title: Determination of neutron flux distribution in an Am-Be irradiator using the MCNP.
abstract: A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.
authors.affiliation: Center of Applied Studies for Nuclear Development, CEADEN, Calle 30, 502, Havana 6122, Cuba.
authors.forename: K
authors.initials: K
authors.lastname: Shtejer-Diaz
authors.forename: C B
authors.initials: CB
authors.lastname: Zamboni
authors.forename: G S
authors.initials: GS
authors.lastname: Zahn
authors.forename: J Y
authors.initials: JY
authors.lastname: Zevallos-Chávez
mesh_terms: D000576:Americium; D001608:Beryllium; D016754:Boron Neutron Capture Therapy / Q000379:methods*; D003198:Computer Simulation; D004867:Equipment Design; D019544:Equipment Failure Analysis; D007553:Isotope Labeling / Q000295:instrumentation / Q000379:methods; D008956:Models, Chemical; D009502:Neutrons; D011829:Radiation Dosage; D011874:Radiometry / Q000295:instrumentation* / Q000379:methods*; D019275:Radiopharmaceuticals / Q000032:analysis / Q000138:chemical synthesis; D011880:Radiotherapy Planning, Computer-Assisted / Q000379:methods*; D015203:Reproducibility of Results; D012680:Sensitivity and Specificity
chemical_list: D019275:Radiopharmaceuticals; D001608:Beryllium; D000576:Americium