Abstract
Fusion neutronics analysis before and after experiments at JET is traditionally performed using Monte Carlo particle transport code Monte Carlo N-Particle. For redundancy and diversity reasons there is a need of an additional Monte Carlo code, such as Serpent 2, capable of fusion neutronics analysis. In order to validate the Serpent code for fusion applications a detailed model of JET was used. Neutron fluxes and reaction rates were calculated and compared for positions outside the tokamak vacuum vessel, in the vacuum vessel above the plasma and next to a limiter inside the vacuum vessel. For all detector positions with DD and DT neutron sources the difference between neutron fluxes calculated with both Monte Carlo codes were within 2σ statistical uncertainty and for most positions (more than 90 % of all studied positions) even within 1σ uncertainty. Fusion neutronics analysis in the JET tokamak with Serpent took on average 10 % longer but this can be improved by changing the threshold value for determination of the transport method used. With the work presented in this paper the Serpent Monte Carlo code was validated to be a viable alternative to MCNP for fusion neutronics analysis for the JET tokamak.
Highlights
Fusion neutronics analysis performed before and after experiments at JET are commonly supported by neutron transport calculations with the Monte Carlo particle transport code Monte Carlo NParticle (MCNP)[1]
The neutron fluxes were calculated in the positions of three fission chambers called KN1 located outside the vacuum vessel next to the transformer limbs (Fig.1), in four indium foils located in irradiation ends, called KN2-3U, located above plasma on the inside of the vacuum vessel commonly used for neutron activation measurements (Fig.1) and in a long term irradiation station called I-LTIS located inside the vacuum vessel next to a limiter (Fig.1)
The difference in the calculated neutron fluxes and reaction rates for Indium-115 between Serpent and MCNP for KN2-3U position is larger for deuterium-deuterium plasma (DD) plasma at around 0.8 % while for deuterium-tritium plasma (DT) plasma the difference is less than 0.2 %
Summary
Fusion neutronics analysis performed before and after experiments at JET are commonly supported by neutron transport calculations with the Monte Carlo particle transport code Monte Carlo NParticle (MCNP)[1]. Its distribution is controlled, limited and subjected to US regulation For this reason there is a need of an additional Monte Carlo code capable of fusion neutronics analysis for redundancy and diversity. In the first step the code validation was performed on a representative simplified model of a fusion tokamak reactor to study the effects of material cross sections on neutron transport and effects of neutron transport parameters (e.g. mean distance for the collision flux estimator, threshold for delta-tracking, etc.) and source definition on results. In the second step, presented in this paper, the code was validated on a detailed model of the tokamak JET. In the first section the detailed MCNP model of the JET tokamak is presented and the process of converting it to Serpent. An analysis of the transport parameter dt in Serpent, defining the threshold for delta-tracking transport method, was performed
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