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Neutron Production in Several
Americium Compounds
Erik F. Shores
American Nuclear Society
A
LosAlamos
M
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a
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Form 836 (8/00)
Neutron Production in Several Americium Compounds
Erik F. Shores
Introduction
Americium, like other alpha emitting actinides, may indirectly produce neutrons when combined
with light target materials. These alpha,n) reaction neutrons, along with well known photon lines,
have been an advantage of the 24 Am isotope for diverse applications such as radiography,
thickness gauges, neutron sources, and even common household smoke detectors.
\
To characterize these sources, the SOURCES code [l
] was used to calculate neutron yields and
spectra from 241Ammetal, americium oxide, and americium aluminum alloys. Such information
may be used as source terms for future transport problems (e.g. shielding calculations).
Assumptions
Several simplifying assumptions were made. For all calculations, 241Am (50 mg) was the sole
nuclide present; contaminants (e.g. Li, Be, 0) were not considered. For the oxide compounds,
americium to oxy en ratios were taken as 1:2 and 2:3 while natural oxygen abundances were
assumed: 0.04% 9 70 and 0.2% l80[2]. Atom fractions were thus readily deduced (e.g. an oxygen
atom fraction of 2/3 in the Am02 case results in an
fraction of 1.33e-3).
Regarding Am atom density, held constant for each scenario, the following calculation was made:
[0.60221367e24 Am atoms/mole] [ l mole/241g] [ l g/l OOOmg] = 2.49881 19el8 Am atoms/mg.
ResuIts
Table 1 contains neutron yields for six americium configurations. The metal, oxides, and alloys
were run as homogeneous problems while the interface case was run in both two- and threeregion interface modes.
Table 1. Calculated Neutron Yields
Neutron Yield (n/s-mg Am-241)
Calculated
Reference
241
Am metal
1.24e-3
1.07e-3 [5], 1.767e-3*40% [3] calc.
241
Am203
2.24
24 1
Am02
2.76
2.78k0.41 141, 2.88~0.07[5] meas.
241
AmAl alloy
20.1
241
Am/AI interface
47.38 n/s-cm2
Note: does not include SF
241
AmAI5 alloy
51.8
241
Am infinitely dilute AI 85.53
92.42 [7] calc.
Compound
Normalized neutron spectra (60 groups between 0 and 6.0 MeV) for these scenarios are shown in
Figure 1. The data points are plotted at the midpoint of each energy bin and curves are shown to
guide the eye. The oxide spectrum is representative of both oxides and that of the alloy
represents both AmAl and AmAI,. The only source of neutrons in purely metallic 241Amis
spontaneous fission (SF) and that curve, typical of a fission spectrum, is readily identified in
Figure 1.
8.E-02
I
I
7.E-02
6.E-02
u
Q)
+AmAl
5.E-02
-SF
.-N 4.E-02
-t-
3.E-02
2
Am02
-x- Interface
2.E-02
1.E-02
O.E+OO
1
0
2
3
4
5
Neutron Energy (MeV)
Figure 1. Normalized neutron spectra for several compounds
In the case of an interface, we can imagine americium metal encapsulated by aluminum. The
three-region-interface(TRI) option in SOURCES allows us to consider a varied aluminum
thickness (region B) between a thick americium layer (region A) and infinite layer of air (region C).
Regardin the latter, if we assume a simple atmospheric composition, the target nuclides of
interest (1%N, "0, and '*O) may be ignored as nitrogen's (alpha,n) threshold is above americium's
alpha energy and production in oxygen is negligible. Thus, the resultant neutrons are due to
(alpha,n) production in the aluminum layer.
Figure 2 is a plot of neutron production as a function of region B thickness. As expected, the yield
increases as AI thickness increases until a maximum, the limit of an infinitely thick target, is
reached. Indeed, a simple two-region interface problem confirmed the TRI results with a 47.38
n/s-cm2yield. Furthermore, the range of a 5.54 MeV alpha particle, the maximum encountered
from americium, is 0.002 cm [6] and higher yields would not be expected at greater thicknesses.
/
h
p 40
35
\
cn 30
.5 25
-0
75 20
15
S
2 10
51
0 '
1.OE-06
/
/
1
+
/
-A,
1.OE-05
I
1.OE-04
,
1.OE-03
,
1.OE-02
<
1.OE-01
AI thickness (cm)
Figure 2. Neutron yields from an Am-AI interface
1.OE+OO
Discussion
The metal’s neutron yield is strictly due to spontaneous fission. While data differences (e.g. half
life) allow slight variations in calculated SF yields, the SOURCES result is reasonable when
compared to two calculations.
The neutron yield for the dioxide is in good agreement with two measurements. While slightly
lower, it bears emphasis that impurities were not considered. Contaminants, of course, could
make a significant contribution and may need to be considered. In the case of Ref. [5]’s sample,
impurities were considered to be e 1% based the absence of (alpha,n) reaction photons. Sample
purity and a fluorine contaminant are also discussed in Ref [4]. In that case, a fluorine
concentration of a few parts per thousand resulted in a neutron contribution nearly equal to that of
oxygen in the AmO, sample. On the other hand, Croft’s analysis of a Pu02sample indicated the
neutron production rate was enhanced by no more than 2.4% over the pure oxide case [3]. The
main light elements in that work, lithium, beryllium, carbon, fluorine, magnesium, and aluminum
were each e 10 ppm by weight relative to the dioxide. Boron (e 20 ppm), sodium (e60),and other
elements were also present. These typical examples are simply presented for consideration.
*Inthe two oxide cases, the Am203has a lower yield because of the reduced oxygen fraction (3/5
vs 2/3). A further reduction is expected if the oxygen fraction dropped to 1/2 (AmO). Recall the
number of americium atoms was held constant for a 50 mg mass. A further consistency check
regarding Figure 1’s americium oxide spectrum is similarity between AmO2and PuO, spectra
implied by similar plutonium alpha energies. To be sure, both spectra peak near 2.5 MeV.
Regarding the aluminum alloys, a higher target fraction in AmAI5 compound results in a greater
yield. The yield should reach a limit as the americium becomes dilute in aluminum. Such a
scenario has been calculated elsewhere to be 92.42 n/s-mg [7].The increasing nature of
SOURCES AmAl yields thus appears reasonable.
For total neutron production, the yields may be scaled accordingly (e.g. multiplied by appropriate
mass). Assuming an aluminum-encapsulatedamericium source (interface case), the total yield for
a thickness of interest must be multiplied by the contact area. An important distinction on the
scenarios considered is that interface yields are strictly due to (alpha,n) sources and do not
include the SF contribution. The SF component, appropriately scaled, could be added as desired.
Finally, Table 2 contains absolute spectra calculated for each case and it’s hoped such tabulation
facilitates further manipulation (e.g. sdef description for MCNPTMtransport calculations).
Measurements would certainly be of interest to further benchmark the code and the author
welcomes suggestions and comments in this regard.
Table 2. Absolute Neutron Spectra
Energy Bin
(MeV)
0.0 - 0.1
0.1 - 0.2
0.2 - 0.3
0.3 - 0.4
0.4 - 0.5
0.5 - 0.6
0.6 - 0.7
0.7 - 0.8
0.8 - 0.9
0.9 - 1.o
1.0- 1.1
1.1 - 1.2
1.2- 1.3
1.3- 1.4
1.4 - 1.5
1.5-1.6
1.6- 1.7
1.7- 1.8
1.8- 1.9
1.9 - 2.0
2.0 - 2.1
2.1 -2.2
2.2 - 2.3
2.3 - 2.4
2.4 - 2.5
2.5 - 2.6
2.6 - 2.7
2.7 - 2.8
2.8 - 2.9
2.9 - 3.0
3.0 - 3.1
3.1 - 3.2
3.2 - 3.3
3.3 - 3.4
3.4 - 3.5
3.5 - 3.6
3.6 - 3.7
3.7 - 3.8
3.8 - 3.9
3.9 - 4.0
4.0 - 4.1
4.1 -4.2
4.2 - 4.3
4.3 - 4.4
4.4 - 4.5
4.5 - 4.6
4.6 - 4.7
4.7 - 4.8
4.8 - 4.9
4.9 - 5.0
5.0 - 5.1
5.1 - 5.2
5.2 - 5.3
5.3 - 5.4
5.4 - 5.5
5.5 - 5.6
5.6 - 5.7
5.7 - 5.8
5.8 - 5.9
5.9 - 6.0
Total
Am
(n/s-mg Am)
1.416E-05
2.470E-05
3.038E-05
3.413E-05
3.667E-05
3.838E-05
3.945E-05
4.003E-05
4.022E-05
4.009E-05
3.972E-05
3.914E-05
3.839E-05
3.751E-05
3.653E-05
3.546E-05
3.434E-05
3.317E-05
3.197E-05
3.075E-05
2.952E-05
2.830E-05
2.708E-05
2.588E-05
2.470E-05
2.354E-05
2.241 E-05
2.131 E-05
2.024E-05
1.921E-05
1.821 E-05
1.725E-05
1.633E-05
1.544E-05
1.459E-05
1.378E-05
1.300E-05
1.226E-05
1.155E-05
1.088E-05
1.024E-05
9.633E-06
9.057E-06
8.51OE-06
7.992E-06
7.502E-06
7.038E-06
6.601 E-06
6.187E-06
5.797E-06
5.429E-06
5.083E-06
4.756E-06
4.449E-06
4.1 60E-06
3.889E-06
3.634E-06
3.394E-06
3.169E-06
2.958E-06
1.240E-03
Am02
(n/s-mg Am)
1.321E-02
2.181 E-02
2.310E-02
2.366E-02
2.705E-02
2.820E-02
2.779E-02
2.882E-02
2.953E-02
3.1 11E-02
3.380E-02
3.671E-02
4.163E-02
4.606E-02
5.339E-02
6.112E-02
6.802E-02
7.505E-02
8.658E-02
9.544E-02
1.030E-01
1.124E-01
1.209E-01
1.248E-01
1.285E-01
1.301E-01
1.264E-01
i.245~-oi
1.219E-01
1.123E-01
1.024E-01
9.500E-02
8.685E-02
7.627E-02
6.704E-02
5.919E-02
5.048E-02
4.268E-02
3.516E-02
2.684E-02
1.917E-02
1.386E-02
9.913E-03
6.046E-03
2.486E-03
1.418E-03
1.282E-03
1.177E-03
1.094E-03
1.009E-03
8.949E-04
7.884E-04
7.012E-04
5.921 E-04
4.745E-04
3.329E-04
1.997E-04
7.470E-05
4.265E-06
2.958E-06
2.761
Am203
(n/s-mg Am)
1.072E-02
1.770E-02
1.874E-02
1.920E-02
2.195E-02
2.288E-02
2.255E-02
2.339E-02
2.397E-02
2.526E-02
2.746E-02
2.983E-02
3.383E-02
3.742E-02
4.338E-02
4.966E-02
5.526E-02
6.097E-02
7.034E-02
7.752E-02
8.368E-02
9.126E-02
9.813E-02
1.013E-01
1.043E-01
1.056E-01
1.026E-01
1.010E-01
9.890E-02
9.109E-02
8.308E-02
7.705E-02
7.044E-02
6.1 85E-02
5.436E-02
4.799E-02
4.092E-02
3.460E-02
2.850E-02
2.175E-02
1.554E-02
1.124E-02
8.037E-03
4.902E-03
2.018E-03
1.152E-03
1.041E-03
9.556E-04
8.882E-04
8.193E-04
7.265E-04
6.401 E-04
5.693E-04
4.808E-04
3.854E-04
2.705E-04
1.625E-04
6.117E-05
4.057E-06
2.958E-06
2.241
Am-AI
(n/s-cm')
6.199E-01
1.388E+00
1.960E+00
2.401E+OO
2.706E+00
2.907E+00
3.037E+00
3.1 07E+00
3.1 47E+00
3.163E+00
3.130E+00
3.052E+00
2.91 8E+00
2.71 8E+00
2.463E+00
2.170E+00
1.822E+00
1.451E+OO
1.122E+00
8.283E-01
5.774E-01
3.732E-01
2.104E-01
8.988E-02
1.970E-02
8.191E-04
3.633E-10
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
0.000E+00
0.000E+00
0.000E+00
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
0.000E+00
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
0.000E+00
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
0.000E+00
47.38
AmAl
(n/s-mg Am)
3.429E-01
6.299E-01
7.749E-01
9.748E-01
1.197E+00
1.187E+00
1.137E+00
1.063E+00
9.842E-01
9.707E-01
1.084E+00
1.100E+00
1.087E+00
1.067E+00
1.041 E+OO
1.014E+00
9.586E-01
7.927E-01
6.745E-01
5.804E-01
4.854E-01
3.902E-01
2.91 9E-01
1.864E-01
7.841E-02
4.234E-03
2.241E-05
2.131E-05
2.024E-05
1.921E-05
1.821E-05
1.725E-05
1.633E-05
1.544E-05
1.459E-05
1.378E-05
1.300E-05
1.226E-05
1.155E-05
1.088E-05
1.024E-05
9.633E-06
9.057E-06
8.510E-06
7.992E-06
7.502E-06
7.038E-06
6.601 E-06
6.1 87E-06
5.797E-06
5.429E-06
5.083E-06
4.756E-06
4.449E-06
4.160E-06
3.889E-06
3.634E-06
3.394E-06
3.1 69E-06
2.958E-06
20.10
AmA15
(nls-mg Am)
8.846E-01
1.624E+00
1.997E+00
2.512E+00
3.086E+00
3.058E+00
2.930E+00
2.737E+00
2.532E+00
2.497E+00
2.791 E+OO
2.833E+00
2.801 E+OO
2.750E+00
2.683E+00
2.615E+00
2.473E+00
2.045E+00
1.741E+OO
1.499E+00
1.254E+00
1.009E+00
7.547E-01
4.821 E-01
2.029E-01
1.092E-02
2.241 E-05
2.131 E-05
2.024E-05
1.921E-05
1.821E-05
1.725E-05
1.633E-05
1.544E-05
1.459E-05
1.378E-05
1.300E-05
1.226E-05
1.155E-05
1.088E-05
1.024E-05
9.633E-06
9.057E-06
8.510E-06
7.992E-06
7.502E-06
7.038E-06
6.601 E-06
6.1 87E-06
5.797E-06
5.429E-06
5.083E-06
4.756E-06
4.449E-06
4.160E-06
3.889E-06
3.634E-06
3.394E-06
3.169E-06
2.958E-06
51.80
References
[ l ] W.B. Wilson, et al, Los Alamos National Lab. Report LA-UR-02-1839(2002).
[2] Chart of the Nuclides, 15'h Ed. (1996).
[3] S.Croft, Ann. Nucl. Energy, 19, 451-457 (1992).
[4] E.W. Lees and D. Lindley, Annals of Nucl. Energy 5, 133-139 (1978).
[5] T. Kimura, et al, Appl. Radiat. Isot. 37, 121-125 (1 986).
[6] E.F. Shores, Trans. Am. Nucl. SOC.87,429-430 (2002).
[7] S. Croft, 23rd Annual Meeting ESARDA (European Safeguards Research and Development
Association) Symposium on Safeguards and Nuclear Material Management, Bruges,
Belgium, 552-560 (8-10 May 2001).