LEU Conversion of the University of Wisconsin Nuclear Reactor
|
|
|
- Junior Lawrence
- 6 years ago
- Views:
Transcription
1 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 Moscow
2 Overview Introducing the UWNR Conversion analysis: neutronics, thermal-hydraulics & accident analyses Methods Challenges Outcomes Lessons learned/opportunities Current status 9 June 2011 P.Wilson/UWNR LEU Conversion 2
3 About the UWNR 1MW TRIGA-type pool reactor Converted MTR with 4 pin square bundle Training as primary mission Irradiation services Neutron research 3 full time staff 1 part-time operator 3-6 student operators hours/week 9 June 2011 P.Wilson/UWNR LEU Conversion 3
4 UWNR History January 1961 First critical (10 kw) December 1964 Power upgrate 250 kw November 1967 TRIGA conversion - 1 MW uprate & pulsing capability June 1979 Completed conversion to GA FLIP TRIGA HEU September 2009 Initial criticality with GA 30/20 TRIGA LEU fuel March 2011 Renewed facility license R-74 issued by U.S. Nuclear Regulatory Commission 9 June 2011 P.Wilson/UWNR LEU Conversion 4
5 Conversion History 1987 First notification of intent to convert October 2005 Confirmation of conversion schedule September 2007 Began in-house conversion analysis August 2008 Submitted conversion report to USNRC February 2009 Requests for Additional Information from USNRC June 2009 Order to convert from USNRC September 2009 First LEU fuel loaded September 2009 First criticality with LEU fuel 9 June 2011 P.Wilson/UWNR LEU Conversion 5
6 Neutronics Analysis Methods MCNP5 with ENDF/B-VII data (UW) Full 3-D core model including axial and radial details REBUS coupling for burnup analysis (ANL) 3 radial x 5 axial burnup zones per fuel pin Some confirmatory analysis with HELIOS (UW) 2-D deterministic analysis with detailed geometry 1-D diffusion approximation 9 June 2011 P.Wilson/UWNR LEU Conversion 6
7 Neutronics Analyses Most analyses repeated for beginning of life (BOL), middle of life (MOL) and end of life (EOL) Power distribution (for T/H analysis) Total pin power/core power distribution Axial/radial distribution in maximum power pin Shutdown margin as a function of burnup Reactivity parameters Delayed neutron fraction, prompt neutron lifetime, control element worth, prompt temperature coefficient 9 June 2011 P.Wilson/UWNR LEU Conversion 7
8 Neutronics Analysis Challenges Lack of data for robust HEU benchmarking of model/bias determination Operational data Material compositions Large CPU resource requirements for some analysis Modest existing capacity with reactor analysis 9 June 2011 P.Wilson/UWNR LEU Conversion 8
9 Neutronics Analysis Outcomes Calculated system reactivity too high with like-for-like replacement Change from symmetric 23 bundle + 10 reflector configuration (H) To symmetric 21 bundle + 14 reflector configuration (X) Slightly reduced core lifetime Additional bundles provided for future reactivity boost 9 June 2011 P.Wilson/UWNR LEU Conversion 9
10 Neutronics Operational Experience Real system substantially less reactive than calculated ~0.7% k/k (~$1) reactivity discrepancy from fuel (including axial reflectors) ~0.7% k/k (~$1) reactivity discrepancy from graphite reflectors Planning near-term core shuffle From X-configuration (21F + 14R) to +-configuration (21F + 6R) 9 June 2011 P.Wilson/UWNR LEU Conversion 10
11 Thermal Analysis Methods RELAP5/Mod3.3 Single channel with highest power channel 20 axial nodes 15 in active fuel region 27 radial nodes 21 in fuel Two channel model for pulsing Hot channel + rest of core RELAP point reactor kinetics model Temperature coefficients from MCNP5 analysis Two channel model of LOCA Three phases with different water levels Axial conduction in fuel 9 June 2011 P.Wilson/UWNR LEU Conversion 11
12 Thermal Analyses Results for high-power channel (assuming no cross-flow) Steady state Low power pulse (transient & maximum) High power pulse (transient & maximum) Flow rate Temperatures Maximum fuel centerline/clad Axial/radial profile Minimum Departure from Nucleate Boiling Ratio (MDNBR) 9 June 2011 P.Wilson/UWNR LEU Conversion 12
13 Thermal Analysis Challenges Sensitivity to gap thickness Discrepancy between critical heat flux correlations Uncertainty in natural convection heat transfer models Appropriate air-cooled temperature safety limits for U-ZrH 2 fuel in SS clad 9 June 2011 P.Wilson/UWNR LEU Conversion 13
14 Thermal Analyses Outcomes Small changes in steady-state operation Average pin power increases with fewer pins Improved definition of fuel temperature Limiting Safety System Setting Pulsing operation from 1 kw Max fuel temperature ~727 o C (MOL) Pulsing operation from 1.3 MW is within technical specifications Max fuel temperature ~1000 o C < 1150 o C (EOL) LOCA fuel temperatures < 700 o C (MOL) 9 June 2011 P.Wilson/UWNR LEU Conversion 14
15 Accident Analysis Methods ORIGEN calculation of fission product inventory Analytic analysis of release fractions and Gaussian plume model MCNP5 calculation of radiation doses 9 June 2011 P.Wilson/UWNR LEU Conversion 15
16 Accident Analysis Methods Fission product release from Maximum Hypothetical Accident Cladding failure in high power pin (25 kw) after continuous full power operation With & without intact pool & operative ventilation system Reactivity insertion (see T/H analysis) Loss of cooling accident Fuel temperature Radiation dose from exposed core 9 June 2011 P.Wilson/UWNR LEU Conversion 16
17 Accident Analysis Outcomes No changes due to LEU conversion Within regulatory limits under all variations to Maximum Hypothetical Accident More detailed understanding of radiation dose Time dependent behavior Spatial distribution 9 June 2011 P.Wilson/UWNR LEU Conversion 17
18 Lessons Learned/Opportunities Capacity building Neutronics analysis performed by post-doc now working at Thai regulator on TRIGA licensing Widespread upgrade in UWNR staff analysis capabilities Ongoing experimental research to better understand natural circulation heat transfer in TRIGA-relevant conditions 100% fresh core provides wide variety of benchmark data for improving analysis capabilities 9 June 2011 P.Wilson/UWNR LEU Conversion 18
Regulatory Challenges and Solutions High-Enriched to Low-Enriched Uranium Fuel Conversion
Regulatory Challenges and Solutions High-Enriched to Low-Enriched Uranium Fuel Conversion Alexander Adams Jr. Senior Project Manager Research and Test Reactor Program Office of Nuclear Reactor Regulation
Safety 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
CHALLENGES WITH THE CONVERSION OF THE MITR
Russian-American Symposium on the Conversion of Research Reactors to Low Enriched Uranium Fuel Moscow, Russia CHALLENGES WITH THE CONVERSION OF THE MITR T. H. Newton, Jr Director of Reactor Operations
COMPARATIVE STUDY OF TRANSIENT ANALYSIS OF PAKISTAN RESEARCH REACTOR-1 (PARR-1) WITH HIGH DENSITY FUEL
COMPARATIVE STUDY OF TRANSIENT ANALYSIS OF PAKISTAN RESEARCH REACTOR-1 (PARR-1) WITH HIGH DENSITY FUEL M. Iqbal, A. Muhammad, T. Mahmood Nuclear Engineering Division, Pakistan Institute of Nuclear Science
Core Management and Fuel Handling for Research Reactors
Core Management and Fuel Handling for Research Reactors A. M. Shokr Research Reactor Safety Section Division of Nuclear Installation Safety International Atomic Energy Agency Outline Introduction Safety
TRANSIENT ANALYSES AND THERMAL-HYDRAULIC SAFETY MARGINS FOR THE GREEK RESEARCH REACTOR (GRRI)*
TRANSIENT ANALYSES AND THERMAL-HYDRAULIC SAFETY MARGINS FOR THE GREEK RESEARCH REACTOR (GRRI)* W. L. Woodruff and J. R. Deen Argonne National Laboratory Argonne, IL USA and C. Papastergiou National Centre
DOSE CALCULATION FOR ACCIDENT SITUATIONS AT TRIGA RESEARCH REACTOR USING LEU FUEL TYPE
Romanian Reports in Physics, Vol. 60, No. 1, P. 57 61, 2008 DOSE CALCULATION FOR ACCIDENT SITUATIONS AT TRIGA RESEARCH REACTOR USING LEU FUEL TYPE S. MÃRGEANU 1, C. A. MÃRGEANU 1, C. PÃUNOIU 1, T. ANGELESCU
PUSPATI TRIGA REACTOR UPGRADING: TOWARDS THE SAFE OPERATION & FEASIBILITY OF NEUTRONIC APPROACH
International Conference on Research Reactors: Safe Management and Effective Utilization, Rabat, Morocco, 14-18 November 2011 PUSPATI TRIGA REACTOR UPGRADING: TOWARDS THE SAFE OPERATION & FEASIBILITY OF
MCNP5 CALCULATIONS COMPARED TO EXPERIMENTAL MEASUREMENTS IN CEA-MINERVE REACTOR
U.P.B. Sci. Bull., Series D, Vol. 74, Iss. 1, 2012 ISSN 1454-2358 MCNP5 CALCULATIONS COMPARED TO EXPERIMENTAL MEASUREMENTS IN CEA-MINERVE REACTOR Mirea MLADIN 1, Daniela MLADIN 21 The paper describes the
Core Management and Fuel handling for Research Reactors
Core Management and Fuel handling for Research Reactors W. Kennedy, Research Reactor Safety Section Division of Nuclear Installation Safety Yogyakarta, Indonesia 23/09/2013 Outline Introduction Safety
Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup
M.H. Altaf and Atom N.H. Badrun Indonesia / Atom Vol. 40 Indonesia No. 3 (2014) Vol. 40107 No. - 112 3 (2014) 107-112 Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering
Nuclear Reactors. 3 Unit Nuclear Power Station - Coastal Bryon Nuclear Plant, Illinois. 3 Unit Nuclear Power Station - Desert
3 Unit Nuclear Power Station - Desert Nuclear Reactors Homeland Security Course February 1, 2007 Tom Gesell 3 Unit Nuclear Power Station - Coastal Bryon Nuclear Plant, Illinois 1 Power Reactor Statistics
MIT NUCLEAR REACTOR LABORATORY. Technical Analysis and Administrative Issues of Criticality Study for Different MITR Facilities
MIT NUCLEAR REACTOR LABORATORY an MIT Interdepartmental Center Technical Analysis and Administrative Issues of Criticality Study for Different MITR Facilities Kaichao Sun, MIT-NRL Group Leader Reactor
Design 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
ANALYSIS OF AN EXTREME LOSS OF COOLANT IN THE IPR-R1 TRIGA REACTOR USING A RELAP5 MODEL
ANALYSIS OF AN EXTREME LOSS OF COOLANT IN THE IPR-R TRIGA REACTOR USING A RELAP MODEL P. A. L. Reis a, A. L. Costa a, C. Pereira a, M. A. F. Veloso a, H. V. Soares a, and A. Z. Mesquita b a Departamento
Conversion of MNSR (PARR-2) from HEU to LEU Fuel
Conversion of MNSR (PARR-2) from HEU to LEU Fuel Malik Tayyab Mahmood Nuclear Engineering Division Pakistan Institute of Nuclear Science & Technology, Islamabad PAKISTAN Pakistan Institute of Nuclear Science
Task 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,
DESIGN AND SAFETY CONSIDERATIONS FOR THE 10 MW(t) MULTIPURPOSE TRIGA REACTOR IN THAILAND. S. Proongmuang Office of Atomic Energy for Peace, Thailand 3
DESIGN AND SAFETY CONSIDERATIONS FOR THE 10 MW(t) MULTIPURPOSE TRIGA REACTOR IN THAILAND J. Razvi 1, J. M. Bolin, J. J. Saurwein, W. L. Whittemore General Atomics 2 S. Proongmuang Office of Atomic Energy
THE NEW 10 MW(T) MULTIPURPOSE TRIGA REACTOR IN THAILAND
THE NEW 10 MW(T) MULTIPURPOSE TRIGA REACTOR IN THAILAND J. RAZVI, J.M. BOLIN, A.R. VECA, W.L. WHITTEMORE TRIGA Reactors Group, General Atomics, San Diego, California, United States of America S. PROONGMUANG
Low Enriched Uranium Core Design for the Massachusetts Institute of Technology Reactor (MITR) with Un-finned 12 mil-thick Clad UMo Monolithic Fuel
Low Enriched Uranium Core Design for the Massachusetts Institute of Technology Reactor (MITR) with Un-finned 12 mil-thick Clad UMo Monolithic Fuel Nuclear Engineering Division About Argonne National Laboratory
A Comparison of the PARET/ANL and RELAP5/MOD3 Codes for the Analysis of IAEA Benchmark Transients
A Comparison of the /ANL and 5/MOD3 Codes for the Analysis of IAEA Benchmark Transients W. L. Woodruff, N. A. Hanan, R. S. Smith and J. E. Matos Argonne National Laboratory Argonne, Illinois 439-4841 U.S.A.
IAEA-TECDOC-643. Research reactor core conversion guidebook
IAEA-TECDOC-643 Research reactor core conversion guidebook Volume RESEARCH REACTOR CORE CONVERSION GUIDEBOOK VOLUME ALL PLEASE BE AWARE THAT FOREWORD In view of the proliferation concerns caused by the
Modelling 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.
Neutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar
Neutronics, Thermal Hydraulics and Safety Parameter Studies of the 3 MW TRIGA Research Reactor at AERE, Savar Md. Quamrul HUDA Energy Institute Atomic Energy Research Establishment Bangladesh Atomic Energy
Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code
Journal of Physics: Conference Series PAPER OPEN ACCESS Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code To cite this article: K Tiyapun and S
A Brief Summary of Analysis of FK-1 and FK-2 by RANNS
A Brief Summary of Analysis of FK- and by RANNS Motoe Suzuki, JAEA. Introduction For the purpose of benchmarking the RANS code, FK- and experiments conducted at NSRR were analyzed. Emphasis was placed
Effect 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
Analysis of Core Physics Test Data and Sodium Void Reactivity Worth Calculation for MONJU Core with ARCADIAN-FBR Computer Code System
FR09 - International Conference on Fast Reactors and Related Fuel Cycles Analysis of Core Physics Test Data and Sodium Void Reactivity Worth Calculation for MONJU Core with ARCADIAN-FBR Computer Code System
SAFETY ANALYSIS AND ASSESSMENT FOR TEMPORARY STORAGE OF 106 IRRADIATED VVR-M2 FUEL TYPE ASSEMBLIES IN DALAT REACTOR
SAFETY ANALYSIS AND ASSESSMENT FOR TEMPORARY STORAGE OF 106 IRRADIATED VVR-M2 FUEL TYPE ASSEMBLIES IN DALAT REACTOR NGUYEN MINH TUAN AND NGUYEN KIEN CUONG Nuclear Research Institute Contents 1. Introduction
Simulation of large and small fast reactors with SERPENT
Simulation of large and small fast reactors with SERPENT Janne Wallenius, Erdenechimeg Suvdantsetseg, Sara Bortot Milan Tesinsky & Youpeng Zhang Reactor Physics Kungliga Tekniska Högskolan R&D activities
Technology Considerations for Deployment of Thorium Power Reactors
Technology Considerations for Deployment of Thorium Power Reactors Matthias Krause International Atomic Energy Agency Email: M.Krause@iaea.org International Atomic Energy Agency Presentation Outline What?
European LEad-Cooled TRAining reactor: structural materials and design issues
Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials 12-14 JUNE 2013 IAEA HQ, VIENNA, AUSTRIA European LEad-Cooled TRAining reactor: structural materials and design
INCREASINGTHENEUTRONFLUXATTHEBEAMTUBE 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
TRANSITIONAL CORES AND FUEL CYCLE ANALYSES IN SUPPORT OF MIT REACTOR LOW ENRICHED URANIUM FUEL CONVERSION
TRANSITIONAL CORES AND FUEL CYCLE ANALYSES IN SUPPORT OF MIT REACTOR LOW ENRICHED URANIUM FUEL CONVERSION Kaichao Sun*, Akshay Dave, and Lin-wen Hu kaichao@mit.edu Nuclear Reactor Laboratory, Massachusetts
Nuclear Engineering and Design
Nuclear Engineering and Design 239 (2009) 45 50 Contents lists available at ScienceDirect Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes Flow blockage analysis of a
Presented by Sorin Margeanu
EMERGENCY INTERVENTION PLAN FOR 14 MW TRIGA - PITESTI RESEARCH REACTOR Sorin Margeanu, Marin Ciocanescu, Constantin Paunoiu Institute for Nuclear Research Pitesti, PO.Box-78, 115400-Mioveni, Romania Presented
Advanced Methods for BWR Transient and Stability Analysis. F.Wehle,S.Opel,R.Velten Framatome ANP GmbH P.O. BOX Erlangen Germany
Advanced Methods for BWR Transient and Stability Analysis F.Wehle,S.Opel,R.Velten Framatome ANP GmbH P.O. BOX 3220 91050 Erlangen Germany Advanced Methods for BWR Transient and Stability Analysis > Background
Argonne National Laboratory 9700 S. Cass Avenue, Bldg. 207 Argonne, Illinois USA. TtiIe: Coordinator for Analysis
. -AN ALTERNATVE LEU DESGN FOR THE FRM-ll N.A. Hanan, S.C.Mo, R.S. Smith and J. E. Matos Argonne National Laboratory 9700 S. Cass Avenue, Bldg. 207 Argonne, llinois 60439-4841 USA Contact J.E. Matos Address:
IRPhEP Benchmark Development Guidance for TREAT-IRP Activities
IRPhEP Benchmark Development Guidance for TREAT-IRP Activities John D. Bess Idaho National Laboratory TREAT-IRP Kickoff Meeting Idaho Falls, ID November 19-20, 2015 Outline TREAT, What s (Neutronically)
Benchmark Evaluation of Reactivity Measurements for Beryllium- Reflected Space Reactor Mockup
Benchmark Evaluation of Reactivity Measurements for Beryllium- Reflected Space Reactor Mockup Margaret A. Marshall Idaho National Laboratory John D. Bess Idaho National Laboratory NETS 2015 NETS 2014 Infinity
Experimental Measurements for Plate Temperatures of MTR Fuel Elements at Sudden Loss of Flow Accident and Comparison with Computed Results
Experimental Measurements for Plate Temperatures of MTR Fuel Elements at Sudden Loss of Flow Accident and Comparison with Computed Results Dr. Bülent SEVDIK TR 2 Reactor Unit Head Turkish Atomic Energy
Fast Reactor Operating Experience in the U.S.
Fast Reactor Operating Experience in the U.S. Harold F. McFarlane Deputy Associate Laboratory Director for Nuclear Science and Technology www.inl.gov 3 March 2010 [insert optional photo(s) here] Thanks
Flexible Conversion Ratio Fast Reactor
American Nuclear Society Student Conference March 29-31, 2007, Oregon State University, Corvallis, OR Flexible Conversion Ratio Fast Reactor Anna Nikiforova Massachusetts Institute of Technology Center
Numerical Modeling and Calculation of the Fuel Cycle for the IRT-Sofia Research Reactor
Bulg. J. Phys. 40 (2013) 281 288 Numerical Modeling and Calculation of the Fuel Cycle for the IRT-Sofia Research Reactor D. Dimitrov, S. Belousov, K. Krezhov, M. Mitev Institute for Nuclear Research and
Design of High Power Density Annular Fuel Rod Core for Advanced Heavy Water. Reactor
Design of High Power Density Annular Fuel Rod Core for Advanced Heavy Water Reactor For the deployment of annular fuel rod cluster in AHWR, whole core calculations with annular fuel rod are necessary.
Experiments 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
Reactivity insertions for the Borax accident in ORPHEE research reactor
Reactivity insertions for the Borax accident in ORPHEE research reactor September 2010, 1X th / IGORR Yacine Chegrani*, Florence Gupta, Franck Bernard IRSN Plan of the Presentation Introduction Context
Calculation of Fuel Temperature Coefficient, Reactivity Loss and Temperature of BAEC TRIGA Research Reactor
Asian Journal of Applied Science and Engineering, Volume 6, No 2/2017 ISSN 2305-915X(p); 2307-9584(e) Calculation of Fuel Temperature Coefficient, Reactivity Loss and Temperature of BAEC TRIGA Research
Primary - 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
Research Article Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5
Hindawi Science and Technology of Nuclear Installations Volume 2017, Article ID 2615409, 10 pages https://doi.org/10.1155/2017/2615409 Research Article Comparative Analysis of the Dalat Nuclear Research
Evaluation of Implementation 18-Month Cycle in NPP Krško
International Conference Nuclear Energy for New Europe 2003 Portorož, Slovenia, September 8-11, 2003 http://www.drustvo-js.si/port2003 Evaluation of Implementation 18-Month Cycle in NPP Krško Martin Novšak,
Douglas Borges Domingos, Antonio Teixeira e Silva, Pedro Ernesto Umbehaun, José Eduardo Rosa da Silva, Thadeu das Neves Conti and Mitsuo Yamaguchi
2009 International Nuclear Atlantic Conference - INAC 2009 Rio de Janeiro,RJ, Brazil, September27 to October 2, 2009 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-03-8 NEUTRONIC, THERMAL-HYDRAULIC
ABSTRACT. 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
Pre-Conceptual Core Design of a LBE-Cooled Fast Reactor (BLESS) Ziguan Wang, Luyu Zhang, Eing Yee Yeoh, Linsen Li, Feng Shen
Pre-Conceptual Core Design of a LBE-Cooled Fast Reactor (BLESS) Ziguan Wang, Luyu Zhang, Eing Yee Yeoh, Linsen Li, Feng Shen State Power Investment Corporation Research Institute, Beijing 102209, P. R.
PARTIAL SAFETY ANALYSIS FOR A REDUCED URANIUM ENRICHMENT CORE FOR THE HIGH FLUX ISOTOPE REACTOR
Joint International Workshop: Nuclear Technology and Society Needs for Next Generation Berkeley, California, January 6-8, 2008, Berkeley Faculty Club, UC Berkeley Campus PARTIAL SAFETY ANALYSIS FOR A REDUCED
Nuclear Research Institute. Nuclear Research Institute CONTENTS
CONTENTS Nuclear Organizational Structure in VN Brief Introduction to the DNRR: - History of the reactor - Operation and Utilization of the reactor Decommissioning Planning 1 THE ORGANIZATION CHART OF
OECD Transient Benchmarks: Preliminary Tinte Results TINTE Preliminary Results
OECD Transient Benchmarks: Preliminary Tinte Results Presentation Overview The use of Tinte at PBMR Tinte code capabilities and overview Preliminary Tinte benchmark results (cases1-6) The use of Tinte
WASHINGTON STATE UNIVERSITY REACTOR LICENSE NO. R-76 DOCKET NO SAFETY ANALYSIS REPORT FOR THE HEU TO LEU CONVERSION OF THE REACTOR
WASHINGTON STATE UNIVERSITY REACTOR LICENSE NO. R-76 DOCKET NO. 50-27 SAFETY ANALYSIS REPORT FOR THE HEU TO LEU CONVERSION OF THE WASHINGTON STATE UNIVERSITY REACTOR REDACTED VERSION SECURITY-RELATED INFORMATION
Trends in Transmutation Performance and Safety Parameters Versus TRU Conversion Ratio of Sodium-Cooled Fast Reactors
Trends in Transmutation Performance and Safety Parameters Versus TRU Conversion Ratio of Sodium-Cooled Fast Reactors The Tenth OECD Nuclear Energy Agency Information Exchange Meeting on Actinide and Fission
Examples of Research Reactor Conversion Assessment of Alternatives
Examples of Research Reactor Conversion Assessment of Alternatives Benoit Dionne, Ph.D. Section Manager - Conversion Analysis and Methods Nuclear Engineering Division, Argonne National Laboratory National
Tools and applications for core design and shielding in fast reactors
Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials, June 12-14, 2013 Tools and applications for core design and shielding in fast reactors Presented by: Reuven Rachamin
Numerical Benchmark Results for 1000MWth Sodium-cooled Fast Reactor
Numerical Benchmark Results for 1000MWth Sodium-cooled Fast Reactor T. K. Kim and T. A. Taiwo Argonne National Laboratory February 13, 2012 Second Meeting of SFR Benchmark Task Force of Working Party on
SERPENT activities at VUJE, a.s.
SERPENT activities at VUJE, a.s. Tomáš Chrebet Amine Bouhaddane František Čajko Michal Sečanský Radoslav Zajac 26-29 September 2016 6th International Serpent User Group Meeting at Milan SERPENT activities
DESIGN AND SAFETY- SUPPORT ANALYSES OF AN IN-PILE MOLTEN SALT LOOP IN THE HFR
DESIGN AND SAFETY- SUPPORT ANALYSES OF AN IN-PILE MOLTEN SALT LOOP IN THE HFR stempniewicz@nrg.eu M.M. Stempniewicz, E.A.R. de Geus, F. Alcaro, P.R. Hania, K. Nagy, N.L. Asquith, J. de Jong, L. Pool, S.
Inspection of PUSPATI 1 Reactor (RTP) Core and Control Rod
TM Assessment of Core Structural Materials and Surveillance Programme for Research Reactors, 14-18 June 2010, Vienna, Austria Inspection of PUSPATI 1 TRIGA Reactor (RTP) Core and Control Rod Zarina Masood
PUSPATI TRIGA REACTOR UPGRADING: TOWARDS THE SAFE OPERATION & FEASIBILITY OF NEUTRONIC APPROACH
Invited Paper B03 PUSPATI TRIGA REACTOR UPGRADING: TOWARDS THE SAFE OPERATION & FEASIBILITY OF NEUTRONIC APPROACH J. ABDUL KARIM, M.P. ABU, Z. MASOOD, M.H. RABIR, M.A.S. SALLEH, M.F ZAKARIA Malaysia Nuclear
PULSED TRIGA REACTOR AS SUBSTITUTE FOR LONG PULSE SPALLATION NEUTRON SOURCE. W. L. Whittemore General Atomics 1
PULSED TRIGA REACTOR AS SUBSTITUTE FOR LONG PULSE SPALLATION NEUTRON SOURCE W. L. Whittemore General Atomics 1 ABSTRACT TRIGA reactor cores have been used to demonstrate various pulsing applications. The
ROLE OF RELAP/SCDAPSIM IN RESEARCH REACTOR SAFETY ABSTRACT
ROLE OF RELAP/SCDAPSIM IN RESEARCH REACTOR SAFETY C. M. ALLISON, J. K. HOHORST, Innovative Systems Software, LLC Idaho Falls, Idaho 83404 USA iss@relap.com A. J. D ARCY South African Nuclear Energy Corporation
Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor
FR13 - TECHNICAL SESSION 3.5: Fast reactor safety: post-fukushima lessons and goals for next-generation reactors Paper n. IAEA-CN-199/260 Safety Analysis Results of Representative DEC Accidental Transients
Sodium 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
Fuel Management Effects on Inherent Safety of Modular High Temperature Reactor
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 26[7], pp. 647~654 (July 1989). 647 Fuel Management Effects on Inherent Safety of Modular High Temperature Reactor Yukinori HIROSEt, Peng Hong LIEM, Eiichi SUETOMI,
MONTE CARLO CALCULATIONS ON THE FIRST CRITICALITY OF THE MULTIPURPOSE REACTOR G.A. SIWABESSY. Liem Peng Hong Center for Multipurpose Reactor - BATAN
MONTE CARLO CALCULATIONS ON THE FIRST CRITICALITY OF THE MULTIPURPOSE REACTOR G.A. SIWASSY Liem Peng Hong Center for Multipurpose Reactor - BATAN ABSTCT MONTE CARLO CALCULATIONS ON THE FIRST CRITICALITY
PARAMETRIC STUDY OF THERMO-MECHANICAL BEHAVIOUR OF 19- ELEMENT PHWR FUEL BUNDLE HAVING AHWR FUEL MATERIAL
PARAMETRIC STUDY OF THERMO-MECHANICAL BEHAVIOUR OF 19- ELEMENT PHWR FUEL BUNDLE HAVING AHWR FUEL MATERIAL R. M. Tripathi *, P. N. Prasad, Ashok Chauhan Fuel Cycle Management & Safeguards, Directorate of
ARTICLE IN PRESS. Nuclear Engineering and Design xxx (2009) xxx xxx. Contents lists available at ScienceDirect. Nuclear Engineering and Design
Nuclear Engineering and Design xxx (2009) xxx xxx Contents lists available at ScienceDirect Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes Thermal hydraulic analysis
Design Study of Innovative Simplified Small Pebble Bed Reactor
Design Study of Innovative Simplified Small Pebble Bed Reactor Dwi Irwanto 1* and Toru OBARA 2 1 Department of Nuclear Engineering, Tokyo Institute of Technology 2 Research Laboratory for Nuclear Reactors,
Full 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
ON-LINE MONITORING OF THE REACTIVITY AND CONTROL RODS WORTH AT THE IPR-R1 TRIGA REACTOR
2007 International Nuclear Atlantic Conference - INAC 2007 Santos, SP, Brazil, September 30 to October 5, 2007 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-02-1 ON-LINE MONITORING
Development of a software-based system for modelling research reactors using heuristics
Development of a software-based system for modelling research reactors using heuristics VERA KOPPERS Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh Boltzmannstraße 14, 85748 Garching bei München,
NOVEL UTILIZATION OF TRIGA REACTORS
Preliminary Safety Analysis Report (PSAR): NOVEL UTILIZATION OF TRIGA REACTORS FOR ISOTOPE PRODUCTION (NUTRIP) MO- 99 PRODUCTION USING TRIGA REACTORS NE 267: Nuclear Reactor Safety Professor Per F. Peterson
Validation of the Monte Carlo Code MVP on the First Criticality of Indonesian Multipurpose Reactor
Validation of the Monte Carlo Code MVP on the First Criticality of Indonesian Multipurpose Reactor T.M. Sembiring, S. Pinem, Setiyanto Center for Reactor Technology and Nuclear Safety,PTRKN-BATAN, Serpong,
Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor
Journal of Physics: Conference Series PAPER OPEN ACCESS Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor To cite this article: Endiah Puji Hastuti et al 2018 J. Phys.:
Preliminary Results of Three Dimensional Core Design in JAPAN
Preliminary Results of Three Dimensional Core Design in JAPAN Information Exchange Meeting on SCWR Development April 29, 2003 Toshiba Corporation The University of Tokyo Scope of SCWR Core Design (in Short
Convert-and-Upgrade Strategies for Research Reactors
Convert-and-Upgrade Strategies for Research Reactors Alexander Glaser Department of Mechanical and Aerospace Engineering and Woodrow Wilson School of Public and International Affairs Princeton University
The new material irradiation infrastructure at the BR2 reactor. Copyright 2017 SCK CEN
The new material irradiation infrastructure at the BR2 reactor The new material irradiation infrastructure at the BR2 reactor Steven Van Dyck, Patrice Jacquet svdyck@sckcen.be Characteristics of the BR2
Evaluation of the Surface Launch of a SingleStage-to-Orbit Nuclear Thermal Rocket
Evaluation of the Surface Launch of a SingleStage-to-Orbit Nuclear Thermal Rocket Jeffrey C. King and Satira I. Labib Nuclear Science and Engineering Program Metallurgical and Materials Engineering Department
Thermal Fluid Characteristics for Pebble Bed HTGRs.
Thermal Fluid Characteristics for Pebble Bed HTGRs. Frederik Reitsma IAEA Course on High temperature Gas Cooled Reactor Technology Beijing, China Oct 22-26, 2012 Overview Background Key T/F parameters
Neutronics Simulations of 237 Np Targets to Support Safety-Basis and 238 Pu Production Assessment Efforts at the High Flux Isotope Reactor
Neutronics Simulations of 237 Np Targets to Support Safety-Basis and 238 Pu Production Assessment Efforts at the High Flux Isotope Reactor David Chandler and R. J. Ellis chandlerd@ornl.gov Oak Ridge National
Use of SERPENT at PSI/EPFL for thermal system analysis.
Paul Scherrer Institut L. Rossinelli, A. Rais, M. Hursin, H. Ferroukhi Wir schaffen Wissen heute für morgen Use of SERPENT at PSI/EPFL for thermal system analysis. Serpent User s Meeting, Cambridge, September
An Overview of the ACR Design
An Overview of the ACR Design By Stephen Yu, Director, ACR Development Project Presented to US Nuclear Regulatory Commission Office of Nuclear Reactor Regulation September 25, 2002 ACR Design The evolutionary
Inventory Prediction and BUC Calculations Related to MEU/LEU IRT Fuels of LVR-15 Research Facility
Inventory Prediction and BUC Calculations Related to MEU/LEU IRT Fuels of LVR-15 Research Facility L. Markova, F. Havluj, M. Marek Nuclear Research Institute at Rez, Czech Republic International Workshop
SAND # SAND C
Annular Core Research Reactor Operational Overview SAND # SAND2016-7720 C TRTR Conference Aug 21-25, 2016 David Clovis, P.E., Facility Supervisor Lonnie Martin, P.E., Reactor Supervisor Sandia is a multiprogram
Analysis of Unprotected Transients in the Lead-Cooled ALFRED Reactor
Analysis of Unprotected Transients in the Lead-Cooled ALFRED Reactor G. Bandini (ENEA/Bologna) E. Bubelis, M. Schikorr (KIT/Karlsruhe) A. Alemberti, L. Mansani (Ansaldo Nucleare/Genova) Consultants Meeting:
UTILIZATION OF ETRR-2 AND COLLABORATION 1. INTRODUCTION
UTILIZATION OF ETRR-2 AND COLLABORATION M.K. SHAAT Nuclear Energy Centre, Egypt s Atomic Energy Authority (EAEA), Cairo, Egypt m_shaat30@hotmail.com 1. INTRODUCTION Owners and operators of many research
Technical University of Sofia, Department of Thermal and Nuclear Power Engineering, 8 Kliment Ohridski Blvd., 1000 Sofia, Bulgaria
BgNS TRANSACTIONS volume 20 number 2 (2015) pp. 143 149 Comparative Analysis of Nodalization Effects and Their Influence on the Results of ATHLET Calculations of VVER-1000 Coolant Transient Benchmark Phase
THE 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
Transmutation of Transuranic Elements and Long Lived Fission Products in Fusion Devices Y. Gohar
Transmutation of Transuranic Elements and Long Lived Fission Products in Fusion Devices Y. Gohar Fusion Power Program Technology Division Argonne National Laboratory 9700 S. Cass Avenue, Argonne, IL 60439,
nuclear science and technology
EUROPEAN COMMISSION nuclear science and technology Co-ordination and Synthesis of the European Project of Development of HTR Technology (HTR-C) Contract No: FIKI-CT-2000-20269 (Duration: November 2000
PROPOSAL OF A GUIDE TO PERFORMANCE ASSESSMENT OF FUEL RODS FOR NUCLEAR POWER PLANTS
2013 International Nuclear Atlantic Conference - INAC 2013 Recife, PE, Brazil, November 24-29, 2013 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-05-2 PROPOSAL OF A GUIDE TO PERFORMANCE
Nuclear Research Reactor TRIGA IAN-R1. Jaime Sandoval Lagos Supervisor Reactor Nuclear October 2014
Nuclear Research Reactor TRIGA IAN-R1 Jaime Sandoval Lagos Supervisor Reactor Nuclear October 2014 INTRODUCTION The Nuclear Reactor IAN-R1 went critical on February 1965. It was run by the Instituto de
Development of Decommissioning Plan in Indonesia
Development of Decommissioning Plan in Indonesia Endang Kurnia and Berthie Isa R²D ²P: Review of a Decommissioning Plan Bucharest, Romania, 4-8 July 2011 IAEA International Atomic Energy Agency Legal and