design study of innovative small pebble bed reactortobara/research/kouonngasuro/... · proposed by...
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Design Study of Innovative Simplified Small
Pebble Bed Reactor
Dwi IRWANTO1), Toru OBARA2)
1)Department of Nuclear Engineering, Tokyo Institute of Technology2)Research Laboratory for Nuclear Reactor, Tokyo Institute of Technology
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Introduction Research Purposes Calculation Procedures Parametric Survey Reference Design Conclusions
Outline
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Introduction
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(Potential) Problem ?• The unloading machinery is a very complex
and high cost system
Pebble Bed Reactor • Peu a Peu fuel loading concept proposed by E.Teuchert et al (1992)
• Pebble bed reactor –based design with fuel unloading devices is removed
• The reactor core subdivided into several fuelling zones
• Startup → lower layers filled → first criticality
• During operation → layer per layer filled → maintain criticality
• The end of the core → unloaded fuel
Peu a Peu Fuel Loading Scheme
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Research Purposes
• To find a means of carrying out the exact calculations needed to analyze the Peu à Peu fuel-loading scheme
• Optimize the fuel design by perfoming parametric survey in the infinite geometry
• Calculate a whole core design by using the optimized fuel design
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Development of a Code for Automate Process of Peu a Peu
Fuel Loading Scheme
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Calculation Procedures
• Some studies have been performed previously, they have used a diffusion-based method but the large empty cavity region in the core, makes accurate calculations is difficult to performed
• The Monte Carlo method is used to perform calculations with high accuracy at the top region of the core near the large cavity
• Unfortunately, the calculation procedures for the Peu à Peu modus using the Monte Carlo method require lot of steps
• Therefore, a computer code to automate the process of the Peu à Peu fuel load scheme has been developed using Fortran-77 and based on the Monte Carlo MVP/MVP-BURN code
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Development of a Code for Automate Process of Peu a Peu Fuel Loading Scheme
• Time needed to prepare the input files, calculate it and sequentially do all the process is very huge
• Huge number of nuclear materials data to edit and/or add to the input files
• In order to avoid mistakes in preparing the input
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Motivation
This code significantly reduce time needed to perform the calculation process of the Peu a Peu fuel load scheme
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Parametric Survey
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Parametric Survey
Parametric Survey235U enrichment 1 – 20 %
Packing Fraction 1 – 20 %
ParametersBurn-Up MWD/Ton
Energy per Ball MWD
235U and 238U used in the core %
Consumed mass of 235U and 238U gram
Critical periods month
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Parametric Survey
Parametric survey of the burn-up (MWD/Ton 235U)
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12 %wt 235U7% packing fraction
of CFP
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Reference Design
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Reference Design
Design Specification
Reactor Power 20 MWth
Fuel TRISO
Core radius 125 cm
Core Height 500 cm
Reflector width 70 cm
Startup fuel layers 85 cm
Initial 235Uenrichment 12 %
Supply fuel 235Uenrichment 12 %
Packing Fraction 7.0 %Schematic view of reactor core design
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Reference Design
Fuel BallDiameter of the ball 6.0 cmDiameter of fuel zone 5.0 cmPacking fraction of Coated FuelParticle (CFP)
7.0 %
Enrichment of 235U 12 %Equivalent natural boron content ofimpurities in uranium
4.0 ppm
Percentages of fuel balls in the core 57 %Packing fraction of fuel and dummyballs in the core
61 %
Fuel Kernel Radius of the kernel 0.250 mm UO2 density 10.4 g/cm3
Boron impurities 4 ppm
Coatings First Buffer Layer (PyC) Thickness 0.09 mm Density 1.1 g/cm3
Boron impurities 1.3 ppm Second Layer (PyC) Thickness 0.04 mm Density 1.9 g/cm3
Boron impurities 1.3 ppm Third Layer (SiC) Thickness 0.035 mm Density 3.18 g/cm3
Boron impurities 1.3 ppm Forth Layer (PyC) Thickness 0.04 mm Density 1.9 g/cm3
Boron impurities 1.3 ppm page 10 of 13
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Reference Design
keff for each fuel-loading step
* The average burnup value of this design is 9.44 x 104 MWD/T-U page 11 of 13
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Conclusions
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Conclusions• Concept of innovative small high temperature gas cooled
pebble bed reactor with possibility to simplify the reactor system by removing the unloading devices has been performed
• A code for criticality analysis of automates Peu a Peufuel load scheme process has been developed and tested
• From the parametric survey in the infinite geometry, the maximum burnup value can be expected if the inserted fuel element is 12 wt% U-235 enrichment with 7% packing fraction
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• A whole-core calculation for the small 20 MWth reactor was performed. This reactor design can maintain its criticality for up to 12 years, with the average burnup is 9.44 x 104 MWD/T-U, which is comparable to that of the conventional PBRs design
• Further analysis such as reduction of the power peak near the top of the reactor core is necessary to performed in order to optimize this design
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Conclusions
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THANK YOU