Modelling of HTRs with Monte Carlo: from a homogeneous to an exact heterogeneous core with microparticles

Plukienė, R.; Ridikas, D.

doi:10.1016/s0306-4549(03)00101-4

Cited by 53 publications

(18 citation statements)

References 1 publication

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“…Monteburns was successfully benchmarked earlier by simulating the uranium based fuel cycle of the high flux reactor at ILL Grenoble [9] and GT-MHR (Gas Turbine -Modular Helium cooled Reactor) in the case of plutonium (uranium free) fuel cycle [10].…”

Section: Modelling Tools and Proceduresmentioning

confidence: 99%

MCNP and ORIGEN codes validation by calculating RBMK spent nuclear fuel isotopic composition

Plukienė¹,

Plukis²,

Remeikis³

et al. 2005

Lith. J. Phys.

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Benchmark calculations of RBMK reactor spent nuclear fuel isotopic composition were performed using the MCNP and ORIGEN codes system. The RBMK assembly is explicitly modelled in order to investigate the differences of spent nuclear fuel burn-up due to different coolant density and the fuel rod position (inner or outer ring in the assembly). The modelling results are compared with experimentally measured data of RBMK-1000 fuel isotopic composition with 2% enrichment 235 U fuel at different fuel burn-up. The comparison of the experimentally measured and calculated isotopic composition for the main actinides and fission products allowed validation of MCNP and ORIGEN codes for RBMK spent nuclear fuel calculations. Variations of RBMK spent nuclear fuel burn-up and plutonium isotopic ratios depending on fuel location in the assembly were determined.

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Section: Modelling Tools and Proceduresmentioning

confidence: 99%

MCNP and ORIGEN codes validation by calculating RBMK spent nuclear fuel isotopic composition

Plukienė¹,

Plukis²,

Remeikis³

et al. 2005

Lith. J. Phys.

Self Cite

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“…It was used for modelling of GT-MHR (Gas Turbine -Modular Helium cooled Reactor) in the case of plutonium (uranium free) fuel cycle and in benchmark calculations of isotopic composition of the RBMK-1500 spent nuclear fuel. 4,5) MCNPX was validated by calculating and measuring carbon activation in graphite samples from the middle of the channel, the upper and bottom reflectors of the RBMK-1500 reactor (i.e. from vertical cross-section of the reactor plateau region).…”

Section: Rbmk Reactor Core Modelling Detailsmentioning

confidence: 99%

“…1,[4][5][6] In this work we applied a Monte Carlo code, MCNPX version 2.6 for calculation of activation of the graphite stack in the RBMK-1500 reactor. The simplified 3D model of the RBMK-1500 reactor core fragment with 14 fuel assemblies and 2 control rods distributed according to the real RBMK-1500 reactor core geometry was created.…”

Section: Introductionmentioning

confidence: 99%

Modelling of Impurity Activation in the RBMK Reactor Graphite Using MCNPX

Plukienė

Plukis

Puzas

et al. 2011

Progress in Nuclear Science and Technology

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Numerical modelling is an important tool for estimation of the radioactive waste generation at the nuclear energy sites. In this work we have applied a Monte Carlo code, MCNPX version 2.6 for calculation of activation of the graphite stack in the RBMK-1500 reactor. The simplified 3D model of the RBMK-1500 reactor core fragment with 14 fuel assemblies and 2 control rods distributed according to the real RBMK-1500 reactor core geometry has been created.Use of the Monte Carlo method for estimation of impurity activation in the reactor core faces a problem of large uncertainties of effective cross-sections if a calculation set is insufficient. Satisfactory results can be obtained by using parallel computing algorithms.The ICP-MS mass spectrometry for identification of the impurity concentration in virgin graphite from the RBMK-1500 reactor has been performed to support MCNPX modelling of the realistic neutron irradiation conditions of the graphite. Simulated radiological characteristics of the graphite have been compared with the previous calculations made with different impurity concentrations obtained by neutron activation analysis and GDMS. The analysis of radioactive impurity content shows that 14 C, 60 Co, 55 Fe, 238 Pu, 241 Am and 244 Cm make the major contribution to graphite activity and radiotoxicity during hundreds of years. The obtained results are important for decommissioning of the Ignalina NPP and other NPP with RBMK type reactors.

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“…The goal of this project was to develop a combined Monte Carlo/integral transport methodology that will enable MCNP5 [7] to perform the resonance absorption calculation within CPM3 (and other lattice physics codes such as HELIOS or CASMO-4) to complement the conventional approach used by these codes for nonresonance energy neutrons. A number of analysts have reported results by modeling the particle fuel as a regular lattice of cubical or hexagonal graphite cells with internal fuel microspheres [8][9][10][11][12]. The size of the graphite cell is determined by the packing fraction of the particle fuel.…”

Section: Background -Analysis Of Triso Particle Fuelmentioning

confidence: 99%

An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for BHTR Analysis

Martin¹,

Lee²

2009

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Modelling of HTRs with Monte Carlo: from a homogeneous to an exact heterogeneous core with microparticles

Cited by 53 publications

References 1 publication

MCNP and ORIGEN codes validation by calculating RBMK spent nuclear fuel isotopic composition

MCNP and ORIGEN codes validation by calculating RBMK spent nuclear fuel isotopic composition

Modelling of Impurity Activation in the RBMK Reactor Graphite Using MCNPX

An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for BHTR Analysis

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