROCOM PTS test case for the use of CFD in reactor design using the CFD-codes ANSYS CFX and TRIOCFD

Höhne, T.; Kliem, S.; Bieder, U.


Neutron Noise Observations in German KWU Built PWR and Analyses with the Reactor Dynamics Code DYN3D

Rohde, U.; Seidl, M.; Kliem, S.; Bilodid, Y.


Unsteady single phase natural circulation flow mixing prediction using 3D thermal-hydraulic system and CFD codes

Bousbia Salah, A.; Ceuca, S. C.; Puragliesi, R.; Mukin, R.; Grahn, A.; Kliem, S.; Vlassenbroeck, J.; Austregesilo, H.


2017

Testing the NURESIM platform on a PWR main steam line break benchmark

Kliem, S.; Kozmenkov, Y.; Hadek, J.; Perin, Y.; Fouquet, F.; Bernard, F.; Sargeni, A.; Cuervo, D.; Sabater, A.; Sanchez-Cervera, S.; Garcia-Herranz, N.; Zerkak, O.; Ferroukhi, H.; Mala, P.


Simulation of an MSLB scenario using the 3D neutron kinetic core model Dyn3D coupled with the CFD software Trio U

Grahn, A.; Gommlich, A.; Kliem, S.; Bilodid, Y.; Kozmenkov, Y.


Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part I: Serpent calculations.

Rachamin, R.; Kliem, S.


Statistical Analysis of the Early Phase of SBO Accident for PWR

Kozmenkov, Y.; Jobst, M.; Kliem, S.; Schaefer, F.; Wilhelm, P.


2016

The reactor Dynamics code DYN3D – models, Validation and applications

Rohde, U.; Kliem, S.; Grundmann, U.; Baier, S.; Bilodid, Y.; Duerigen, S.; Fridman, E.; Gommlich, A.; Holt, L.; Grahn, A.; Kozmenkov, Y.; Mittag, S.


Hybrid microscopic depletion model in nodal code DYN3D

Bilodid, Y.; Kotlyar, D.; Shwageraus, E.; Fridman, E.; Kliem, S.


Investigation of Feedback on Neutron Kinetics and Thermal Hydraulics from Detailed Online Fuel Behavior Modelling during a Boron Dilution Transient in a PWR with the Two-way Coupled Code System DYN3D-TRANSURANUS

Holt, L.; Rohde, U.; Kliem, S.; Baier, S.; Seidl, M.; Macían-Juan, R.


2015

Validation and verification of the coupled neutron kinetic/thermalhydraulic system code DYN3D/ATHLET

Kozmenkov, Y.; Kliem, S.; Rohde, U.


Boron dilution transient simulation analyses in a PWR with neutronics/thermal-hydraulics coupled codes in the NURISP project

Jimenez, G.; Herrero, J.; Gommlich, A.; Kliem, S.; Cuervo, D.; Jimenez, J.


Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project

Chanaron, B.; Ahnert, C.; Crouzet, N.; Sanchez, V.; Kolev, N.; Marchand, O.; Kliem, S.; Papukchiev, A.


WASA-BOSS: ATHLET-CD model for severe accident analysis for a generic KONVOI reactor

Tusheva, P.; Schäfer, F.; Kozmenkov, Y.; Kliem, S.; Hollands, T.; Trometer, A.; Buck, M.

  • Contribution to proceedings
    Jahrestagung Kerntechnik/Annual Meeting on Nuclear Technology, 05.-07.05.2015, Berlin, Germany
  • Lecture (Conference)
    Jahrestagung Kerntechnik/Annual Meeting on Nuclear Technology, 05.-07.05.2015, Berlin, Germany
  • atw - International Journal for Nuclear Power 60(2015)7, 442

Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS CFX

Grahn, A.; Kliem, S.; Rohde, U.

  • Contribution to proceedings
    ASME 2014 22nd International Conference on Nuclear Engineering (ICONE22), 07.-11.07.2014, Prag, Tschechische Republik
    Proceedings of ICONE22
  • Lecture (Conference)
    ASME 2014 22nd International Conference on Nuclear Engineering (ICONE22), 07.-11.07.2014, Prag, Tschechische Republik
  • Annals of Nuclear Energy 84(2015), 197-203
    DOI: 10.1016/j.anucene.2014.12.015

2014

Implementation of a fast running full core pin power reconstruction method in DYN3D

Gomez, A.; Sanchez Espinosa, V. H.; Kliem, S.; Gommlich, A.


Assessment of accident management measures on early in-vessel station blackout sequence at VVER-1000 pressurized water reactors

Tusheva, P.; Schäfer, F.; Reinke, N.; Kamenov, A.; Mladenov, I.; Kamenov, K.; Kliem, S.


Coolant mixing experiments in the upper plenum of the ROCOM test facility

Prasser, H.-M.; Kliem, S.


Fuel cycle advantages and dynamics features of liquid fueled MSR

Krepel, J.; Hombourger, B.; Fiorina, C.; Mikityuk, K.; Rohde, U.; Kliem, S.; Pautz, A.


Extension and application of the reactor dynamics code DYN3D for Block-type High Temperature Reactors

Baier, S.; Fridman, E.; Kliem, S.; Rohde, U.

  • Contribution to proceedings
    6th International Topical Meeting on High Temperature Reactor Technology HTR2012, 28.10.-01.11.2012, Tokyo, Japan
  • Nuclear Engineering and Design 271(2014), 431-436
    DOI: 10.1016/j.nucengdes.2013.12.013

2013

Numerical simulation of the insulation material transport in a PWR core under loss of coolant conditions

Höhne, T.; Grahn, A.; Kliem, S.; Rohde, U.; Weiss, F.-P.


Overview of major HZDR developments for fast reactor analysis

Merk, B.; Glivici-Cotruţă, V.; Duerigen, S.; Rohde, U.; Kliem, S.


2012

Research on the reactor physics and reactor safety of VVER reactors – Selected contributions to the XXIst Symposium of the Atomic Energy Research organization

Aszodi, A.; Kliem, S.

  • Kerntechnik 77(2012)4, 212-212

Study on severe accidents and countermeasures for VVER-1000 reactors using the integral code ASTEC

Tusheva, P.; Schäfer, F.; Reinke, N.; Altstadt, E.; Kliem, S.

  • Kerntechnik 77(2012)4, 271-277

Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

Rohde, U.; Baier, S.; Duerigen, S.; Fridman, E.; Kliem, S.; Merk, B.


Use of Zirconium-Based Moderators to Enhance Feedback Coefficients in a MOX-Fueled Sodium-Cooled Fast Reactor

Merk, B.; Weiß, F.-P.; Fridman, E.; Kliem, S.

  • Nuclear Science and Engineering 171(2012)2, 136-149

2011

Buoyancy driven mixing studies of natural circulation flows using ROCOM experiments and CFD

Höhne, T.; Kliem, S.; Rohde, U.


Development of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

Rohde, U.; Baier, S.; Duerigen, S.; Fridman, E.; Kliem, S.; Merk, B.

  • Kerntechnik 76(2011)3, 166-173

CFD simulation of fibre material transport in a PWR under loss of coolant conditions

Höhne, T.; Grahn, A.; Kliem, S.; Weiss, F.-P.

  • Kerntechnik 76(2011), 39-45

Development of Multi-Physics Code Systems based on the Reactor Dynamics Code DYN3D

Kliem, S.; Gommlich, A.; Grahn, A.; Rohde, U.; Schütze, J.; Frank, T.; Gomez, A.; Sanchez, V.

  • Invited lecture (Conferences)
    Fachtag der KTG: "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland", 07.-08.10.2010, Dresden, Deutschland
  • Contribution to proceedings
    Fachtag der KTG: "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland", 07.-08.10.2010, Dresden, Deutschland
    Tagungsband des Fachtages der KTG: "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland", CDROM: FZ Dresden-Rossendorf
  • Kerntechnik 76(2011)3, KT100569

Pu recycling in a full Th-MOX PWR core: Part I - steady state analysis

Fridman, E.; Kliem, S.


Burning plutonium and minimizing radioactive waste in existing PWRs

Mittag, S.; Kliem, S.


2010

Bedeutung von Experimenten für die Reaktorsicherheit

Teschendorff, V.; Glaeser, H.; Kliem, S.

  • Invited lecture (Conferences)
    Jahrestagung Kerntechnik 2009, 12.-14.05.2009, Dresden, Deutschland
  • Contribution to proceedings
    Jahrestagung Kerntechnik 2009, 12.-14.05.2009, Dresden, Deutschland
    Tagungsband der Jahrestagung Kerntechnik 2009; Fachsitzung: Thermohydraulische Experimente für Reaktoren der Generation II - III, Berlin: INFORUM GmbH
  • atw - International Journal for Nuclear Power 55(2010)03, 163-173

Experiments on slug mixing under natural circulation conditions at the ROCOM test facility using high resolution measurement technique and numerical modeling

Kliem, S.; Höhne, T.; Rohde, U.; Weiß, F.-P.

  • Lecture (Conference)
    XCFD4NRS - Experiments and CFD Codes Application to Nuclear Reactor Safety, 10.-12.09.2008, Grenoble, France
  • Contribution to proceedings
    XCFD4NRS - Experiments and CFD Codes Application to Nuclear Reactor Safety, 10.-12.09.2008, Grenoble, Frankreich
    Proceedings of the XCFD4NRS - Experiments and CFD Codes Application to Nuclear Reactor Safety, CDROM
  • Nuclear Engineering and Design 240(2010)9, 2271-2280
    DOI: doi:10.1016/j.nucengdes.2009.11.015

Application of CFD codes in nuclear reactor safety analysis

Rohde, U.; Höhne, T.; Krepper, E.; Kliem, S.

  • Invited lecture (Conferences)
    TOPSAFE 2008, 01.-03.10.2008, Dubrovnik, Kroatia
  • Contribution to proceedings
    TOPSAFE 2008, 01.-03.10.2008, Dubrovnik, Kroatia
    CD-ROM. paper 099
  • Open Access Logo Science and Technology of Nuclear Installations 2010(2010), Article ID 198758
    DOI: doi:10.1155/2010/198758

2009

ATWS analysis for PWR using the coupled code system DYN3D/ATHLET

Kliem, S.; Mittag, S.; Rohde, U.; Weiß, F.-P.


Simulation von ATWS-Transienten in Druckwasserreaktoren

Kliem, S.; Mittag, S.; Rohde, U.; Grundmann, U.; Weiß, F.-P.

  • Contribution to proceedings
    Jahrestagung Kerntechnik 2008, 27.-29.05.2008, Hamburg, Germany
    Tagungsband der Jahrestagung Kerntechnik 2008, CDROM, Berlin: INFORUM GmbH
  • Invited lecture (Conferences)
    Jahrestagung Kerntechnik 2008, 27.-29.05.2008, Hamburg, Germany
  • atw - International Journal for Nuclear Power 54(2009)2, 100-110

Experimental and numerical modeling of transition matrix from momentum to buoyancy-driven flow in a pressurized water reactor

Höhne, T.; Kliem, S.; Vaibar, R.

  • Contribution to proceedings
    16th International Conference on Nuclear Engineering ICONE16, 11.-15.05.2008, Orlando, USA
    CD_ROM, 48490
  • Lecture (Conference)
    16th International Conference on Nuclear Engineering ICONE16, 11.-15.05.2008, Orlando, USA
  • Journal of Engineering for Gas Turbines and Power - Transactions of the ASME 131(2009)1, 012906
    DOI: DOI: 10.1115/1.2983137

2008

Jahrestagung Kerntechnik 2008 - Sektionsbericht Sektion: Thermo- und Fluiddynamik

Schaffrath, A.; Kliem, S.

  • atw - International Journal for Nuclear Power 53(2008)8/9, 559-561

Experimental determination of the boron concentration distribution in the primary circuit of a PWR after a postulated cold leg small break loss-of-coolant-accident with cold leg safety injection

Kliem, S.; Prasser, H.-M.; Sühnel, T.; Weiss, F.-P.; Hansen, A.


CFX simulations of ROCOM slug mixing experiments

Moretti, F.; Melideo, D.; D’Auria, F.; Höhne, T.; Kliem, S.

  • Contribution to proceedings
    15th International Conference on Nuclear Engineering (ICONE15), 22.-26.04.2007, Nagoya, Japan
    ICONE15-10461
  • Lecture (Conference)
    15th International Conference on Nuclear Engineering (ICONE15), 22.-26.04.2007, Nagoya, Japan
  • Open Access Logo Journal of Power and Energy Systems 2(2008)2, 720-733
    DOI: 10.1299/jpes.2.720

Boron Dilution Transients during natural circulation flow in PWR – experiments and CFD simulations

Höhne, T.; Kliem, S.; Rohde, U.; Weiss, F.-P.


Experiments at the mixing test facility ROCOM for benchmarking of CFD-codes

Kliem, S.; Sühnel, T.; Rohde, U.; Höhne, T.; Prasser, H.-M.; Weiß, F.-P.

  • Contribution to proceedings
    OECD/NEA & IAEA Workshop: "Benchmarking of CFD Codes for Application to Nuclear Reactor Safety", 05.-07.09.2006, Garching, Germany
    Proceedings CD-ROM paper A4-17, Issy-les-Moulineaux, France: OECD NEA
  • Lecture (Conference)
    OECD/NEA & IAEA Workshop: "Benchmarking of CFD Codes for Application to Nuclear Reactor Safety", 05.-07.09.2006, Garching, Germany
  • Nuclear Engineering and Design 238(2008), 566-576
    DOI: 10.1016/j.nucengdes.2007.02.053

2007

DYN3D - Advanced Reactor Simulations in 3D

Rohde, U.; Grundmann, U.; Kliem, S.

  • Nuclear Energy Review 2(2007), 28-30

Jahrestagung Kerntechnik - Sektionsbericht Sektion: Thermo- und Fluiddynamik

Stieglitz, R.; Kliem, S.

  • atw - International Journal for Nuclear Power 52(2007)10, 652-654

Comparative evaluation of coolant mixing experiments at the ROCOM, Vattenfall, and Gidropress test facilities

Kliem, S.; Hemström, B.; Bezrukov, Y.; Höhne, T.; Rohde, U.

  • Open Access Logo Science and Technology of Nuclear Installations 2007(2007), 25950
    DOI: doi:10.1155/2007/25950
  • Lecture (Conference)
    Annual Meeting of the AER Working Group D, 08.-09.05.2007, Paris, France

Modeling of a Buoyancy-Driven Flow experiment in Pressurized Water Reactors using CFD-Methods

Höhne, T.; Kliem, S.

  • Open Access Logo Nuclear Engineering and Technology 39(2007)4, 327-336

Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of main steam line break benchmarks in a NPP with VVER-440 reactor

Kliem, S.; Danilin, S.; Hämäläinen, A.; Hadek, J.; Kereszturi, A.; Siltanen, P.

  • Nuclear Science and Engineering 157(2007)3, 280-298

Fluid mixing and flow distribution ín a primary circuit of a nuclear pressurized water reactor – Validation of CFD codes

Rohde, U.; Höhne, T.; Kliem, S.; Hemström, B.; Scheuerer, M.; Toppila, T.; Aszodi, A.; Boros, I.; Farkas, I.; Muehlbauer, P.; Vyskocil, V.; Klepac, J.; Remis, J.; Dury, T.


Hydrodynamic phenomena in the downcomer during flow rate transients in the primary circuit of a PWR

Cartland Glover, G. M.; Höhne, T.; Kliem, S.; Rohde, U.; Weiss, F.-P.; Prasser, H.-M.


Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD-Code ANSYS CFX

Höhne, T.; Kliem, S.; Weiß, F.-P.

  • atw - International Journal for Nuclear Power 3(2007), 168-174

Calculation of the VVER-1000 Coolant Transient Benchmark using the Coupled Code Systems DYN3D/RELAP5 and DYN3D/ATHLET

Kozmenkov, Y.; Kliem, S.; Grundmann, U.; Rohde, U.; Weiss, F.-P.


Experimente an der Versuchsanlage ROCOM zur Kühlmittelvermischung bei Wiederanlauf der Naturzirkulation

Kliem, S.; Sühnel, T.; Prasser, H.-M.; Weiß, F.-P.

  • Contribution to proceedings
    KTG-Fachtag "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland", 03.-04.04.2006, Dresden, Germany
    Tagungsband FZR-455, 1437-322X, I-6
  • atw - International Journal for Nuclear Power (2007), 352-360
  • Invited lecture (Conferences)
    KTG-Fachtag "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland", 03.-04.04.2006, Dresden, Germany

2006

Simulating turbulent mixing in nuclear reactor pressure vessels

Rohde, U.; Höhne, T.; Kliem, S.

  • ANSYS Solutions 7(2006)2, 27-28

Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled thermal-hydraulic 3D kinetics code

Bousbia Salah, A.; Kliem, S.; Rohde, U.; D’Auria, F.; Petruzzi, A.


Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations

Kliem, S.; Kozmenkov, Y.; Höhne, T.; Rohde, U.


Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD-codes CFX-5 and TRIO_U

Höhne, T.; Kliem, S.; Bieder, U.


2005

Validation of coupled codes using VVER plant measurements

Vanttola, T.; Hämäläinen, A.; Kliem, S.; Kozmenkov, Y.; Weiß, F.-P.; Kereszturi, A.; Hadek, J.; Strmensky, C.; Stefanova, S.; Kuchin, A.

  • Nuclear Engineering and Design 235(2005), 507-519 (2005)

Comprehensive uncertainty and sensitivity analysis for coupled code calculations of VVER plant transients

Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Hegyi, G.; Kereszturi, A.; Kliem, S.; Hadek, J.; Danilin, S.; Nikonov, S.; Kuchin, A.

  • Nuclear Engineering and Design 235(2005), 521-540

Fluid mixing and flow distribution in the reactor circuit - Part 1: Measurement data base

Rohde, U.; Kliem, S.; Höhne, T.; Karlsson, R.; Hemström, B.; Lillington, J.; Toppila, T.; Elter, J.; Bezrukov, Y.

  • Nuclear Engineering and Design, 235(2005), 421-443

2004

Core response of a PWR to a slug of under-borated water

Kliem, S.; Rohde, U.; Weiß, F.-P.


Analysis of the Boiling Water Reactor Turbine Trip Benchmark with the Codes DYN3D and ATHLET/DYN3D

Grundmann, U.; Kliem, S.; Rohde, U.

  • Nuclear Science and Engineering 148(2004), 226-234

2003

Experimental and Numerical Investigation of Boron Dilution Transients in Pressurized Water Reactors

Hertlein, R.; Umminger, K.; Kliem, S.; Prasser, H.-M.; Höhne, T.; Weiß, F.-P.

  • Nuclear Technology, vol. 141,January 2003, pp. 88-107

Coolant mixing in a PWR - deboration transients, steam line breaks and emergency core cooling injection - experiments and analyses

Prasser, H.-M.; Grunwald, G.; Höhne, T.; Kliem, S.; Rohde, U.; Weiss, F.-P.

  • Lecture (Conference)
    International Congress on Advanced Nuclear Power Plants (ICAPP), June 9-13, 2002 - Hollywood Florida, USA, Proc. CD-ROM, paper #1214.
  • Nuclear Technology 143 (2003) 37-56
  • Contribution to proceedings
    International Congress on Advanced Nuclear Power Plants (ICAPP), June 9-13, 2002 - Hollywood Florida, USA, Proc. CD-ROM, paper #1214.

Analyses of the OECD Main Steam Line Break Benchmark with the Codes DYN3D and ATHLET

Grundmann, U.; Kliem, S.

  • Nuclear Technology 142(2003) 146-153

2002

Validation of coupled neutron kinetic / thermal-hydraulic codes Part 2: Analysis of a VVER-440 transient (Loviisa-1)

Hämäläinen, A.; Kyrki-Rajamäki, R.; Mittag, S.; Kliem, S.; Weiss, F.-P.; Langenbuch, S.; Danilin, S.; Hadek, J.; Hegyi, G.

  • Annals of Nuclear Energy 29 (2002) 255-269

2001

Validation of coupled neutron kinetic / thermal-hydraulic codes Part 1: Analysis of a VVER-1000 transient (Balakovo-4)

Mittag, S.; Kliem, S.; Weiß, F.-P.; Kyrki-Rajamäki, R.; Hämäläinen, A.; Langenbuch, S.; Danilin, S.; Hadek, J.; Hegyi, G.; Kuchin, A.; Panayotov, D.

  • Annals of Nuclear Energy 28/9 (2001) 857-873

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