NUCLEONICS DATA SHEET No. 28 
Neutron Physics 
Fission-Neutron Cross Sections for Threshold Reactions 
By ROBERT S. ROCHLIN 
General Engineering Laboratory, General Electric Company, Schenectady, New York 
THIS DATA SHEET presents published and 
unpublished data on fission-spectrum- 
neutron activation by the (n,p), (n,a) 
and (n,2n) reactions. We shall refer 
to these three reactions as threshold 
reactions, although many of them are 
exoergic. Although many (n,7) cross 
sections have been measured (1) for 
both thermal and _fission-spectrum 
neutrons, data on _ reactor-neutron 
activation by the threshold reactions 
are comparatively scanty. 
Cross sections for threshold reactions 
are needed for calculating activation 
levels of reactor coolants and com- 
ponents, especially when these re- 
actions produce longer-lived activities 
then the corresponding (n,7) reactions. 
For example, in the case of aluminum 
the (n,y) product decays with a 2.3-min 
half-life, while the (n,a) product has a 
half-life of 15 hr. 
An example of the usefulness of 
threshold reactions for neutron-activa- 
tion analysis of materials is the detec- 
tion of aluminum in silicon, using the 
Al?7(n,a)Na*4 reaction. The Al??(n,7) 
Al?® reaction cannot be used in this 
case, because Al?* is also produced 
from silicon by the Si%8(n,p)Al?* re- 
action. Na?4, on the other hand, 
cannot be produced from silicon by 
reactor neutrons. 
Threshold reactions are also useful 
for measuring fast-neutron fluxes in 
nuclear reactors. By using several 
elements that have different activation 
thresholds, information can be obtained 
about the neutron energy spectrum. 
Neutrons produced by fission of 
uranium-235 have a distribution of 
energies (2) given by 
N(E) dE = (2/ne)¥e-¥ sinh (2E)¥ dE 
The cross section for a neutron-induced 
reaction depends on the energy of the 
incident neutrons. For fission-spec- 
246 
trum neutrons, an ‘‘effective cross 
section” is defined as (3) 
o. = P/NF 
where P is the number of product 
nuclei formed per unit time, N is the 
number of target nuclei, and F is the 
total flux density of fission neutrons 
of all energies, whether or not they 
contribute to the reaction. 
Because of the steep slope of the 
fission spectrum above 2 Mey, the 
effect of moderated neutrons on the 
spectrum can usually be neglected in 
the energy range above 2 Mey. 
Neutron energy thresholds for en- 
doergic reactions are given by Er = 
—Q(A + 1)/A, where A is the atomic 
weight of the target nuclide, and the Q 
of the reaction can be calculated from 
isotopic-mass tables (4). Threshold 
energies are listed in the table. Nega- 
tive “thresholds” indicate exoergic 
reactions, for which even thermal 
neutrons may have a cross section 
greater than zero. However, even for 
exoergic reactions, the potential barrier 
against particle emission (3) usually 
causes the cross sections to be small for 
thermal neutrons. 
Inthoff (5) has prepared a nomogram 
for fission-neutron (n,p) and (n,qa) 
cross sections, based upon the theory 
described by Hughes (3). This theory 
gives rough agreement with many 
experimental values, but in some cases 
(e.g., Ti, Cs, Tl) there are disagree- 
ments by factors of 100 or more (6-8). 
Experimental Techniques 
In the present work, samples of Al, 
Cd, Co, Cu, Fe, Ge, Ni, Se and Zn 
were irradiated in the Brookhaven 
National Laboratory graphite reactor. 
The fission-neutron-flux calibration was 
based upon an assumed effective cross 
section of 0.60 mb for Al??(n,a)Na*4, 
To reduce interference from impuri- 
ties, Matthey ‘‘Specpure” oxide sam- 
ples were used. To check on inter- 
ference from activation of impurities 
by thermal neutrons, samples of each 
element were irradiated with and 
without a tight wrapping of cadmium. 
Since the thermal-neutron flux inside 
the cadmium wrapping was negligible, 
the ratio of measured activities of 
samples with and without cadmium 
would depart from unity if thermal- 
neutron activation were appreciable. 
In the present work, calculation of this 
ratio revealed no activities due to 
thermal-neutron activation of im- 
purities in any of the samples. 
Chemical separations were made to 
facilitate measurement of weak activi- 
ties, which would otherwise have been 
masked by stronger ones. 
To obtain unique identification of 
each reaction product, both gamma- 
ray energies and half-lives were meas- 
ured. Values of photons/disintegra- 
tion for each observed gamma peak 
were taken from references 9, 10 or 11. 
The relative counting efficiency of the 
Nal scintillation spectrometer as a 
function of gamma-ray energy was ob- 
tained by comparing observed peaks 
of Mn*® and Br®? gamma rays with 
their known relative intensities (9). 
* * * 
I want to thank S I. Friedman for his 
diligent assistance in operating the spectrom- 
eter and calculating data, Miss B. A. 
Thompson and A. Eldridge for making the 
chemical separations, and W. W. Schultz for 
continual encouragement and support. 
BIBLIOGRAPHY 
1. D. J. Hughes, R. B. Schwartz, ‘‘ Neutron 
Cross Sections,’’ BNL-325, 2nd ed. (1958) 
2. Handbook 63, Fig. 16 and Table 7 (National 
Bureau of Standards, U. S. Department of 
Commerce, Washington, D. C., 1957); also 
B. E. Watt, Phys. Rev. 87, 1037 (1952) 
8. D. J. Hughes, ‘Pile Neutron Research,” 
Ch. 4 (Addison-Wesley Publishing Co., Cam- 
