Biomedical Engineering Reference
In-Depth Information
ture gamma rays, these shields can acquire induced radioactivity through neutron
capture or other reactions.
Examples of neutron attenuation in a hydrogenous material are provided by the
depth-dose curves in Chapter 12 for monoenergetic neutrons normally incident on
tissue slabs. Figure 12.10 for 5-MeV neutrons, for instance, shows that the absorbed
dose decreases by an order of magnitude over 30 cm. The energy spectrum of the
neutrons changes with the penetration depth as the original 5-MeV neutrons are
moderated. The relative number of thermal neutrons at different depths can be
seen from the dose curve labeled
E
γ
for the
1
H(n,
γ
)
2
H thermal-neutron capture
reaction. The thermal-neutron density builds up to a maximum at about 10 cm
and thereafter falls off as the total density of neutrons decreases by absorption. In
paraffin, the half-value layer for 1-MeV neutrons is about 3.2 cm and that for 5-MeV
neutrons is about 6.9 cm.
Neutron shielding can sometimes be estimated by a simple “one-velocity” model
that employs neutron removal cross sections. Such shielding must be sufficiently
thick and the neutron source energies so distributed that only the most penetrat-
ing neutrons in a narrow energy band contribute appreciably to the dose beyond
the shield. The neutron dose can then be represented by an exponential function
of shield thickness. Conditions must also be such that the slowing-down distance
from the most penetrating energies down to 1 MeV is short. In addition, the shield
must contain enough hydrogen to assure a short average transport distance from
1 MeV down to thermal energy and the point of absorption. The removal cross sec-
tions for various elements are roughly three-quarters of the total cross sections (ex-
cept
∼0.9
for hydrogen). Most measurements of removal cross sections have been
made with fission-neutron sources and shields of such a thickness that the princi-
pal component of dose arises from source neutrons in the energy range 6-8 MeV.
Table 15.5 gives macroscopic removal cross sections,
r
, and attenuation lengths,
1/
r
, in some shielding and reactor materials.
Table 15.5
Macroscopic Neutron Removal Cross Sections and
Attenuation Lengths in Several Materials
Macroscopic Removal
Cross Section
Attenuation Length
r
(cm
-1
)
Material
1/
r
(cm)
Water
0.103
9.7
Paraffin
0.106
9.4
Iron
0.1576
6.34
Concrete (6% H
2
O by weight)
0.089
11.3
Graphite (density 1.54 g cm
-3
)
0.0785
12.7
Source
: Data in part from
Protection Against Neutron Radiation
,
NCRP Report No. 38, National Council on Radiation
Protection and Measurements, Washington, D.C. (1971).
