The fast neutron fluence and the activation monitor activity calculations using fine multigroup and continuous nuclear data

Result code in IS VaVaI

<a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F49777513%3A23220%2F24%3A43973885" target="_blank" >RIV/49777513:23220/24:43973885 - isvavai.cz</a>

Result on the web

<a href="https://www.epj-conferences.org/articles/epjconf/abs/2024/18/epjconf_isrd-17_02007/epjconf_isrd-17_02007.html" target="_blank" >https://www.epj-conferences.org/articles/epjconf/abs/2024/18/epjconf_isrd-17_02007/epjconf_isrd-17_02007.html</a>

DOI - Digital Object Identifier

<a href="http://dx.doi.org/10.1051/epjconf/202430802007" target="_blank" >10.1051/epjconf/202430802007</a>

Result language

angličtina

Original language name

The fast neutron fluence and the activation monitor activity calculations using fine multigroup and continuous nuclear data

Original language description

The fast neutron fluence and the activation monitor activities are routinely calculated with TORT deterministic code and BUGLE-B7 nuclear data library with 47 broad energy groups. The objective of the paper is to analyse options to improve reactor dosimetry transport calculations. There are two paths to improve reactor dosimetry calculations. Increasing geometry, angular and energy mesh size is applicable for TORT code while using newer nuclei data libraries is relevant for both deterministic and Monte Carlo codes. Two new calculation options (improved TORT and Monte Carlo MCNP6) were compared with the standard TORT calculation for VVER-440 Dukovany Unit 3 Cycle 31. The fast neutron fluence with 0.5 MeV threshold as well as activity of Fe, Ni, Ti, Cu, Mn and Nb monitors were evaluated. Standard TORT calculations were improved from S16P3 to S30P3 with three times finer axial mesh size, 120° core symmetry r-9 mesh size with 0.5° step and fine multigroup libraries VITAMIN-B7 with 199 neutron energy groups and ENDF/B-VII.1 with 200 neutron energy groups. Both EMDF/B and IRDFF activation cross sections were used The drawback of expanded mesh size is raised calculation runtime since TORT deterministic code is not parallelized and one calculation can require multiple weeks of CPU time An alternative option of using MCNP6 Monte Carlo code with continuous ENDF/B-VE.1 nuclear data with detailed 3-D geometry and pin-wise effective neutron source prepared by MOBY-DICK diffusion code reactor analysis was explored. It was found that using finer mesh size affects reactor dosimetry tallies less than the choice of nuclear data library. BUGLE-B7 and VITAMIN-B7 produce results typically within 1% difference. ENDF/B-VII.1 calculations with 200 neutron energy groups with TORT code are even in better agreement with MCNP6 calculations with continuous nuclear data libraries The largest differences of around 2% were observed between VITAMIN-B7 library based on ENDF/B-WO nuclear data and ENDF/B-VII.l library. Nuclear data library has larger impact on the results with up to 7 % difference between all 0.5 MeV fast neutron fluence calculations. The largest impact of nuclear data was observed for Mn(n,2n) monitor.

Czech name

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Czech description

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Type

D - Article in proceedings

CEP classification

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OECD FORD branch

20305 - Nuclear related engineering; (nuclear physics to be 1.3);

Project

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Continuities

I - Institucionalni podpora na dlouhodoby koncepcni rozvoj vyzkumne organizace

Publication year

2024

Confidentiality

S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Article name in the collection

EPJ Web of Conferences

ISBN

978-2-7598-9127-6

ISSN

2101-6275

e-ISSN

2100-014X

Number of pages

8

Pages from-to

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Publisher name

EDP Sciences

Place of publication

Les Ulis

Event location

Lausanne, Switzerland

Event date

May 21, 2023

Type of event by nationality

WRD - Celosvětová akce

UT code for WoS article

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