)233Th, 98Mo(n,)99Mo, 186W(n,)187W, 115In(n,)116m1In, and 92Mo(n,p)92mNb Cross Sections in the Energy Range of 1.6 to 3.7 MeV">
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Measurement of 232Th(n,)233Th, 98Mo(n,)99Mo, 186W(n,)187W, 115In(n,)116m1In, and 92Mo(n,p)92mNb Cross Sections in the Energy Range of 1.6 to 3.7 MeV

Megha Bhike, B. J. Roy, A. Saxena, R. K. Choudhury, S. Ganesan

Nuclear Science and Engineering / Volume 170 / Number 1 / January 2012 / Pages 44-53

Technical Papers

Neutron-induced reaction cross sections for the reaction 232Th(n, )233Th have been measured at neutron energies of 1.6 ± 0.03 MeV, 2.2 ± 0.03 MeV, 3.0 ± 0.03 MeV, and 3.7 ± 0.03 MeV. We have also measured cross sections for the reactions 98Mo(n, )99Mo, 186W(n, )187W, 115In(n, )116m1In, and 92Mo(n, p)92mNb at a neutron energy of 3.2 ± 0.03 MeV. The 7Li(p, n)7Be reaction was used as the neutron source with the proton beam from the 14-MV Pelletron accelerator, Mumbai, and the standard off-line gamma counting method was followed for activation measurement. The present measurements supplement the existing data and provide new data in the neutron energy range where no results are available. While the cross-section values for the 98Mo(n, )99Mo and 186W(n, )187W reactions are reported for the first time, the data for 92Mo(n, p)92mNb exists with a large discrepancy between the two available data sets. For the 115In(n, )116m1In reaction, our measurement at 3.2 MeV is an additional data point where there exists significant disagreement among the data measured by different groups. The measurements are performed relative to the 115In(n, n)115mIn and 197 Au(n, )198 Au cross sections of International Reactor Dosimetry File 2002. Detailed theoretical calculations using the statistical model code EMPIRE-II (latest version EMPIRE-2.19) have been performed. Good agreement with the present data along with the existing data set has been obtained by suitable adjustment of the level density parameter for all the systems. The experimental and theoretical results have been compared with the recent evaluations of ENDF/B-VII.0, JENDL-4.0, and JEFF-3.1.

 
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