Delayed Neutron Emission Measurements from Fast Fission of U²³⁵ and Np²³⁷

Conference · Mon Sep 16 04:00:00 EDT 1996

OSTI ID:397104

Charlton, William S. ^[1]; Parish, Theodore A. ^[1]; Raman, Subramanian ^[2]; Shinohara, Nubuo ^[3]; Andoh, Masaki ^[3]

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1. Texas A & M Univ., College Station, TX (United States)
2. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
3. Japan Atomic Energy Research Inst. (JAERI), Tokai (Japan)

Experiments have been designed and conducted to measure the periods and yields of delayed neutrons from fast fission of ²³⁵U and ²³⁷Np. These measurements were performed in a pool type reactor using a fast flux in-core irradiation device. The energy dependent neutron flux spectrum within the irradiation device was characterized using a foil activation technique and the SAND-II unfolding code. Five delayed neutron groups were measured. The total yield (sum of the five group yields) for ²³⁵U was found to be 0.0141 ± 0.0009. The total yield for ²³⁷Np was found to be 0.0102 ± 0.0008. The total delayed neutron yield data were found to be in good agreement with previous measurements. The individual group yields reported here are preliminary and are being further refined.

• Research Organization: Texas A & M Univ., College Station, TX (United States); Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Japan Atomic Energy Research Inst. (JAERI), Tokai (Japan)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); Japan Atomic Energy Research Inst. (JAERI), Tokai (Japan)
• DOE Contract Number: AC05-96OR22464
• OSTI ID: 397104
• Report Number(s): CONF-960924--6; ON: DE96015052
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
97 MATHEMATICS AND COMPUTING
ACTINIDE BURNER REACTORS
DELAYED NEUTRON ANALYSIS
DELAYED NEUTRONS
End-Of-Irradiation (EOI)
Evaluated Nuclear Data File (ENDF)
FAST FISSION
FLIP Fuel
Fast Flux Pneumatic Receiver (FFPR)
Foil Activation Method
Gamma Spectroscopy
HIGH-LEVEL RADIOACTIVE WASTES
Monte Carlo Code
NEPTUNIUM 237
Nuclear Criticality Safety Program (NCSP)
SAND-II
SPENT FUEL ELEMENTS
Spent Nuclear Fuel
TRANSMUTATION
TRIGA Reactor
Thermal Neutron Flux Spectrum
URANIUM 235