Kontakt

Dr. Sören Kliem

Leiter
Reaktorsicherheit
s.kliemAthzdr.de
Tel.: +49 351 260 2318

Persönliche Webseite von Sören Kliem

Wissenschaftlicher Werdegang

  • Seit 2010:
    Abteilungsleiter "Reaktorsicherheit" am Helmholtz-Zentrum Dresden- Rossendorf, Institut für Ressourcenökologie
  • 2010:
    Promotion an der Fakultät Maschinenwesen der Technischen Universität Dresden
  • 1994-2010:
    Wissenschaftlicher Mitarbeiter am Forschungszentrum Rossendorf, Institut für Sicherheitsforschung
  • 1992:
    Master of Techn. Sciences an der Energiewirtschaftlichen Hochschule Moskau (MEI - Technische Universität)

Forschungsschwerpunkte

  • Reaktordynamik
  • Störfallanalyse von Kernreaktoren
  • Kopplung von neutronenkinetischen und thermohydraulischen Rechenprogrammen
  • Experimentelle und numerische Untersuchungen der Kühlmittelvermischung in Druckwasserreaktoren

Projekte

  • 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, 2019 - 2023
  • Horizon 2020 McSAFER: High-Performance Advanced Methods and Experimental Investigations for the Safety Evaluation of Generic Small Modular Reactors, 2020 - 2023

Veröffentlichungen

2024

NuScale-like SMR Model Development and Applied Safety Analyses with the Code Chain Serpent-DYN3D-ATHLET

Diaz Pescador, E.; Bilodid, Y.; Jobst, M.; Kliem, S.


2023

Extension of the X2 VVER-1000 benchmark by a control rod cluster ejection exercise

Bilodid, Y.; Fischer, M.; Zilly, M.; Aures, A.; Henry, R.; Kilger, R.; Kliem, S.


Extension of the DYN3D/ATHLET code system to SFR applications: models description and initial validation

Fridman, E.; Nikitin, E.; Ponomarev, A.; Di Nora, A.; Kliem, S.; Mikityuk, K.

Verknüpfte Publikationen


Numerical Analysis Related to the ROCOM Pressurized Thermal Shock Benchmark

Höhne, T.; Kliem, S.


2022

On the validation of ATHLET 3-D features for the simulation of multidimensional flows in horizontal geometries under single-phase subcooled conditions

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


Verification of the code DYN3D for calculations of neutron flux fluctuations

Viebach, M.; Lange, C.; Kliem, S.; Demaziere, C.; Rohde, U.; Hennig, D.; Hurtado, A.


2021

The H2020 McSAFER project: Main goals, technical work program, and status

Sanchez-Espinoza, V. H.; Gabriel, S.; Suikkanen, H.; Telkkä, J.; Valtavirta, V.; Bencik, M.; Kliem, S.; Queral, C.; Farda, A.; Abéeguilée, F.; Smith, P.; van Uffelen, P.; Ammirabile, L.; Seidl, M.; Schneidesch, C.; Grishchenko, D.; Lestani, H.


SERPENT2-SUBCHANFLOW-TRANSURANUS pin-by-pin depletion calculations for a PWR fuel assembly

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


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.


Modelling of multidimensional effects in thermal-hydraulic system codes under asymmetric flow conditions – Simulation of ROCOM Tests 1.1 and 2.1 with ATHLET 3D-Module

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


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.

  • Beitrag zu Proceedings
    CFD4NRS-8 : Computational Fluid Dynamics for Nuclear Reactor Safety - OECD/NEA Workshop, 25.-27.11.2020, Palaiseau, Frankreich
  • Vortrag (Konferenzbeitrag)
    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
    Online First (2020) 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.

Verknüpfte Publikationen


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.

  • Beitrag zu Proceedings
    Jahrestagung Kerntechnik/Annual Meeting on Nuclear Technology, 05.-07.05.2015, Berlin, Germany
  • Vortrag (Konferenzbeitrag)
    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.

  • Beitrag zu Proceedings
    ASME 2014 22nd International Conference on Nuclear Engineering (ICONE22), 07.-11.07.2014, Prag, Tschechische Republik
    Proceedings of ICONE22
  • Vortrag (Konferenzbeitrag)
    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
    Cited 26 times in Scopus

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.


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
    ISSN: 0932-3902

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.

  • Eingeladener Vortrag (Konferenzbeitrag)
    Fachtag der KTG: "Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland", 07.-08.10.2010, Dresden, Deutschland
  • Beitrag zu 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.

  • Eingeladener Vortrag (Konferenzbeitrag)
    Jahrestagung Kerntechnik 2009, 12.-14.05.2009, Dresden, Deutschland
  • Beitrag zu 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
    ISSN: 1431-5254

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.

  • Vortrag (Konferenzbeitrag)
    XCFD4NRS - Experiments and CFD Codes Application to Nuclear Reactor Safety, 10.-12.09.2008, Grenoble, France
  • Beitrag zu 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: 10.1016/j.nucengdes.2009.11.015
    Cited 30 times in Scopus

Application of CFD codes in nuclear reactor safety analysis

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

  • Eingeladener Vortrag (Konferenzbeitrag)
    TOPSAFE 2008, 01.-03.10.2008, Dubrovnik, Kroatia
  • Beitrag zu 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: 10.1155/2010/198758
    Cited 24 times in Scopus

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.

  • Beitrag zu Proceedings
    Jahrestagung Kerntechnik 2008, 27.-29.05.2008, Hamburg, Germany
    Tagungsband der Jahrestagung Kerntechnik 2008, CDROM, Berlin: INFORUM GmbH
  • Eingeladener Vortrag (Konferenzbeitrag)
    Jahrestagung Kerntechnik 2008, 27.-29.05.2008, Hamburg, Germany
  • atw - International Journal for Nuclear Power 54(2009)2, 100-110
    ISSN: 1431-5254

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

Höhne, T.; Kliem, S.; Vaibar, R.

  • Beitrag zu Proceedings
    16th International Conference on Nuclear Engineering ICONE16, 11.-15.05.2008, Orlando, USA
    CD_ROM, 48490
  • Vortrag (Konferenzbeitrag)
    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: 10.1115/1.2983137
    Cited 15 times in Scopus

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.

  • Beitrag zu Proceedings
    15th International Conference on Nuclear Engineering (ICONE15), 22.-26.04.2007, Nagoya, Japan
    ICONE15-10461
  • Vortrag (Konferenzbeitrag)
    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.

  • Beitrag zu 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
  • Vortrag (Konferenzbeitrag)
    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
    ISSN: 0029-5493
    Cited 80 times in Scopus

2007

DYN3D - Advanced Reactor Simulations in 3D

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

  • Nuclear Energy Review 2(2007), 28-30
    ISSN: 1753-3910

Jahrestagung Kerntechnik - Sektionsbericht Sektion: Thermo- und Fluiddynamik

Stieglitz, R.; Kliem, S.

  • atw - International Journal for Nuclear Power 52(2007)10, 652-654
    ISSN: 1431-5254

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: 10.1155/2007/25950
    ISSN: 1687-6075
  • Vortrag (Konferenzbeitrag)
    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
    ISSN: 1431-5254

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.

  • Beitrag zu 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
    ISSN: 1431-5254
  • Eingeladener Vortrag (Konferenzbeitrag)
    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.

  • Vortrag (Konferenzbeitrag)
    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
  • Beitrag zu 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