Neutronic Study of Fuel Depletion for the MNSR Research Reactor Using DRAGON5 Code

Zain, Jamal Al; Hajjaji, O. El; Bardouni, T. El

doi:10.3103/s0027134919060031

Cited by 1 publication

(2 citation statements)

References 4 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Over a relatively short period of time for days or weeks, the atomic density of the fuel is also relatively constant 28 . As the density of the atom and microscopic cross‐section 31 are constant, the macroscopic cross‐section(XS) should also be constant 28 . It is evident that if the macroscopic cross‐section and the pin‐cell volume are constant, 31 then the pin‐cell power and neutron flux are directly proportional.…”

Section: Resultsmentioning

confidence: 99%

“…As the density of the atom and microscopic cross‐section 31 are constant, the macroscopic cross‐section(XS) should also be constant 28 . It is evident that if the macroscopic cross‐section and the pin‐cell volume are constant, 31 then the pin‐cell power and neutron flux are directly proportional. The value of the neutron flux for a certain power level will increase very slowly during a period of months due to fuel burn‐up and the resulting decrease in the density of the atom and the macroscopic cross‐section 28 …”

Section: Resultsmentioning

confidence: 99%

See 1 more Smart Citation

Neutronic and burn‐up calculations of the ( ThO2‐UO2 ) pin cell benchmark using DRAGON5 and MCNP6 .2 codes with ENDF / B‐VIII .0 nuclear data library

et al. 2021

Self Cite

View full text Add to dashboard Cite

Summary A (ThO2‐UO2) fuel pin benchmark, consisting of 75 w/o Th and 25 w/o U, was utilized for analysis of depletion and inventory of the fuel for the pressurized water reactor (PWR) pin‐cell model. To perform this study and evaluate cross‐section data, the new version VIII.0 of the ENDF/B nuclear data library was utilized to generate a data set in ACE (A Compact ENDF) format at different temperatures (583.1, 600, 621.1, and 900 K), utilizing NJOY2016 and MAKXSF6.2 codes to process the data at the specific temperatures. In this paper, the infinite multiplication factor (kinf), neutron energy flux spectrum (utilizing 172 energy groups), the concentrations and activities of the most important products of fission, the actinide radionuclides accumulated in the pin‐cell, and the total radioactivity were calculated and compared, utilizing the MCNP6.2 and DRAGON5 codes. Furthermore, the burn‐up (BU) dependent behavior of the PWR thorium pin‐cell was validated and compared with the reference results obtained utilizing the CASMO‐4, MIT‐MOCUP, and INEEL‐MOCUP codes. The results of this study show good agreements between all codes analysed.

show abstract