The need for cross section measurements for neutroninduced reactions
• If no cross section measurement exists, alternative strategies are:
• The cross section for the corresponding proton-induced reaction is used.
• Theoretical models are used to estimate the needed cross sections.
• The cross section is inferred from analysis of the results from irradiating thick
target stacks with protons.
• None of these strategies is as good as an actual measurement!
• To remedy this situation
We are measuring cross sections for neutron induced reactions:
• LANSCE to make an energy integrated (average) cross section measurement
using ‘white’ neutron beams 0 – 750 MeV.
• In the first year we will measure cross sections for the production of:
10Be, 14C, Ne, 26Al
from
O and Si
The aim of the experiment in 2005
2 or 3 irradiations.
These times are calculated assuming 1.8 microsec spacing and ~4-5
nA protons on the W target.
50 x 50 mm SiO2 and/or 50 mm diameter Si targets.
10 days using 3 mm thick targets:
SiO2(n,x)10Be
Si or SiO2(n,x)20,21,22Ne
1 day using 1 mm targets;
SiO2(n,x)26Al or Si(n,x)26Al
SiO2(n,x)14C
1/2 day using 1 mm thick targets:
SiO2(n,x)3He and Si(n,x)3He.
Total target irradiation time requested = 15 days
In addition, we need ~4 days with a long micropulse spacing to
characterize the low energy flux.
Experimental Procedure at LANSCE
• Neutron beams cover the whole target stack.
• Total stack thickness is designed to attenuate <10% of the beam at all neutron
energies.
• Irradiation times are designed to produce the optimum number of product atoms
for determination using AMS or MS by appropriate collaborators.
• Short-lived radionuclides are measured using non-destructive gamma-ray
spectroscopy.
• AMS and MS determinations will be made later.
Experiments at LANSCE
LANSCE: 4FP15R 2002
Target in target holder
• The energy spectrum ranges from 0.1 – 750 MeV.
• The neutron fluence is monitored directly using an uranium fission chamber.
E vs. flux (corrected)
MCNP
1.00E+00
Run 5
original neutron flux
corrected flux n/MeV/inc.n/cm^2
1.00E-01
1.00E-02
1.00E-03
1.00E-04
1.00E-05
1.00E-06
1.00E-01
1.00E+00
1.00E+01
E upper bin lim it [MeV]
1.00E+02
1.00E+03
Average cross sections measured at LANSCE 1998-2003 include:
C
7Be
7.1 
0.8
SiO2
6.7  0.8
Mg
1.4  0.2
Al
1.5  0.3
22Na
9.1  1.0
20.4  2.3
7.8  1.0
24Na
6.9  0.8
27.1  3.1
23.6  2.7
Ni
Cu
Ti
46Sc
49.8 
5.9
Fe
4.3  0.6
1.4  0.2
1.0  0.1
5.8  0.7
14.9  1.7
48V
10.0 
1.1
51Cr
40.4  4.6
24.7  2.8
8.2  0.9
52Mn
9.6  1.1
7.7  0.9
1.7  0.2
54Mn
69.7  8.0
17.5  2.3
9.8  1.2
56Co
29.4  3.3
3.4  0.4
57Co
130.0  15.0
19.2  2.3
58Co
110.0  13.1 28.4  3.2
Au
194Au
145.0  17.0
196Au
271.0  31.0
198Au
15.1  1.8
7-27-2004
100
nat
60
Cu(n,x) Co
Cross section (mb)
10
1
MC-ALICE
This work
LANSCE average (1.25-750 MeV)
KI99
0.1
nat
60
Cu(p,x) Co MI92
Other measurements
0.01
1
10
100
1000
Incident neutron energy (MeV)
natCu(p,x)60Co
from S. J. Mills, G. F. Steyn and F. M. Nortier, Appl. Rad. Isot. 43, 1019, 1992
MC-ALICE calculations courtesy of Mark Chadwick.
27
22
27
22
Al(p,x) Na and Al(n,x) Na
Cross section (mb)
100
10
1
27
22
27
Al(n,x) Na iTL
Al(p,x) Na
derived excitation function
Imamura et al.
0.1
22
27
22
average Al(n,x) Na LANSCE
0.01
10
100
1000
Neutron energy (MeV)
The excitation function was constructed from the measured values and the adopted
values of W. S. Gilbert et al (1968) of 10 mb for En>60 MeV and ‘tweaked’ to get
reasonable agreement with the average value measured at LANSCE..