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Serpent - A Monte Carlo Reactor Physics Burnup Calculation Code

Traditional reactor physics applications, including spatial homogenization, criticality calculations, fuel cycle studies, research reactor modeling, validation of deterministic transport codes, etc. Multi-physics simulations, i.e. coupled calculations with thermal hydraulics, CFD and fuel performance codes. Neutron and photon transport simulations for radiation dose rate calculations, shielding, fusion research and medical physics. And the Serpent Wiki. Acts as an on-line user manual. High-temperature ga...

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Serpent - A Monte Carlo Reactor Physics Burnup Calculation Code | montecarlo.vtt.fi Reviews

<META> DESCRIPTION

Traditional reactor physics applications, including spatial homogenization, criticality calculations, fuel cycle studies, research reactor modeling, validation of deterministic transport codes, etc. Multi-physics simulations, i.e. coupled calculations with thermal hydraulics, CFD and fuel performance codes. Neutron and photon transport simulations for radiation dose rate calculations, shielding, fusion research and medical physics. And the Serpent Wiki. Acts as an on-line user manual. High-temperature ga...

<META> KEYWORDS

1 overview

2 woodcock delta tracking method

3 reactor geometries

4 mesh file format

5 interaction physics

6 photon transport mode

7 burnup calculation

8 coupled multi physics simulations

9 variance reduction

10 parallelization

CONTENT

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overview,woodcock delta tracking method,reactor geometries,mesh file format,interaction physics,photon transport mode,burnup calculation,coupled multi physics simulations,variance reduction,parallelization,results and output,discontinuity factors,opened

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Serpent - A Monte Carlo Reactor Physics Burnup Calculation Code | montecarlo.vtt.fi Reviews

https://montecarlo.vtt.fi

Traditional reactor physics applications, including spatial homogenization, criticality calculations, fuel cycle studies, research reactor modeling, validation of deterministic transport codes, etc. Multi-physics simulations, i.e. coupled calculations with thermal hydraulics, CFD and fuel performance codes. Neutron and photon transport simulations for radiation dose rate calculations, shielding, fusion research and medical physics. And the Serpent Wiki. Acts as an on-line user manual. High-temperature ga...

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montecarlo.vtt.fi

1

Serpent - A Monte Carlo Reactor Physics Burnup Calculation Code

http://montecarlo.vtt.fi/links.htm

Serpent thermal scattering data library. Serpent JEF-2.2 ACE format data library with ures ptables. Serpent JEFF-3.1 ACE format data library with ures ptables. Serpent JEFF-3.1.1 ACE format data library with ures ptables. Serpent ENDF/B-VI.8 ACE format data library with ures ptables. Serpent ENDF/B-VII ACE format data library with ures ptables. Cross Section Libraries for Serpent 1.1.7. PSG Status Report (March 2008). Serpent Progress Report 2009. Serpent Progress Report 2010. Serpent Progress Report 2011.

2

Serpent User Group Meeting 2011

http://montecarlo.vtt.fi/mtg/2012_Madrid/index.htm

The Second International Serpent User Group Meeting was hosted by the Universidad Politécnica de Madrid in Madrid, Spain, on September 19-21, 2012. The meeting brought together 40 Serpent users from 16 Organizations in Europe and the U.S. The Serpent developer team wants to thank all participants for interesting presentations and discussion, and UPM for organizing and hosting the event. Presentations - Day 1. J Leppänen (VTT) - Welcome and a very short introduction to Serpent. Presentations - Day 2.

3

Serpent - A Monte Carlo Reactor Physics Burnup Calculation Code

http://montecarlo.vtt.fi/publications.htm

Publications and reports related to Serpent development. Aufiero, M., Cammi, A., Fiorina, C., Leppänen, J., Luzzi, L., and Ricotti, M. (2013a). An extended version of the Serpent-2 code to investigate fuel burn-up and core material evolution of the molten salt fast reactor. J Nucl. Mat., 441 (2013) 473-486. Calculating the effective delayed neutron fraction in the molten salt fast reactor: Analytical, deterministic and Monte Carlo approaches. Ann Nucl. Energy, 65 (2014) 78-90. In proc. PHYSOR 2016. Direc...

4

Serpent - A Monte Carlo Reactor Physics Burnup Calculation Code

http://montecarlo.vtt.fi/development.htm

Why the Monte Carlo method? Spatial homogenization (Koebke, 1978. History of the Serpent project. The development of Serpent started at VTT in 2004, under the working title "Probabilistic Scattering Game", or PSG. All publications dated before the official pre-release in October 2008 refer to the code using this name. The name was later changed to Serpent, due to the various ambiguities related to the acronym. The predecessor of Serpent was used as an in-house code at VTT. The project also spawned th...

5

Serpent User Group Meeting 2011

http://montecarlo.vtt.fi/mtg/2011_Dresden/index.htm

The Serpent developer team wants to thank all participants for interesting presentations and discussion, and HZDR for organizing and hosting the event. Presentations - Day 1. J Leppänen - Introduction to Serpent. M Pusa - Solving Burnup Equations in Serpent: Matrix Exponential Method CRAM. R Dagan - Introduction of the Resonance dependent scattering kernel in SERPENT. E Fridman - Some remarks on XS preparation with SERPENT. H Suikkanen and V. Rintala - Pebble bed reactor modeling using Serpent. Presented...

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Serpent-to-MCNP input file converter | high flux

https://highflux.wordpress.com/2010/11/07/serpent-to-mcnp-input-file-converter

Thanks for dropping by my blog! Here I post some ideas about reactor modeling and software development. Take a look around and grab the RSS feed. To stay updated. See you around! Serpent-to-MCNP input file converter. Filed under: Monte Carlo. Mdash; Leave a comment. November 7, 2010. I believe everyone would agree, that it’s quite hard and time-consuming to make a MCNP input file for a more-or-less complex model. Many tools were developed for simplifying the procedure, including visual editors. In this r...

serpent.vtt.fi

Serpent Wiki

http://serpent.vtt.fi/mediawiki

Welcome to Serpent Wiki. This platform, set up in November 2015, serves as the on-line user Manual for Serpent 2, as well as a repository for input files, validation reports and other resources for Serpent users. A more general description of the code is found at the project website. This Wiki covers the features and capabilities of the most recent version of Serpent 2. The content is still very far from complete, and additional information can be found from the Serpent 1 Manual. And the discussion forum.

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montecarlo.vtt.fi

Serpent - A Monte Carlo Reactor Physics Burnup Calculation Code

Traditional reactor physics applications, including spatial homogenization, criticality calculations, fuel cycle studies, research reactor modeling, validation of deterministic transport codes, etc. Multi-physics simulations, i.e. coupled calculations with thermal hydraulics, CFD and fuel performance codes. Neutron and photon transport simulations for radiation dose rate calculations, shielding, fusion research and medical physics. And the Serpent Wiki. Acts as an on-line user manual. High-temperature ga...

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