1
Evaluation of neutron cross sections for 244CM, 246CM, and 248CM : [E-Book]
Other Personal Name(s): “...McCrosson, F. J....”

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ENDF/B-IV thermal data testing : methods, results, and recommendations : an invited paper to be presented to the 1976 annual meeting of the American Nuclear Society at -Toronto, Ontario, Canada, on June 13 - 18, 1976 [E-Book] /
Other Personal Name(s): “...McCrosson, F. J....”

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Evaluation and production testing of cross sections for 244, 246, 248Cm : abstract of an invited paper for presentation at the 1976 annual meeting of the American Nuclear Society Toronto, Canada June 13 - 18, 1976 [E-Book] /
Other Personal Name(s): “...McCrosson, F. J....”

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Thermal data testing of ENDF/B-III and prognosis for ENDF/B-IV : proposed for presentation at the American Nuclear Society annual meeting, Philadelphia, Pennsylvania, June 23 - 28, 1974 [E-Book] /
Other Personal Name(s): “...McCrosson, F. J....”

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5
Calculated enhancement of thermal neutron flux from a 252CF source : [E-Book]
Other Personal Name(s): “...McCrosson, F. J....”

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6
Testing of ENDF/B - thermos cross sections for H"O, D"O, C, ZRH2, (C2H4)X, BE, BEO, C6H6, and UO2 : [E-Book]
Other Personal Name(s): “...McCrosson, F. J....”

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7
New production reactor project-management plan : [E-Book]
Other Personal Name(s): “...McCrosson, F. J....”

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