Activities for Safety Assessment of Fast Spectrum Systems
|
|
- Catherine Austin
- 6 years ago
- Views:
Transcription
1 Activities for Safety Assessment of Fast Spectrum Systems A. Seubert Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh Forschungszentrum, D Garching, Germany 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors IAEA HQ, Wien, June 2015 Work supported by German Federal Ministry of Economic Affairs and Energy due to a resolution of the German Bundestag.
2 Content R&D for the safety assessment of liquid metal cooled fast spectrum systems Thermal hydraulics of liquid metals (ATHLET) Simulation of time-dependent distributed neutron sources (PARCS) Modeling of radial and axial core thermal expansion (FEM/PARCS) Simulation of spallation neutron sources (MCNPX) MAXSIMA (MYRRHA) ESNII+ (ASTRID) neutronic and thermal-hydraulic simulation 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
3 Generic ASTRID Core Design Radial core layout: Axial core layout: 301cm 248cm 210cm 166cm 146cm 121cm 91cm Flat-to-flat assembly pitch at nominal operation: 17,611 cm 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
4 ASTRID Neutronic Modeling with HELIOS HELIOS energy group library, unadjusted 112 energy group library 8 energy group structure: Energy group index Lower limit (ev) Energy group index Lower limit (ev) E E E E E E E E-4 For selected comparative test calculations Serpent v JEFF-3.1 nuclear data 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
5 HELIOS Models (examples) Fissile/fertile assemblies: CSD/DSD assemblies: 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
6 p''' in W/cm^3 Evaluation of Core Safety Parameters (HELIOS/PARCS/FEM) Reactivities (pcm) for 9 voiding scenarios 350 Radially integrated axial power density Institute A, Code 1 Institute A, Code 2 Institute B GRS HELIOS, FEM Institute B Institute C, Code 1 Institute C, Code PARCS v32m16co 50 FEM-Diff-3D Axial elevation (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
7 Evaluation of Core Safety Parameters Radial power distributions 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
8 ASTRID Thermal-Hydraulic Modeling with ATHLET 3.0b Three single SAs Inner hot SA CORE-IH Inner average SA CORE-I1 Outer SA CORE-O1 Mass flow controlled by common fill FILL-IN entering DIAGRID Mass flow distribution in SAs according to individual flow loss coefficients Pressure controlled by time dependent volume BOUND SAs Geometric data, mass flows, diagrid porosity etc. according to ESNII+ specifications One representative heated rod group for each SA pipe One unheated rod group for each SA pipe in lower section (-IH, -I1, -O1) (only structure of cladding) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
9 ASTRID Steady State TH Simulation with ATHLET 3.0b 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
10 Diagrid Thermal Expansion Modeling In an SFR, increasing coolant inlet temperature causes thermal expansion of the diagrid plate, i.e. the assembly core support structure. V non-deformed state state with expanded diagrid Enlarged spacing between adjacent subassembly wrappers spacing is filled by coolant. Associated reactivity changes may present a large (negative) contribution to the total reactivity feedback. Assembly pitch changes affect the radial spatial meshing of the core simulator. Aim: Treat pitch changes with the core simulator s fixed radial meshing. 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
11 Diagrid Thermal Expansion Modeling Mapping the meshing of the radially deformed core (x, y, z) to the meshing of the non-deformed core (,, z): (K. Azekura, T. Hayase, J. Nucl. Sci. Techn. 26 (1989) 374) Diffusion equation of the non-deformed core: Approximate diffusion equation of the deformed core: D g r 2 x 2 g 2 D g r 2 y 2 g 2 + σ g r r g r = q g r q g r = r g r s σ gg D g r g r + σ g r r g r = q g r Change of cross sections to account for enlarged inter-assembly gap g g + g k g g Different multiplication factor σ f g r g r 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
12 Diagrid Thermal Expansion Modeling p(t) Thermal expansion correlation for diagrid Subassembly pitch thermal expansion: p T = p 20 C 1 + ε SS316 T Thermal expansion of SS316 ε SS316 T = a T 20 C + b T 20 C 2 +c T 20 C 3 with coefficients a, b, c given within ESNII+ project WP6 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
13 Multiplication factor Results for Diagrid Expansion of the ASTRID Generic Design Implementation: Fortran module Made available first to an FEM few-group diffusion code Application to ASTRID generic design within ESNII+ project Ref.: A. SEUBERT ET AL., ANNUAL MEETING ON NUCLEAR TECHNOLOGY, BERLIN, GERMANY, 2015 Could be also implemented in PARCS in future 1,005 1,004 1,003 1,002 1,001 1,000 Exact 0,999 Approximate 0, Diagrid temperature (K) Diagid temperature 400 K (nom.) Flat-to-flat pitch (cm) exact k eff Reactivity change ρ (pcm) approx k eff Deviation from exact ρ (pcm) K (+0.36%) K (+0.74%) K (+1.1%) th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
14 p'''approx./p'''exact - 1 p''' in W/cm^3 Results for Diagrid Expansion of the ASTRID Generic Design Axial power density profiles in individual subassemblies Axial power density profiles Outer zone close to reflector Inner zone close to center Axial elevation (cm) 0,15% Axial power density deviations 0,10% 0,05% 0,00% -0,05% -0,10% -0,15% Outer zone close to reflector Inner zone close to center -0,20% Axial elevation (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
15 p'''approx./p'''exact - 1 p''' in W/cm^3 Results for Diagrid Expansion of the ASTRID Generic Design Axial power density profiles in individual subassemblies compared with nominal diagrid temperature Axial power density profiles Outer zone close to reflector Inner zone close to center Outer zone at Tdiagrid nominal Inner zone at Tdiagrid nominal Axial elevation (cm) 14,0% Axial power density deviations 12,0% 10,0% 8,0% 6,0% 4,0% 2,0% Outer zone close to reflector Inner zone close to center Outer zone at Tdiagrid nominal Inner zone at Tdiagrid nominal 0,0% -2,0% Axial elevation (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
16 Axial fuel/cladding expansion modeling Problem: Keep axial nodalization of the core simulator unchanged Assembly pitch unchanged, simultaneous radial pin expansion treated by cross section library parameter Not expanded (nominal state) z 4 axially expanded Expanded state mapped on nodalization of nominal state z 4 z 4 z 4 σ 4 σ 4 z 3 z 3 z 3 z 3 σ 3 σ 3 σ 3 = f σ 2, σ 3 σ 2 z 2 z 2 σ 2 z 2 z 2 Interpolation using either: Linear, quadratic, cubic neutron flux FMFD diffusion neutron flux solution z 1 z' 1 z 1 z' 1 σ 1 σ 1 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
17 Axial fuel/cladding expansion modeling Test calculations of axial expansion modeling for ASTRID Nuclear material properties fixed consistent changes in nuclide density and radial pin dimensions will be treated by a few-group nuclear cross section library parameter Linear thermal expansion according to ESNII+ specifications 1,0075 1,007 1,0065 Multiplication factor exact T/ C k r r/t 75 1, , , , , , ,311 1,006 1,0055 1,005 1,0045 exact Linear interpolation Quadr. Interpol. (F'li as calc.) Quadr. Interpol. (F'li = F're = 0) FMFD diffusion solution (model 900 C) T ( C) FMFD diffusion solution (model 900 C) delta_z = 1,0 cm, nz_min = 2, nz_max = 4 k r w.r.t. exact 1, , , , th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
18 MYRRHA Critical Core Modeling with SERPENT and PARCS Reflector Assembly Withdrawn Safety Rod Inserted Safety Rod Fuel Assembly (MOX) 26 different materials Hexagonal geometry Complex geometry for safety and control rod Dummy Assembly (Lead) Withdrawn Control Rod Inserted Control Rod 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
19 Reference calculation: Monte-Carlo calculation Preparation of XS for PARCS calculations JEFF31 library (continuous energy) Exact 3D geometry (No approximation) SERPENT: 3D modelling Macro XS k-eff, Power distribution Number of source neutrons per cycle: Number of active cycles run: 2000 Number of inactive cycles run: 200 Calculated with the 3D neutron flux Collapsed into 8 energy groups for every type of material 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
20 2 methods to generate few-group cross sections for PARCS Serpent with JEFF31 library (continuous energy) Method 1: 3D whole core modelling Method 2: 2D single assembly modelling Realistic 3D geometry 3D neutron flux Macro XS collapsed into 8 groups Macro XS collapsed into 8 groups infinite lattice periodic boundary condition 8 energy groups structure Lower energy limits (ev) 2.23E E E E E E E E-04 PARCS k-eff, Power distribution PARCS k-eff, Power distribution 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
21 K-eff Multiplication factor results (preliminary) 1,05 1,013381, , , , ,96893 Serpent 3D PARCS Method 1 PARCS Method 2 0,95 0, , , , , , ,9 0,85 0,8 Control Rods position Withdrawn Inserted Withdrawn Inserted Safety Rods position Withdrawn Withdrawn Inserted Inserted 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
22 Average power per ring (in % of the total power) Radial power distribution comparison (preliminary) 3 Radial power distribution 2,5 2 PARCS 1,5 SERPENT 1 0,5 Relative difference between PARCS (Method 1) and Serpent results for each FA (in %) Distance from center [ring number] 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
23 Extension of PARCS for External Neutron Source Simulation 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June Objective: Simulation of ADS Experience from implementation in TORT-TD and validation by YALINA-Thermal Transport equation: Example for a single localized source pulse: l l l d gl g g fg g g g gg g g tot g g t r c t r r t r r d t r q t r r t v,, ) (1,,,,,,, ˆ 1 ' ' ' ' 4 ' normalized to steady state power Time in sec Power Level
24 Neutron source strength (a.u.) Neutron source strength (a.u.) Neutron source strength (a.u.) Simulation of Spallation Neutron Sources MCNPX model of the MYRRHA spallation target 0,08 0,07 0,06 0,05 0,04 0,03 0,02 0,01 0,00 0,001 0,01 0, Energy (ev) 9,0E-02 8,0E-02 7,0E-02 6,0E-02 5,0E-02 4,0E-02 3,0E-02 6,0E-05 5,0E-05 4,0E-05 3,0E-05 2,0E-05 2,0E-02 1,0E-05 1,0E-02 Energy group 8 Energy group 7 0,0E+00 0,0E Axial position (cm) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
25 Summary Activities at GRS for Safety Assessment of Fast Spectrum Systems Systems: Sodium cooled fast reactors (e.g. ASTRID) Liquid metal cooled fast spectrum systems, including source driven sub-critical systems (e.g. MYRRHA) Codes: ATHLET, PARCS, HELIOS, SERPENT, MCNPX Modeling extensions: Thermal hydraulics of liquid metals (ATHLET) Simulation of time-dependent distributed neutron sources (PARCS) Modeling of radial and axial core thermal expansion Simulation of spallation neutron sources (MCNPX) Few-group nuclear cross-section generation (HELIOS, SERPENT) Whole core neutronics modeling (PARCS, SERPENT) 5th Joint IAEA-GIF Technical Meeting/Workshop on Safety of Sodium-Cooled Fast Reactors, IAEA HQ, Wien, June
INVESTIGATION OF VOID REACTIVITY BEHAVIOUR IN RBMK REACTORS
INVESTIGATION OF VOID REACTIVITY BEHAVIOUR IN RBMK REACTORS M. Clemente a, S. Langenbuch a, P. Kusnetzov b, I. Stenbock b a) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS)mbH, Garching, E-mail:
More informationHELIOS-2: Benchmarking Against Hexagonal Lattices
HELIOS-2: Benchmarking Against Hexagonal Lattices Teodosi Simeonov a and Charles Wemple b a Studsvik Scandpower, GmbH., Hamburg,Germany b Studsvik Scandpower, Inc., Idaho Falls, ID, USA ABSTRACT The critical
More informationTask 1 Progress: Analysis of TREAT Minimum Critical and M8CAL Cores with SERPENT and SERPENT/PARCS
Task 1 Progress: Analysis of TREAT Minimum Critical and M8CAL Cores with SERPENT and SERPENT/PARCS Volkan Seker, Matt Neuman, Nicholas Kucinski, Hunter Smith, Tom Downar University of Michigan May 24,
More informationNatural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactor
Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactor Ade Gafar Abdullah 1,2,*, Zaki Su ud 2, Rizal Kurniadi 2, Neny Kurniasih 2, Yanti Yulianti 2,3 1 Electrical
More informationDesign and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Core Configurations
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 32[9], pp. 834-845 (September 1995). Design and Safety Aspect of Lead and Lead-Bismuth Cooled Long-Life Small Safe Fast Reactors for Various Core Configurations
More informationLFR core design. for prevention & mitigation of severe accidents
LFR core design for prevention & mitigation of severe accidents Giacomo Grasso UTFISSM Technical Unit for Reactor Safety and Fuel Cycle Methods Coordinator of Core Design Work Package in the EURATOM FP7
More informationFinal Results: PWR MOX/UO 2 Control Rod Eject Benchmark
Final Results: PWR MOX/UO 2 Control Rod Eject Benchmark T. Kozlowski T. J. Downar Purdue University January 25, 2006 This work has been sponsored by the U.S. Nuclear Regulatory Commission. The views expressed
More informationKAPROS-E: A Modular Program System for Nuclear Reactor Analysis, Status and Results of Selected Applications.
KAPROS-E: A Modular Program System for Nuclear Reactor Analysis, Status and Results of Selected Applications. C.H.M. Broeders, R. Dagan, V. Sanchez, A. Travleev Forschungszentrum Karlsruhe Institut für
More informationAnalytical support to experiment QUENCH-17 and first post-test calculations with ATHLET-CD
Analytical support to experiment QUENCH-17 and first post-test calculations with ATHLET-CD C. Bals, T. Hollands, H. Austregesilo Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), Germany Content Short
More informationWhy CMS5? Arthur S DiGiovine 2012 International Users Group Meeting Charlotte, NC, USA May 2-3, 2012
Why CMS5? Arthur S DiGiovine 2012 Charlotte, NC, USA What Problem are We trying to Solve? Economics of Core Design Fuel enrichment Number of assemblies Cycle Length Economics of Core Operation Plant availability
More informationA Parametric Study on Core Performance of Sodium Fast Reactors Using SERPENT Code RUBÉN GARCÍA MORENO
A Parametric Study on Core Performance of Sodium Fast Reactors Using SERPENT Code RUBÉN GARCÍA MORENO Master of Science Thesis Division of Nuclear Safety Royal Institute of Technology Stockholm, Sweden
More informationCalculation of Pellet Radial Power Distributions with Monte Carlo and Deterministic Codes
Progress in NUCLEAR SCIENCE and TECHNOLOGY, Vol. 2, pp.301-305 (2011) TECHNICAL MATERIAL Calculation of Pellet Radial Power Distributions with Monte Carlo and Deterministic Codes Motomu SUZUKI *, Toru
More informationBenchmark Specification for HTGR Fuel Element Depletion. Mark D. DeHart Nuclear Science and Technology Division Oak Ridge National Laboratory
I. Introduction Benchmark Specification for HTGR Fuel Element Depletion Mark D. DeHart Nuclear Science and Technology Division Oak Ridge National Laboratory Anthony P. Ulses Office of Research U.S. Nuclear
More informationSodium versus Lead-Bismuth Coolants for the ENHS (Encapsulated Nuclear Heat Source) Reactor
Proceedings of the Korean Nuclear Society Autumn Meeting Yongpyong, Korea, October 2002 Sodium versus Lead-Bismuth Coolants for the ENHS (Encapsulated Nuclear Heat Source) Reactor Ser Gi Hong a, Ehud Greenspan
More informationEnglish - Or. English NUCLEAR ENERGY AGENCY NUCLEAR SCIENCE COMMITTEE. Benchmark Specification for HTGR Fuel Element Depletion
Unclassified NEA/NSC/DOC(2009)13 NEA/NSC/DOC(2009)13 Unclassified Organisation de Coopération et de Développement Économiques Organisation for Economic Co-operation and Development 16-Jun-2009 English
More informationSafety Analysis of the MIT Nuclear Reactor for Conversion to LEU Fuel
Global Threat Reduction Initiative Safety Analysis of the MIT Nuclear Reactor for Conversion to LEU Fuel Erik H. Wilson, Floyd E. Dunn Argonne National Laboratory Thomas H. Newton Jr., Lin-wen Hu MIT Nuclear
More informationEvolution of Nuclear Energy Systems
ALLEGRO Project 2 Evolution of Nuclear Energy Systems 3 General objectives Gas cooled fast reactors (GFR) represent one of the three European candidate fast reactor types. Allegro Gas Fast Reactor (GFR)
More informationLEU Conversion of the University of Wisconsin Nuclear Reactor
LEU Conversion of the University of Wisconsin Nuclear Reactor Paul Wilson U. Wisconsin-Madison Russian-American Symposium on the Conversion of Research Reactors to Low Enriched Uranium Fuel 8-10 June 2011
More informationKIPT ACCELERATOR DRIVEN SYSTEM DESIGN AND PERFORMANCE
KIPT ACCELERATOR DRIVEN SYSTEM DESIGN AND PERFORMANCE Yousry Gohar 1, Igor Bolshinsky 2, Ivan Karnaukhov 3 1 Argonne National Laboratory, USA 2 Idaho National Laboratory, USA 3 Kharkov Institute of Physics
More informationCFD Analysis of Decay Heat Removal Scenarios of the Lead cooled ELSY reactor. Michael Böttcher
CFD Analysis of Decay Heat Removal Scenarios of the Lead cooled ELSY reactor Michael Böttcher Institut für Neutronenphysik und Reaktortechnik (INR), Karlsruher Institut für Technologie, KIT Abstract In
More informationSafety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors
Journal of Physics: Conference Series PAPER OPEN ACCESS Safety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors To cite this article: Zaki Su'ud et al 2017 J. Phys.: Conf. Ser. 799 012013 View
More informationApplication of CANDLE Burnup to Block-Type High Temperature Gas Cooled Reactor for Incinerating Weapon Grade Plutonium
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1079 Application of CANDLE Burnup to Block-Type High Temperature Gas Cooled Reactor for Incinerating Weapon Grade Plutonium Yasunori Ohoka * and Hiroshi
More information1. INTRODUCTION. Corresponding author. Received December 18, 2008 Accepted for Publication April 9, 2009
DEVELOPMENT OF A SIMPLIFIED MODEL FOR ANALYZING THE PERFORMANCE OF KALIMER-600 COUPLED WITH A SUPERCRITICAL CARBON DIOXIDE BRAYTON ENERGY CONVERSION CYCLE SEUNG-HWAN SEONG *, TAE-HO LEE and SEONG-O KIM
More informationEffect of U-9Mo/Al Fuel Densities on Neutronic and Steady State Thermal Hydraulic Parameters of MTR Type Research Reactor
International Conference on Nuclear Energy Technologies and Sciences (2015), Volume 2016 Conference Paper Effect of U-9Mo/Al Fuel Densities on Neutronic and Steady State Thermal Hydraulic Parameters of
More informationInvestigation of Surface Vortex Formation at Pump Intakes in PWR
Investigation of Surface Vortex Formation at Pump Intakes in PWR P. Pandazis 1, A. Schaffrath 1, F. Blömeling 2 1 Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh, Munich 2 TÜV NORD SysTec GmbH
More informationABSTRACT. 1. Introduction
Improvements in the Determination of Reactivity Coefficients of PARR-1 Reactor R. Khan 1*, Muhammad Rizwan Ali 1, F. Qayyum 1, T. Stummer 2 1. DNE, Pakistan Institute of Engineering and Applied Sciences
More informationActivities of Helmholtz Association research centers on fast reactors
Activities of Helmholtz Association research centers on fast reactors A. Rineiski, KIT, Karlsruhe, Germany 50 th IAEA TWG-FR Meeting, Vienna, May, 2017 INSTITUTE FOR NUCLEAR AND ENERGY TECHNOLOGIES KIT
More informationDEVELOPMENT AND VERIFICATION OF DYNAMICS CODE FOR MOLTEN SALT REACTORS
Proceedings of ICONE 2: 2 th International Conference on Nuclear Engineering April 25-29, 24, Arlington, Virginia, USA ICONE 2-493 DEVELOPMENT AND VERIFICATION OF DYNAMICS CODE FOR MOLTEN SALT REACTORS
More informationPrimary - Core Performance Branch (CPB) Reactor Systems Branch (SRXB) 1
U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION NUREG-0800 (Formerly NUREG-75/087) 4.3 NUCLEAR DESIGN REVIEW RESPONSIBILITIES Primary - Core Performance Branch
More informationAN INVESTIGATION OF TRU RECYCLING WITH VARIOUS NEUTRON SPECTRUMS
AN INVESTIGATION OF TRU RECYCLING WITH VARIOUS NEUTRON SPECTRUMS Yong-Nam Kim, Hong-Chul Kim, Chi-Young Han and Jong-Kyung Kim Hanyang University, South Korea Won-Seok Park Korea Atomic Energy Research
More informationBenchmark for Neutronic
Nuclear Science NEA/NSC/R(2015)9 February 2016 www.oecd-nea.org Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes Unclassified NEA/NSC/R(2015)9
More informationKIPT ADS Facility. Yousry Gohar 1, Igor Bolshinsky 2, Ivan Karnaukhov 3. Argonne National Laboratory, USA 2. Idaho National Laboratory, USA 3
KIPT ADS Facility Yousry Gohar 1, Igor Bolshinsky 2, Ivan Karnaukhov 3 1 Argonne National Laboratory, USA 2 Idaho National Laboratory, USA 3 Kharkov Institute of Physics & Technology, Ukraine EuCARD 2
More informationJournal of American Science 2014;10(2) Burn-up credit in criticality safety of PWR spent fuel.
Burn-up credit in criticality safety of PWR spent fuel Rowayda F. Mahmoud 1, Mohamed K.Shaat 2, M. E. Nagy 3, S. A. Agamy 3 and Adel A. Abdelrahman 1 1 Metallurgy Department, Nuclear Research Center, Atomic
More informationA Research Reactor Simulator for Operators Training and Teaching. Abstract
Organized and hosted by the Canadian Nuclear Society. Vancouver, BC, Canada. 2006 September 10-14 A Research Reactor Simulator for Operators Training and Teaching Ricardo Pinto de Carvalho and José Rubens
More informationSelf-Sustaining Thorium-Fueled BWR
Self-Sustaining Thorium-Fueled BWR Jeffrey E. Seifried, Guanheng Zhang, Christopher R. Varela, Phillip M. Gorman, Ehud Greenspan, Jasmina L. Vujic University of California, Berkeley, Department of Nuclear
More informationCOMPARISON BETWEEN EXPERIMENTAL RESULTS AND CALCULATIONS DURING THE COMMISSIONING OF THE ETRR2
COMPAISON BETWEEN EXPEIMENTAL ESULTS AND CALCULATIONS DUING THE COMMISSIONING OF THE ET2 Eduardo Villarino 1, Carlos Lecot 1, Ashraf Enany 2 and Gustavo Gennuso 3. This work presents the comparison between
More informationWir schaffen Wissen heute für morgen
Paul Scherrer Institut Wir schaffen Wissen heute für morgen Spallation Target Developments B. Riemer (ORNL), H. Takada (JAEA), N. Takashi (JAEA) and M. Wohlmuther (PSI) Thorium Energy Conference 2013 (ThEC13),
More informationModelling an Unprotected Loss-of-Flow Accident in Research Reactors using the Eureka-2/Rr Code
Journal of Physical Science, Vol. 26(2), 73 87, 2015 Modelling an Unprotected Loss-of-Flow Accident in Research Reactors using the Eureka-2/Rr Code Badrun Nahar Hamid, 1* Md. Altaf Hossen, 1 Sheikh Md.
More informationMonte Carlo analysis of the battery-type high temperature gas cooled reactor
archives of thermodynamics Vol. 38(2017), No. 4, 209 227 DOI: 10.1515/aoter-2017-0032 Monte Carlo analysis of the battery-type high temperature gas cooled reactor MARCIN GRODZKI PIOTR DARNOWSKI GRZEGORZ
More informationSystem Analysis of Pb-Bi Cooled Fast Reactor PEACER
OE-INES-1 International Symposium on Innovative Nuclear Energy Systems for Sustainable Development of the World Tokyo, Japan, October 31 - November 4, 2004 System Analysis of Pb-Bi ooled Fast Reactor PEAER
More informationREACTIVITY EFFECTS OF TEMPERATURE CHANGES THIS SECTION IS NOT REQUIRED FOR MECHANICAL MAINTAINERS
REACTIVITY EFFECTS OF TEMPERATURE CHANGES THIS SECTION IS NOT REQUIRED FOR MECHANICAL MAINTAINERS OBJECTIVES At the conclusion of this lesson the trainee will be able to: 1. Define: a) temperature coefficient
More informationRadiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I 2 S-LWR)
EPJ Web of Conferences 106, 03004 (2016) DOI: 10.1051/epjconf/201610603004 C Owned by the authors, published by EDP Sciences, 2016 Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral
More information3-D fuel shuffling for reduced peak burnup and increased uranium utilization of breed-and-burn reactors
3-D fuel shuffling for reduced peak burnup and increased uranium utilization of breed-and-burn reactors Jason Hou, a Staffan Qvist, b and Ehud Greenspan a a Department of Nuclear Engineering, University
More informationTHE FUEL BURN UP DETERMINATION METHODOLOGY AND INDICATIVE DEPLETION CALCULATIONS IN THE GREEK RESEARCH REACTOR M. VARVAYANNI
THE FUEL BURN UP DETERMINATION METHODOLOGY AND INDICATIVE DEPLETION CALCULATIONS IN THE GREEK RESEARCH REACTOR M. VARVAYANNI Nuclear Research Reactor Laboratory Institute of Nuclear Technology & Radiation
More informationTechnical Safety Organisation (TSO) safety assessments own research and development Post-Fukushima Research simulation codes
Victor Teschendorff Dipl.-Ing. Mechanical Engineering (Technical University of Aachen, Germany) 1973-2010: Employee of GRS in Garching near Munich Last Position: Head of Reactor Safety Research Division
More informationSpecification for Phase VII Benchmark
Specification for Phase VII Benchmark UO 2 Fuel: Study of spent fuel compositions for long-term disposal John C. Wagner and Georgeta Radulescu (ORNL, USA) November, 2008 1. Introduction The concept of
More informationWorkgroup Thermohydraulics. The thermohydraulic laboratory
Faculty of Mechanical Science and Engineering Institute of Power Engineering Professorship of Nuclear Energy and Hydrogen Technology Workgroup Thermohydraulics The thermohydraulic laboratory Dr.-Ing. Christoph
More informationFENDL NEUTRONICS BENCHMARK: NEUTRON MULTIPLICATION MEASUREMENTS IN BERYLLIUM, BERYLLIUM OXIDE AND LEAD WITH 14-MEV NEUTRONS
fflfornohonoi Atomic cn^oy AQGHCY INDCfNDSl-314 Distrib.: G+F I N DC INTERNATIONAL NUCLEAR DATA COMMITTEE FENDL NEUTRONICS BENCHMARK: NEUTRON MULTIPLICATION MEASUREMENTS IN BERYLLIUM, BERYLLIUM OIDE AND
More informationRELAP 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-07
Fifth International Seminar on Horizontal Steam Generators 22 March 21, Lappeenranta, Finland. 5 ANALYSIS OF PACTEL PRIMARY-TO-SECONDARY LEAKAGE EXPERIMENT PSL-7 József Bánáti Lappeenranta University of
More informationThe GUINEVERE-project at VENUS
The GUINEVERE-project at VENUS P. Baeten, H. Aït Abderrahim, G. Vittiglio, B. Verboomen, G. Bergmans, F. Vermeersch On behalf of the ECATS community 1 Structure of IP-EUROTRANS IP Co-ordinator J.U. Knebel,
More informationNUCLEAR FUEL AND REACTOR
NUCLEAR FUEL AND REACTOR 1 Introduction 3 2 Scope of application 3 3 Requirements for the reactor and reactivity control systems 4 3.1 Structural compatibility of reactor and nuclear fuel 4 3.2 Reactivity
More informationBatch Annealing Model for Cold Rolled Coils and Its Application
China Steel Technical Report, No. 28, pp.13-20, (2015) Chun-Jen Fang and Li-Wen Wu 13 Batch Annealing Model for Cold Rolled Coils and Its Application CHUN-JEN FANG and LI-WEN WU New Materials Research
More informationGERMAN APPROACH TO ESTIMATE POTENTIAL RADIOLOGICAL CONSEQUENCES FOLLOWING A SABOTAGE ATTACK AGAINST NUCLEAR INTERIM STORAGES
GERMAN APPROACH TO ESTIMATE POTENTIAL RADIOLOGICAL CONSEQUENCES FOLLOWING A SABOTAGE ATTACK AGAINST NUCLEAR INTERIM STORAGES Gunter Pretzsch* Ralph Maier** *Gesellschaft für Anlagen und Reaktorsicherheit
More informationWM2014 Conference, March 2 6, 2014, Phoenix, Arizona, USA
Integrity Study of Spent PWR Fuel under Dry Storage Conditions 14236 Jongwon Choi *, Young-Chul Choi *, Dong-Hak Kook * * Korea Atomic Energy Research Institute ABSTRACT Technical issues related to long-term
More informationSelf-Sustaining Thorium Boiling Water Reactors
Sustainability 2012, 4, 2472-2497; doi:10.3390/su4102472 Article OPEN ACCESS sustainability ISSN 2071-1050 www.mdpi.com/journal/sustainability Self-Sustaining Thorium Boiling Water Reactors Francesco Ganda
More informationSpecification for Phase IID Benchmark. A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA)
Specification for Phase IID Benchmark PWR-UO 2 Assembly: Study of control rods effects on spent fuel composition A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA) 1. Introduction
More informationThermal Hydraulic Simulations of the Angra 2 PWR
Thermal Hydraulic Simulations of the Angra 2 PWR Javier González-Mantecón, Antonella Lombardi Costa, Maria Auxiliadora Fortini Veloso, Claubia Pereira, Patrícia Amélia de Lima Reis, Adolfo Romero Hamers,
More informationExperiments Carried-out, in Progress and Planned at the HTR-10 Reactor
Experiments Carried-out, in Progress and Planned at the HTR-10 Reactor Yuliang SUN Institute of Nuclear and New Energy Technology, Tsinghua University Beijing 100084, PR China 1 st Workshop on PBMR Coupled
More informationCFD on Small Flow Injection of Advanced Accumulator in APWR
54 CFD on Small Flow Injection of Advanced Accumulator in APWR TOMOSHIGE TAKATA TAKAFUMI OGINO TAKASHI ISHIBASHI TADASHI SHIRAISHI The advanced accumulator in the advanced pressurized-water reactor is
More information40-Ton Articulated Truck Cooling System Modelling Using STAR-CCM+
40-Ton Articulated Truck Cooling System Modelling Using STAR-CCM+ Gary Yu, Martin Timmins and Mario Ciaffarafa DENSO Marston Ltd, Bradford, BD17 7JR, UK DENSO Marston Founded in 1904 Acquired by DENSO
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
NUCLEAR DESIGN AND SAFETY ANALYSIS OF ACCIDENT TOLERANT FUEL CANDIDATES IN OPR1000 Wang-Kee In 1, Ser-Gi Hong 2, Tae-Wan Kim 3, Tae-Hyun Chun 1, Chang-Hwan Shin 1 1 Korea Atomic Energy Research Institute:
More informationINCREASINGTHENEUTRONFLUXATTHEBEAMTUBE POSITIONS OF THE FRG-1. P. Schreiner, W. Krull and W. Feltes*
XA04C1707 INCREASINGTHENEUTRONFLUXATTHEBEAMTUBE POSITIONS OF THE FRG-1 P. Schreiner, W. Krull and W. Feltes* GKSS-Forschungszentrum Geesthacht GmbH Max-Planck-StraBe D21502 Geesthacht * Siemens AG, KWU
More informationTwo Methods for Converting Iran s IR-40 Reactor to Use Low-Enriched-Uranium Fuel to Improve Proliferation Resistance After Startup
Two Methods for Converting Iran s IR-40 Reactor to Use Low-Enriched-Uranium Fuel to Improve Proliferation Resistance After Startup R.S. Kemp March 2014 Abstract This article demonstrates the feasibility
More informationSimulation of Fatigue relevant thermal Loads of Components in Piping Networks
Simulation of Fatigue relevant thermal Loads of Components in Piping Networks Dr. Jan Leilich Dr. Gerhard Schlicht NSSS Primary Circuit and Ageing Management Dresden, Oct. 15 th, 2014 Thermal Loads in
More informationWWER-440 CONTROL ASSEMBLY LOCAL POWER PEAKING INVESTIGATION ON LR-0 REACTOR
International Conference Nuclear Energy for New Europe 2002 Kranjska Gora, Slovenia, September 9-12, 2002 www.drustvo-js.si/gora2002 WWER-440 CONTROL ASSEMBLY LOCAL POWER PEAKING INVESTIGATION ON LR-0
More informationGENERAL CONTENTS SECTION I - NUCLEAR ISLAND COMPONENTS
- June 2013 Addendum GENERAL CONTENTS SECTION I - NUCLEAR ISLAND COMPONENTS SUBSECTION "A" : GENERAL RULES SUBSECTION "B" : CLASS 1 COMPONENTS SUBSECTION "C" : CLASS 2 COMPONENTS SUBSECTION "D" : CLASS
More informationAn Investigation of Oxide Layer Impact on Heat Transfer in a Fuel Element of the MARIA Reactor
Open Access Journal Journal of Power Technologies 93 (4) (2013) 247 256 journal homepage:papers.itc.pw.edu.pl An Investigation of Oxide Layer Impact on Heat Transfer in a Fuel Element of the MARIA Reactor
More informationXENON-INDUCED POWER OSCILLATIONS IN A GENERIC SMALL MODULAR REACTOR
XENON-INDUCED POWER OSCILLATIONS IN A GENERIC SMALL MODULAR REACTOR A Dissertation by EVANS DAMENORTEY KITCHER Submitted to the Office of Graduate and Professional Studies of Texas A&M University in partial
More informationMaterial Properties Measurement
Material Properties Measurement Nondestructive Characterization of Neutron Induced Embrittlement in Nuclear Pressure Vessel Steel Microstructure by Using Electromagnetic Testing I. Altpeter, G. Dobmann,
More informationResource Evaluation of Heavy Rare Earth Derived from the Spent Gd 2 O 3 Burnable Poison in LWRs
Journal of Energy and Power Engineering 1 (216) 237-241 doi: 1.17265/1934-8975/216.4.4 D DAVID PUBLISHING Resource Evaluation of Heavy Rare Earth Derived from the Spent Gd 2 O 3 Burnable Poison in LWRs
More informationStudies on the Recriticality Potential during 1F3 Reflooding
8 TH CONFERENCE ON SEVERE ACCIDENT RESEARCH ERMSAR 2017 Studies on the Recriticality Potential during 1F3 Reflooding Piotr Darnowski, Kacper Potapczyk, Konrad Świrski Institute of Heat Engineering, Warsaw
More informationModule 06 Boiling Water Reactors (BWR)
Module 06 Boiling Water Reactors (BWR) 1.10.2015 Prof.Dr. Böck Vienna University oftechnology Atominstitute Stadionallee 2 A-1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Contents BWR Basics
More informationThe PARAMETER test series
The PARAMETER test series V. Nalivaev 1, A. Kiselev 2, J.-S. Lamy 3, S. Marguet 3, V. Semishkin 4, J. Stuckert, Ch. Bals 6, K. Trambauer 6, T. Yudina 2, Yu. Zvonarev 7 1 Scientific Manufacturer Centre,
More informationResearch Article Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR Assemblies
Science and Technology of Nuclear Installations Volume, Article ID 77, 8 pages http://dx.doi.org/.//77 Research Article Void Reactivity Coefficient Analysis during Void Fraction Changes in Innovative BWR
More informationA Review of Suitability for PWHT Exemption Requirements in the Aspect of Residual Stresses and Microstructures
Transactions, SMiRT-23 Division IX, Paper ID 612 (inc. assigned division number from I to X) A Review of Suitability for PWHT Exemption Requirements in the Aspect of Residual Stresses and Microstructures
More informationGIF Lead-cooled Fast Reactor Development Status Alessandro Alemberti (EURATOM / Ansaldo Nucleare)
GIF Lead-cooled Fast Reactor Development Status Alessandro Alemberti (EURATOM / Ansaldo Nucleare) on behalf of GIF LFR provisional System Steering Committee 10th GIF-IAEA Interface Meeting IAEA Headquarters,
More informationA Nuclear Characteristics Study of Inert Matrix Fuel for MA Transmutation in Thermal Spectrum
Proceeding of the Korean Nuclear Autumn Meeting Yongpyong, Korea, Octorber 2002 A Nuclear Characteristics Study of Inert Matrix Fuel for MA Transmutation in Thermal Spectrum Jae-Yong Lim, Myung-Hyun Kim
More information6th European Review Meeting on Severe Accident Research (ERMSAR-2013) Avignon (France), Palais des Papes, 2-4 October, 2013
Analytical support to experiment QUENCH-17 and first post-test calculations with ATHLET-CD C Bals, H Austregesilo, T Hollands Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), Garching (GE) ABSTRACT
More informationFLOW & HEAT TRANSFER IN A PACKED BED - TRANSIENT
FLOW & HEAT TRANSFER IN A PACKED BED - TRANSIENT This case study demonstrates the transient simulation of the heat transfer through a packed bed with no forced convection. This case study is applicable
More informationCFD analysis of coolant flow in the nuclear reactor VVER440
Applied and Computational Mechanics 1 (27) 499-56 CFD analysis of coolant flow in the nuclear reactor VVER44 J. Katolický a, *, M. Bláha b, J. Frelich b, M. Jícha a a Brno University of Technology, Brno,
More informationEFIT: THE EUROPEAN FACILITY FOR INDUSTRIAL TRANSMUTATION OF MINOR ACTINIDES. II.A. Core
EFIT: THE EUROPEAN FACILITY FOR INDUSTRIAL TRANSMUTATION OF MINOR ACTINIDES Barbensi Andrea a, Corsini Giovanni a, Mansani Luigi a, Artioli Carlo b, Glinatsis Georgios b a Ansaldo Nucleare S.p.A., Corso
More informationPHENIX AND SUPERPHENIX FEEDBACK EXPERIENCE. GUIDEZ Joel C E A / France
PHENIX AND SUPERPHENIX FEEDBACK EXPERIENCE GUIDEZ Joel C E A / France Meet the presenter Joël Guidez began his career in the field of sodium-cooled fast reactors, after graduating from the Ecole Centrale
More informationVALIDATION OF BATAN'S STANDARD DIFFUSION CODES ON IAEA BENCHMARK STATIC CALCULATIONS
VALIDATION OF BATAN'S STANDARD DIFFUSION CODES ON IAEA BENCHMARK STATIC CALCULATIONS Sembiring, T.M. and Liem P.H." ABSTRACT VALIDATION OF BATAN'S STANDARD CODES ON laea BENCHMARK STATIC CALCULATIONS.
More informationCOUPLING MICROSTRUCTURE EVOLUTION MODEL WITH FE CODE FOR NUMERICAL SIMULATION OF ROLLING COOLING SEQUENCE FOR RAILS
COUPLING MICROSTRUCTURE EVOLUTION MODEL WITH FE CODE FOR NUMERICAL SIMULATION OF ROLLING COOLING SEQUENCE FOR RAILS Grzegorz SMYK *, Martin FRANTZKE **, Roman KUZIAK ***, Maciej PIETRZYK * * AGH University
More informationA NEUTRONIC FEASIBILITY STUDY OF THE AP1000 DESIGN LOADED WITH FULLY CERAMIC MICRO-ENCAPSULATED FUEL
Engineering (M&C 2013), Sun Valley, Idaho, USA, May 5-9, 2013, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2013) A NEUTRONIC FEASIBILITY STUDY OF THE AP1000 DESIGN LOADED WITH FULLY CERAMIC
More informationCriticality Safety Study for the Disposal of Spent Nuclear Fuel in Water-Saturated Geologic Repository. Xudong Liu
Criticality Safety Study for the Disposal of Spent Nuclear Fuel in Water-Saturated Geologic Repository By Xudong Liu A dissertation submitted in partial satisfaction of the requirements for the degree
More informationDry storage systems and aging management
Dry storage systems and aging management H.Issard, AREVA TN, France IAEA TM 47934 LESSONS LEARNED IN SPENT FUEL MANAGEMENT Vienna, 8-10 July 2014 AREVA TN Summary Dry storage systems and AREVA Experience
More informationTexas A&M University, Department of Nuclear engineering, Ph.D. Qualifying Examination, Fall 2016
Part 2 of 2 100 points of the total exam worth of 200 points Research Area Specific Problems Select and answer any 4 problems from the provided 15 problems focusing on the topics of research tracks in
More informationLEAD-COOLED FAST-NEUTRON REACTOR BREST. Yu.G. Dragunov, V.V. Lemekhov, A.V. Moiseyev, V.S. Smirnov (NIKIET, Moscow, Russia)
LEAD-COOLED FAST-NEUTRON REACTOR BREST Yu.G. Dragunov, V.V. Lemekhov, A.V. Moiseyev, V.S. Smirnov (NIKIET, Moscow, Russia) Large-scale nuclear power based on fast-neutron reactors operating in a closed
More informationFull MOX Core Design in ABWR
GENES4/ANP3, Sep. -9, 3, Kyoto, JAPAN Paper 8 Full MOX Core Design in ABWR Toshiteru Ihara *, Takaaki Mochida, Sadayuki Izutsu 3 and Shingo Fujimaki 3 Nuclear Power Department, Electric Power Development
More informationGerman contribution on the safety assessment of research reactors
German contribution on the safety assessment of research reactors S. Langenbuch J. Rodríguez Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mh. Schwertnergasse 1, D-50667 Köln, Federal Republic
More informationHideout of Sodium Phosphates in Steam Generator Crevices
Hideout of Sodium Phosphates in Steam Generator Crevices By Gwendy Harrington Department of Chemical Engineering, University of New Brunswick, P.O. Box 4400, Fredericton, New Brunswick, E3B 5A3 Abstract
More informationAnalyses of Unprotected Transients in the Lead/Bismuth-Cooled Accelerator Driven System PDS-XADS
Analyses of Unprotected Transients in the Lead/Bismuth-Cooled Accelerator Driven System PDS-XADS Tohru Suzuki, Xue-Nong Chen, Andrei Rineiski, and Werner Maschek Forschungszentrum Karlsruhe, Institute
More informationMaterials Challenges for the Supercritical Water-cooled Reactor (SCWR)
Materials Challenges for the Supercritical Water-cooled Reactor (SCWR) http://ottawapolicyresearch.ca sbaindur@ottawapolicyresearch.ca CNS 2007 Saint John, NB. Outline of Talk Introduction Talk aimed at
More informationGas-cooled Fast Reactor Status and program. Pascal ANZIEU Commissariat à l énergie atomique Atomic Energy Commission France
Gas-cooled Fast Reactor Status and program Pascal ANZIEU Commissariat à l énergie atomique Atomic Energy Commission France Nuclear Energy Division P. Anzieu - GFR Status 1 GFR: an alternative Fast Neutrons
More informationPreliminary Study for Design Core of Nuclear Research Reactor of TRIGA Bandung Using Fuel Element Plate MTR
Article Preliminary Study for Design Core of Nuclear Research Reactor of TRIGA Bandung Using Fuel Element Plate MTR Anwar Ilmar Ramadhan 1,3, *, Aryadi Suwono 1, Efrizon Umar 2, and Nathanael Panagung
More informationInternational Thorium Energy Conference 2015 (ThEC15) BARC, Mumbai, India, October 12-15, 2015
International Thorium Energy Conference 2015 (ThEC15) BARC, Mumbai, India, October 12-15, 2015 Feasibility and Deployment Strategy of Water Cooled Thorium Breeder Reactors Naoyuki Takaki Department of
More informationNUCLEAR ENERGY MATERIALS AND REACTORS - Vol. II - Advanced Gas Cooled Reactors - Tim McKeen
ADVANCED GAS COOLED REACTORS Tim McKeen ADI Limited, Fredericton, Canada Keywords: Advanced Gas Cooled Reactors, Reactor Core, Fuel Elements, Control Rods Contents 1. Introduction 1.1. Magnox Reactors
More informationSimulating Pellet and Clad Mechanical Interactions of Nuclear Fuel Rod for Pressure Water Reactors
14 th International LS-DYNA Users Conference Session: Simulation Simulating Pellet and Clad Mechanical Interactions of Nuclear Fuel Rod for Pressure Water Reactors W. Zhao, D. Mitchell, R. Oelrich Westinghouse
More informationThermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes
EPJ Nuclear Sci. Technol. 2, 16 (2016) J.H. Park and Y.M. Song, published by EDP Sciences, 2016 DOI: 10.1051/epjn/2016010 Nuclear Sciences & Technologies Available online at: http://www.epj-n.org REGULAR
More information