VVER-1000 Coolant Transient Benchmark - Overview and Status of Phase 2
|
|
- Rodger Richardson
- 5 years ago
- Views:
Transcription
1 International Conference Nuclear Energy for New Europe 2005 Bled, Slovenia, September 5-8, 2005 VVER-1000 Coolant Transient Benchmark - Overview and Status of Phase 2 Nikola Kolev, Nikolay Petrov Institute for Nuclear Research and Nuclear Energy 72 Tzarigradsko chaussee, Blvd., BG-1784 Sofia, Bulgaria npkolev@inrne.bas.bg, nlp@ mail.bg Sylvie Aniel, Eric Royer Commissariat à l Energie Atomique DM2S/SFME, Centre de Scalay F Gif sur Yvette Cedex, France sylvie.aniel@cea.fr, eric.royer@cea.fr Ulrich Bieder Commissariat à l Energie Atomique DER/SSTH, Centre de Grenoble 17 rue des Martyrs, F Grenoble Cedex 9, France ulrich.bieder@cea.fr ABSTRACT The Main Steam Line Break (MSLB) is identified as a Design Basis Accident (DBA) exhibiting significant localized space-time effects. There is a need to validate the models used in thermal-hydraulics and neutronics, and particularly the flow mixing in the core vessel. For this purpose, a consistent approach is defined with three exercises. First, a Coolant Mixing test carried out at the Kozloduy Nuclear Power Plant (KNPP) Unit 6 is used to validate the thermal-hydraulics of the vessel. Several types of models are possible, from Large Eddy Scale to multi-channels. Then two scenarios of MSLB are specified to check the coupling between core vessel thermal-hydraulics and neutronics on one hand, and between vessel and plant on the other hand. The paper presents an overview of the V1000CT-2 benchmark exercises and currently available results. 1 INTRODUCTION In the framework of joint effort between the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD), the United States Department of Energy (US DOE), and the Commissariat à l Energie Atomique (CEA), France a coupled 3-D neutron kinetics/thermal hydraulics benchmark was defined [2]. The benchmark is based on data from the Unit 6 of the Bulgarian Kozloduy NPP. In performing this work the Pennsylvania State University (PSU), USA and CEA-Saclay, France have collaborated with Bulgarian organizations, in particular with KNPP and the Institute for 116.1
2 116.2 Nuclear Research and Nuclear Energy (INRNE). The benchmark consists of two phases: Phase 1: Main Coolant Pump Switching On; Phase 2: Coolant Mixing Tests and MSLB. Since the previous coupled code benchmarks [1] indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and MSLB transients were selected for simulation in Phase 2 of the benchmark. The MSLB event is characterized by a large asymmetric cooling of the core, stuck rods and a large primary coolant flow variation. Two scenarios are defined: the first scenario is taken from the current licensing practice and the second is derived from the original one using aggravating assumptions to enhance the code-to-code comparisons. Three workshops have taken place since the starter benchmark workshop in Dresden. The first workshop was held in Saclay, France in May 2003, the second workshop was in Sofia, Bulgaria in April 2004, and the latest one in Garching, Germany in April During the third workshop, the final results for V1000CT-1 were discussed, and preliminary results for V1000CT-2 Exercise 1 were presented. This paper aims at presenting the three exercises of the benchmark, discuss the modelling issues of temperature mixing in the core vessel, and the application to a MSLB analysis. 2 DESCRIPTION OF V1000CT-2 EXERCISES 2.1 Exercise 1: Computation of coolant mixing experiments This exercise is based on a comparison with a mixing experiment conducted at Kozloduy-6 as part of the plant-commissioning phase. The experiment includes isolation of a steam generator at 9.3% of the nominal power causing single loop heat-up, with all MCP in operation. It is characterized by temperature rise of about 14 degrees and a decrease of mass flow rate by 3.4% in the disturbed loop, affecting the neighbouring loops as well. It will be used to test and validate vessel-mixing models (CFD, coarse-mesh and mixing matrix). Vessel boundary conditions and core power distribution along with pressure above the core will be part of the exercise specification. A particular effort was necessary to define the vessel geometry, because plant specific data appeared to be sensitive on the flow mixing. The task is to calculate the core inlet and outlet distributions. 2.2 Exercises 2 and 3: Main Steam-Line Break (MSLB) modelling The transient to be analyzed is initiated by a main steam line break in a VVER-1000 between the steam generator (SG) and the steam isolation valve (SIV), outside the containment. A mechanical failure of the main feed water regulation valve is assumed. This event is characterized by a large asymmetric cooling of the core, stuck control rods and a large primary coolant flow variation. Two scenarios will be defined: the first scenario is taken from the current licensing practice and the second is derived from the original one using aggravating assumptions to enhance the code-to-code comparison. The main objective of the study is to clarify the local 3-D feedback effects depending on the vessel mixing. Special emphasis is put on testing 3-D vessel thermal-hydraulic (T-H) models and the coupling of 3-D neutronics/vessel thermal hydraulics. The MSLB is thus divided in two exercises (to be done for the two scenarios): Exercise 2 consists of coupled 3-D neutronics/vessel thermal-hydraulic simulations using specified vessel T-H boundary conditions and Exercise 3 consists of best estimate coupled plant simulations (plant, 3-D vessel and core).
3 STATUS AND PRELIMINARY RESULTS 3.1 Exercise 1 Specifications for Exercise 1 have been delivered in March, 2004, after a comprehensive verification of the geometrical data of the pressure vessel of KNPP. These data are available both in spreadsheet format with tables and drawings, and as a CAD geometry file. The geometry of both, the reactor vessel and of the lower plenum with the 163 support columns is presented in Figure 1. Cold leg nozzle Zoom of the Lower Plenum Downcomer Consols Support Column Core Plate Elliptic RPV Narrowing Gap Support Column Perforated Barrel Figure 1: Geometry of the VVER-1000 reactor There are currently six participants who have already obtained preliminary results, and in addition six other participants (in italic in Table 1) who have intention to perform the calculations. Table 1: List of participants for V1000CT-2 Exercise 1 Company Country Code CEA France Trio_U FZR Germany CFX 5 FZK Germany CFX 5 Kurchatov Institut Russia ATHLET Penn State and ORNL USA RELAP 3D INRNE Bulgaria CATHARE GRS Germany ATHLET University of Pisa Italy RELAP VTT Finland Porflo NRI Czech Republic Fluent Technical University of Budapest Hungary CFX 5 EREC, Electrogorsk Russia In-house CFD code
4 116.4 Experimental data show a net counter clock-wise shift (rotation) of "loop flow centres" in certain VVER-1000 V320 flow patterns. The swirl intensity is depending on the considered unit. The origin of the swirl is not clearly identified, however the asymmetry of the actual vessel and internals due to the fabrication process as well as local flow disturbances are expected to be the main sources of the swirl. CFD computations with the Trio_U code [4] have been carried out by CEA [3]. The experimentally detected swirl can be reproduced with the Trio_U code by using a tetrahedral mesh of about control volumes and an under-resolved LES turbulence modelling approach. The calculated and measured temperature distribution at the core inlet is given in Figure 2. Here, the stabilized thermal hydraulic situation about 20 minutes after the isolation of SG-1 is shown. The locations of the cold legs are also added to this Figure. The calculated asymmetric temperature field and the counter clockwise rotation of 24 of the temperature maximums with respect to the axis of cold leg 1 are in excellent agreement with the experimental data. Experiment Trio_U Calculation Leg 1 Leg 1 Figure 2 : Temperature distribution at the core inlet at the end of the test The experimental values include the mixing in the core region, what is not taken into account in the calculation. This explains the enhanced mixing in the experiment at the periphery of the zone affected by leg 1.
5 Trio_U Experiment Temperature Assembly Figure 3: Assembly by assembly comparison of the fluid temperature ( C) A more quantitative evaluation is given in Figure 3. The measured and calculated mean temperature of the 163 assemblies are compared, where the numbering goes line by line from the lower left to the upper right assembly (see Figure 2). Exercises 2 and 3 MSLB situations have been computed by INRNE with the CATHARE 1.3 code [5], in order to specify the scenarios. The core vessel model is based on four channels, and has been validated against the pump start-up problem (Phase 1 of the benchmark) and the coolant mixing problem (Phase 2, Exercise 1). The results have been discussed during the third workshop in Garching. The initial conditions for the accident correspond to the end of Cycle 8 at nominal power. The hypotheses for the realistic scenario are: a double ended break on Steam Line 4, upstream of the isolation valve, the SG water inventory is maximal, the SG-4 feedwater regulating valve is stuck in fully open position, the most effective control rod is not inserted when scram occurs, and boron of HPSI is not taken into account. This scenario, labelled Scenario 1, is very close to what is used in the current licensing practice. A preliminary calculation of the reactor vessel boundary conditions is shown in Figures 4 through 6. There is no return to power, because the Main Coolant Pump 4 trips, resulting in a reverse flow in loop 4 (Figure 5) and a limited cooling of the core (Figure 4). In order to enhance the reactivity insertion (return to power after scram, Figure 7) and the multi-dimensional effects, two extra hypotheses are considered for the pessimistic scenario labelled Scenario 2: Main Coolant Pump 4 does not trip on signal, and a second control rod remains stuck out of the core. Preliminary results for the reactor power history and vessel boundary conditions are shown in Figures 7 through 10. The temperature at cold leg 4 drops to 210 C (Figure 8).The preliminary calculations are used to derive the vessel boundary
6 116.6 conditions for Exercise 2, and also to define the range of parameters for the cross-section libraries: fuel temperature, moderator density and temperature. Cross-section libraries for the reference core are being prepared with the Helios 1.7 lattice code in PSU-INRNE-KNPP collaboration. Figure 4: History of cold leg temperatures for Scenario 1 (realistic) Figure 5: History of vessel inlet flows for Scenario 1 (realistic)
7 116.7 Figure 6: History of vessel outlet pressures for Scenario 1 (realistic) Figure 7: History of relative power for Scenario 2 (pessimistic)
8 116.8 Figure 8: History of cold leg temperatures for Scenario 2 (pessimistic) Figure 9: History of vessel inlet flows for Scenario 2 (pessimistic)
9 116.9 Figure 10: History of vessel outlet pressure for Scenario 2 (pessimistic) 4 CONCLUSIONS AND PERSPECTIVES V1000CT-2 benchmark consists of three exercises. The first exercise, dedicated to the computation of a flow mixing experiment in the core vessel, is under way for the participants. Several types of thermal-hydraulic models are used, ranging from Large Eddy Scale to multi- 1D channel. The final results are expected by September 2005, in order to be analysed by the benchmark team, and finally presented during the fourth workshop in April The second and third exercises are devoted to the simulation of a Main Steam Line Break accident. Two scenarios are currently under definition for the participants: the realistic one is close to the licensing practice, whereas the pessimistic one assumes aggravating events: the Main Coolant Pump of the faulted loop does not trip on signal and a second control rod remains stuck-out of the core when the scram occurs. The pessimistic scenario results in a significant return to power and well pronounced multi-dimensional effects in the core. Participants results for exercises 2 and 3 of V1000CT-2 are expected in 2006 in order to have a comparative analysis. The conclusions will be reported in OECD documents. ACKNOWLEDGMENTS The authors thank all the members of the benchmark team for their support. REFERENCES [1] Ivanov, K., Beam, T., Baratta, A., Irani, A., and Trikorous, N., PWR MSLB Benchmark. Volume 1: Final Specifications, NEA/NSC/DOC (99) 8 [2] Ivanov, B., Ivanov, K., Royer, E., Aniel, U., Kolev, N. and Groudev, P. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark for assessing coupled neutronics/thermal-hydraulics system codes for VVER-1000 RIA analysis, Proc. PHYSOR 2004, April 25-29, 2004, Chicago, IL USA
10 [3] Bieder, U., Fauchet, G., Betin, S., Kolev, N., Popov, D., Simulation of mixing effects in a VVER-1000 reactor, Proc. NURETH 11 conference, October 2-6, 2005, Avignon, France [4] Bieder, U., Calvin C., Mutelle H., Detailed thermal-hydraulic analysis of induced break severe accidents using the massively parallel CFD code Trio_U. Int. Conf. on Supercomputing in Nuclear Applications, Paris September 2003 [5] Bestion, D., The physical closure laws in the CATHARE code, Nuclear Engineering and Design, vol. 124, Dec. 1990, pp
QUALIFICATION OF A CFD CODE FOR REACTOR APPLICATIONS
QUALIFICATION OF A CFD CODE FOR REACTOR APPLICATIONS Ulrich BIEDER whole TrioCFD Team DEN-STMF, CEA, UNIVERSITÉ PARIS-SACLAY www.cea.fr SÉMINAIRE ARISTOTE, NOVEMBER 8, 2016 PAGE 1 Outline Obective: analysis
More informationANALYSIS OF THE OECD MSLB BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE RELAP5/PARCS
ANALYSIS OF THE OECD MSLB BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE RELAP5/PARCS T. Kozlowski, R. M. Miller, T. Downar School of Nuclear Engineering Purdue University United States
More informationDEVELOPMENT OF A COUPLED CODE SYSTEM BASED ON SPACE SAFETY ANALYSIS CODE AND RAST-K THREE-DIMENSIONAL NEUTRONICS CODE
DEVELOPMENT OF A COUPLED CODE SYSTEM BASED ON SPACE SAFETY ANALYSIS CODE AND RAST-K THREE-DIMENSIONAL NEUTRONICS CODE Seyun Kim, Eunki Lee, Yo-Han Kim and Dong-Hyuk Lee Central Research Institute, Korea
More informationSafety Analyses for Dynamical Events (SADE) SAFIR2018 Interim Seminar Ville Sahlberg
VTT TECHNICAL RESEARCH CENTRE OF FINLAND LTD Safety Analyses for Dynamical Events (SADE) SAFIR2018 Interim Seminar Ville Sahlberg SADE Project SADE produces more reliable answers to safety requirements
More informationDepartment of Engineering and System Science, National Tsing Hua University,
3rd International Conference on Materials Engineering, Manufacturing Technology and Control (ICMEMTC 2016) The Establishment and Application of TRACE/CFD Model for Maanshan PWR Nuclear Power Plant Yu-Ting
More informationTitle: Development of a multi-physics, multi-scale coupled simulation system for LWR safety analysis
Title: Development of a multi-physics, multi-scale coupled simulation system for LWR safety analysis Author: Yann Périn Organisation: GRS Introduction In a nuclear reactor core, different fields of physics
More informationAPPLICATION OF THE COUPLED THREE DIMENSIONAL THERMAL- HYDRAULICS AND NEUTRON KINETICS MODELS TO PWR STEAM LINE BREAK ANALYSIS
APPLICATION OF THE COUPLED THREE DIMENSIONAL THERMAL- HYDRAULICS AND NEUTRON KINETICS MODELS TO PWR STEAM LINE BREAK ANALYSIS Michel GONNET and Michel CANAC FRAMATOME Tour Framatome. Cedex 16, Paris-La
More informationANALYSIS OF THE OECD PEACH BOTTOM TURBINE TRIP 2 TRANSIENT BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE TRAC-M/PARCS
ANALYSIS OF THE OECD PEACH BOTTOM TURBINE TRIP 2 TRANSIENT BENCHMARK WITH THE COUPLED NEUTRONIC AND THERMAL-HYDRAULICS CODE TRAC-M/PARCS Deokjung Lee and Thomas J. Downar School of Nuclear Engineering
More informationDevelopments and Applications of TRACE/CFD Model of. Maanshan PWR Pressure Vessel
Developments and Applications of TRACE/CFD Model of Maanshan PWR Pressure Vessel Yu-Ting Ku 1, Yung-Shin Tseng 1, Jung-Hua Yang 1 Shao-Wen Chen 2, Jong-Rong Wang 2,3, and Chunkuan Shin 2,3 1 : Department
More informationTHERMAL HYDRAULIC REACTOR CORE CALCULATIONS BASED ON COUPLING THE CFD CODE ANSYS CFX WITH THE 3D NEUTRON KINETIC CORE MODEL DYN3D
THERMAL HYDRAULIC REACTOR CORE CALCULATIONS BASED ON COUPLING THE CFD CODE ANSYS CFX WITH THE 3D NEUTRON KINETIC CORE MODEL DYN3D A. Grahn, S. Kliem, U. Rohde Forschungszentrum Dresden-Rossendorf, Institute
More informationPRESSURISED WATER REACTOR MAIN STEAM LINE BREAK (MSLB) BENCHMARK
Nuclear Science ISBN 92-64-02152-3 NEA/NSC/DOC(2003)21 NEA NUCLEAR SCIENCE COMMITTEE NEA COMMITTEE ON SAFETY OF NUCLEAR INSTALLATIONS PRESSURISED WATER REACTOR MAIN STEAM LINE BREAK (MSLB) BENCHMARK Volume
More informationSensitivity Analyses of the Peach Bottom Turbine Trip 2 Experiment
International Conference Nuclear Energy for New Europe 2003 Portorož, Slovenia, September 8-11, 2003 http://www.drustvo-js.si/port2003 Sensitivity Analyses of the Peach Bottom Turbine Trip 2 Experiment
More informationUNCERTAINTY AND SENSITIVITY ANALYSIS OF THE OECD/NEA KALININ-3 BENCHMARK. {Ihor.Pasichnyk, Winfried.Zwermann,
UNCERTAINTY AND SENSITIVITY ANALYSIS OF THE OECD/NEA KALININ-3 BENCHMARK I. Pasichnyk 1, S. Nikonov 2, W. Zwermann 1, K. Velkov 1 1 Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh, Garching,
More informationPWR CONTROL ROD EJECTION ANALYSIS WITH THE MOC CODE DECART
PWR CONTROL ROD EJECTION ANALYSIS WITH THE MOC CODE DECART Mathieu Hursin and Thomas Downar University of California Berkeley, USA mhursin@nuc.berkeley.edu,downar@nuc.berkeley.edu ABSTRACT During the past
More informationStratification issues in the primary system. Review of available validation experiments and State-of-the-Art in modelling capabilities.
Stratification issues in the primary system. Review of available validation experiments and State-of-the-Art in modelling capabilities. (StratRev) NKS seminar, Armémuseum, 2009-03-26 Johan Westin and Mats
More informationCFX CODE APPLICATION TO THE FRENCH REACTOR FOR INHERENT BORON DILUTION SAFETY ISSUE
CFX CODE APPLICATION TO THE FRENCH REACTOR FOR INHERENT BORON DILUTION SAFETY ISSUE Estelle Graffard, Frédéric Goux Institute for Radiological Protection and Nuclear Safety, France Abstract Inherent boron
More informationInstability Analysis in Peach Bottom NPP Using a Whole Core Thermalhydraulic-Neutronic Model with RELAP5/PARCS v2.7
Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol., pp.10-18 (011) ARTICLE Instability Analysis in Peach Bottom NPP Using a Whole Core Thermalhydraulic-Neutronic Model with RELAP/PARCS v. Agustín ABARCA,
More informationResearch Article Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code
Hindawi Science and Technology of Nuclear Installations Volume 2017, Article ID 5151890, 8 pages https://doi.org/10.1155/2017/5151890 Research Article Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal
More informationEUROPEAN COMMISSION 5th EURATOM FRAMEWORK PROGRAMME KEY ACTION : NUCLEAR FISSION FLOMIX-R FIKS-CT
EUROPEAN COMMISSION 5th EURATOM FRAMEWORK PROGRAMME 1998-2002 KEY ACTION : NUCLEAR FISSION FLOMIX-R FIKS-CT-2001-00197 Deliverable D13 Final Summary report (extended version) Dissemination level : PU:
More informationValidation of Traditional and Novel Core Thermal- Hydraulic Modeling and Simulation Tools
Validation of Traditional and Novel Core Thermal- Hydraulic Modeling and Simulation Tools Issues in Validation Benchmarks: NEA OECD/US NRC NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) Benchmark Maria
More informationA Hybrid Deterministic / Stochastic Calculation Model for Transient Analysis
A Hybrid Deterministic / Stochastic Calculation Model for Transient Analysis A. Aures 1,2, A. Pautz 2, K. Velkov 1, W. Zwermann 1 1 Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh Boltzmannstraße
More informationBUOYANCY DRIVEN MIXING STUDIES OF NATURAL CIRCULATION FLOWS AT THE ROCOM FACILITY USING THE ANSYS CFX CODE
BUOYANCY DRIVEN MIXING STUDIES OF NATURAL CIRCULATION FLOWS AT THE ROCOM FACILITY USING THE ANSYS CFX CODE 1. Introduction Thomas Höhne, Sören Kliem, Ulrich Rohde, and Frank-Peter Weiss A small break loss
More informationTOWARDS A COUPLED SIMULATION OF THERMAL HYDRAULICS AND NEUTRONICS IN A SIMPLIFIED PWR WITH A 3x3 PIN ASSEMBLY
TOWARDS A COUPLED SIMULATION OF THERMAL HYDRAULICS AND NEUTRONICS IN A SIMPLIFIED PWR WITH A 3x3 PIN ASSEMBLY Anni Schulze, Hans-Josef Allelein Institute for Reactor Safety and Reactor Technology, RWTH
More informationSafety Analysis of Loss of Flow Transients in a Typical Research Reactor by RELAP5/MOD3.3
International Conference Nuclear Energy for New Europe 23 Portorož, Slovenia, September 8-11, 23 http://www.drustvo-js.si/port23 Safety Analysis of Loss of Flow Transients in a Typical Research Reactor
More informationPRESSURISED WATER REACTOR MAIN STEAM LINE BREAK (MSLB) BENCHMARK
NEA/NSC/DOC(2002)12 NEA NUCLEAR SCIENCE COMMITTEE NEA COMMITTEE ON SAFETY OF NUCLEAR INSTALLATIONS PRESSURISED WATER REACTOR MAIN STEAM LINE BREAK (MSLB) BENCHMARK Volume III: Results of Phase 2 on 3-D
More informationEUROPEAN COMMISSION 5th EURATOM FRAMEWORK PROGRAMME KEY ACTION : NUCLEAR FISSION
EUROPEAN COMMISSION 5th EURATOM FRAMEWORK PROGRAMME 1998-2002 KEY ACTION : NUCLEAR FISSION FLOMIX-R FLUID MIXING AND FLOW DISTRIBUTION IN THE PRIMARY CIRCUIT CO-ORDINATOR Dr. Ulrich Rohde Forschungszentrum
More informationThe Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit
The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit ABSTRACT Peter Hermansky, Marian Krajčovič VUJE, Inc. Okružná
More informationATLAS Facility Description Report
KAERI/TR-3754/2009 기술보고서 ATLAS Facility Description Report ATLAS 실험장치기술보고서 한국원자력연구원 제출문 한국원자력연구원장귀하 본보고서를 2009 연도 APR1400/OPR1000 핵심사고열수력종합 효과실험 과제의기술보고서로제출합니다. 2009. 4. 주저자 : 강경호공저자 : 문상기박현식조석최기용 ATLAS
More informationLoads on RPV Internals in a PWR due to Loss-of-Coolant Accident considering Fluid-Structure Interaction
Loads on RPV Internals in a PWR due to Loss-of-Coolant Accident considering Fluid-Structure Interaction Dr. P. Akimov, Dr. M. Hartmann, L. Obereisenbuchner Fluid Dynamics Stuttgart, May 24, 2012 Content
More informationCalculation of a Reactivity Initiated Accident with a 3D Cell-by-Cell Method: Application of the SAPHYR System to a Rod Ejection Accident in TMI1
Calculation of a Reactivity Initiated Accident with a 3D Cell-by-Cell Method: Application of the SAPHYR System to a Rod Ejection Accident in TMI1 S. Aniel-Buchheit 1, E. Royer 2, P. Ferraresi 3 1 S. Aniel
More informationOECD/NEA Transient Benchmark Analysis with PARCS - THERMIX
OECD/NEA Transient Benchmark Analysis with PARCS - THERMIX Volkan Seker Thomas J. Downar OECD/NEA PBMR Workshop Paris, France June 16, 2005 Introduction Motivation of the benchmark Code-to-code comparisons.
More informationANALYSIS OF THE OECD MAIN STEAM LINE BREAK BENCHMARK PROBLEM USING THE REFINED CORE THERMAL-HYDRAULIC NODALIZATION FEATURE OF THE MARS0MASTER CODE
ANALYSIS OF THE OECD MAIN STEAM LINE BREAK BENCHMARK PROBLEM USING THE REFINED CORE THERMAL-HYDRAULIC NODALIZATION FEATURE OF THE MARS0MASTER CODE THERMAL HYDRAULICS KEYWORDS: MARS/MASTER code, coupled
More informationPresenters: E.Keim/Dr.R.Trewin (AREVA GmbH) WP6.9 Task leader: Sébastien Blasset (AREVA-G) NUGENIA+ Final Seminar, Helsinki August, 2016
NUGENIA+ WP6.9 DEFI-PROSAFE DEFInition of reference case studies for harmonized PRObabilistic evaluation of SAFEty margins in integrity assessment for LTO of RPV/DEFI-PROSAFE Presenters: E.Keim/Dr.R.Trewin
More informationThermal Hydraulic System Codes Performance in Simulating Buoyancy Flow Mixing Experiment in ROCOM Test Facility
Thermal Hydraulic System Codes Performance in Simulating Buoyancy Flow Mixing Experiment in ROCOM Test Facility ABSTRACT Eugenio Coscarelli San Piero a Grado Nuclear Research Group (GRNSPG), University
More informationInternational Benchmark on
Nuclear Science NEA/NSC/R(2015)7 March 2016 www.oecd-nea.org International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests Volume III: Departure from nucleate boiling Nuclear Science
More information3D CFD and FEM Evaluations of RPV Stress Intensity Factor during PTS Loading
E-Journal of Advanced Maintenance Vol.9-2 (2017) 84-90 Japan Society of Maintenology 3D CFD and FEM Evaluations of RPV Stress Intensity Factor during PTS Loading Xiaoyong Ruan 1,*, Toshiki Nakasuji 1 and
More informationBoiling Water Reactor Turbine Trip (TT) Benchmark
Nuclear Science ISBN 978-92-64-99137-8 NEA/NSC/DOC(2010)11 NEA Nuclear Science Committee NEA Committee on the Safety of Nuclear Installations US Nuclear Regulatory Commission Boiling Water Reactor Turbine
More informationStatus and Future Challenges of CFD for Liquid Metal Cooled Reactors
Status and Future Challenges of CFD for Liquid Metal Cooled Reactors IAEA Fast Reactor Conference 2013 Paris, France 5 March 2013 Ferry Roelofs roelofs@nrg.eu V.R. Gopala K. Van Tichelen X. Cheng E. Merzari
More informationVVER-1000 Reflooding Scenario Simulation with MELCOR Code in Comparison with MELCOR Simulation
VVER-1000 Reflooding Scenario Simulation with MELCOR 1.8.6 Code in Comparison with MELCOR 1.8.3 Simulation Jiří Duspiva Nuclear Research Institute Řež, plc. Nuclear Power and Safety Division Dept. of Reactor
More informationCourse on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors
SMR/1848-T16 Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors T16 - The CSNI Separate Effects Test and Integral Test Facility Matrices for Validation of Best-Estimate
More informationApplication of System Codes to Void Fraction Prediction in Heated Vertical Subchannels
Application of System Codes to Void Fraction Prediction in Heated Vertical Subchannels Taewan Kim Incheon National University, 119 Academy-ro, Yeonsu-gu, Incheon 22012, Republic of Korea. Orcid: 0000-0001-9449-7502
More informationKeywords: PTS, CFD, Thermalhydraulics, safety, Fracture Mechanics.
FRACTURE MECHANICS ANALYSIS FOR VVER1000 REACTOR PRESSURE VESSEL D. Araneo, G. Agresta, F. D Auria GRNSPG -University of Pisa, Pisa, Italy This document deals with a research activity aimed at calculating
More informationEnglish text only NUCLEAR ENERGY AGENCY COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS
Unclassified NEA/CSNI/R(2008)6/VOL2 NEA/CSNI/R(2008)6/VOL2 Unclassified Organisation de Coopération et de Développement Économiques Organisation for Economic Co-operation and Development 26-Nov-2008 English
More informationSteady-State and Transient Neutronic and Thermal-hydraulic Analysis of ETDR using the FAST code system
Steady-State and Transient Neutronic and Thermal-hydraulic Analysis of ETDR using the FAST code system Sandro Pelloni, Evaldas Bubelis and Paul Coddington Laboratory for Reactor Physics and Systems Behaviour,
More informationThe Pennsylvania State University. The Graduate School. College of Engineering
The Pennsylvania State University The Graduate School College of Engineering TRACE/PARCS ASSESSMENT BASED ON PEACH BOTTOM TURBINE TRIP AND LOW FLOW STABILITY TESTS A Thesis in Nuclear Engineering by Boyan
More informationExtension of the Simulation Capabilities of the 1D System Code ATHLET by Coupling with the 3D CFD Software Package ANSYS CFX
Extension of the Simulation Capabilities of the 1D System Code ATHLET by Coupling with the 3D CFD Software Package ANSYS CFX Angel Papukchiev and Georg Lerchl Gesellschaft fuer Anlagen und Reaktorischerheit
More informationTRACE/SIMULATE-3K Analysis of the NEA/OECD Oskarshamn-2 Stability Benchmark
TRACE/SIMULATE-3K Analysis of the NEA/OECD Oskarshamn-2 Stability Benchmark ABSTRACT Abdelhamid Dokhane, Omar Zerkak, and Hakim Ferroukhi Paul Scherrer Institute CH-5232 Villigen, Switzerland Abdelhamid.dokhane@psi.ch;
More informationScaling Analysis as a part of Verification and Validation of Computational Fluid Dynamics and Thermal-Hydraulics software in Nuclear Industry
Scaling Analysis as a part of Verification and Validation of Computational Fluid Dynamics and Thermal-Hydraulics software in Nuclear Industry M. Dzodzo 1), A. Ruggles 2), B. Woods 3), U. Rohatgi 4), N.
More informationNPP Simulators for Education Workshop - Passive PWR Models
NPP Simulators for Education Workshop - Passive PWR Models Wilson Lam (wilson@cti-simulation.com) CTI Simulation International Corp. www.cti-simulation.com Sponsored by IAEA Learning Objectives Understand
More informationReactor radiation skyshine calculations with TRIPOLI-4 code for Baikal-1 experiments
DOI: 10.15669/pnst.4.303 Progress in Nuclear Science and Technology Volume 4 (2014) pp. 303-307 ARTICLE Reactor radiation skyshine calculations with code for Baikal-1 experiments Yi-Kang Lee * Commissariat
More informationBEST ESTIMATE PLUS UNCERTAINTY SAFETY STUDIES AT THE CONCEPTUAL DESIGN PHASE OF THE ASTRID DEMONSTRATOR
BEST ESTIMATE PLUS UNCERTAINTY SAFETY STUDIES AT THE CONCEPTUAL DESIGN PHASE OF THE ASTRID DEMONSTRATOR M. Marquès CEA, DEN, DER F-13108, Saint-Paul-lez-Durance, France Advanced simulation in support to
More informationCFD STUDIES IN THE PREDICTION OF THERMAL STRIPING IN AN LMFBR
CFD STUDIES IN THE PREDICTION OF THERMAL STRIPING IN AN LMFBR K. Velusamy, K. Natesan, P. Selvaraj, P. Chellapandi, S. C. Chetal, T. Sundararajan* and S. Suyambazhahan* Nuclear Engineering Group Indira
More informationCANDU Safety #3 - Nuclear Safety Characteristics Dr. V.G. Snell Director Safety & Licensing
CANDU Safety #3 - Nuclear Safety Characteristics Dr. V.G. Snell Director Safety & Licensing 24/05/01 CANDU Safety - #3 - Nuclear Safety Characteristics.ppt Rev. 0 vgs 1 What Makes A Safe Nuclear Design?
More informationInvestigation of falling control rods in deformed guiding tubes in nuclear reactors using multibody approaches
The th Joint International Conference on Multibody System Dynamics June 8, 8, Lisboa, Portugal Investigation of falling control rods in deformed guiding tubes in nuclear reactors using multibody approaches
More informationTHE USE OF PB-BI EUTECTIC AS THE COOLANT OF AN ACCELERATOR DRIVEN SYSTEM. Joint research Centre of the European Commission Ispra, Italy.
THE USE OF PB-BI EUTECTIC AS THE COOLANT OF AN ACCELERATOR DRIVEN SYSTEM Alberto Peña 1, Fernando Legarda 1, Harmut Wider 2, Johan Karlsson 2 1 University of the Basque Country Nuclear Engineering and
More informationDetailed Modeling of Passive Auto-Catalytic Recombiner Operational Behavior with the Coupled REKODIREKT-CFX Approach
Mitglied der Helmholtz-Gemeinschaft Detailed Modeling of Passive Auto-Catalytic Recombiner Operational Behavior with the Coupled REKODIREKT-CFX Approach S. Kelm, E.-A.Reinecke, *Hans-Josef Allelein *Institute
More informationCFD-NEUTRONIC COUPLED CALCULATION OF A PWR FUEL ASSEMBLY CONSIDERING PRESSURE DROP AND TURBULENCE PRODUCED BY SPACER GRIDS
CFD-NEUTRONIC COUPLED CALCULATION OF A PWR FUEL ASSEMBLY CONSIDERING PRESSURE DROP AND TURBULENCE PRODUCED BY SPACER GRIDS C. Peña-Monferrer a, F. Pellacani d, S. Chiva b,, T. Barrachina c, R. Miró c,
More informationEnglish - Or. English NUCLEAR ENERGY AGENCY NUCLEAR SCIENCE COMMITTEE
Unclassified NEA/NSC/DOC(200)15 NEA/NSC/DOC(200)15 Unclassified Organisation de Coopération et de Développement Economiques Organisation for Economic Co-operation and Development 05-Nov-200 English - Or.
More informationMulti-physics (SP3) requires a software environment that supports code coupling: SALOME. First presentation given by. Nicolas Crouzet Overview on SP5
Multi-physics (SP3) requires a software environment that supports code coupling: SALOME First presentation given by Nicolas Crouzet Overview on SP5 1 NURISP SP3 Multi-Physics CEA, HZDR, GRS, KIT-G, KIT-U,
More informationVVER-1000 MOX Core Computational Benchmark
Nuclear Science ISBN 92-64-01081-5 NEA/NSC/DOC(2005)17 VVER-1000 MOX Core Computational Benchmark Specification and Results Expert Group on Reactor-based Plutonium Disposition Eugeny Gomin, Mikhail Kalugin,
More informationSCWR Research in Korea. Yoon Y. Bae KAERI
SCWR Research in Korea Yoon Y. ae KAERI Organization President Dr. In-Soon Chnag Advanced Reactor Development Dr. Jong-Kyun Park Nuclear Engineering & Research Dr. M. H. Chang Mechanical Engineering &
More informationA Dummy Core for V&V and Education & Training Purposes at TechnicAtome: In and Ex-Core Calculations
A Dummy Core for V&V and Education & Training Purposes at TechnicAtome: In and Ex-Core Calculations S. Nicolas, A. Noguès, L. Manifacier, L. Chabert TechnicAtome, CS 50497, 13593 Aix-en-Provence Cedex
More informationAuthors : Eric CHOJNACKI IRSN/DPAM/SEMIC Jean-Pierre BENOIT IRSN/DSR/ST3C. IRSN : Institut de Radioprotection et de Sûreté Nucléaire
WORKSHOP ON THE EVALUATION OF UNCERTAINTIES IN RELATION TO SEVERE ACCIDENTS AND LEVEL II PROBABILISTIC SAFETY ANALYSIS CADARACHE, FRANCE, 7-9 NOVEMBER 2005 TITLE The use of Monte-Carlo simulation and order
More informationA PWR HOT-ROD MODEL: RELAP5/MOD3.2.2Y AS A SUBCHANNEL CODE I.C. KIRSTEN (1), G.R. KIMBER (2), R. PAGE (3), J.R. JONES (1) ABSTRACT
FR0200515 9 lh International Conference on Nuclear Engineering, ICONE-9 8-12 April 2001, Nice, France A PWR HOT-ROD MODEL: RELAP5/MOD3.2.2Y AS A SUBCHANNEL CODE I.C. KIRSTEN (1), G.R. KIMBER (2), R. PAGE
More informationSteam Condensation Induced Water Hammer Phenomena, a theoretical study
Steam Condensation Induced Water Hammer Phenomena, a theoretical study Imre Ferenc Barna and György Ézsöl Hungarian Academy of Sciences, KFKI Atomic Energy Research Institute(AEKI) Thermohydraulic Laboratory
More informationCode Strategy for Simulating Severe Accident Scenario
Code Strategy for Simulating Severe Accident Scenario C. SUTEAU, F. SERRE, J.-M; RUGGIERI, F. BERTRAND -CEA- March 4-7, 2013, Paris, France christophe.suteau@cea.fr OULINES INTRODUCTION AND CONTEXT REFERENCE
More informationThe Effect of Burnup on Reactivity for VVER-1000 with MOXGD and UGD Fuel Assemblies Using MCNPX Code
Journal of Nuclear and Particle Physics 2016, 6(3): 61-71 DOI: 10.5923/j.jnpp.20160603.03 The Effect of Burnup on Reactivity for VVER-1000 with MOXGD and UGD Fuel Assemblies Using MCNPX Code Heba K. Louis
More informationNumerical Simulation of the MYRRHA reactor: development of the appropriate flow solver Dr. Lilla Koloszár, Philippe Planquart
Numerical Simulation of the MYRRHA reactor: development of the appropriate flow solver Dr. Lilla Koloszár, Philippe Planquart Von Karman Institute, Ch. de Waterloo 72. B-1640, Rhode-St-Genese, Belgium,
More informationA Novel Multi-Scale Domain Overlapping CFD/STH Coupling Methodology for Multi-Dimensional Flows Relevant to Nuclear Applications. Timothy P.
A Novel Multi-Scale Domain Overlapping CFD/STH Coupling Methodology for Multi-Dimensional Flows Relevant to Nuclear Applications by Timothy P. Grunloh A dissertation submitted in partial fulfillment of
More informationRole of the Halden Reactor Project for TVEL nuclear fuels & materials development. B. Volkov IFE/HRP (Norway) Sochi, May 14-16
Role of the Halden Reactor Project for TVEL nuclear fuels & materials development B. Volkov IFE/HRP (Norway) Sochi, May 14-16 1 International OECD Halden Reactor Project foundation and history organisation
More informationFinal Report on the Reactor Pressure Vessel Pressurized-Thermal- Shock International Comparative Assessment Study (RPV PTS ICAS)
Gesellschaft für Anlagenund Reaktorsicherheit (GRS) mbh Final Report on the Reactor Pressure Vessel Pressurized-Thermal- Shock International Comparative Assessment Study (RPV PTS ICAS) GRS - 152 Gesellschaft
More informationANALYSIS OF A REACTOR PRESSURE VESSEL SUBJECTED TO PRESSURIZED THERMAL SHOCKS
M. Niffenegger et al., Int. J. Comp. Meth. and Exp. Meas., Vol. 4, No. 3 (2016) 288 300 ANALYSIS OF A REACTOR PRESSURE VESSEL SUBJECTED TO PRESSURIZED THERMAL SHOCKS M. NIFFENEGGER 1, G. QIAN 1, V.F. GONZALEZ-ALBUIXECH
More informationStudy of boron dilution phenomenon in the core and fuel assemblies of Bushehr VVER-1000 reactor in normal operating conditions
NUKLEONIKA 2010;55(3:323 330 ORIGINAL PAPER Study of boron dilution phenomenon in the core and fuel assemblies of Bushehr VVER-1000 reactor in normal operating conditions Yashar Rahmani, Ehsan Zarifi,
More informationFLOW CHARACTERIZATION WITHIN A SPHERE-PACKED BED USING PIV MEASUREMENT
FLOW CHARACTERIZATION WITHIN A SPHERE-PACKED BED USING PIV MEASUREMENT J. H.ZHANG, L.XIONG, N.X.WANG and W ZHOU Department of reactor physics, Shanghai institute of applied physics, Chinese academy of
More informationCOMPARATIVE ANALYSIS OF WWER-440 REACTOR CORE WITH PARCS/HELIOS AND PARCS/SERPENT CODES
COMPARATIVE ANALYSIS OF WWER-440 REACTOR CORE WITH PARCS/HELIOS AND PARCS/SERPENT CODES S. Bznuni, A. Amirjanyan, N. Baghdasaryan Nuclear and Radiation Safety Center Yerevan, Armenia Email: s.bznuni@nrsc.am
More informationCoolant flow field in a real geometry of PWR downcomer and lower plenum
Available online at www.sciencedirect.com annals of NUCLEAR ENERGY Annals of Nuclear Energy 35 (2008) 610 619 www.elsevier.com/locate/anucene Coolant flow field in a real geometry of PWR downcomer and
More informationAdvanced Simulation: applications for fast reactors
Advanced Simulation: applications for fast reactors Andrew Siegel Argonne National Laboratory 12/18/2009 FR09 1 Two approaches to reactor modeling Yesterday s computers Device modeling "I would rather
More informationSUB-CHAPTER D.1. SUMMARY DESCRIPTION
PAGE : 1 / 12 CHAPTER D. REACTOR AND CORE SUB-CHAPTER D.1. SUMMARY DESCRIPTION Chapter D describes the nuclear, hydraulic and thermal characteristics of the reactor, the proposals made at the present stage
More informationMONTE CARLO POWER ITERATION: ENTROPY AND SPATIAL CORRELATIONS
MONTE CARLO POWER ITERATION: ENTROPY AND SPATIAL CORRELATIONS ANDREA ZOIA, M. NOWAK (CEA/SACLAY) E. DUMONTEIL, A. ONILLON (IRSN) J. MIAO, B. FORGET, K. S. SMITH (MIT) NEA EGAMCT meeting Andrea ZOIA DEN/DANS/DM2S/SERMA/LTSD
More informationApplied Reactor Technology and Nuclear Power Safety, 4A1627; 4 cp. Course Description
Applied Reactor Technology and Nuclear Power Safety, 4A1627; 4 cp Course Objectives Course Description The purpose of the course is to provide a general knowledge on the physical processes that take place
More informationABSTRACT 1 INTRODUCTION
A NODAL SP 3 APPROACH FOR REACTORS WITH HEXAGONAL FUEL ASSEMBLIES S. Duerigen, U. Grundmann, S. Mittag, B. Merk, S. Kliem Forschungszentrum Dresden-Rossendorf e.v. Institute of Safety Research P.O. Box
More informationDEVELOPMENT AND ASSESSMENT OF A METHOD FOR EVALUATING UNCERTAINTY OF INPUT PARAMETERS
DEVELOPMENT AND ASSESSMENT OF A METHOD FOR EVALUATING UNCERTAINTY OF INPUT PARAMETERS A. Kovtonyuk, S. Lutsanych, F. Moretti University of Pisa, San Piero a Grado Nuclear Research Group Via Livornese 1291,
More informationRELAP5 to TRACE model conversion for a Pressurized Water Reactor
RELAP5 to TRACE model conversion for a Pressurized Water Reactor Master s thesis Federico López-Cerón Nieto Department of Physics Division of Subatomic and Plasma Physics Chalmers University of Technology
More informationResearch Article Analysis of Subchannel and Rod Bundle PSBT Experiments with CATHARE 3
Science and Technology of Nuclear Installations Volume 22, Article ID 23426, pages doi:.55/22/23426 Research Article Analysis of Subchannel and Rod Bundle PSBT Experiments with CATHARE 3 M. Valette CEA
More informationComputational and Experimental Benchmarking for Transient Fuel Testing: Neutronics Tasks
Computational and Experimental Benchmarking for Transient Fuel Testing: Neutronics Tasks T. Downar W. Martin University of Michigan C. Lee Argonne National Laboratory November 19, 2015 Objective of Neutronics
More informationFuel Element Burnup Determination in HEU - LEU Mixed TRIGA Research Reactor Core
Fuel Element Burnup Determination in HEU - LEU Mixed TRIGA Research Reactor Core Tomaž Žagar, Matjaž Ravnik Institute "Jožef Stefan", Jamova 39, Ljubljana, Slovenia Tomaz.Zagar@ijs.si Abstract This paper
More informationResults from the Application of Uncertainty Methods in the CSNI Uncertainty Methods Study (UMS)
THICKET 2008 Session VI Paper 16 Results from the Application of Uncertainty Methods in the CSNI Uncertainty Methods Study (UMS) Horst Glaeser Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbh,
More informationIntroduction to Reactivity and Reactor Control
Introduction to Reactivity and Reactor Control Larry Foulke Adjunct Professor Director of Nuclear Education Outreach University of Pittsburgh IAEA Workshop on Desktop Simulation October 2011 Learning Objectives
More informationCFD and Thermal Stress Analysis of Helium-Cooled Divertor Concepts
CFD and Thermal Stress Analysis of Helium-Cooled Divertor Concepts Presented by: X.R. Wang Contributors: R. Raffray and S. Malang University of California, San Diego ARIES-TNS Meeting Georgia Institute
More informationNuclear Data for Innovative Fast Reactors: Impact of Uncertainties and New Requirements
Nuclear Data for Innovative Fast Reactors: Impact of Uncertainties and New Requirements G.Palmiotti 1, M.Salvatores 1, 2, M. Assawaroongruengchot 1 1 Idaho National Laboratory, 2525 Fremont Ave. P.O. Box
More informationCONFORMITY BETWEEN LR0 MOCK UPS AND VVERS NPP PRV ATTENUATION
CONFORMITY BETWEEN LR MOCK UPS AND VVERS NPP PRV ATTENUATION D. Kirilova, K. Ilieva, S. Belousov Institute for Nuclear Research and Nuclear Energy, Bulgaria Email address of main author: desi.kirilova@gmail.com
More informationEUROPEAN COMMISSION 5th EURATOM FRAMEWORK PROGRAMME KEY ACTION : NUCLEAR FISSION FIKS-CT Final Report (D14)
EUROPEAN COMMISSION th EURATOM FRAMEWORK PROGRAMME -00 KEY ACTION : NUCLEAR FISSION FIKS-CT-00-00 Final Report (D) VALIDATION OF COUPLED NEUTRONIC / THERMAL-HYDRAULIC CODES FOR VVER REACTORS S. Mittag,
More informationAnalysis and interpretation of the LIVE-L6 experiment
Analysis and interpretation of the LIVE-L6 experiment A. Palagin, A. Miassoedov, X. Gaus-Liu (KIT), M. Buck (IKE), C.T. Tran, P. Kudinov (KTH), L. Carenini (IRSN), C. Koellein, W. Luther (GRS) V. Chudanov
More informationReactivity Coefficients
Reactivity Coefficients B. Rouben McMaster University Course EP 4D03/6D03 Nuclear Reactor Analysis (Reactor Physics) 2015 Sept.-Dec. 2015 September 1 Reactivity Changes In studying kinetics, we have seen
More informationULOF Accident Analysis for 300 MWt Pb-Bi Coolled MOX Fuelled SPINNOR Reactor
ULOF Accident Analysis for 300 MWt Pb-Bi Coolled MOX Fuelled SPINNOR Reactor Ade afar Abdullah Electrical Engineering Department, Faculty of Technology and Vocational Education Indonesia University of
More informationAn Integrated Approach for Characterization of Uncertainty in Complex Best Estimate Safety Assessment
An Integrated Approach for Characterization of Uncertainty in Complex Best Estimate Safety Assessment Presented By Mohammad Modarres Professor of Nuclear Engineering Department of Mechanical Engineering
More informationBOILING WATER REACTOR TURBINE TRIP (TT) BENCHMARK
Nuclear Science ISBN 92-64-02331-3 NEA/NSC/DOC(2006)23 NEA NUCLEAR SCIENCE COMMITTEE NEA COMMITTEE ON SAFETY OF NUCLEAR INSTALLATIONS BOILING WATER REACTOR TURBINE TRIP (TT) BENCHMARK Volume III: Summary
More informationDemonstration of Full PWR Core Coupled Monte Carlo Neutron Transport and Thermal-Hydraulic Simulations Using Serpent 2/ SUBCHANFLOW
Demonstration of Full PWR Core Coupled Monte Carlo Neutron Transport and Thermal-Hydraulic Simulations Using Serpent 2/ SUBCHANFLOW M. Daeubler Institute for Neutron Physics and Reactor Technology (INR)
More informationNeutronic analysis of nanofluids as a coolant in the Bushehr VVER-1000 reactor
NUKLEONIKA 2012;57(3):375 381 ORIGINAL PAPER Neutronic analysis of nanofluids as a coolant in the Bushehr VVER-1000 reactor Ehsan Zarifi, Gholamreza Jahanfarnia, Farzad Veisy Abstract. The main goal of
More informationVHTR Thermal Fluids: Issues and Phenomena
VHTR Thermal Fluids: Issues and Phenomena www.inl.gov Technical workshop at PHYSOR 2012: Advanced Reactor Concepts April 15, 2012 Knoxville, TN Gerhard Strydom Idaho National Laboratory (INL) Overview
More information