Monte Carlo simulation of Feynman-α and Rossi-α techniques for calculation of kinetic parameters of Tehran Research Reactor

Hosseini, Seyed Abolfazl; Vosoughi, Naser; Hosseini, Mohammad

doi:10.1016/j.anucene.2011.06.025

Cited by 8 publications

(4 citation statements)

References 12 publications

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“…The exact time when the neutron is captured in the neutron detector cannot be obtained The MCNP6 calculation methodology illustrated in this Section can also be used to calculate the Feynman-α distribution. [2,5] For the latter distribution, the pulse superimposition methodology (discussed in Sections 3 and 4) is not necessary because delayed neutrons come from radio-active decay, a process which follows the Poisson distribution. [5] 3.…”

Section: B Comparison Between Rossi-α Distribution and Reaction Ratmentioning

confidence: 99%

“…This study presents the methodology to calculate the Rossi-α distribution [1][2][3] obtained from the signal (e.g. 3 He(n,p) reactions) of a neutron detector placed in a subcritical nuclear reactor core (assembly) driven by a pulsed neutron source, assuming that the signal is obtained from computer simulations modeling a single neutron pulse.…”

Section: Introductionmentioning

confidence: 99%

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CUDA and MPI performances on the computation of Rossi-α distribution from pulsed neutron sources

Talamo¹

2015

Annals of Nuclear Energy

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This study presents the methodology to calculate the Rossi-α distribution for a subcritical nuclear reactor core (assembly) driven by an external pulsed neutron source through MCNP6 computer simulations. The Rossi-α distribution is obtained from the signal (e.g. 3 He(n,p) reactions) of a neutron detector placed in one experimental channel of the subcritical assembly. The detector signal is obtained from MCNP6 computer simulations modeling a single pulse of the external neutron source. The MCNP6 output data for a single pulse of the external neutron source must be processed to take into account the effect of delayed neutrons born in all previous source pulses. The calculation of the Rossi-α distribution from a pulsed neutron source is time consuming because of the huge amount of processed data. Consequently, the algorithms introduced in this work use parallel computing platforms, e.g. CUDA (which costs up to a few thousand dollars) or MPI (which costs from a few ten thousand to a few hundred thousand dollars), to reduce the computing time.

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Section: B Comparison Between Rossi-α Distribution and Reaction Ratmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

CUDA and MPI performances on the computation of Rossi-α distribution from pulsed neutron sources

Talamo¹

2015

Annals of Nuclear Energy

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“…Another approach for determination of gamma ray population (fluence) is based on the probabilistic algorithms like Monte Carlo, in which the physics of transport of gamma rays is simulated [10,11].…”

Section: Introductionmentioning

confidence: 99%

Extraction of theoretical equation for the gamma ray buildup factor of the three-layered spherical shield

Rabiee

Hosseini

2019

J. Inst.

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In the previous researches, several theoretical equations were presented for calculation of buildup factor of the single layer shields. Also, the theoretical equations were offered for the two-layered shield that consist of the known materials. For other possible modes of the multilayered shield, the buildup factor are usually calculated via experimental or simulated data. The purpose of present study is the extraction of the new theoretical equation for the gamma ray buildup factor of three-layered spherical shield made of water, concrete and iron based on the Monte Carlo calculation. To this end, the gamma ray buildup factor of the three-layered spherical shield is calculated via simulation by MCNPX computational code. Finally, the theoretical equation for gamma ray buildup factor of three-layered spherical shield vs. gamma ray buildup factors and thickness of the single-layered shields is extracted through curve fitting and regression. K: Dosimetry concepts and apparatus; Models and simulations; Radiation damage evaluation methods 1Corresponding author.

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“…Kinetic parameters for both HEU and LEU fuels in the first TRR core have been calculated by using WIMS-D4 and CITATION ( Zaker, 2003 ). In a new study, the results of MTR_PC for the first TRR operating core were compared with the noise analysis techniques and experimental data ( Hosseini et al, 2011 ). The results of this study are in good agreement with the results from MTR_PC, noise analysis and experimental data.…”

Section: Introductionmentioning

confidence: 99%

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

Lashkari

Khalafi

Kazeminejad

2013

Annals of Nuclear Energy

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Highlights► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis.

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Monte Carlo simulation of Feynman-α and Rossi-α techniques for calculation of kinetic parameters of Tehran Research Reactor

Cited by 8 publications

References 12 publications

CUDA and MPI performances on the computation of Rossi-α distribution from pulsed neutron sources

CUDA and MPI performances on the computation of Rossi-α distribution from pulsed neutron sources

Extraction of theoretical equation for the gamma ray buildup factor of the three-layered spherical shield

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

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