Data library of irradiated fuel salt and off-gas tank composition for a molten salt reactor concept produced with Serpent2 and SOURCES 4C codes.
Burnup
Molten salt reactors
Nuclear safeguards
SOURCES 4C
Safeguards verification
Serpent
Spent fuel
Journal
Data in brief
ISSN: 2352-3409
Titre abrégé: Data Brief
Pays: Netherlands
ID NLM: 101654995
Informations de publication
Date de publication:
Jun 2024
Jun 2024
Historique:
received:
26
01
2024
revised:
05
03
2024
accepted:
06
03
2024
medline:
29
3
2024
pubmed:
29
3
2024
entrez:
29
3
2024
Statut:
epublish
Résumé
This paper describes the methodology used to create a fuel data library comprising safeguards-relevant quantities that may be useful for verification of spent nuclear fuel (SNF) produced by simulating a concept Molten Salt Reactor (MSR). The Monte-Carlo particle transport code, Serpent2 and the calculation code SOURCES 4C were used to compile this fuel data library. The data library is based on the Compact Molten Salt Reactor (CMSR) concept being developed by Seaborg Technologies (based in Copenhagen, Denmark). The library includes data such as nuclide mass densities for a total of 1398 nuclides (in g/cm
Identifiants
pubmed: 38550234
doi: 10.1016/j.dib.2024.110314
pii: S2352-3409(24)00283-X
pmc: PMC10973575
doi:
Types de publication
Journal Article
Langues
eng
Pagination
110314Informations de copyright
© 2024 The Author(s).