Radiant energy – Invisible radiant energy responsive electric signalling – With radiant energy source
Reexamination Certificate
1999-02-26
2001-11-20
Hannaher, Constantine (Department: 2878)
Radiant energy
Invisible radiant energy responsive electric signalling
With radiant energy source
C250S395000, C250S358100, C378S053000
Reexamination Certificate
active
06320193
ABSTRACT:
TECHNICAL FIELD
The present invention relates to nuclear densitometry, and, in particular, to a method for non-intrusively identifying a target material enclosed within a container by detecting on-axis, uncollided radiation transmissions of mono-energetic penetrating radiation through the container at first and second radiation beam energies, and applying developed algorithms to estimate either an identifying ratio or to solve an identifying minimization formula, wherein the ratio and the minimization formula are based on characteristic macroscopic neutron cross-sections or linear attenuation coefficients for the target material at the first and second neutron or gamma-ray beam energies, respectively.
BACKGROUND OF INVENTION
Diagnostic nuclear techniques generally involve the use of highly penetrating radiation comprised of nuclear particles to identify a concealed or unknown target material by detecting and measuring the interaction between the nuclear particles of the penetrating radiation and the target material nuclei, and analyzing the absorptive and/or scattering patterns that result from the interaction. For example, Thermal Neutron Activation (TNA) and Fast Neutron Activation (FNA) are neutron transmission techniques for identifying a target material by measuring the spectrum of gamma rays emitted from the target material, as a result of neutron bombardment and the subsequent absorption of neutrons by the target material. Thermal and Fast Neutron Activation methods are characterized by the energy level of the interrogating neutron beam, i.e., TNA uses a neutron beam having a low energy of about 0.025 eV, while FNA involves a very high energy neutron beam of about 14 MeV.
Although a beam of penetrating radiation transmitted through a target material may interact with the target material nuclei to produce identifiable signatures, a number of the nuclear particles comprising the penetrating radiation may also have no interaction with the target material nuclei, such that they pass through the target material, maintaining their initial speed and trajectory. The intensity of the non-interacting or uncollided radiation flux exiting the target material is less than the initial beam intensity, diminished by the fraction of interacting nuclear particles, either absorbed or scattered. For a neutron transmission, the measurement of the number of uncollided neutrons per unit time is quantified as the uncollided neutron flux or intensity, I. Analogously, the detection and measurement of the number of non-interacting photons per unit time, resulting from interrogation of the target material by a gamma-ray beam, is quantified as the uncollided gamma-ray flux or intensity, I.
With respect to neutron transmission measurements, each chemical element has a microscopic parameter, referred to as the neutron cross-section, that represents the probability of a neutron interaction with a nucleus of the target material, depending upon the velocity of the neutron. The sum of the cross-sections at a given neutron velocity is the total microscopic neutron cross-section, expressed in units of effective target area per atom (cm
2
/atom). The corresponding macroscopic neutron cross-section (cm
2
/cm
3
) is the product of the total microscopic cross-section (cm
2
/atom) and the atomic density of the target material (atoms/cm
3
). For a chemical compound composed of n different elements, the macroscopic neutron cross-section &Sgr;
C
is described by Equation (1), and Equation (2) is an expression of the molecular density for a compound, C.
Σ
C
=
N
C
∑
i
=
1
n
υ
i
σ
i
(
1
)
N
C
=
N
(
ρ
M
W
)
C
(
2
)
where &Sgr;
C
is the total macroscopic neutron cross-section for compound C;
N
C
is the molecular density, molecules of compound C/cm
3
;
N is Avogadro's number (6.02e23 molecules/gm mole);
&ngr;
i
is the number of atoms of element i per molecule of compound C; and
&sgr;
i
is the total microscopic neutron cross-section for element i at a given neutron velocity.
Tables I and II below list physical characteristics of six chemical agents and their neutron kinetic energy dependence (½ MV
2
), respectively.
TABLE I
Physical Characteristics of Selected Chemical Agents
Mass
Chemical
Molecular
Attenuation
Agent
Type
Formula
Weight
Density
Coefficient
GB
Nerve
C
4
H
10
O
2
PF
140.1
1.09
0.159
GA
Nerve
C
5
H
11
O
2
PN
2
162.3
1.07
0.156
GD
Nerve
C
7
H
16
O
2
PF
182.2
1.02
0.151
VX
Nerve
C
11
H
26
O
2
PSN
267.4
1.008
0.152
TABLE II
Neutron Kinetic Energy Dependence of Selected Chemical Agents
Energy
&sgr;
GA
&Sgr;
GA
&sgr;
GB
&Sgr;
GB
&sgr;
GD
&Sgr;
GD
&sgr;
VX
&Sgr;
VX
keV
cm
2
/mol
cm
2
/cm
3
cm
2
/mol
cm
2
/cm
3
cm
2
/mol
cm
2
/cm
3
cm
2
/mol
cm
2
/cm
3
150
0.596
0.63772
0.609
0.66381
0.728
0.74256
0.789
0.79689
220
0.513
0.54891
0.528
0.57552
0.63
0.6426
0.677
0.68377
272
0.475
0.50625
0.525
0.57225
0.596
0.60996
0.643
0.64943
350
0.433
0.46331
0.464
0.50576
0.541
0.55182
0.559
0.56459
450
0.477
0.51039
0.511
0.55699
0.557
0.56814
0.555
0.56055
550
0.346
0.37022
0.382
0.39458
0.428
0.43656
0.451
0.45551
660
0.324
0.34666
0.334
0.36408
0.393
0.40086
0.415
0.41915
750
0.302
0.32314
0.319
0.34771
0.373
0.38048
0.391
0.39491
850
0.283
0.30281
0.299
0.32591
0.35
0.357
0.367
0.37067
1000
0.31
0.3317
0.323
0.35207
0.358
0.36516
0.366
0.36968
For distance &dgr;x within compound C, the probability of a neutron-nucleus collision is &Sgr;&dgr;x, provided that this product is much less than one (<<1.0). The probability that a neutron travels distance X
C
without undergoing an interaction is a Posison distribution described by e
−&Sgr;cXc
. If a neutron beam passes in series through m sucessive compounds, then the probability of zero neutron-nucleus interactions occurring over the total path length is described by Equation (3).
ⅇ
-
∑
j
=
1
m
Σ
j
x
j
(
3
)
Finally, the uncollided neutron intensity, I, is represented in the attenuation equation, Equation (4).
I
=
ϵ
A
d
4
π
R
2
S
0
ⅇ
-
∑
j
=
1
m
Σ
j
x
j
(
4
)
where &egr; is the detector efficiency at the original neutron velocity, representing the fraction of uncollided neutrons at a certain velocity that produce counted pulses;
S
o
is the source strength of neutrons per second, at a certain velocity, emitted in all directions;
A
d
4
π
R
2
is the solid angle subtended by a detector of sensitive area A
d
located at a distance R from the point source, representing the fraction of S
o
neutrons leaving the source in a direction toward the detector; and
e
−&Sgr;
&Sgr;jXj
is the fraction of uncollided neutrons incident on the detector area A
d
.
An analogous attenuation equation describes the uncollided gamma-ray intensity from gamma-ray transmissions, except that the variable &Sgr;
C
, the total macroscopic neutron cross-section for compound C, is replaced by the variable &mgr;
C
, the linear attenuation coefficient for compound C. The gamma-ray linear attenuation coefficient is a function of the compound's molecular structure and the gamma photon energy.
Importantly, where the parameters of Equation (4) are known, a measurement of the uncollided neutron intensity, I, allows calculation of the macroscopic neutron cross-section for the compound, &Sgr;
C
. Since the macroscopic neutron cross-section, &Sgr;
C
, is unique to each compound, the compound is identifiable by the calculation of &Sgr;
C
. Alternatively, measurement of the uncollided gamma-ray beam intensity, I, allows calculation of the linear attenuation coefficient, &mgr;
C
, an identifying characteristic unique to each compound.
Unfortunately, defining the parameters of the attenuation equation, Equation (4), to enable calculation of &Sgr;
C
or &mgr;
C
is difficult, since the parameters are largely dependent on geometry. In addition, where the target material, for example, is a chemical agent contained within a chemical munition made of a thick steel shell, the use of neutron transmiss
Grover S. Blaine
Morrison John L.
Stephens Alan G.
Caress Virginia B.
Dvorscak Mark P.
Gagliardi Albert
Hannaher Constantine
Smith Bradley W.
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