IMPROVED BWR CORE DESIGN USING HYDRIDE FUEL
|
|
- Megan Nicholson
- 6 years ago
- Views:
Transcription
1 Joint Reactor Seminar University of Tokyo (GoNERI) and UC Berkeley March 5, 2009 IMPROVED BWR CORE DESIGN USING HYDRIDE FUEL Massimiliano Fratoni, Alessandro Piazza, Ehud Greenspan University of California, Berkeley Paolo Ferroni, Neil Todreas Massachusetts Institute of Technology
2 OUTLINE Work motivation Hydride fuel BWR core designs Analysis methodology Hydride fuel BWRs performance Neutronic analysis Thermo-hydraulic analysis Stability analysis 2
3 OXIDE FUEL BWR CORE IS HIGHLY HETEROGENEOUS Two phase moderator Reference model GE11 BWR/5 9x9 Need for extra moderation Water rods Partial length fuel rods E6 Water gaps in between bundles High level of heterogeneity Axial and radial enrichment distribution E1 E2/E3 E4/E5 E5/E6 Partial E5 Partial E6 Fuel rods enrichment levels E1<E2< <E6 Gd E4 3
4 HYDRIDE FUEL FOR BWRS Hydride fuel examined U(45w/o)-ZrH1.6 High hydrogen density similar to liquid water at 0.7 g/cm 3 High thermal conductivity about 5 times that of UO 2 Low fission gas release But relatively large swelling use LM bonding Incentives for using hydride fuel in BWR No need for the extra moderation Eliminate water rods and partial length rods to insert fuel rods Minimize thickness of water gaps for larger bundles Reduce heterogeneity of BWR core Reduce number of enrichments per bundle 4
5 HYDRIDE FUEL BUNDLE DESIGN FOR UPPER BOUND ANALYSIS Upper bound estimate on improvement possibilities are obtained by examining a design that is not practical to implement in existing BWR designs Same core volume per bundle as oxide fuel Same OD as in oxide fuel 10x10 array 5% enrichment uniform Dispersed control rods Optimal P/D to maximize BU from 1-D analysis 5
6 FUEL BUNDLE DESIGNS COMPATIBLE WITH THE CURRENT BWR CORE LAYOUT Corner cruciform control rods CCCR 2 cruciform corner control rods 1 square corner position for instrumentations 93 full length fuel rods Control blades CB 1 control blade every 4 bundles Reduced canisters distance 100 full length fuel rods 6
7 NEUTRONIC AND THERMAL-HYDRAULIC ANALYSIS Bundle-level neutronic analysis (3-D) Fix axial water density distribution (24 points) 9 zones depletion analysis (3 axial and 3 radial) 4-batch refueling scheme with batch dependent power (5.40, 4.73, 4.17, 2.95 MW per bundle) Uncoupled from thermal-hydraulic analysis Sub-channel thermal-hydraulic analysis Stability analysis 7
8 ATTAINABLE BURN-UP FOR THE UPPER BOUND DESIGN Hydride compared to oxide fuel Achieves similar burn-up Decreases the cycle length Parameter Oxide 9x9 Hydride 10x10 Number of fuel rods ~71 96 Number of control rods Control blades 4 P/D Pellet diameter (cm) Average enrichment 3.90% 5% Initial HM mass ratio Single batch BOC k eff Burn-up (GWd/tHM) Fuel residence time (EFPD)
9 POWER PEAKING FACTOR FOR THE UPPER BOUND DESIGN Pin-by-pin power distribution in hydride fuel is extremely flat 1.04 peak 9
10 CONTROL SYSTEM FOR THE UPPER BOUND DESIGN Dispersed control rods PWR s alike Parameters: Number of rods Rods diameter Requirement: Shut-down reactivity margin at least as with control blades in the reference oxide fuel core Parameter Oxide 9x9 Hydride 10x10 Reactivity control system Control blades Control rods 4 control rods (B 4 C) per bundle Control system reactivity worth at hot full power ($) Control system reactivity worth at cold shut-down ($) Diameter slightly larger than fuel rods 10
11 HYDRIDE CORE DESIGNS COMPARISON Parameter Hydride 10x10 (upper bound) CCCR CB Total bundle unit width (cm) Bundles distance (cm) /1.22 Fuel rod OD (cm) P/D Number of fuel rods Control system 4 dispersed control rods 2 corner cruciform control rods Control blades Neutron absorber Natural B 90% enriched B Natural B BOC zero power cold shutdown margin relative to oxide -0.56% +0.00% -0.22% HM load (kg/bundle) Burnup (GWd/tHM) Fuel residence time (EFPD) Radial peaking factor
12 HYDRIDE BUNDLE DESIGNS ARE OPTIMIZED BY ENRICHMENT DISTRIBUTION Constraints from reference oxide design Cycle length Control system worth Power peaking factor Methodology Three axial enrichments As many as needed radial enrichments 12
13 HYDRIDE BUNDLE DESIGNS ARE OPTIMIZED BY ENRICHMENT DISTRIBUTION Cruciform corner control rods Axial Zone Rod A Rod B Rod C Rod D Rod E % 8.27% 6.53% 5.44% 4.79% % 7.40% 5.66% 4.57% 3.92% % 6.53% 4.88% 3.70% 3.05% 13
14 HYDRIDE BUNDLE DESIGNS ARE OPTIMIZED BY ENRICHMENT DISTRIBUTION Control blades Axial Zone Rod A Rod B Rod C Rod D Rod E Rod F Rod G Rod H % 8.15% 6.34% 5.43% 5.43% 4.98% 6.80% 5.89% % 7.70% 5.89% 4.98% 4.53% 4.08% 6.57% 5.44% % 7.25% 5.43% 4.53% 4.08% 3.62% 6.34% 4.98% 14
15 CYCLE LENGTH AND PEAK POWER Parameter Oxide GE14 10x10 CCCB Average enrichment (%) CB Burnable poison Gadolinia IFBA IFBA Average burnable poison load (wt%) 5 (12 rods) 0.50 (uniform) 0.48 (uniform) Cycle length (EFPD) 1,465 1,465 1,465 BOL radial peaking factor BOL axial peaking factor
16 CONTROL SYSTEM Parameter Control system Maximum excess reactivity ($) Reactivity worth at hot full power ($) Reactivity worth at cold shut-down ($) Oxide GE14 10x10 Control blades CCCB Cruciform control rods CB Control blades Shut-down margin ($)
17 REACTIVITY COEFFICIENTS Coolant void (pcm/void%) Time Oxide GE14 10x10 CCCB CB BOC EOC Fuel temperature (pcm/k) Time Oxide GE14 10x10 CCCB CB BOC EOC
18 METHODOLOGY FOR THE THERMAL- HYDRAULIC ANALYSIS Part I: whole core analysis in search for D and P/D giving maximum core power in the range 0.6 < D < 1.6 cm 1.1 < P/D < 1.6 Part II: sub-channel analysis of selected geometries Constraints Parameter Oxide bundle Hydride bundles MCHFR (via EPRI-1 correlation) Fuel Centerline T (ºC) at Steady State Fuel Average T (ºC) at Steady State 1400 Not Applied Clad Surface T (ºC) at Steady State Bundle ΔP (MPa) Subchannel Average Exit Quality (%)
19 MAXIMUM POWER ACHIEVABLE WITH A MPa LIMIT ON THE PRESSURE DROP Parameter Bundle Active Flow Rate (kg/s) MCHFR Max Peak Fuel Temperature ( o C) Max Average Fuel Temperature ( o C) Max Surface Clad Temperature ( o C) Bundle Pressure Drop (MPa) Subchannel-averaged Exit Quality (%) Bundle Power (kw) % difference vs. Oxide 10x10 % difference vs. Hyd. 10x10 Oxide Hydride Hydride CCCR Hydride CB (1.213) 1,568 (2,805) 1,046 (1,400) 310 (349) (0.147) (1.213) 497 (750) Not Applied 306 (349) (0.147) (1.213) 477 (750) Not Applied 305 (349) (0.147) (1.213) 462 (750) Not Applied 304 (349) (0.147) , , , ,
20 MAXIMUM POWER ACHIEVABLE WITH A MPa LIMIT ON THE PRESSURE DROP Parameter Bundle Active Flow Rate (kg/s) MCHFR Max Peak Fuel Temperature ( o C) Max Average Fuel Temperature ( o C) Max Surface Clad Temperature ( o C) Bundle Pressure Drop (MPa) Subchannel-averaged Exit Quality (%) Bundle Power (kw) % difference vs. Oxide 10x10 % difference vs. Hyd. 10x10 Oxide Hydride Hydride CCCR Hydride CB (1.213) 1,873 (2,805) 1,200 (1,400) 309 (349) (0.220) (1.213) 551 (750) Not Applied 310 (349) (0.220) (1.213) 527 (750) Not Applied 309 (349) (0.220) (1.213) 509 (750) Not Applied 308 (349) (0.220) , , , ,
21 MAXIMUM POWER ACHIEVABLE WITH NO LIMIT ON THE PRESSURE DROP Parameter Bundle Active Flow Rate (kg/s) MCHFR Max Peak Fuel Temperature ( o C) Max Average Fuel Temperature ( o C) Max Surface Clad Temperature ( o C) Oxide Hydride Hydride CCCR Hydride CB (1.213) 1,873 (2,805) 1,200 (1,400) 309 (349) (1.213) 551 (750) Not Applied 310 (349) (1.213) 563 (750) Not Applied 311 (349) (1.213) 527 (750) Not Applied 309 (349) Bundle Pressure Drop (MPa) Subchannel-averaged Exit Quality (%) Bundle Power (kw) % difference vs. Oxide 10x10 % difference vs. Hyd. 10x , , , ,
22 FLOW STABILITY WITHOUT PARTIAL LENGTH FUEL RODS Hydride fuel bundles w/o PLFR have better stability than the oxide fuel bundles with PLFR Due to higher thermal conductivity of hydride fuel and liquid metal compared to oxide fuel and helium Parameter Reference oxide 9x9 p limit 0.156/0.234 MPa Hydride 10x10 p limit 0.156/0.234 MPa Decay Ratio 0.111/ /0.094 Decay Ratio if Ox and He /0.126 Decay Ratio if Ox and LM /
23 CONCLUSIONS Use of hydride fuel in BWR s can ideally Greatly simplify the fuel bundle design Eliminating water rods and partial length fuel rods Reducing number of enrichment levels Increasing number of fuel rods per unit core volume by 35% Increase core power density by ~40% By increasing the reactor power level by ~40% By reducing the core height (volume) by ~40% May improve stability Designs compatible with the current BWR core design still offer ~20% power increase compared to oxide fuel 23
24 BACK-UP SLIDES
25 HP-BWR FEATURES CONTROL RODS INSERTED FROM THE VESSEL HEAD Frigyes Reisch, Concept of a future high pressure-boiling water reactor, HP-BWR, International Topical Meeting on Safety of Nuclear Installations, Dubrovnik, Croatia,
Neutronics of Prismatic Fluoride Salt Cooled High Temperature Reactors
Neutronics of Prismatic Fluoride Salt Cooled High Temperature Reactors Workshop on Advanced Reactors Concepts PHYSOR 2012 Knoxville, TN April 15, 2012 Dan Ilas IlasD@ornl.gov Main Neutronic Design Characteristics
More information1: CANDU Reactor. B. Rouben McMaster University Nuclear Reactor Physics EP 4D03/6D Sept-Dec September 1
1: CANDU Reactor B. Rouben McMaster University Nuclear Reactor Physics EP 4D03/6D03 2015 Sept-Dec 2015 September 1 Outline A quick look at the design of CANDU Reactors: Reactor Assembly Pressure Tubes
More informationSPERT-III: REACTIVITY INSERTION ANALYSIS WITH SIMULATE-3K
SPERT-III: REACTIVITY INSERTION ANALYSIS WITH SIMULATE-3K Gerardo Grandi 2012 International Users Group Meeting Charlotte, NC, USA May 2-3, 2012 2 Overview Introduction Special Power Excursion Reactor
More informationA Helium Cooled Particle Fuelled Reactor for Fuel Sustainability. T D Newton, P J Smith, Y Askan. Serco Assurance. Work Sponsored by BNFL
Serco Assurance A Helium Cooled Particle Fuelled Reactor for Fuel Sustainability T D Newton, P J Smith, Y Askan Work Sponsored by BNFL Background New generation of reactor designs to meet the needs of
More informationSingle-phase Coolant Flow and Heat Transfer
22.06 ENGINEERING OF NUCLEAR SYSTEMS - Fall 2010 Problem Set 5 Single-phase Coolant Flow and Heat Transfer 1) Hydraulic Analysis of the Emergency Core Spray System in a BWR The emergency spray system of
More informationFuel cladding stress evaluation for the control rods pattern change at the Shimane Nuclear Power Station Unit No1 and No2
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1086 Fuel cladding stress evaluation for the control rods pattern change at the Shimane Nuclear Power Station Unit No1 and No2 Saemi Shimatani 1*, Masayuki
More informationModule 09 Heavy Water Moderated and Cooled Reactors (CANDU)
Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Module 09 Heavy Water Moderated and Cooled Reactors (CANDU) 1.10.2016
More informationDesign and Performance of South Ukraine NPP Mixed Cores with Westinghouse Fuel
National Science Center "Kharkov Institute of Physics and Technology (NSC KIPT) Design and Performance of South Ukraine NPP Mixed Cores with Westinghouse Fuel A.M. Abdullayev, V.Z. Baidulin, A.I. Zhukov
More informationAbout Reasonably Achievable Balance between Economy and Safety indices in WWERs
IAEA INPRO DF8, Vienna 26-29 August 2014 About Reasonably Achievable Balance between Economy and Safety indices in WWERs Grigory Ponomarenko OKB GIDROPRESS Podolsk, Russian Federation Contents 1. Safety
More informationFission gas release and temperature data from instrumented high burnup LWR fuel
Fission gas release and temperature data from instrumented high burnup LWR fuel XA0202217 T. Tverberg, W. Wiesenack Institutt for Energiteknikk, OECD Halden Reactor Project, Norway Abstract.The in-pile
More informationStatus of HPLWR Development
Status of HPLWR Development Thomas Schulenberg SCWR System Steering Committee Karlsruhe Institute of Technology Germany What is a Supercritical Water Cooled Reactor? PH HP IP LP PH Produces superheated
More informationReport No. IDO APPENDIX B ML-1 PLANT CHARACTERISTICS 1. GENERAL Mu to gas; 3.41 Mw total Cycle efficiency. Gross elect. wr. Net elect.
... Report No. IDO-28653 APPENDIX B ML-1 PLANT CHARACTERISTICS 0 Design performance at 100 F Gross electrical output Net electrical output 1. GENERAL Reactor thermal power 2.98 Mu to gas; 3.41 Mw total
More informationACCIDENT TOLERANT FUEL AND RESULTING FUEL EFFICIENCY IMPROVEMENTS
Advances in Nuclear Fuel Management V (ANFM 2015) Hilton Head Island, South Carolina, USA, March 29 April 1, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) ACCIDENT TOLERANT FUEL AND
More informationAP1000 European 4. Reactor Design Control Document CHAPTER 4 REACTOR. 4.1 Summary Description
CHAPTER 4 REACTOR 4.1 Summary Description This chapter describes the mechanical components of the reactor and reactor core, including the fuel rods and fuel assemblies, the nuclear design, and the thermal-hydraulic
More informationIntroduction and Summary
6 Chapter Core and Fuel Design Introduction and Summary The design of the Advanced Boiling Water Reactor (ABWR) core and fuel is based on the proper combination of many design variables and operating experience.
More informationPOWER RAMPING AND CYCLING TESTING OF VVER FUEL RODS IN THE MIR REACTOR
POWER RAMPING AND CYCLING TESTING OF VVER FUEL RODS IN THE MIR REACTOR A.G. Eshcherkin*, V.A. Ovchinnikov, E.E. Shakhmut, E.E. Kuznetsova, A.L. Izhutov, V.V. Kalygin INTRODUCTION A series of experiments
More informationThermal Hydraulics Design Limits Class Note II. Professor Neil E. Todreas
Thermal Hydraulics Design Limits Class Note II Professor Neil E. Todreas The following discussion of steady state and transient design limits is extracted from the theses of Carter Shuffler and Jarrod
More informationFBR and ATR fuel developments in JNC
International Seminar on MOX Utilization February 19, 2002 FBR and ATR fuel developments in JNC Design and performance of MOX fuel in safety view points Plutonium Fuel Center, Tokai Works, Japan Nuclear
More informationCNS Fuel Technology Course: Fuel Design Requirements
4525 Lakeshore Road Burlington, Ontario L7L 1B3 Phone: 905-639-4090 FAX: 905-639-9506 CNS Fuel Technology Course: Fuel Design Requirements Al Manzer, B.Sc., M. Eng. Senior Fuel Specialist CANTECH Associates
More informationTREAT Startup Update
Resumption of Transient Testing Program TREAT Startup Update John D. Bumgardner Director, Resumption of Transient Testing Program December 5 th, 2018 1 Facility Location 2 Nuclear Fuels Development Requires
More informationNATIONAL NUCLEAR SECURITY ADMINISTRATION GLOBAL THREAT REDUCTION INITIATIVE Core Modifications to address technical challenges of conversion
NATIONAL NUCLEAR SECURITY ADMINISTRATION GLOBAL THREAT REDUCTION INITIATIVE Core Modifications to address technical challenges of conversion G17 G18 G19 G20 G21 G16 F15 F16 G22 F17 F14 9.76 9.85 9.91 F18
More informationRecent Predictions on NPR Capsules by Integrated Fuel Performance Model
Massachusetts Institute of Technology Department of Nuclear Engineering Advanced Reactor Technology Pebble Bed Project Recent Predictions on NPR Capsules by Integrated Fuel Performance Model Jing Wang
More informationSMR multi-physics calculations with Serpent at VTT
VTT TECHNICAL RESEARCH CENTRE OF FINLAND LTD SMR multi-physics calculations with Serpent at VTT Serpent UGM 2016 Riku Tuominen, VTT Outline Serpent-COSY coupling Future work 18/10/2016 2 COSY Three-dimensional
More informationModule 03 Pressurized Water Reactors (PWR) Generation 3+
Module 03 Pressurized Water Reactors (PWR) Generation 3+ 1.3.2017 Prof.Dr. Böck Technical University Vienna Atominstitut Stadionallee 2, 1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at Flow
More informationImpact of configuration variations on small modular reactor core performance
Scholars' Mine Masters Theses Student Research & Creative Works Spring 2015 Impact of configuration variations on small modular reactor core performance William Kirby Compton Follow this and additional
More informationSFR CORE DESIGN PERFORMANCE AND SAFETY
SFR CORE DESIGN PERFORMANCE AND SAFETY A. VASILE European Nuclear Education Network Association Gen IV - INSTN Alfredo Vasile 19 SEPTEMBER 2012 13 SEPTEMBRE 2012 CEA 10 AVRIL 2012 PAGE 1 OUTLINE GEN IV
More informationSUB-CHAPTER B.3 COMPARISON TABLE COMPARISON WITH REACTORS OF SIMILAR DESIGN (N4 AND KONVOI)
CHAPTER B: INTRODUCTION AND GENERAL DESCRIPTION OF THE SUB-CHAPTER: B.3 SECTION : - PAGE : 1 / 9 SUB-CHAPTER B.3 COMPARISON TABLE COMPARISON WITH REACTORS OF SIMILAR DESIGN (N4 AND KONVOI) A comparison
More informationThermal analysis of IRT-T reactor fuel elements
Thermal analysis of IRT-T reactor fuel elements A Naymushin, Yu Chertkov, I Lebedev and M Anikin National Research Tomsk Polytechnic University, TPU, Tomsk, Russia E-mail: agn@tpu.ru Abstract. The article
More informationTHE ROMANIAN IRRADIATION TESTS PREPARED FOR IAEA/OECD DATABASE Author: M. C. Paraschiv, Institute for Nuclear Research Pitesti, Romania
THE ROMANIAN IRRADIATION TESTS PREPARED FOR IAEA/OECD DATABASE Author: M. C. Paraschiv, Institute for Nuclear Research Pitesti, Romania Abstract The main features of two Romanian experimental fuel elements
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
FUEL ROD WITH E110 CLADDING OVERPRESSURE/LIFT-OFF EXPERIMENT AT HALDEN. IN-PILE DATA AND MODELING WITH START-3A CODE D. A. Chulkin 1, P.G. Demianov 1, V. I. Kuznetsov 1, V. V. Novikov 1, Yu. V. Pimenov
More informationSTATUS OF HIGHLY LOADED URANIUM-ZIRCONIUM HYDRIDE LOW ENRICHED URANIUM (LEU) FUEL PROGRAMS. G. B. West GA Technologies Inc.
STATUS OF HIGHLY LOADED URANIUM-ZIRCONIUM HYDRIDE LOW ENRICHED URANIUM (LEU) FUEL PROGRAMS G. B. West GA Technologies Inc. San Diego, CA The long term test program for U-ZrH LEU (less than 20$ enrichment)
More informationModule 03 Pressurized Water Reactors (PWR) Generation 3+
Module 03 Pressurized Water Reactors (PWR) Generation 3+ Status 1.10.2013 Prof.Dr. Böck Vienna University of Technology Atominstitute Stadionallee 2 A-1020 Vienna, Austria ph: ++43-1-58801 141368 boeck@ati.ac.at
More informationCurrent and Prospective Tests in Reactor MIR.M1
The 18th IGORR Conference 3-7 December 2017, The International Conference Centre, Darling Harbour, Sydney, Australia Current and Prospective Tests in Reactor MIR.M1 Alexey IZHUTOV INTRODUCTION Research
More informationFUNDAMENTAL SAFETY OVERVIEW VOLUME 2: DESIGN AND SAFETY CHAPTER D: REACTOR AND CORE
PAGE : 1 / 12 SUB-CHAPTER D.2 FUEL DESIGN This sub-chapter lists the safety requirements related to the fuel assembly design. The main characteristics of the fuel and control rod assemblies which have
More informationCurrent Status of the Melcor Nodalization for Atucha I Nuclear Power Plant. Zárate. S.M. and Valle Cepero, R.
Current Status of the Melcor Nodalization for Atucha I Nuclear Power Plant Zárate. S.M. and Valle Cepero, R. presentado en USNRC Cooperative Severe Accident Research Program (CSARP), Maryland, EE. UU.,
More informationDual-Cooled Blanket Modular Replacement Design Approach
Dual-Cooled Blanket Modular Replacement Design Approach Presented by X.R. Wang Contributors: S. Malang, A.R. Raffray, and The ARIES Team ARIES Meeting General Atomics, San Diego February 24-25, 2005 Outline
More informationAccident-related fuel experiments in Halden --- HRP LOCA Test Series IFA-650
Accident-related fuel experiments in Halden --- HRP LOCA Test Series IFA-650 TWGFPT Orientation 24 April 2014 W. Wiesenack 1 LOCA test overview Issues to be addressed with the HRP LOCA tests Fuel fragmentation,
More information1. INTRODUCTION XA SLIGHTLY ENRICHED URANIUM FUEL FOR A PHWR
SLIGHTLY ENRICHED URANIUM FUEL FOR A PHWR XA9846610 C. NOTARI, A. MARAJOFSKY Centra Atomico Constituyentes, Comision Nacional de Energia Atomica, Buenos Aires, Argentina Abstract An improved fuel element
More informationRe evaluation of Maximum Fuel Temperature
IAEA Technical Meeting on on Re evaluation of Maximum Operating Temperatures and Accident Conditions for High Temperature Reactor Fuel and Structural Materials, 10 12 July 2012, Vienna, Austria Re evaluation
More informationApplication of Serpent in EU FP7 project FREYA: Fast Reactor Experiments for hybrid Applications
Application of Serpent in EU FP7 project FREYA: Fast Reactor Experiments for hybrid Applications E. Fridman Text optional: Institutsname Prof. Dr. Hans Mustermann www.fzd.de Mitglied der Leibniz-Gemeinschaft
More informationNuclear Fuel Industry in China. Sao Paulo, Brazil Oct, 2015
Nuclear Fuel Industry in China Sao Paulo, Brazil Oct, 2015 Content 1. Nuclear Fuel Cycle System 2. Nuclear Fuel 3. Experience in Fuel Product Exporting 2 1. Nuclear Fuel Cycle in China 3 A completed nuclear
More informationSchool on Physics, Technology and Applications of Accelerator Driven Systems (ADS) November 2007
1858-2 School on Physics, Technology and Applications of Accelerator Driven Systems (ADS) 19-30 November 2007 Engineering Design of the MYRRHA. Part II Didier DE BRUYN Myrrha Project Coordinator Nuclear
More informationThermal Conductivity Change in High Burnup MOX Fuel Pellet
Journal of Nuclear Science and Technology ISSN: 22-3131 (Print) 1881-1248 (Online) Journal homepage: https://www.tandfonline.com/loi/tnst2 Thermal Conductivity Change in High Burnup MOX Fuel Pellet Jinichi
More informationThe role of CVR in the fuel inspection at Temelín NPP
The role of CVR in the fuel inspection at Temelín NPP Marek Mikloš, Martina Malá Research Centre Řež, Ltd. Daniel Ernst NPP Temelín IAEA TM on Hot Cell Post-Irradiation Examination and Pool-Side Inspection
More informationSpent Fuel Transport Container C-30
Spent Fuel Transport Container C-30 Part II. Using of the old type transport cask C-30 for an improved fuel VVER-440 Vladimír Chrapčiak, Radoslav Zajac Pavol Lipták VUJE, a.s, Slovakia VUJE, Inc., Okružná
More informationThe further development of WWER-440 fuel design performance. Authors: V.B.Lushin, I.N.Vasilchenko, J.A.Ananjev, G.V.Abashina OKB "GIDROPRESS".
The further development of WWER-440 fuel design performance Authors: V.B.Lushin, I.N.Vasilchenko, J.A.Ananjev, G.V.Abashina OKB "GIDROPRESS". 1 Introduction VVER fuel development is determined by two main
More informationREDACTED PUBLIC VERSION HPC PCSR3 Sub-chapter 4.2 Fuel Design NNB GENERATION COMPANY (HPC) LTD REDACTED PUBLIC VERSION HPC PCSR3: { PI Removed }
Page No.: i / iii NNB GENERATION COMPANY (HPC) LTD HPC PCSR3: CHAPTER 4 REACTOR AND CORE DESIGN SUB-CHAPTER 4.2 FUEL DESIGN { PI Removed } uncontrolled. 2017 Published in the United Kingdom by NNB Generation
More informationTypes, Problems and Conversion Potential of Reactors Produced in Russia
Types, Problems and Conversion Potential of Reactors Produced in Russia Moscow, Russian-American symposium on Conversion of the Research Reactors to LEU Fuel, 8-10 June 2011 Director, General Designer
More informationBohunice V-2 power plant mixed core licensing and operation experiences Ondrej Grežďo
operation experiences Ondrej Grežďo TM Vienna, 12/2011 Information about our NPP BOHUNICE NPP TYPE: 2 * VVER 440-213 in operation 2* VVER 440-230 in decomisioning 1* A-1 in decomisioning 2 Contents Why?
More informationPreliminary Neutronics Assessment of Molten Salt Blanket Concepts
Preliminary Neutronics Assessment of Molten Salt Blanket Concepts Mohamed Sawan Fusion Technology Institute University of Wisconsin, Madison, WI ITER TBM Meeting UCLA Feb. 23-25, 2004 1 Preliminary Neutronics
More informationSuper-Critical Water-cooled Reactors
Super-Critical Water-cooled Reactors (SCWRs) SCWR System Steering Committee T. Schulenberg, H. Matsui, L. Leung, A. Sedov Presented by C. Koehly GIF Symposium, San Diego, Nov. 14-15, 2012 General Features
More informationSerpent Code Using in ALLEGRO Project
Serpent Code Using in ALLEGRO Project 4 th Annual Serpent User Group Meeting Radoslav ZAJAC Department of Nuclear Design and Fuel Management University of Cambridge Cambridge, 17 th 19 th September 2014
More informationDEMO FUSION CORE ENGINEERING: Blanket Integration and Maintenance
DEMO FUSION CORE ENGINEERING: Blanket Integration and Maintenance 1.) Overview on European Blanket Concepts and Integration principles 2.) Large Module Integration 3.) Multi Module Segment (MMS) Integration
More informationNEA-WPFC/FCTS Benchmark for Fuel Cycle Scenarios Study with COSI6
NEA-WPFC/FCTS Benchmark for Fuel Cycle Scenarios Study with COSI6 G. Grasso, S. Monti, M. Sumini Report RSE/2009/136 Ente per le Nuove tecnologie, l Energia e l Ambiente RICERCA SISTEMA ELETTRICO NEA-WPFC/FCTS
More informationBy: Eugenijus Uspuras Algirdas Kaliatka Sigitas Rimkevicius ASME 2012 Verification & Validation Symposium May 2-4, 2012, Las Vegas, NV
Computer codes validation for transient analysis at nuclear power plants with RBMK-1500 reactor By: Eugenijus Uspuras Algirdas Kaliatka Sigitas Rimkevicius ASME 2012 Verification & Validation Symposium
More informationExperimental study of DHC. cladding and implications. dry storage conditions
17th ASTM International Symposium Zirconium in the Nuclear Industry 03-07 February, 2013, Hyderabad, India Experimental study of DHC of unirradiated d and irradiated d fuel cladding and implications to
More informationThe Establishment and Application of TRACE/FRAPTRAN Model for Kuosheng Nuclear Power Plant
The Establishment and Application of /FRAPTRAN Model for Kuosheng Nuclear Power Plant S. W. Chen, W. K. Lin, J. R. Wang, C. Shih, H. T. Lin, H. C. Chang, W. Y. Li Abstract Kuosheng nuclear power plant
More informationANALYSIS OF BREST-OD-300 SAFETY DURING ANTICIPATED OPERATIONAL OCCURRENCES
ANALYSIS OF BREST-OD-300 SAFETY DURING ANTICIPATED OPERATIONAL OCCURRENCES D.V. Didorin, V.A. Kogut, A.G. Muratov, V.V. Tyukov, A.V. Moiseev (NIKIET, Moscow, Russia) 1. Brief description of the aim and
More informationPost Irradiation Examinations of High Performance Research Reactor Fuels
Post Irradiation Examinations of High Performance Research Reactor Fuels www.inl.gov National Academy of Science Technical Review Francine Rice, Walter Williams, Daniel Wachs, Mitchell Meyer, Adam Robinson
More informationCARA design criteria for HWR fuel burnup extension
CARA design criteria for HWR fuel burnup extension P.C. Florido, R.O. Cirimello, J.E. Bergallo, A.C. Marino, D.F. Delmastro, D.O. Brasnarof, J.H. Gonzalez, L.A. Juanico Centro Atomico Bariloche, Comision
More informationA.L. Izhutov, A.V. Burukin, Current and prospective fuel test programmes in the MIR reactor
A.L. Izhutov, A.V. Burukin, S.A. Iljenko,, V.A. Ovchinnikov, V.N. Shulimov,, V.P. Smirnov State Scientific Centre of Russia Research Institute of Atomic Reactors, 433510, Dimitrovgrad, Ulyanovsk region,
More informationIAEA REPORT 2007 NEUTRONICS DESIGN OF THE FBNR. Submitted to the INTERNATIONAL ATOMIC ENERGY AGENCY. Principal investigator Farhang Sefidvash
IAEA REPORT 2007 NEUTRONICS DESIGN OF THE FBNR Submitted to the INTERNATIONAL ATOMIC ENERGY AGENCY Principal investigator Farhang Sefidvash Collaborator Robson Silva da Silva Federal University of Rio
More informationCoupling of SERPENT and OpenFOAM for MSR analysis
Coupling of SERPENT and OpenFOAM for MSR analysis Olga Negri Supervisor Prof. Tim Abram, University of Manchester Co-supervisor Dr. Hywel Owen, University of Manchester Industrial supervisor Steve Curr,
More informationB. HOLMQVIST Nuclear Fuel Division, ABB Atom AB, Vasteras, Sweden
I Iflllll IPIBM1I IHtl!!!! Blini Vllll! «! all REDUCTION OF COST OF POOR QUALITY IN NUCLEAR FUEL MANUFACTURING XA0055764 B. HOLMQVIST Nuclear Fuel Division, ABB Atom AB, Vasteras, Sweden Abstract Within
More informationEXTENDED STUDY OF COOLABILITY OF VVER BUNDLE WITH BALLOONED REGION
EXTENDED STUDY OF COOLABILITY OF VVER BUNDLE WITH BALLOONED REGION Imre Nagy, Zoltán Hózer, Tamás Novotny, Péter Windberg, András Vimi Centre for Energy Research, Hungarian Academy of Sciences H-1525 Budapest,
More informationINSPECTION TECHNIQUE FOR BWR CORE SPRAY THERMAL SLEEVE WELD
More Info at Open Access Database www.ndt.net/?id=18479 INSPECTION TECHNIQUE FOR BWR CORE SPRAY THERMAL SLEEVE WELD ABSTRACT J.L. Fisher, G. Light, Jim Crane, Albert Parvin, Southwest Research Institute,
More informationAdvanced Cooling Technologies, Inc. Low-Cost Radiator for Fission Power Thermal Control NETS Conference
Advanced Cooling Technologies, Inc. Low-Cost Radiator for Fission Power Thermal Control 2015 NETS Conference Advanced Cooling Technologies, Inc. Taylor Maxwell Calin Tarau Bill Anderson Vanguard Space
More informationOPAL : Commissioning a New Research Reactor. IAEA Conference, Sydney, November 2007
OPAL : Commissioning a New Research Reactor IAEA Conference, Sydney, November 2007 Project Timeline Government announcement 1997 Design and licence application 2000/2001 Construction Licence April 2002
More informationAUG you inform have taken maximum Order.
-'- -1 AUG 2 4 1973 Docket No. 50-293 Boston Edison Company ATTN: Mr. James M. Carroll Vice President and General Counsel 800 Boylston Street Boston, Massachusetts 02199 Subject: FUEL DENSIFICATION Gentlemen:
More informationА Е Ц К О З Л О Д У Й - Е А Д N P P K O Z L O D U Y P L C
А Е Ц К О З Л О Д У Й - Е А Д N P P K O Z L O D U Y P L C 16 th Symposium of AER Bratislava, September 25-29, 2006 STATIONARY TVSA FUEL CYCLES AT KOZLODUY NPP WWER-1000 REACTORS K. Kamenov, NPP Kozloduy,
More informationImprovement of Irradiation Capability in the Experimental Fast Reactor Joyo
IAEA Technical Meeting November, 2008 Improvement of Irradiation Capability in the Experimental Fast Reactor Joyo Tomonori Soga Fast Reactor Technology Section Experimental Fast Reactor Department O-arai
More informationTEMPERATURE AND STRESS IN ALCATOR C-MOD DUE TO THE DIVERTOR UPGRADE
THERMAL ANALYSIS TO CALCULATE THE VESSEL TEMPERATURE AND STRESS IN ALCATOR C-MOD DUE TO THE DIVERTOR UPGRADE Han Zhang 1, Peter H. Titus 1, Robert Ellis 1, Soren Harrison 1,2, Rui Vieira 2 1 Princeton
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
TEST IRRADIATION OF ENHANCED NUCLEAR FUEL AND CLADDING Klara Insulander Björk 1, Cheuk Wah Lau 1, Carlo Vitanza 1, Hyun-Gil Kim 2, Dong-Joo Kim 2 1 Thor Energy, Karenslyst Allé 9C, 0278 Oslo, Norway, post@scatec.no
More informationRosatom Seminar on Russian Nuclear Energy Technologies and Solutions
ROSATOM STATE ATOMIC ENERGY CORPORATION ROSATOM VVER-100 Reactor Plant and Safety Systems Rosatom Seminar on Russian Nuclear Energy Technologies and Solutions N.S. Fil Chief Specialist, OKB GIDROPRESS
More informationCANDU Fuel Bundle Deformation Model
CANDU Fuel Bundle Deformation Model L.C. Walters A.F. Williams Atomic Energy of Canada Limited Abstract The CANDU (CANada Deuterium Uranium) nuclear power plant is of the pressure tube type that utilizes
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
Plant and Cycle Specific Fuel Assembly Bow Evolution Assessment Yuriy Aleshin 1, Jorge Muñoz Cardador 2 1 Westinghouse Electric Company LLC, PWR Fuel Technology: 5801 Bluff Road, Hopkins, SC 29061 - USA
More informationEXAMINATION OF NOZZLE INNER RADIUS AND PIPING FROM THE OUTER SURFACE
More Info at Open Access Database www.ndt.net/?id=18560 ABSTRACT NEW DEVELOPMENTS FOR AUTOMATED NOZZLE INNER RADIUS AND PIPING INSPECTIONS D. Eargle,WesDyne International, USA WesDyne has recently engaged
More informationNEW TYPES OF NUCLEAR FUEL D. Krylov JSC TVEL
NEW TYPES OF NUCLEAR FUEL D. Krylov JSC TVEL International Forum ATOMEXPO 2011 Moscow, 6 8 June 2011 1 Objective To supply Customer with the fuel providing: Safe and reliable operation Economic efficiency
More informationImpact of Medium-Temperature Magnet and 2-FP Configuration on Radial Build
Impact of Medium-Temperature Magnet and 2-FP Configuration on Radial Build Laila El-Guebaly Fusion Technology Institute University of Wisconsin - Madison With input from: L. Bromberg (MIT), J. Lyon (ORNL),
More informationALLEGRO Project EUROSAFE Branislav HATALA Petr DAŘÍLEK Radoslav ZAJAC. 2 nd - 3 rd November 2015 Brussels
ALLEGRO Project EUROSAFE 2015 Branislav HATALA Petr DAŘÍLEK Radoslav ZAJAC 2 nd - 3 rd November 2015 Brussels Content ALLEGRO Project Introduction / ALLEGRO Demonstrator ALLEGRO Consortium V4G4 Centre
More informationFRM II Converter Facility
FRM II Converter Facility Volker Zill Heiko Gerstenberg Axel Pichmaier Technical University of Munich Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II) Lichtenbergstrasse 1, 85748 Garching - Federal
More informationAn Investigation on the Fuel Assembly Structural Performance for the PLUS7 TM Fuel Design
2th International Conference on Structural Mechanics in Reactor Technology (SMiRT 2) Espoo, Finland, August 9-14, 29 SMiRT 2-Division 6, Paper 1824 An Investigation on the Fuel Assembly Structural Performance
More informationCONTROL SYSTEM DESIGN FOR A SMALL PRESSURIZED WATER REACTOR
CONTROL SYSTEM DESIGN FOR A SMALL PRESSURIZED WATER REACTOR Peiwei Sun and Jianmin Zhang Xi'an Jiaotong University No. 28 Xianing Road West, Xi'an, Shaanxi 710049, China sunpeiwei@mail.xjtu.edu.cn; zhangjm@mail.xjtu.edu.cn
More informationNuclear Thermal Propulsion (NTP) Engine Component Development
Nuclear Thermal Propulsion (NTP) Engine Component Development Presented to the NETS 2015 Conference O. Mireles, K. Benenski, J. Buzzell, D. Cavender, J. Caffrey, J. Clements, W. Deason, C. Garcia, C. Gomez,
More informationTHE CURRENT STATUS ON DUPIC FUEL TECHNOLOGY DEVELOPMENT
THE CURRENT STATUS ON DUPIC FUEL TECHNOLOGY DEVELOPMENT Song K.C., Choi H., Kim H.D., Park J.J., Park G.I., Kang K.H., Lee J.W., Yang M.S. Korea Atomic Energy Research Institute, Daejeon, Korea 1. Introduction
More informationMULTI-PARAMETER OPTIMIZATION OF BRAKE OF PISTON
3 2 1 MULTI-PARAMETER OPTIMIZATION OF BRAKE OF PISTON Á. Horváth 1, I. Oldal 2, G. Kalácska 1, M. Andó 3 Institute for Mechanical Engineering Technology, Szent István University, 2100 Gödöllő, Páter Károly
More informationTerraPower s Molten Chloride Fast Reactor Program. August 7, 2017 ANS Utility Conference
TerraPower s Molten Chloride Fast Reactor Program August 7, 2017 ANS Utility Conference Molten Salt Reactor Features & Options Key Molten Salt Reactor (MSR) Distinguishing Features Rather than using solid
More informationControllability of MSR-FUJI
Controllability of MSR-FUJI Ritsuo Yoshioka(*), Koshi Mitachi International Thorium Molten-Salt Forum (*):e-mail: ritsuo.yoshioka@nifty.com http://msr21.fc2web.com/english.htm 1 Table of contents (1) Molten
More informationFrom MYRRHA to XT-ADS: lessons learned and towards implementation
From MYRRHA to XT-ADS: lessons learned and towards implementation Didier De Bruyn On behalf of the EUROTRANS DM1 partners AccApp 09 Satellite meeting 1 Summary More than 40 partners have started the FP6
More informationAP1000 European 5. Reactor Coolant System and Connected Systems Design Control Document
CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1 Summary Description This section describes the reactor coolant system (RCS) and includes a schematic flow diagram of the reactor coolant system
More informationDesign and Performance of PWR and BWR Fuel Workshop on Modeling and Quality Control for Advanced and Innovative Fuel Technologies
Design and Performance of PWR and BWR Fuel Workshop on Modeling and Quality Control for Advanced and Innovative Fuel Technologies Lecture given by Hans G. Weidinger At International Centre of Theoretical
More informationNUCLEAR FUEL RELIABILITY IN NPP KRŠKO
International Conference Nuclear Energy in Central Europe 21 Hoteli Bernardin, Portorož, Slovenia, September 1-13, 21 www: http://www.drustvo-js.si/port21/ e-mail: PORT21@ijs.si tel.:+ 386 1 588 5247,
More informationVirtual Testing for Automotive Components and its Integration into the OEM s Product Creation Process. Dr. Gerald Seider Dr.
Virtual Testing for Automotive Components and its Integration into the OEM s Product Creation Process Dr. Gerald Seider Dr. Fabiano Bet Orlando, 18 March, 2013 Company Profile Consulting, Engineering Services
More informationPresentation Outline
Institutt for Energiteknikk OECD Halden Reactor Project Joint International Program - VVER Fuel Presented by: Dr Margaret McGrath Presentation Outline The OECD Halden Reactor Project Capabilities for Fuel
More informationRecommendations for a demonstrator of Molten Salt Fast Reactor
Recommendations for a demonstrator of Molten Salt Fast Reactor E. MERLE-LUCOTTE, D. HEUER, M. ALLIBERT, M. BROVCHENKO, V. GHETTA, P. RUBIOLO, A. LAUREAU merle@lpsc.in2p3.fr Professor at Grenoble INP/PHELMA
More informationTOPIC 5 EXPERIMENTS FOR IMPROVING NUCLEAR FUELS MODELS AND PERFORMANCE
The 2008 Frédéric JOLIOT & Otto HAHN Summer School August 20 August 29, 2008 Aix-en-Provence, France TOPIC 5 EXPERIMENTS FOR IMPROVING NUCLEAR FUELS MODELS AND PERFORMANCE The Importance of In-Pile Measurements
More informationForsmark 12. S3K Applications. Thomas Smed US User Group Meeting Arizona, October 2008
Forsmark 12 S3K Applications Thomas Smed US User Group Meeting Arizona, October 2008 Introduction It is well-known that we have vast experience in providing S3R (and RAMONA) to training simulators It may
More informationEl'Eh.centro Ricerche
El'Eh.centro Ricerche Titolo Bologna Sigla di identificazione Distrib. Pago di FPN - P9LU- 034 L l 60 NEA-WPFC/FCTS benchmark for fuel cycle scenarios study with COSI6 Descrittori Tipologia del documento:
More informationDESCRIPTION OF THE DELPHI SUBCRITICAL ASSEMBLY AT DELFT UNIVERSITY OF TECHNOLOGY
IRI-131-2003-008 DESCRIPTION OF THE DELPHI SUBCRITICAL ASSEMBLY AT DELFT UNIVERSITY OF TECHNOLOGY J.L. Kloosterman October, 2003 INTRODUCTION For educational purposes, the Reactor Physics Department of
More informationNeutronic Performance Issues of the Breeding Blanket Options for the European DEMO Fusion Power Plant
Neutronic Performance Issues of the Breeding Blanket Options for the European DEMO Fusion Power Plant U. Fischer, KIT Contributors C. Bachmann EUROfusion/PMU, Garching, Germany J. C. Jaboulay CEA Saclay,
More information