Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code

Leppänen, Jaakko; Pusa, Maria; Fridman, E.

doi:10.1016/j.anucene.2016.06.007

Cited by 95 publications

(23 citation statements)

References 17 publications

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“…Firstly, Serpent is capable of generating few-group homogenized macroscopic cross section data typically required by 3D core simulators (Leppänen et al, 2016). Moreover, starting from the version 2.1.29 Serpent was extended with a capability to conveniently generate also fewgroup microscopic cross sections and fission yield data.…”

Section: Generation Of the Few-group Macro-and Microscopic Cross Sectmentioning

confidence: 99%

Explicit decay heat calculation in the nodal diffusion code DYN3D

Bilodid

Fridman

Kotlyar

et al. 2018

Annals of Nuclear Energy

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Section: Generation Of the Few-group Macro-and Microscopic Cross Sectmentioning

confidence: 99%

Explicit decay heat calculation in the nodal diffusion code DYN3D

Bilodid

Fridman

Kotlyar

et al. 2018

Annals of Nuclear Energy

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“…It con tains data for a num ber of tem per a tures -300, 600, 900, 1200, 1500, and 1800 K for all the iso topes. The tem per a ture range for light wa ter for the ther mal scat ter ing li brary has the form: 296, 350, 400, 450, 500, and 600 K [34].…”

Section: Ser Pent Code As An In Stru Ment For Re Ac Tor Cal Cu La Tiomentioning

confidence: 99%

“…This is be cause the used nu clear data li brary for Ser pent calcu la tions does not con tain data for the tem per a ture of 1027 K. In that case, the cross-sec tions with a tem per ature of 900 K were used. Ac cord ing to [37], in the second ver sion of Ser pent code, there is an op tion that allows ad just ing the mi cro scopic cross-sec tions to the re quired tem per a ture val ues for fuel and other ma te rials. Ta ble 4 also shows that Ser pent code gives a good match com pared to the MCU, WIMS8A, and HELIOS codes.…”

Section: Neu Tron Mul Ti Pli Ca Tion Fac Tor Cal Cu La Tionsmentioning

confidence: 99%

Computational benchmark for fuel assembly of VVER-1000 using the Monte Carlo Serpent code

Galchenko¹,

Shlapak²,

Gulik³

2018

NUCL TECH RAD PROT

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The use of a new Monte Carlo Ser pent code for the cal cu la tion of wa ter-cooled re ac tors is presented and a cal cu la tion scheme of the fuel as sem bly for VVER-1000 re ac tors de vel oped. The cal cu la tion of neu tron-phys i cal char ac ter is tics for the fuel as sem bly of VVER-1000 is car ried out for dif fer ent states and the re sults ob tained by the Ser pent model com pared with the results of other re ac tor codes. The anal y ses of these re sults are pre sented in the pa per sub mit ted here. Based on this ar ti cle, the Monte Carlo Ser pent code could be used for neu tron-phys i cal cal cu la tions of VVER-1000 re ac tors.

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“…Both blanket zones are made of made of ThO 2, and the fissile zone is made of a mixture of 233 UO 2 and ThO 2 . Since a boiling water cooled reactor has a robust axial variation in the coolant density, the coolant channel of the calculation model is divided into 9 axial zones, and the axial zone-dependent coolant density is calculated based on the assumption of cosine axial power distribution [9]. The exit void fraction is 0.43, inlet water density is approximately 0.74 g/cm 3 , and water density at the exit is approximately 0.08 g/cm 3 .…”

Section: Methodsmentioning

confidence: 99%

“…The neutronic calculations in this study were performed using SERPENT [8,9], which is a continuous energy Monte Carlo (MC) neutron transport code developed at VTT research center in Finland. In these calculations, the nuclear data library based on the evaluated file ENDF/B-VII were used.…”

Section: Methodsmentioning

confidence: 99%

Neutronic Analysis of a Thorium-Fueled Reduced Moderation Boiling Water Reactor

Falaakh¹,

Agung²,

Harto³

2018

International Journal on Advanced Science, Engineering and Information Technology

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Abstract-A study about Resource Renewable Boiling Water Reactor (RBWR) core, a reduced moderation boiling water reactor that features the breeding ratio larger than 1 was conducted. This study focuses on the neutronic performances of the core and aims to investigate the core sustainability when using thorium as the main fertile fuel. A fuel-self-sustaining core with high burnup set as the design target. 233 U+Th were used as the initial fuel, and the impact of initial fissile ( 233 U) content in the core fissile zone on the core neutronic performances was evaluated. Parameters related to the neutronic performances such as the core burnup, fissile breeding, and fissile inventory ratio (FIR) are considered in this study. From these results, it was confirmed that it is feasible to create a selfsustaining fuel cycle system using thorium fueled RBWR. However, there was a trade-off between the core burnup and fissile breeding that can be a significant challenge in the development of this system. Evaluating the other design variables may be considered to address this challenge. The further study to analyze the safety performances of the core is required to arrive at a safe and reliable reactor system.

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Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code

Cited by 95 publications

References 17 publications

Explicit decay heat calculation in the nodal diffusion code DYN3D

Explicit decay heat calculation in the nodal diffusion code DYN3D

Computational benchmark for fuel assembly of VVER-1000 using the Monte Carlo Serpent code

Neutronic Analysis of a Thorium-Fueled Reduced Moderation Boiling Water Reactor

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