Modeling and simulation of a TRIGA MARK-II research reactor using WIMSD-5B and CITATION codes. (June 2019)

• Record Type: Journal Article
• Authors: Benaalilou, K.; El Bardouni, T.; Boulaich, Y.; El Yaakoubi, H.; Chham, E.; Lahdour, M.
• Is Part Of: Applied radiation and isotopes. Volume 148(2019:Jun.)
• Page Start: 64
• Page End: 75
• Publication Date: 2019-06
• Subjects: ENDFB-VII.1 -- JEFF-3.1.1 -- JENDL-4.0 -- NJOY-99 -- WIMSD-5B -- CITATION; Radiology -- Periodicals; Radiation -- Industrial applications -- Periodicals; Nuclear chemistry -- Periodicals; Internet resource; Periodical; 660.298
• DOI: 10.1016/j.apradiso.2019.01.034
• Languages: English
• ISSNs: 0969-8043

Abstract

Abstract: The main objective of this study is to analyse neutronic safety parameters of the Moroccan TRIGA Mark-II research reactor using the WIMSD-5B and CITATION computer codes. New 172-group libraries of multi-group constants for the lattice code WIMSD-5B have been generated for all isotopes presented in the TRIGA reactor core by processing nuclear data from ENDFB-VII.1, JENDL-4.0 and JEFF-3.1.1 using NJOY99. The lattice code WIMSD-5B was employed to generate multi-group cross sections in the suitable format that will be used by the 3-dimensional diffusion code CITATION. This later was used to calculate various neutronic safety parameters of the TRIGA Mark-II research reactor, such as reactivity excess and neutron fluxes profiles. The results of these calculations are compared to the results of Monte Carlo calculation based on MCNP code. A good agreement is achieved and the current computation scheme will be adopted for our further coupling neutronic/thermal-hydraulic study of the Moroccan TRIGA reactor. Highlights: Simulation of the Moroccan TRIGA Mark-II research reactor. WIMSD-5B code and CITATION have been utilized to model the core. Calculations are performed by use of ENDFB-VII.1, JENDL-4.0 and JEFF-3.1.1 libraries. Calculation results are compared to experiment and MCNP simulation results.