Recent Developments of the
|
|
- Todd Chase
- 5 years ago
- Views:
Transcription
1 emeinschaft der Helmholtz-Ge Mitglied d Recent Developments of the HTR Code Package (HCP) Forschungszentrum Jülich, Germany Technical Meeting on Re-evaluation of Maximum Operating Temperatures g p g p and Accident Conditions for HTR Fuel and Structural Materials IAEA Headquarters,Vienna, June
2 Motivation for HCP Developement Overcome current drawbacks and limitations of individual legacy codes Conservation of knowledge in a contemporary way for future demands Speed up implementation times for necessary (future) model extensions Reduce software maintance costs by avoiding code duplication Increase the numerical stability by applying recent Fortran coding standards and new implementations in C++ to use objects and templates Design, implementation ti and validation of software for the simulation of an HTR reactor core applying the latest programming techniques and standards 2
3 Overview of the HCP 3
4 Input Concept 4
5 Input Concept - Data Library One consistent data library for all HCP modules: cross sections, scattering matrices, decay data, FP release data,... One master file DataLibrary.xml (ENDF/B-VII.0, JEFF 3.1.1,...) Each nuclide has its own data file linked to the master (e.g. Th-232.xml) endfb70.xml Th-232.xml U-235.xml Pu-239.xml 5
6 Input Concept - Data Library A new code HCPLibGen is being developed to build the data library for the HCP Nuclide objects are filled by the code step by step with different kind of data sets Basic interface in HCPLibGen therefore is the HcpNuclide object, this is the same object (same piece of code) used in HCP later on basic data decay data cross section data Fission yield data other data Program flow 6
7 Input Concept - Model and Scenario Model: Definition of the reactor model Mesh grid Material assignment Flow curves (in case of fuel shuffling of pebble bed) Basic data for decay data for cross sections for fission yields for other data 3842 nuclides 3842 nuclides 393 nuclides 31 nuclides Scenario: Definition of the program calculations and boundary conditions program flow Definition of the program flow Module specific (MGT-N/-T, TNT, SHUFLE, STACY, STAR) Program specific (e.g. output) 7
8 Physics Modules 8
9 Physics Modules - Overview Physics Model Former Code HCP Module 3D Neutronics 3D Fluid Dynamics MGT-3D Graphite Corrosion by Air and Water MGT-T MGT-N Depletion and Energy Release VSOP, Origen-Juel NAKURE TNT Fuel Management VSOP SHUFLE FP Release / Fuel Performance Graphite Dust Deposition and Resuspension FRESCO I/II PANAMA STACY --- STAR 9
10 Physics Modules - MGT-N and MGT-T Multi Group TINTE (Time dependent Neutronics and Temperatures) Main Features: Basic data for 3842 nuclides Time dependent neutronics and fluid dynamics for 3D reactor models Feedback between neutronics and fluid dynamics I/Xe dynamics decay data for cross sections for fission yields for other data 3842 nuclides 393 nuclides 31 nuclides Basic graphite corrosion chemistry Local and non-local nuclear heat sources Decay heat calculations according to DIN standard (German standard) program flow Spectrum calculation code with (so far) up to 43 energy groups Gas flow and mixing Heat transport 10
11 MGT-T - Simulation Example: Unit Cell Temperature distribution in representative unit cell of prismatic block Coolant bore hole Fuel 11
12 MGT-T - Simulation Example: Unit Cell Comparison with CFX calculation Maximum temperature difference : 10 C CFX MGT-3D (Unit-cell) 1150 Tem mperature [ C C] Distance to fuel rod center [cm] 12
13 Benchmark of Prismatic Block Calculation FZJ participates in OECD/NEA LOFC Project Accident tests are being / will be performed in HTTR: 1. LOFC at part load, 9 MW (post calculation) 2. LOFC at full load, 30 MW (predictive calculation) MGT-3D will be used to calculate temperature distribution and eventually the time point of recriticality Monte-Carlo neutronics Code SERPENT is being used to simulate the state t of fhttr at tthe beginning i of fthe LOFC as a boundary condition for MGT-3D Newly developed interface code generates mesh data out of block/ pin-wise data 13
14 Physics Modules - TNT TNT : Topological Nuclide Transmutation Main Features: Calculation of time-dependent nuclide amounts due to decay or particle-induced reactions by using the graph theory and a topological solver Calculation of thermal power separately by fission, decay and Basic data capture for decay processes data for cross sections for fission yields for other data 3842 nuclides 3842 nuclides 393 nuclides 31 nuclides Usage of energy-dependent fission yields Determination of burnup measures like FIMA or MWd/kgHM Application of graph theory to model complex nuclide chains program flow Optimized for performance ( minimum graph approach ) Can be runned parallel for different batches with OpenMP 14
15 TNT - Verification of Decay Heat Calculation LWR, 1300 MW class 200 th] Decay po ower [MW Way-Wigner Glasstone 2nd TNT 0 1E+01 1E+02 1E+03 1E+04 1E+05 1E+06 1E+07 Time [s] 15
16 TNT - New Visualisation Features Colored edges for different reactions Different edge types for decay (solid) and neutron induced reactions (dashed) d) Stable nuclides have thicker vertex Edges with different rates have different thickness ( scaled logarithmically) Visualisation of short- and long-lived nuclides Grap ph for data a set C
17 Physics Modules - SHUFLE Software for Handling Universal FueL Elements Main Features: Shuffling of both pebbles and blocks simulation of conical piles Basic data for 3842 nuclides mesh-specific filling factors (e.g. simulation of seismic effects) decay data for cross sections for fission yields for 3842 nuclides 393 nuclides 31 nuclides azimuthally differing flow velocities (3D fuel shuffling) Measured flow paths [1] fuel shuffling includes external facilities (e.g. fuel storages, fuel production plant, ) same mesh grid for pebble flow and other code modules Silo drainage [2] [1] Yang, X.: Experimental Investigation on Feasibility of Two-Region-Designed Pebble-Bed High-Temperature Gas-Cooled Reactor, INET, 2008 [2] Kamrin K., Rycrof C., Bazant M., The stochastic flow rule: a multi-scale model for granular plasticity, MIT. Cambridge,
18 SHUFLE - Graph Theory Graph connecting fuel collections Graph connecting meshes in fuel collection Core 18
19 SHUFLE First Validation Step 100 [%] of tagged pebbles ANABEK SHUFLE ANABEK facility Discharged fraction Recycled fraction of core volume (RCV) [%] 19
20 Output Concept 20
21 Output Concept 21
22 Output Concept - Examples Message example: ***************************************************************** 11 :04:12 INFO TNT TNT calcburnupmeasures ***************************************************************** number of fissions: e 20 average power W : FIMA % : MWd /kg HM : Data field example: time y ; H 001 ; He 004 ; V 053 ; e 00; e 00; e 00; e 00; e 00; e 07; 5,4062e 08; e 22;... 22
23 HCP Backbone (Main Program) 23
24 Summary Existing programs have been refactored, new modules were developed As a next step they will be coupled to the HCP backbone Basic data for 3842 nuclides decay data for cross sections for fission yields for other data 3842 nuclides 393 nuclides 31 nuclides A first prototype will be finished in 2014 (beta testers are welcome) Modules can also be used as stand alone codes and be applied to different physics problems (reactors, radionuclide production, waste repositories) program flow 24
25 Memory Requirements for Datamodel 25
Calculation of the Fission Product Release for the HTR-10 based on its Operation History
Calculation of the Fission Product Release for the HTR-10 based on its Operation History A. Xhonneux 1, C. Druska 1, S. Struth 1, H.-J. Allelein 1,2 1 Forschungszentrum Jülich 52425 Jülich, Germany phone:
More informationSerpent Monte Carlo Neutron Transport Code
Serpent Monte Carlo Neutron Transport Code NEA Expert Group on Advanced Monte Carlo Techniques, Meeting September 17 2012 Jaakko Leppänen / Tuomas Viitanen VTT Technical Research Centre of Finland Outline
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 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 informationTarget accuracy of MA nuclear data and progress in validation by post irradiation experiments with the fast reactor JOYO
Target accuracy of MA nuclear data and progress in validation by post irradiation experiments with the fast reactor JOYO Shigeo OHKI, Kenji YOKOYAMA, Kazuyuki NUMATA *, and Tomoyuki JIN * Oarai Engineering
More informationNeutronic analysis of SFR lattices: Serpent vs. HELIOS-2
Neutronic analysis of SFR lattices: Serpent vs. HELIOS-2 E. Fridman 1, R. Rachamin 1, C. Wemple 2 1 Helmholtz Zentrum Dresden Rossendorf 2 Studsvik Scandpower Inc. Text optional: Institutsname Prof. Dr.
More informationInvestigation of Nuclear Data Accuracy for the Accelerator- Driven System with Minor Actinide Fuel
Investigation of Nuclear Data Accuracy for the Accelerator- Driven System with Minor Actinide Fuel Kenji Nishihara, Takanori Sugawara, Hiroki Iwamoto JAEA, Japan Francisco Alvarez Velarde CIEMAT, Spain
More informationA COMPARISON OF PEBBLE MIXING AND DEPLETION ALGORITHMS USED IN PEBBLE-BED REACTOR EQUILIBRIUM CYCLE SIMULATION
International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2009) Saratoga Springs, New York, May 3-7, 2009, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2009) A COMPARISON
More informationKr-85m activity as burnup measurement indicator in a pebble bed reactor based on ORIGEN2.1 Computer Simulation
Journal of Physics: Conference Series PAPER OPEN ACCESS Kr-85m activity as burnup measurement indicator in a pebble bed reactor based on ORIGEN2.1 Computer Simulation To cite this article: I Husnayani
More informationCriticality analysis of ALLEGRO Fuel Assemblies Configurations
Criticality analysis of ALLEGRO Fuel Assemblies Configurations Radoslav ZAJAC Vladimír CHRAPČIAK 13-16 October 2015 5th International Serpent User Group Meeting at Knoxville, Tennessee ALLEGRO Core - Fuel
More informationNuclear Data Uncertainty Quantification for Applications in Energy, Security, and Isotope Production
Nuclear Data Uncertainty Quantification for Applications in Energy, Security, and Isotope Production I. Gauld M. Williams M. Pigni L. Leal Oak Ridge National Laboratory Reactor and Nuclear Systems Division
More informationStudy on SiC Components to Improve the Neutron Economy in HTGR
Study on SiC Components to Improve the Neutron Economy in HTGR Piyatida TRINURUK and Assoc.Prof.Dr. Toru OBARA Department of Nuclear Engineering Research Laboratory for Nuclear Reactors Tokyo Institute
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 informationComparison of PWR burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results. Abstract
Comparison of PWR burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results Ph.Oberle, C.H.M.Broeders, R.Dagan Forschungszentrum Karlsruhe, Institut for Reactor Safety Hermann-von-Helmholtz-Platz-1,
More informationModeling and Simulation of Dispersion Particle Fuels in Monte Carlo Neutron Transport Calculation
18 th IGORR Conference 2017 Modeling and Simulation of Dispersion Particle Fuels in Monte Carlo Neutron Transport Calculation Zhenping Chen School of Nuclear Science and Technology Email: chzping@yeah.net
More informationNEUTRON PHYSICAL ANALYSIS OF SIX ENERGETIC FAST REACTORS
NEUTRON PHYSICAL ANALYSIS OF SIX ENERGETIC FAST REACTORS Peter Vertes Hungarian Academy of Sciences, Centre for Energy Research ABSTRACT Numerous fast reactor constructions have been appeared world-wide
More informationIAEA-TECDOC Nuclear Fuel Cycle Simulation System (VISTA)
IAEA-TECDOC-1535 Nuclear Fuel Cycle Simulation System (VISTA) February 2007 IAEA-TECDOC-1535 Nuclear Fuel Cycle Simulation System (VISTA) February 2007 The originating Section of this publication in the
More informationWorking Party on Pu-MOX fuel physics and innovative fuel cycles (WPPR)
R&D Needs in Nuclear Science 6-8th November, 2002 OECD/NEA, Paris Working Party on Pu-MOX fuel physics and innovative fuel cycles (WPPR) Hideki Takano Japan Atomic Energy Research Institute, Japan Introduction(1)
More informationChallenges in Prismatic HTR Reactor Physics
Challenges in Prismatic HTR Reactor Physics Javier Ortensi R&D Scientist - Idaho National Laboratory www.inl.gov Advanced Reactor Concepts Workshop, PHYSOR 2012 April 15, 2012 Outline HTR reactor physics
More informationCore Physics Second Part How We Calculate LWRs
Core Physics Second Part How We Calculate LWRs Dr. E. E. Pilat MIT NSED CANES Center for Advanced Nuclear Energy Systems Method of Attack Important nuclides Course of calc Point calc(pd + N) ϕ dn/dt N
More informationMOx Benchmark Calculations by Deterministic and Monte Carlo Codes
MOx Benchmark Calculations by Deterministic and Monte Carlo Codes G.Kotev, M. Pecchia C. Parisi, F. D Auria San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa via Diotisalvi 2, 56122
More informationREVIEW OF RESULTS FOR THE OECD/NEA PHASE VII BENCHMARK: STUDY OF SPENT FUEL COMPOSITIONS FOR LONG-TERM DISPOSAL
REVIEW OF RESULTS FOR THE OECD/NEA PHASE VII BENCHMARK: STUDY OF SPENT FUEL COMPOSITIONS FOR LONG-TERM DISPOSAL Georgeta Radulescu John Wagner (presenter) Oak Ridge National Laboratory International Workshop
More informationActivation Calculation for a Fusion-driven Sub-critical Experimental Breeder, FDEB
Activation Calculation for a Fusion-driven Sub-critical Experimental Breeder, FDEB K. M. Feng (Southwestern Institute of Physics, China) Presented at 8th IAEA Technical Meeting on Fusion Power Plant Safety
More informationDevelopment of depletion models for radionuclide inventory, decay heat and source term estimation in discharged fuel
Development of depletion models for radionuclide inventory, decay heat and source term estimation in discharged fuel S. Caruso, A. Shama, M. M. Gutierrez National Cooperative for the Disposal of Radioactive
More informationChapter 5: Applications Fission simulations
Chapter 5: Applications Fission simulations 1 Using fission in FISPACT-II FISPACT-II is distributed with a variety of fission yields and decay data, just as incident particle cross sections, etc. Fission
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 informationEvaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste 13187
Evaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste 13187 Ivan Fast, Yuliya Aksyutina and Holger Tietze-Jaensch* Product Quality Control Office for Radioactive
More informationTRANSMUTATION OF CESIUM-135 WITH FAST REACTORS
TRANSMUTATION OF CESIUM-3 WITH FAST REACTORS Shigeo Ohki and Naoyuki Takaki O-arai Engineering Center Japan Nuclear Cycle Development Institute (JNC) 42, Narita-cho, O-arai-machi, Higashi-Ibaraki-gun,
More informationConsistent Code-to-Code Comparison of Pin-cell Depletion Benchmark Suite
Consistent Code-to-Code Comparison of Pin-cell Depletion Benchmark Suite September 27, 26 Jinsu Park, Deokjung Lee * COmputational Reactor Physics & Experiment lab Contents VERA depletion benchmark suite
More informationStudy on Nuclear Transmutation of Nuclear Waste by 14 MeV Neutrons )
Study on Nuclear Transmutation of Nuclear Waste by 14 MeV Neutrons ) Takanori KITADA, Atsuki UMEMURA and Kohei TAKAHASHI Osaka University, Graduate School of Engineering, Division of Sustainable Energy
More informationEquilibrium core depletion and criticality analysis of the HTR-10 for Uranium and Thorium fuel cycles
Equilibrium core depletion and criticality analysis of the HTR-10 for Uranium and Thorium fuel cycles Godart van Gendt May 8th August 17th 2006 Supervisors Dr. Ir. Jan Leen Kloosterman Ir. Brian Boer TU
More informationSPentfuel characterisation Program for the Implementation of Repositories
SPentfuel characterisation Program for the Implementation of Repositories WP2 & WP4 Development of measurement methods and techniques to characterise spent nuclear fuel Henrik Widestrand and Peter Schillebeeckx
More informationDEVELOPMENT AND VALIDATION OF SCALE NUCLEAR ANALYSIS METHODS FOR HIGH TEMPERATURE GAS-COOLED REACTORS
DEVELOPMENT AND VALIDATION OF SCALE NUCLEAR ANALYSIS METHODS FOR HIGH TEMPERATURE GAS-COOLED REACTORS Jess Gehin, Matthew Jessee, Mark Williams, Deokjung Lee, Sedat Goluoglu, Germina Ilas, Dan Ilas, Steve
More informationFuel cycle studies on minor actinide transmutation in Generation IV fast reactors
Fuel cycle studies on minor actinide transmutation in Generation IV fast reactors M. Halász, M. Szieberth, S. Fehér Budapest University of Technology and Economics, Institute of Nuclear Techniques Contents
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 informationCalculation of dose rate, decay heat and criticality for verifying compliance with transport limits for steel packages
2005, April 14th & 15th Radioactivity, radionuclides & radiation Lars Niemann Forschungszentrum Karlsruhe in der Helmholtz-Gemeinschaft Nuclear Facilities Decommissioning Division Calculation of dose rate,
More informationHTR Spherical Super Lattice Model For Equilibrium Fuel Cycle Analysis. Gray S. Chang. September 12-15, 2005
INEEL/CON-05-02655, Revision 3 PREPRINT HTR Spherical Super Lattice Model For Equilibrium Fuel Cycle Analysis Gray S. Chang September 12-15, 2005 Mathematics And Computation, Supercomputing, Reactor Physics
More informationNew Capabilities for the Chebyshev Rational Approximation method (CRAM)
New Capabilities for the Chebyshev Rational Approximation method (CRAM) A. Isotaloa,b W. Wieselquista M. Pusac aoak Ridge National Laboratory PO Box 2008, Oak Ridge, TN 37831-6172, USA baalto University
More informationIntroduction to Nuclear Data
united nations educational, scientific and cultural organization the abdus salam international centre for theoretical physics international atomic energy agency SMR.1555-34 Workshop on Nuclear Reaction
More informationUSA HTR NEUTRONIC CHARACTERIZATION OF THE SAFARI-1 MATERIAL TESTING REACTOR
Proceedings of HTR2008 4 th International Topical Meeting on High Temperature Reactors September 28-October 1, 2008, Washington, D.C, USA HTR2008-58155 NEUTRONIC CHARACTERIZATION OF THE SAFARI-1 MATERIAL
More informationNeutronic Issues and Ways to Resolve Them. P.A. Fomichenko National Research Center Kurchatov Institute Yu.P. Sukharev JSC Afrikantov OKBM,
GT-MHR Project High-Temperature Reactor Neutronic Issues and Ways to Resolve Them P.A. Fomichenko National Research Center Kurchatov Institute Yu.P. Sukharev JSC Afrikantov OKBM, GT-MHR PROJECT MISSION
More informationSensitivity and Uncertainty Analysis Methodologies for Fast Reactor Physics and Design at JAEA
Sensitivity and Uncertainty Analysis Methodologies for Fast Reactor Physics and Design at JAEA Kick off meeting of NEA Expert Group on Uncertainty Analysis for Criticality Safety Assessment IRSN, France
More informationAssessment of the MCNP-ACAB code system for burnup credit analyses
Assessment of the MCNP-ACAB code system for burnup credit analyses N. García-Herranz, O. Cabellos, J. Sanz UPM - UNED International Workshop on Advances in Applications of Burnup Credit for Spent Fuel
More informationAdaptation of Pb-Bi Cooled, Metal Fuel Subcritical Reactor for Use with a Tokamak Fusion Neutron Source
Adaptation of Pb-Bi Cooled, Metal Fuel Subcritical Reactor for Use with a Tokamak Fusion Neutron Source E. Hoffman, W. Stacey, G. Kessler, D. Ulevich, J. Mandrekas, A. Mauer, C. Kirby, D. Stopp, J. Noble
More informationMonte Carlo neutron transport and thermal-hydraulic simulations using Serpent 2/SUBCHANFLOW
Monte Carlo neutron transport and thermal-hydraulic simulations using Serpent 2/SUBCHANFLOW M. Knebel (Presented by V. Valtavirta) Institute for Neutron Physics and Reactor Technology (INR) Reactor Physics
More informationDETERMINATION OF THE EQUILIBRIUM COMPOSITION OF CORES WITH CONTINUOUS FUEL FEED AND REMOVAL USING MOCUP
Supercomputing in Nuclear Applications (M&C + SNA 2007) Monterey, California, April 15-19, 2007, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2007) DETERMINATION OF THE EQUILIBRIUM COMPOSITION
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 informationTechnical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 1
Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 1 Technical Meeting on Priorities in Modeling and Simulation for Fast Neutron Systems 2 Established in 1937, Bachelor,
More informationNeutron Dose near Spent Nuclear Fuel and HAW after the 2007 ICRP Recommendations
Neutron Dose near Spent Nuclear Fuel and HAW after the 2007 ICRP Recommendations Gunter Pretzsch Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbh Radiation and Environmental Protection Division
More informationIMPACT OF THE FISSION YIELD COVARIANCE DATA IN BURN-UP CALCULATIONS
IMPACT OF THE FISSION YIELD COVARIANCE DATA IN BRN-P CALCLATIONS O. Cabellos, D. Piedra, Carlos J. Diez Department of Nuclear Engineering, niversidad Politécnica de Madrid, Spain E-mail: oscar.cabellos@upm.es
More informationQuestion to the class: What are the pros, cons, and uncertainties of using nuclear power?
Energy and Society Week 11 Section Handout Section Outline: 1. Rough sketch of nuclear power (15 minutes) 2. Radioactive decay (10 minutes) 3. Nuclear practice problems or a discussion of the appropriate
More informationPEBBLE BED REACTORS FOR ONCE THROUGH NUCLEAR TRANSMUTATION.
PEBBLE BED REACTORS FOR ONCE THROUGH NUCLEAR TRANSMUTATION. Pablo León, José Martínez-Val, Alberto Abánades and David Saphier. Universidad Politécnica de Madrid, Spain. C/ J. Gutierrez Abascal Nº2, 28006
More informationOn the Use of Serpent for SMR Modeling and Cross Section Generation
On the Use of Serpent for SMR Modeling and Cross Section Generation Yousef Alzaben, Victor. H. Sánchez-Espinoza, Robert Stieglitz INSTITUTE for NEUTRON PHYSICS and REACTOR TECHNOLOGY (INR) KIT The Research
More information3.12 Development of Burn-up Calculation System for Fusion-Fission Hybrid Reactor
3.12 Development of Burn-up Calculation System for Fusion-Fission Hybrid Reactor M. Matsunaka, S. Shido, K. Kondo, H. Miyamaru, I. Murata Division of Electrical, Electronic and Information Engineering,
More informationENHANCEMENT OF COMPUTER SYSTEMS FOR CANDU REACTOR PHYSICS SIMULATIONS
ENHANCEMENT OF COMPUTER SYSTEMS FOR CANDU REACTOR PHYSICS SIMULATIONS E. Varin, M. Dahmani, W. Shen, B. Phelps, A. Zkiek, E-L. Pelletier, T. Sissaoui Candu Energy Inc. WORKSHOP ON ADVANCED CODE SUITE FOR
More informationPWR AND WWER MOX BENCHMARK CALCULATION BY HELIOS
PWR AND WWER MOX BENCHMARK CALCULATION BY HELIOS Radoslav ZAJAC 1,2), Petr DARILEK 1), Vladimir NECAS 2) 1 VUJE, Inc., Okruzna 5, 918 64 Trnava, Slovakia; zajacr@vuje.sk, darilek@vuje.sk 2 Slovak University
More informationLesson 14: Reactivity Variations and Control
Lesson 14: Reactivity Variations and Control Reactivity Variations External, Internal Short-term Variations Reactivity Feedbacks Reactivity Coefficients and Safety Medium-term Variations Xe 135 Poisoning
More informationPower Installations based on Activated Nuclear Reactions of Fission and Synthesis
Yu.V. Grigoriev 1,2, A.V. Novikov-Borodin 1 1 Institute for Nuclear Research RAS, Moscow, Russia 2 Joint Institute for Nuclear Research, Dubna, Russia Power Installations based on Activated Nuclear Reactions
More informationOn-the-fly Doppler Broadening in Serpent
On-the-fly Doppler Broadening in Serpent 1st International Serpent User Group Meeting 16.9.2011, Dresden Tuomas Viitanen VTT Technical Research Centre of Finland Outline Fuel temperatures in neutronics
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 informationSpectral History Correction of Microscopic Cross Sections for the PBR Using the Slowing Down Balance. Abstract
Organized and hosted by the Canadian Nuclear Society. Vancouver, BC, Canada. 2006 September 10-14 Spectral History Correction of Microscopic Cross Sections for the PBR Using the Slowing Down Balance Nathanael
More informationFusion/transmutation reactor studies based on the spherical torus concept
FT/P1-7, FEC 2004 Fusion/transmutation reactor studies based on the spherical torus concept K.M. Feng, J.H. Huang, B.Q. Deng, G.S. Zhang, G. Hu, Z.X. Li, X.Y. Wang, T. Yuan, Z. Chen Southwestern Institute
More informationNumerical analysis on element creation by nuclear transmutation of fission products
NUCLEAR SCIENCE AND TECHNIQUES 26, S10311 (2015) Numerical analysis on element creation by nuclear transmutation of fission products Atsunori Terashima 1, and Masaki Ozawa 2 1 Department of Nuclear Engineering,
More informationTechnical workshop : Dynamic nuclear fuel cycle
Technical workshop : Dynamic nuclear fuel cycle Reactor description in CLASS Baptiste LENIAU* Institut d Astrophysique de Paris 6-8 July, 2016 Introduction Summary Summary The CLASS package : a brief overview
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 informationDecay heat calculations. A study of their validation and accuracy.
Decay heat calculations A study of their validation and accuracy. Presented by : Dr. Robert W. Mills, UK National Nuclear Laboratory. Date: 01/10/09 The UK National Nuclear Laboratory The NNL (www.nnl.co.uk)
More informationA.BIDAUD, I. KODELI, V.MASTRANGELO, E.SARTORI
SENSITIVITY TO NUCLEAR DATA AND UNCERTAINTY ANALYSIS: THE EXPERIENCE OF VENUS2 OECD/NEA BENCHMARKS. A.BIDAUD, I. KODELI, V.MASTRANGELO, E.SARTORI IPN Orsay CNAM PARIS OECD/NEA Data Bank, Issy les moulineaux
More informationASSESSMENT OF THE EQUILIBRIUM STATE IN REACTOR-BASED PLUTONIUM OR TRANSURANICS MULTI-RECYCLING
ASSESSMENT OF THE EQUILIBRIUM STATE IN REACTOR-BASED PLUTONIUM OR TRANSURANICS MULTI-RECYCLING T.K. Kim, T.A. Taiwo, J.A. Stillman, R.N. Hill and P.J. Finck Argonne National Laboratory, U.S. Abstract An
More informationPreventing xenon oscillations in Monte Carlo burnup calculations by forcing equilibrium
Preventing xenon oscillations in Monte Carlo burnup calculations by forcing equilibrium Aarno Isotaloa), Jaakko Leppänenb), Jan Dufekcc) a) Aalto University, Finland b) VTT Technical Research Centrte of
More informationIAEA-TECDOC-1349 Potential of thorium based fuel cycles to constrain plutonium and reduce long lived waste toxicity
IAEA-TECDOC-1349 Potential of thorium based fuel cycles to constrain plutonium and reduce long lived waste toxicity Final report of a co-ordinated research project 1995 2001 April 2003 The originating
More informationFirst ANDES annual meeting
First ANDES Annual meeting 3-5 May 011 CIEMAT, Madrid, Spain 1 / 0 *C.J. Díez e-mail: cj.diez@upm.es carlosjavier@denim.upm.es UNCERTAINTY METHODS IN ACTIVATION AND INVENTORY CALCULATIONS Carlos J. Díez*,
More informationCONFIGURATION ADJUSTMENT POTENTIAL OF THE VERY HIGH TEMPERATURE REACTOR PRISMATIC CORES WITH ADVANCED ACTINIDE FUELS. A Thesis DAVID E.
CONFIGURATION ADJUSTMENT POTENTIAL OF THE VERY HIGH TEMPERATURE REACTOR PRISMATIC CORES WITH ADVANCED ACTINIDE FUELS A Thesis by DAVID E. AMES II Submitted to the Office of Graduate Studies of Texas A&M
More informationQUADRATIC DEPLETION MODEL FOR GADOLINIUM ISOTOPES IN CASMO-5
Advances in Nuclear Fuel Management IV (ANFM 2009) Hilton Head Island, South Carolina, USA, April 12-15, 2009, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2009) QUADRATIC DEPLETION MODEL FOR
More informationBURNUP CALCULATION CAPABILITY IN THE PSG2 / SERPENT MONTE CARLO REACTOR PHYSICS CODE
International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 29) Saratoga Springs, New York, May 3-7, 29, on CD-ROM, American Nuclear Society, LaGrange Park, IL (29) BURNUP CALCULATION
More informationImproved time integration methods for burnup calculations with Monte Carlo neutronics
Improved time integration methods for burnup calculations with Monte Carlo neutronics Aarno Isotalo 13.4.2010 Burnup calculations Solving time development of reactor core parameters Nuclide inventory,
More informationAbstract. STAHALA, MIKE PETER. High Level Nuclear Waste Repository Thermal Loading Analysis. (Under the direction of Man-Sung Yim and David McNelis)
Abstract STAHALA, MIKE PETER. High Level Nuclear Waste Repository Thermal Loading Analysis. (Under the direction of Man-Sung Yim and David McNelis) A spent nuclear fuel (SNF) decay heat model was developed
More informationREACTOR PHYSICS ASPECTS OF PLUTONIUM RECYCLING IN PWRs
REACTOR PHYSICS ASPECTS OF PLUTONIUM RECYCLING IN s Present address: J.L. Kloosterman Interfaculty Reactor Institute Delft University of Technology Mekelweg 15, NL-2629 JB Delft, the Netherlands Fax: ++31
More informationSome thoughts on Fission Yield Data in Estimating Reactor Core Radionuclide Activities (for anti-neutrino estimation)
Some thoughts on Fission Yield Data in Estimating Reactor Core Radionuclide Activities (for anti-neutrino estimation) Dr Robert W. Mills, NNL Research Fellow for Nuclear Data, UK National Nuclear Laboratory.
More informationCurrent studies of neutron induced reactions regard essentially two mass regions, identified in the chart of nuclides: isotopes in the region from Fe
The talk gives an overview of the current reseach activity with neutron beams for fundamental and applied Nuclear Physics. In particular, it presents the status and perspectives of neutron studies in the
More informationRequests on Nuclear Data in the Backend Field through PIE Analysis
Requests on Nuclear Data in the Backend Field through PIE Analysis Yoshihira Ando 1), Yasushi Ohkawachi 2) 1) TOSHIBA Corporation Power System & Services Company Power & Industrial Systems Research & Development
More informationTitle. Author(s)Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi. CitationAnnals of nuclear energy, 65: Issue Date Doc URL.
Title Photon transport effect on intra-subassembly thermal Author(s)Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi CitationAnnals of nuclear energy, 65: 41-46 Issue Date 2014-03 Doc URL http://hdl.handle.net/2115/55139
More informationUSE OF LATTICE CODE DRAGON IN REACTOR CALUCLATIONS
USE OF LATTICE CODE DRAGON IN REACTOR CALUCLATIONS ABSTRACT Dušan Ćalić ZEL-EN razvojni center Hočevarjev trg 1 Slovenia-SI8270, Krško, Slovenia dusan.calic@zel-en.si Andrej Trkov, Marjan Kromar J. Stefan
More informationTransmutation of Minor Actinides in a Spherical
1 Transmutation of Minor Actinides in a Spherical Torus Tokamak Fusion Reactor Feng Kaiming Zhang Guoshu Fusion energy will be a long-term energy source. Great efforts have been devoted to fusion research
More informationG. S. Chang. April 17-21, 2005
INEEL/CON-04-02085 PREPRINT MCWO Linking MCNP and ORIGEN2 For Fuel Burnup Analysis G. S. Chang April 17-21, 2005 The Monte Carlo Method: Versatility Unbounded In A Dynamic Computing World This is a preprint
More informationScope and Objectives. Codes and Relevance. Topics. Which is better? Project Supervisor(s)
Development of BFBT benchmark models for sub-channel code COBRA-TF versus System Code TRACE MSc Proposal 1 Fuel Assembly Thermal-Hydraulics Comparative Assessment of COBRA-TF and TRACE for CHF Analyses
More informationNuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like Sodium Fast Reactor
Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like Sodium Fast Reactor García-Herranz Nuria 1,*, Panadero Anne-Laurène 2, Martinez Ana 1, Pelloni
More informationCiclo combustibile, scorie, accelerator driven system
Ciclo combustibile, scorie, accelerator driven system M. Carta, C. Artioli ENEA Fusione e Fissione Nucleare: stato e prospettive sulle fonti energetiche nucleari per il futuro Layout of the presentation!
More informationON THE EVALUATION OF PEBBLE BEAD REACTOR CRITICAL EXPERIMENTS USING THE PEBBED CODE
ON THE EVALUATION OF PEBBLE BEAD REACTOR CRITICAL EXPERIMENTS USING THE PEBBED CODE Hans D. Gougar, R. Sonat Sen Idaho National Laboratory 2525 N. Fremont Avenue, Idaho Falls, ID 83415-3885 phone: +1-208-5261314,
More informationCALCULATION OF TEMPERATURE REACTIVITY COEFFICIENTS IN KRITZ-2 CRITICAL EXPERIMENTS USING WIMS ABSTRACT
CALCULATION OF TEMPERATURE REACTIVITY COEFFICIENTS IN KRITZ-2 CRITICAL EXPERIMENTS USING WIMS D J Powney AEA Technology, Nuclear Science, Winfrith Technology Centre, Dorchester, Dorset DT2 8DH United Kingdom
More informationError Estimation for ADS Nuclear Properties by using Nuclear Data Covariances
Error Estimation for ADS Nuclear Properties by using Nuclear Data Covariances Kasufumi TSUJIMOTO Center for Proton Accelerator Facilities, Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, Ibaraki-ken
More informationFULL CORE POWER AND ISOTOPIC OSCILLATIONS WITH VARIOUS DEPLETION SCHEMES
FULL CORE POWER AND ISOTOPIC OSCILLATIONS WITH VARIOUS DEPLETION SCHEMES A N D R E W J O H N S O N C O M P U TAT I O N A L R E A C T O R E N G I N E E R I N G OUTLINE Governing depletion equations Summary
More informationFRENDY: A new nuclear data processing system being developed at JAEA
FRENDY: A new nuclear data processing system being developed at JAEA Kenichi Tada a, Yasunobu Nagaya, Satoshi Kunieda, Kenya Suyama, and Tokio Fukahori Japan Atomic Energy Agency, Tokai, Japan Abstract.
More informationChemistry 500: Chemistry in Modern Living. Topic 5: The Fires of Nuclear Fission. Atomic Structure, Nuclear Fission and Fusion, and Nuclear.
Chemistry 500: Chemistry in Modern Living 1 Topic 5: The Fires of Nuclear Fission Atomic Structure, Nuclear Fission and Fusion, and Nuclear Weapons Chemistry in Context, 2 nd Edition: Chapter 8, Pages
More informationAnalytical Validation of Uncertainty in Reactor Physics Parameters for Nuclear Transmutation Systems
Journal of Nuclear Science and Technology ISSN: 22-3131 (Print) 1881-1248 (Online) Journal homepage: http://www.tandfonline.com/loi/tnst2 Analytical Validation of Uncertainty in Reactor Physics Parameters
More informationSENSITIVITY ANALYSIS OF ALLEGRO MOX CORE. Bratislava, Iľkovičova 3, Bratislava, Slovakia
SENSITIVITY ANALYSIS OF ALLEGRO MOX CORE Jakub Lüley 1, Ján Haščík 1, Vladimír Slugeň 1, Vladimír Nečas 1 1 Institute of Nuclear and Physical Engineering, Slovak University of Technology in Bratislava,
More informationA PWR ASSEMBLY COMPUTATIONAL SCHEME BASED ON THE DRAGON V4 LATTICE CODE
Supercomputing in Nuclear Applications (M&C + SNA 2007) Monterey, California, April 15-19, 2007, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2007) A PWR ASSEMBLY COMPUTATIONAL SCHEME BASED
More informationTreatment of Implicit Effects with XSUSA.
Treatment of Implicit Effects with Friederike Bostelmann 1,2, Andreas Pautz 2, Winfried Zwermann 1 1 Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh, Boltzmannstraße 14, 85748 Garching, Germany
More informationEnglish text only NUCLEAR ENERGY AGENCY NUCLEAR SCIENCE COMMITTEE
Unclassified NEA/NSC/DOC(2007)9 NEA/NSC/DOC(2007)9 Unclassified Organisation de Coopération et de Développement Economiques Organisation for Economic Co-operation and Development 14-Dec-2007 English text
More informationARTICLE. EASY-II(12): a system for modelling of n, d, p, γ, α activation and transmutation processes
DOI: 10.15669/pnst.4.349 Progress in Nuclear Science and Technology Volume 4 (2014) pp. 349-353 ARTICLE EASY-II(12): a system for modelling of n, d, p, γ, α activation and transmutation processes Jean-Christophe
More informationParametric study of thorium fuel cycles in a 100MWth pebble bed high temperature reactor
Parametric study of thorium fuel cycles in a 100MWth pebble bed high temperature reactor F Panday Mini-dissertation submitted in partial fulfilment of the requirements for the degree Master of Science
More information