TOPIK RISET NUKLIR ITB

Reactor Physics LaboratoryITB

SISTEMATIKA RISET NUKLIR DI ITB1. SPINNORs2. MODIFIED CANDLE, including GCFR3. Code Development4. Nuclear Data5. Fuel Cycle and Waste Management6. Nuclear Physics Fundamental research

Long Life Pb-Bi Cooled Fast Reactors

SMALL SIZE Pb-Bi COOLED NUCLEAR POWER REACTORSPower Range 25MWe ~ 100MWeLong life operation without refuelingIdeal for remote area (islands): especially outside Java-Bali AreaCurrent status : Final Optimization

especially in safety, thermal system, etc.Inherent safetyNon proliferationFissile self sustain

Very Small Size Pb-Bi COOLED NUCLEAR POWER REACTORS Power Range 5MWe ~ 25MWeLong life operation without refuelingIdeal for remote area (islands): especially outside Java-Bali Area, special purposeCurrent status : Final Optimization

especially in safetyInherent safetyNon proliferationFissile self sustain

Medium & Large Size Pb-Bi COOLED

NUCLEAR POWER REACTORS Power Range 100MWe ~ 2000MWeFew years operation without refuelingIdeal for Java-Bali Area, special

purpose:Hydrogen ProductionCurrent status : Optimization in Neutronic design

, safety and thermal systemInherent safetyNon proliferationBreedingEconomicalLoad followerCogeneration

ADS (Accelerator Driven System)Power range : 100KWe~50MWeFast and thermalHigh safety performanceOptimization of neutron source design and

configurationOptimization of thermal systemSafety analysis

Pb-Bi Corrosion InvestigationClasical and Quantum Mechanical Based

simulationBased on Ab initio ModelComparation with existing experimental dataSearching for better fit structural material

Hydrogen Production reactorsFast: Pb-Bi Cooled, Thermal : HTGR Based Selection of Best chemical mechanismThermal configuration optimization Material feasibilitySimulation system

MODIFIED CANDLE REACTORRegion 1

Region 10

Region 9

Region 8

Region 7

Region 6

Region 5

Region 4

Region 3

Region 2

Region 1

Region 10

Region 9

Region 8

Region 7

Region 6

Region 5

Region 4

Region 3

Region 2

OUT

Modified Candle ReactorsIn this study conceptual design study of Pb-Bi

cooled fast reactors which fuel cycle need only natural uranium input has been performed. In this case CANDLE burn-up strategy is slightly modified by introducing discreet regions.

In this design the reactor cores are subdivided into several parts with the same volume in the axial directions.

The natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel.

This concept is basically applied to all regions, i.e. shifted the core of I’th region into I+1 region after the end of 10 years burn-up cycle .

Long Life Reactor With Natural Uranium as Fuel Cycle input

D:X

3C:X

2B

:X1A:0

C:X

3B:X

2A

:X1Ura

niu

m

ala

m

BOC EOC

input

Long Life Reactor With Natural Uranium as Fuel Cycle input

time (y unit)10987654321

Ke

ff

1.05

1.045

1.04

1.035

1.03

1.025

1.02

1.015

1.01

1.005

Thorium(Th) and Protactinium (231Pa) Based Fuel for Tight Lattice Long Life BWR

Keff Vs Time

1.001

1.0015

1.002

1.0025

1.003

0 3 6 9 12 15 18 21 24Time (Month)

Active Core Volume(minus reflector)17635.8 Liter

Thermal Power 620 Mwatt

Average Power Density 35.2 Watt/cc

Enrichment Uranium-233 8.1% and 11%

Percentage Protactinium-231 6.7%dan12.5%

Reactor operation time 30 year

Excess-reactivity 0.384%

SHIP BASED NUCLEAR POWER REACTORPb-Bi Based and Water cooled basedSmall and very small sizedIdeal for remote area, emergency and

temporary developmentStatus: Final optimization and safety analysis

Group Contant ProcessingFast group constant : general geometryThermal system: implementation & toward

general geometryInterface to other codeParalel computation

Neutronic DesignThree dimensional system analysisAdditional featureBetter user interfaceTransport analysisSpecial investigation

Safety AnalysisThree dimensional modelLocal blockage analysisOther Hypothetical accident analysisADS safety analysisParalel Computation

Monte Carlo SimulationFor shielding and neutronic calculationDevelopment of generic subroutineParalel Computation

Paralel ComputationBased on ehternet and dedicated systemBased on Socket programming or specially

developped systemDevelopment of new algorithm better fit to

paralel computation

Intelligent computationBased on AI or JSTTo help better convergence in special

system : thermal hydraulic, and other optimization

Safety system prediction

TOPIK BESAR: INTEGRATED SYSTEM ANALYSIS CODETAHAP I : 1.CELL HOMOGENIZATION CODE2.MULTI GROUP DIFFUSION CALCULATION3.BURNUP ANALYSIS

MAIN PROJECT 2009: INTEGRATED NEUTRONIC CODETAHAP I :Cell Calculation CodeMultigroup Diffusion CodeBurnup Analysis code

Analisa BurnupStandar untuk fast reactorSetara COREBURN untuk termal reactorMetoda Semi analitikClose fuel cycle

STATUS SAAT INIIndividual code telah adaFokus IntegrasiPerluasan cakupan libraryGeneralisasi geometriAdvanced simulation method mengejar

akurasi tinggi secara ekonomis: Misal FP treatment untuk LLFP

Cross section calculationPhenomenological and microscopic basedPhenomenological: optical modelMicroscopic model: Generating coordinat

method and Hartree Fock MethodFP Yield Calculation DWBA CodeIntermediate Energy