Geodesic
An approximative method of solving the flat thermal neutron flux problem for an infinite cylindrical homogenized reactor fueled with natural uranium
Acta Universitatis Carolinae. Mathematica et Physica, Tome 13 (1972) no. 2, pp. 61-71 Cet article a éte moissonné depuis la source Czech Digital Mathematics Library

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     author = {Mjasnikov, A. and Zezula, R.},
     title = {An approximative method of solving the flat thermal neutron flux problem for an infinite cylindrical homogenized reactor fueled with natural uranium},
     journal = {Acta Universitatis Carolinae. Mathematica et Physica},
     pages = {61--71},
     year = {1972},
     volume = {13},
     number = {2},
     language = {en},
     url = {http://geodesic.mathdoc.fr/item/AUCMP_1972_13_2_a6/}
}
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Mjasnikov, A.; Zezula, R. An approximative method of solving the flat thermal neutron flux problem for an infinite cylindrical homogenized reactor fueled with natural uranium. Acta Universitatis Carolinae. Mathematica et Physica, Tome 13 (1972) no. 2, pp. 61-71. http://geodesic.mathdoc.fr/item/AUCMP_1972_13_2_a6/