Note: Descriptions are shown in the official language in which they were submitted.
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COMPOSITE GAMMA-NEUTRON DETECTION SYSTEM
TECHNICAL FIELD
The present specification generally relates to the field of detection of
radioactive
materials, specifically to systems and techniques for detecting neutrons and
gamma rays and
more specifically to a neutron and gamma-ray based detection system and method
that is cost-
effective, compact, and fabricated from readily available materials.
BACKGROUND
Physical shipment of materials, including the shipment of mail, merchandise,
raw
materials, and other goods, is an integral part of any economy. Typically, the
materials are
shipped in a type of shipping containment or cargo box. Such containments or
boxes include
semi-trailers, large trucks, and rail cars as well as inter-modal containers
that can be carried on
container ships or cargo planes. However, such shipping or cargo containers
can be used for
illegal transportation of contraband such as nuclear and radioactive
materials. Detection of these
threats require a rapid, safe and accurate inspection system for determining
the presence of
hidden nuclear materials, especially at state and national borders, along with
transit points such
as airports and shipping ports.
Currently, both passive and active detection techniques are employed for the
detection of
concealed nuclear materials. Passive detection techniques are based on the
principle that nuclear
and radiological threats emit gamma, and in some cases neutron, radiation that
can be detected.
Although passive detection systems can be easily deployed, they suffer from a
number of
drawbacks, including high rates of false positives and misdetections caused by
unavoidable
factors such as depression of the natural background by the vehicle being
scanned and its
contents, variation in natural background spectrum due to benign cargo such as
clay tiles,
fertilizers, etc., and the presence of radio therapeutic isotopes in the cargo
with gamma lines at or
near threat lines. Further, many gamma sources are self-shielded and/or can
readily be externally
shielded, which makes them difficult to detect, since the radiation is
absorbed in the shielding.
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Also, in general, gamma detectors make poor neutron detectors and good neutron
detectors tend
to be poor gamma detectors.
Other detection techniques employ uncharged particles, such as neutrons and
photons
(gamma rays) to irradiate suspicious containers. Uncharged particles have the
potential to
penetrate relatively large dense objects to identify particular elements of
interest; thus, some
detection devices utilize the absorption and/or scattering patterns of
neutrons or photons as they
interact with certain elements present in the object being inspected. Examples
of such devices
can be found in U.S. Patent Nos. 5,006,299 and 5,114,662, which utilize
thermal neutron
analysis (TNA) techniques for scanning luggage for explosives, and in U.S.
Pat. No. 5,076,993
which describes a contraband detection system based on pulsed fast neutron
analysis (PFNA).
Active detection techniques, such as Differential Dieaway Analysis (DDA) and
measurements of delayed gamma-ray and neutrons following either neutron- or
photon-induced
fission, can be used to detect the presence of fissile materials. The
radiation is measured with
neutron and gamma-ray detectors, preferentially insensitive to each other's
radiation. Detection
of delayed neutrons is an unequivocal method to detect fissile materials even
in the presence of
shielding mechanism(s) to hide the nuclear materials and notwithstanding the
low background
compared to delayed gamma rays. Because the number of delayed neutrons is two
orders of
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magnitude lower than the number of delayed gamma rays, efficient and large
area detectors are
required for best sensitivity in neutron detection.
Each of the detector systems described above is not without drawbacks. In
particular,
these devices generally utilize accelerators that produce high energy neutrons
with a broad
spectrum of energies. The absorption/scattering of neutrons traveling at
specific energies is
difficult to detect given the large number of neutrons that pass through the
object without
interaction. Thus, the "fingerprint" generated from the device is extremely
small, difficult to
analyze, and often leads to significant numbers of false positive or false
negative test results.
In addition, known prior art detection systems have limitations in their
design and method
that prohibit them from achieving low radiation doses, which poses a risk to
the personnel
involved in inspection as well as to the environment, or prevent the
generation of high image
quality, which are prerequisites for commercial acceptance.
While the use of both passive and active detection techniques is desirable,
what is needed
is a neutron and gamma-ray based detection system and method that is cost-
effective, compact,
and wherein the neutron detector is fabricated from readily available
materials.
The most commonly used neutron detector is a He-3 gas proportional chamber.
Here, He-
3 interacts with a neutron to produce a He-4 ion. This ion is accelerated in
the electric field of the
detector to the point that it becomes sufficiently energetic to cause
ionisation of other gas atoms.
If carefully controlled, an avalanche breakdown of the gas can be generated,
which results in a
measurable current pulse at the output of the detector. By pressurizing the
gas, the probability of
a passing thermal neutron interacting in the gas can be increased to a
reasonable level. However,
He-3 is a relative scarce material and it does not occur naturally. This makes
the availability and
future supply of such detectors somewhat uncertain. Further, a special permit
is required to
transport pressurized He-3 tubes, which can be cumbersome and potentially
problematic.
The most common globally deployed passive radioactive material detectors
employ a
neutron moderator 105 in an upper portion, having a plurality of He-3 detector
tubes 116
embedded therein covered by a lead shield 108 and a lower portion comprising a
plastic
scintillator and moderator 110 with a PMT (Photo Multiplier Tube) 115 embedded
therein, as
shown in FIG. 1A. This detector configuration, however, still employs the
scarce He-3. In
addition, another commonly deployed detector where the gamma-ray and neutron
detectors are
separate is shown in FIG. 1B. As shown in FIG. 1B, neutron moderator 105,
comprising a
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plurality of He-3 detector tubes 116 is positioned adjacent to plastic
scintillator 110, comprising
a PMT 115 and a lead shield 108. This detector configuration, however, still
employs the scarce
He-3 and takes up a larger footprint.
Several alternative detectors to replace He-3 detectors have been identified.
However,
many of these detectors are also sensitive to gamma rays, which is not
acceptable in applications
where neutrons must be discriminated from gamma rays.
Therefore, what is needed is a neutron and gamma-ray based detection system
and
method that is cost-effective, compact, and wherein the neutron detector is
fabricated from
readily available materials. In addition, what is needed is a cost-effective
and compact detection
system in which neutron and gamma-ray detectors are separate.
SUMMARY OF THE INVENTION
The present specification describes, in one embodiment, a thinly-coated 1013
flat-panel
ionization chamber neutron detector, which can be deployed as a direct drop-in
replacement for
current Radiation Portal Monitor (RPM) 3He detectors.
In one embodiment, the detector of the present specification comprises an
argon gas cell
sandwiched between boron-coated anode and cathode electrode plates.
In one embodiment, multiple cells are stacked together to increase the
intrinsic efficiency
of the detector. In one embodiment, the detector is multi-layered and includes
greater than 20
layers.
In one embodiment, multiple detector unit cells are "tiled" to achieve areas
of us to 1
square meter. In one embodiment, large detector units are folded for ease of
transportation.
In one embodiment, parallel plate geometry is employed, which allows for
integration of
neutron moderating sheets, such as polyethylene, on the back of the electrode
plates to
thermalize the neutrons and then detect them with high efficiency. Optionally,
the moderator can
be replaced with plastic scintillator sheets that can be viewed with a large
area photomultiplier
tube to detect gamma-rays in addition to neutrons, as is the case with
existing RPMs.
The present specification further describes a large-area detector that is
simple in its
construction and manufacture, easily scalable with respect to the unit cell
detector, easily
adaptable to a variety of applications, and low cost.
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In one embodiment, the present specification is directed towards a neutron
unit cell
detector, comprising: a first and a second layer, comprising a polyethylene,
for moderating a fast
neutron; a third and a fourth layer comprising B-10, for capturing a moderated
fast neutron,
wherein the third and fourth layers are positioned between the first and
second layers; and a gas
cell layer positioned between the third and fourth layers, which, when a
neutron is captured, emit
charged particles that ionize the gas in the gas cell layer creating free
electron and ion pairs.
In one embodiment, the neutron detector comprises a plurality of unit cell
detectors,
which are stacked, thereby increasing detector efficiency.
In another embodiment, the present specification is directed towards a gamma-
neutron
unit cell detector, comprising: a first and a second layer comprising gamma
sensitive plastic
scintillators for moderating a fast neutron and detecting gamma rays; a third
and a fourth layer
comprising B-10 for capturing a moderated fast neutron, wherein the third and
fourth layers are
positioned between the first and second layers; and a gas cell layer
positioned between the third
and fourth layers, which, when a neutron is captured, emit charged particles
that ionize the gas in
the gas cell layer creating free electron and ion pairs.
In one embodiment, the gamma-neutron detector comprises a plurality of unit
cell
detectors, which are stacked, thereby increasing detector efficiency.
In one embodiment, the plastic scintillator comprises at least one of an
organic solid
scintillator, an inorganic solid scintillator, or a liquid scintillator
positioned between glass layers.
In another embodiment, the present specification is directed towards a method
for
manufacturing a scalable, low-cost, large-area boron substrate for use in a
detector comprising:
employing a thin copper foil sheet as a metallic base; attaching the copper
foil to a rigid layer to
form a composite base for providing large areal structural strength; etching a
tile pattern and
individual electrical lines into the composite base by immersing the composite
base in a ferric-
chloride solution; mounting the composite base onto a drum for vacuum
deposition; and
depositing boron onto a surface of the copper foil to form the said boron
substrate, wherein a
mask is used to block the deposition of boron onto the electrical lines. In
one embodiment, the
thickness of the copper foil ranges from 50 to 100 um. In one embodiment, the
rigid layer
comprises Kapton.
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In one embodiment, the method of manufacturing the large area boron substrate
optionally comprises the step of fabricating a fast neutron detector by
laminating the boron
substrate onto a sheet of polyethylene.
The aforementioned and other embodiments of the present shall be described in
greater
depth in the drawings and detailed description provided below.
BRIEF DESCRIPTION OF THE DRAWINGS
These and other features and advantages of the present invention will be
appreciated, as
they become better understood by reference to the following detailed
description when
considered in connection with the accompanying drawings, wherein:
FIG. 1A illustrates a prior art radioactive material detector comprising a
neutron
moderator and a plastic scintillator, in which He-3 is employed;
FIG. 1B illustrates a prior art radioactive material detector comprising a
neutron
moderator and a plastic scintillator, in which He-3 is employed;
FIG. 1C is a schematic layout of the composite gamma-neutron detector
according to one
embodiment of the present invention;
FIG. 2 illustrates an exemplary neutron detector based on mixtures of silver
activated
zinc sulfide;
FIG. 3 illustrates an exemplary neutron detector based on mixtures of silver
activated
zinc sulfide that also uses a plastic scintillator for gamma ray detection;
FIG. 4 illustrates experimental results with the silver activated zinc sulfide
based neutron
detector;
FIG. 5 illustrates pulse signals as a function of time for gamma interactions
and neutron
interactions, respectively;
FIG. 6 illustrates discrimination between gamma ray and neutron measurement
signals;
FIG. 7A illustrates one embodiment of the detector of present invention with
multiple
layers of gamma and neutron detector materials to increase neutron
sensitivity;
FIG. 7B illustrates another embodiment of the detector of present invention
with angled
detector slabs to increase neutron detection efficiency;
FIG. 8 illustrates an exemplary readout circuit used with the detection system
of the
present invention;
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FIG. 9 illustrates an exemplary application of the gamma-neutron detector of
the present
invention in a drive-by vehicle;
FIG. 10 illustrates another exemplary application of gamma-neutron detectors
in a drive-
thru scanning configuration;
FIG. 11 illustrates yet another exemplary application of the gamma-neutron
detector
combined with a mobile X-ray scanner for generating composite gamma-neutron X-
ray images;
FIG. 12 illustrates another embodiment of the combined gamma-neutron detector
and
based X-ray imaging system in a portal or gantry configuration;
FIG. 13 illustrates the gamma-neutron detector in a portable configuration,
according to
one embodiment of the present invention;
FIG. 14 illustrates a parallel plate based Boron-10 (B-10) detector, according
to one
embodiment of the present invention;
FIGS. 15A illustrates a fast neutron detector geometry, in a first embodiment;
FIGS. 15B illustrates a fast neutron detector geometry, in a second
embodiment;
FIG. 16A illustrates an exemplary manner in which scalability can be achieved
for
manufacturing the B-10 detector of the present specification;
FIG. 16B illustrates an exemplary manner in which scalability can be achieved
for
manufacturing the B-10 detector of the present specification;
FIG. 16C illustrates an exemplary manner in which scalability can be achieved
for
manufacturing the B-10 detector of the present specification;
FIG. 17 is a graph illustrating detection efficiency of the B-10 detector of
the present
specification;
FIG. 18 is a graph showing the fast neutron detection efficiency of the 10B
neutron
detector of the present specification compared with a 3He-based Differential
Die-Away Analysis
(DDAA) detector;
FIG. 19A illustrates a first manufacturing step for fabricating the large area
boron
substrate of the present specification;
FIG. 19B illustrates a second manufacturing step for fabricating the large
area boron
substrate of the present specification;
FIG. 19C illustrates a third manufacturing step for fabricating the large area
boron
substrate of the present specification;
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FIG. 19D illustrates a fourth manufacturing step for fabricating the large
area boron
substrate of the present specification;
FIG. 19E illustrates a fifth manufacturing step for fabricating the large area
boron
substrate of the present specification; and
FIG. 19F illustrates a sixth manufacturing step for fabricating the large area
boron
substrate of the present specification.
DETAILED DESCRIPTION OF EXAMPLE EMBODIMENTS
The present specification discloses systems and methods for detecting
radiological
threats using a composite gamma-neutron detector which can be configured to
have a high
sensitivity for both gamma and neutron detection, with a sufficient separation
of the gamma and
neutron signatures. The system of the present specification allows for maximum
threat detection
with minimum false alarms, and thus increased throughput.
Further, the present specification is directed towards a composite gamma-
neutron
detection system and method that is cost-effective, compact, and wherein the
neutron detector is
fabricated from readily available materials.
The present specification is directed towards multiple embodiments. The
following
specification is provided in order to enable a person having ordinary skill in
the art to practice
the invention of the present specification. Language used in this
specification should not be
interpreted as a general disavowal of any one specific embodiment or used to
limit the claims
beyond the meaning of the terms used therein. The general principles defined
herein may be
applied to other embodiments and applications without departing from the scope
of the present
specification. Also, the terminology and phraseology used is for the purpose
of describing
exemplary embodiments and should not be considered limiting. Numerous
alternatives,
modifications and equivalents consistent with the principles and features
disclosed are possible.
For purpose of clarity, details relating to technical material that is known
in the technical fields
related to the invention have not been described in detail so as not to
unnecessarily obscure the
present invention.
Several nuclei have a high cross-section for detection of thermal neutrons.
These nuclei
include He, Gd, Cd and two particularly high cross-section nuclei: Li-6 and B-
10. In each case,
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after the interaction of a high cross-section nucleus with a thermal neutron,
the result is an
energetic ion and a secondary energetic charged particle.
For example, the interaction of a neutron with a B-10 nucleus can be
characterized by the
following equation:
Equation 1: n + B-10 Li-7 + He-4 (945 barns, Q = 4.79 MeV)
Here, the cross section and the Q value, which is the energy released by the
reaction, are
shown in parenthesis.
Similarly, the interaction of a neutron with a Li-6 nucleus is characterized
by the
following equation:
Equation 2: n + Li-6 H-3 + He-4 (3840 barn, Q = 2.79 MeV)
It is known that charged particles and heavy ions have a short range in
condensed matter,
generally travelling only a few microns from the point of interaction.
Therefore, there is a high
rate of energy deposition around the point of interaction. In the present
invention, molecules
containing nuclei with a high neutron cross section are mixed with molecules
that provide a
scintillation response when excited by the deposition of energy. Thus, neutron
interaction with
Li-6 or B-10, for example, results in the emission of a flash of light when
intermixed with a
scintillation material. If this light is transported via a medium to a
photodetector, it is then
possible to convert the optical signal to an electronic signal, where that
electronic signal is
representative of the amount of energy deposited during the neutron
interaction.
Further, materials such as Cd, Gd and other materials having a high thermal
capture cross
section with no emission of heavy particles produce low energy internal
conversion electrons,
Auger electrons, X-rays, and gamma rays ranging in energy from a few keV to
several MeV
emitted at substantially the same time. Therefore, a layer of these materials,
either when mixed
in a scintillator base or when manufactured in a scintillator, such as
Gadolinium Oxysulfide
(GOS) or Cadmium Tungstate (CWO) will produce light (probably less than
heavier particles).
GOS typically comes with two activators, resulting in slow (on the order of 1
ms) and fast (on
the order of 5 iLts) decays. CWO has a relatively fast decay constant.
Depending on the overall
energy, a significant portion of the energy will be deposited in the layer,
while some of the
electrons will deposit the energy in the surrounding scintillator. In
addition, the copious X-rays
and gamma rays produced following thermal capture will interact in the
surrounding scintillator.
Thus, neutron interactions will result in events with both slow and fast decay
constants. In many
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cases, neutron signals will consist of a signal with both slow and fast
components (referred to as
"coincidence") due to electron interlacing in the layer and gamma rays
interacting in the
surrounding scintillator.
The scintillation response of the material that surrounds the Li-6 or B-10
nuclei can be
tuned such that this light can be transported through a second scintillator,
such as a plastic
scintillator in one embodiment, with a characteristic which is selected to
respond to gamma
radiation only. In another embodiment, the material that surrounds the Li-6 or
B-10 is not a
scintillator, but a transparent non-scintillating plastic resulting in a
detector that is only sensitive
to neutrons.
Thus, the plastic scintillator is both neutron and gamma sensitive. When a
neutron is
thermalized and subsequently captured by the H in the detector, a 2.22 MeV
gamma ray is also
emitted and often detected. In this manner, the present invention achieves a
composite gamma-
neutron detector capable of detecting neutrons as well as gamma radiation with
high sensitivity.
Further, the composite detector of the present invention also provides an
excellent separation of
the gamma and neutron signatures. It should be noted herein that in addition
to charged particles,
B-10 produces gamma rays. Therefore, in using materials that produce gamma
rays following
neutron capture, the result may be a detection that looks like gamma rays.
Most applications,
however, want to detect neutrons; thus, the detector of the present invention
is advantageous in
that it also detects the neutrons.
FIG. 1C illustrates a schematic layout of the composite gamma-neutron detector
100
according to one embodiment of the present invention. Referring to FIG. 1C,
the detector design
employs two gamma-sensitive scintillation panels (gamma-detectors) 101 and 102
that surround
a single neutron detector 103. The neutron detector 103 further comprises a
single slab of
neutron sensitive composite scintillator, in which nuclei of a neutron
sensitive material such as
Li-6 or B-10 are mixed with a scintillation material such as ZnS. In one
embodiment, a density
of 20 ¨ 30% by volume can be achieved for the neutron sensitive material (such
as Li-6) while
maintaining an efficient scintillation response from ZnS.
In one embodiment, gamma detector panels can be fabricated from solid
scintillation
materials (without a substrate) such as, but not limited to organic
scintillators, including solid
plastic scintillators (e.g. NE102) and anthracene; inorganic scintillators
including NaI(T1),
CsI(T1), CsI(Na), and BaF2.
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In another embodiment, it is possible to position liquid scintillators between
glass sheets
to act as the gamma detector. These tend to use organic solvents formed with
the anthracene
molecule as their base with organometallic compounds to enhance scintillation
efficiency and
therefore are generally less easy to use than solid scintillators.
In one embodiment, the neutron detector may be comprised of binder molecules
such as,
but not limited to styrenes dissolved in suitable solvents as the base
substrate. As the solvent
evaporates, a plastic film forms which, once dry, is quite stable and self-
supporting. The
scintillation material (for example ZnS) and the neutron specific element
(i.e. Gd, Li, B, etc.) are
intermixed with the solvent and binder prior to solvent evaporation. As the
solvent evaporates, an
intimate mixture of all three components is formed.
In an alternative embodiment, a Gd, Li or B loaded liquid scintillator
(generally based on
the anthracene molecule with suitable organometallic compounds to increase
scintillation
efficiency) can be sealed in the gap between the gamma scintillation panels.
Advantageously, a
thin glass barrier will be placed between the neutron scintillator and the
gamma-detector to
prevent chemical interaction between the two scintillator materials.
In one embodiment, a typical panel size ranges from 0.1 m x 0.1 m for handheld
applications up to 2mx 1 m for large fixed site installations. Above this
maximum size, light
collection starts to become an issue as does physical handling and packaging.
Below the
minimum size, detection efficiency will start to drop below useful levels,
resulting in
increasingly long measurement times.
In one embodiment, the gamma detector is thicker than the neutron detector.
The gamma
detector thickness will advantageously be no less than 0.01 m (for hand held
applications) up to
0.2 m for large fixed site systems. The front gamma detector may be optimized
to a different
thickness compared to the back gamma detector in order to maximize overall
gamma and
neutron detection efficiency. For example, a front gamma detector thickness of
0.05 m and a rear
gamma detector thickness of 0.1 m would be applicable to a large fixed site
system. The neutron
detector will generally be thin to minimize gamma interaction probability and
to maximize the
chance of light escape from the scintillator. A typical neutron detector based
on a solid screen
scintillator would be in the range of 0.5-1 mm thick while a liquid neutron
scintillator may be in
the range of 0.01 to 0.05 m thick.
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Optical signals from both the gamma detectors 101, 102 and the neutron
detector 103 are
readout by one or more photodetectors, which in one embodiment are
photomultiplier tubes
(PMTs) 104. The optical signals are thus converted to electronic signals which
are then
processed by a pulse processor 105 which assigns interactions separately due
to gamma and
neutron interactions 106 and 107, respectively.
In one embodiment, the gamma-sensitive 101 and 102 panels are advantageously
fabricated from a plastic scintillator with a fast decay time, such as less
than 0.1 las. Further, the
Li-6 or B-10 nuclei of the neutron detector 103 are advantageously mixed with
a scintillation
material having a slower decay time, such as ZnS. In one embodiment, the decay
time for the
scintillation material is greater than 1 las. The difference in decay times
for scintillators in
gamma detectors and in neutron detector contributes to provide a significant
separation between
the gamma and neutron signatures 106 and 107. In general, it is desirable to
select a scintillation
material with low atomic number so as to minimise the probability of direct
excitation by a
passing gamma ray which causes enhanced gamma-neutron rejection.
In another embodiment, the Li-6 or B-10 is mixed with a material with very
fast response
(-10 ns) and surrounded by a material with slow response (-1 s).
It may be noted that if material used around Li-6 is a very fast scintillator,
the detector
can measure neutrons at a very high counting rate, in particular when no
scintillator is used to
surround it.
One of ordinary skill in the art would appreciate that scintillation materials
such as ZnS
can absorb their own light and therefore there is a limit to the thickness of
a scintillation based
detector in ZnS. It may be noted that this thickness is typically only a few
millimetres. Further,
since light is emitted isotropically during each scintillation event, it is
efficient to form the
scintillator into a wide area screen where light emission can be captured from
both sides of the
screen simultaneously. Therefore, in one embodiment the scintillator based
neutron detector 103
is designed as a screen with a wide area, such that light may be collected
with a high efficiency
from both sides of the screen.
It may be noted that the detection efficiency of a lmm thick Li-6/ZnS screen
is of the
same order as that of a pressurised He-3 gas proportional tube several cm in
diameter. That is,
the Li-6/ZnS based neutron detector of the present invention offers equivalent
or greater
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detection efficiency as compared to the pressurised He-3 gas tube detector, at
a much reduced
size.
Therefore, in one embodiment, a neutron detector is based on mixtures of
silver activated
zinc sulfide, ZnS(Ag), with the mixtures containing materials with high
thermal neutron-capture
cross section with emission of heavy particles, such as 6Li or 10B. That is,
the mixtures consist of
thermal neutron absorbers that produce heavy-particle emission following
thermal capture. FIG.
2 illustrates one such exemplary neutron detector 200. Referring to FIG. 2,
the detector 200
consists of one or more thin screens 201, comprising the ZnS(Ag) based
mixtures, as described
above. The screens 201, in one embodiment, have a thickness of about 0.5 mm
and are
embedded in a transparent hydrogenous light guide 202. Light guide 202 also
serves as a
neutron moderator. The light produced by neutron interaction in the ZnS(Ag)
phosphorus screen
is collected by the light guide 202 into a photodetector, such as a
photomultiplier tube (PMT)
203, which produces a signal from which the neutrons are counted, using the
counter 204.
The technology described above can also be implemented with simultaneous gamma-
ray
detection with the same basic electronics. Thus, the detector 200 further
comprises a plastic
scintillator 205, which serves as a gamma-ray detector and moderator. The
plastic scintillator
may be made up polyvinyl toluene or PVT, or any other suitable plastic
scintillator material
known in the art. Light produced by gamma-ray interactions in the scintillator
205 is detected by
another PMT 206, which produces a signal from which the gamma-ray events are
counted, using
the counter 207. In one embodiment, counter 207 is a Multi-Channel Analyzer
(MCA) that is
used to measure the spectra of the gamma rays.
A reflector foil 208 is placed between the plastic scintillator 205 and the
screen(s) 201 to
prevent cross-contamination between optical signals from the neutron and gamma
detection
materials. Thus, the reflector is used to prevent light produced from the
gamma rays to be
collected with the same PMT as light produced by the neutrons. This prevents
appearance of
false neutron counts from gamma rays. Due to the reflector 208, some of the
light produced by
neutron interactions in the screen will be reflected back into the light
guide.
The design of FIG. 2 provides a compact gamma-ray/neutron detector with the
advantages
of standard electronics and significantly high gamma-ray rejection. A small
fraction of gamma
rays will interact with the Li-6 sheet and will produce a low-intensity
signal. This signal can be
removed by thresholding, at the expense of some neutron detection. In one
embodiment, a pulse
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shape discriminator can be employed within neutron channel 204 to enhance
gamma-ray
rejection.
Another exemplary detector 300 for simultaneous neutron and gamma-ray
detection is
shown in FIG. 3. In this case, the light guide material is replaced by a
plastic scintillator 301,
which serves as the gamma-ray detector, moderator and light guide. The
detector 300 also
includes screens 302, which are preferably thin and fabricated from ZnS(Ag)
based mixtures for
neutron detection. The neutrons and gamma-ray events are separated employing a
Pulse-Shape
Discrimination (PSD) circuit 303 between the pulses 304 generated from the
ZnS(Ag) and
plastic scintillator (PVT). Additionally, gamma-ray rejection is obtained as
the light produced by
electron interaction in the screen have similar decay time as the PVT's and
will be eliminated
with PSD. The light produced is transported via the transparent and neutron
moderating medium
301 to a Photomultiplier Tube (PMT) 305 where the light is converted to a
measurable signal to
measure gamma as well as neutron events. The advantage of this hybrid
neutron/gamma-ray
detector approach is that the same PMT can be employed to measure the neutron
as well as
gamma events.
FIG. 4 illustrates the performance of an exemplary detector with a
6LiF:ZnS(Ag) screen
embedded in a light-guide with two 6LiF concentrations and thickness. The
results in FIG. 4
show the signal for the 1:2 weight ratio and screen thickness of 0.45 mm.
Similar results were
obtained with simulations employing 1, 2 and 3 6LiF:ZnS(Ag) screens embedded
in polyethylene,
and detection efficiencies ranging from around 12% to 22% were obtained. One
of ordinary skill
in the art would appreciate that this efficiency is comparable to the highest
efficiency achievable
with closely-packed three rows 3He detectors, which is around 25%.
The signal distribution in FIG. 4 shows that not all the particle energy
absorption is
converted to light and that some of the light may be absorbed by the screen.
This demonstrates
the need for a comprehensive optimization where the right concentration of 6Li
is obtained to
produce high neutron absorption, while still having sufficient interactions in
the scintillator to
produce a sizeable light output. The screen thickness, the number of screens
and moderator
thickness are also important optimization parameters.
For applications focused on neutron detection, a major advantage of ZnS(Ag)
phosphorus
is the large light output for heavy particles compared with electrons produced
by gamma-ray
interactions. Also, due to the small thickness of the screen, the gamma-ray
detection efficiency is
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low. Further, since the time-decay of the PVT light is ¨3 ns, similar to that
of the light produced
by electrons in the ZnS(Ag) screen, PSD will also reject gamma rays
interacting in the PVT.
As known to persons of ordinary skill in the art, neutrons generated by
radioactive
materials of interest have a range of energies, and that the efficiency of
neutron interaction in the
detector will generally increase markedly as the energy of the interacting
neutron decreases. For
this reason, most He-3 detectors are located within a hydrogen rich moderating
material, such as
polythene, whose function is to promote neutron scattering of high energy
neutrons such that
they lose substantial amounts of energy in order to increase the probability
of detection in the
He-3 gas proportional counter. In the present invention, the gamma detector is
advantageously
designed to provide a dual function of gamma detection and neutron moderation
to further
improve the detection efficiency for neutrons. A plastic scintillator material
is quite an efficient
moderator as this feature is incorporated in the overall detector design.
FIG. 5 illustrates pulse signals, as a function of time corresponding to gamma
interactions
and neutron interactions in the composite detector of the present invention.
Referring to FIG. 5,
the scintillation characteristics curve 502 of the neutron sensitive
scintillator is very different
from the characteristics 501 of the surrounding gamma sensitive detector.
These two
characteristic signals 501 and 502, can be further tuned to exhibit a
significant difference. This
can be done by using appropriate pulse shape discrimination methods. Thus, in
one embodiment
of the present invention, both the total energy deposited in the detector and
the types of
interaction are determined. While the total energy can be determined by
analysing the peak
magnitude of the pulse signal, the type of interaction is determined by
analysing the rate of decay
of the scintillation pulse.
FIG. 6 illustrates the discrimination between gamma rays and neutrons for
252Cf and
60Co source, when analog Pulse-Shape Discrimination is applied to separate
gamma rays from
neutron events. While curve 601 reflects measurement of gamma rays emitted
from 60Co source,
curve 602 reflects measurement of neutrons emitted from 252Cf source. It would
be apparent to
those of ordinary skill in the art that the two curves are separate and
distinctly identifiable.
In one embodiment, the gamma-ray rejection is improved by subtracting a
calibrated
fraction of gamma-ray counts from the measured neutron counts.
In one embodiment, the digital pulse processing is advantageously performed
directly at
the output of the detector. Since data rates can be quite high, processing at
the detector helps
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filter the data down to a low bandwidth for transmission on to other
processing systems. This
data can be used to monitor the amount of radioactivity that is detected and
to raise suitable
alarms and/or display data by a number of means.
In yet another aspect of this invention, it is noted that the neutron reaction
may also
create an associated gamma-ray emission. For example in the reaction of a
neutron with Gd-157,
the excited Gd-158 nucleus decays with the emission of a gamma-ray. This gamma-
ray is
produced within a finite time of the neutron interaction and, therefore, it is
possible to include the
gamma-ray response that is measured in the surrounding gamma-detector in
combination with
the neutron scintillator response to produce a combined signal using the
principle of pulse shape
discrimination and time domain correlation.
While FIG. 1C illustrates an exemplary configuration for a composite detector,
alternative detector configurations may be established in order to further
enhance neutron and
gamma detection efficiency. Two exemplary alternative configurations are
illustrated in FIGS.
7A and 7B. As shown in FIG. 7A, a first configuration combines multiple layers
of gamma
sensitive scintillator slabs 701 and neutron sensitive scintillator slabs 702
placed alternately with
each other, in a direction substantially perpendicular to the direction of
arrival of incident
radiation 705. In this configuration, the efficiency of the gamma-neutron
detector scales in
proportion to the number of slabs of detector material; although this is a
diminishing effect due
to preferential absorption of radiation in the first layers of the detector
compared to the later
layers of the detector. Neutron sensitivity is significantly enhanced when the
detector slabs are
arranged in this configuration.
In another configuration shown in FIG. 7B, multiple layers of gamma detector
materials
710 and neutron detector materials 720 are placed alternately with each other
and are oriented at
an angle to the direction of the incoming radiation 715. That is, layers 710
and 720 are not
parallel to the direction of the incoming radiation 715. Such a detector
configuration with angled
detector slabs significantly increases neutron detection efficiency. This is
because a neutron or
photon in this case has a longer path length through each detector slab, which
contributes to
detection efficiency, as compared to the arrangement of slabs shown in FIG.
7A. However, this
arrangement of detectors is also more expensive to fabricate and requires more
extensive readout
circuits.
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One of ordinary skill in the art would appreciate that other configurations of
scintillator
materials and photo-detectors are possible, and any configuration may be
selected depending
upon its suitability to the application. Therefore, the composite gamma-
neutron detector of the
present invention described with reference to FIGS. 1, 7A and 7B is not
limited to plastic
scintillator gamma detector with Li-6/ZnS neutron detector. In one embodiment
for example, the
composite detector may be configured using NaI(T1) as the gamma detector,
along with a lithium,
boron or gadolinium based liquid scintillator with a very fast decay time.
Here, the NaI(T1)
gamma detector will provide significant pulse height information about the
gamma ray
interaction while the neutron detector will continue to provide information
about the incident
neutron flux.
It shall be appreciated that the use of light reflective coatings with
suitable optical
coupling materials will improve overall light collection efficiency and hence
the uniformity of
response of the detector. It should also be understood that optical light
guides and shaping of the
scintillator materials may also be used to improve light collection efficiency
of the detection
system. Further, it should also be understood that the addition of radiation
shielding materials
such as lead, polythene and cadmium foil around the scintillation materials
may be used to
reduce the response of the detection system to naturally occurring background
radiation.
In a further embodiment of the invention, a neutron scintillator can be used
which
provides different pulse shapes due to fast and thermal neutron interactions,
where each pulse
shape is different to that selected for the gamma detector.
FIG. 8 illustrates an exemplary detector readout circuit architecture.
Referring to FIG. 8,
the circuit 800 comprises a photomultiplier tube (PMT) 801, which is operated
with its cathode
802 held at negative high voltage with a grounded anode 803. The anode 803 is
AC coupled
using a transformer 804 to a high speed sampling analogue-to-digital converter
(ADC) 805. The
ADC 805 forms a time domain sample of the incoming signal from the PMT 801. In
one
embodiment, the ADC operates at a clock speed of 100 MHz or more to provide at
most 10 ns
sampling periods for accurate measurement of peak height and of the rise and
fall decay times. In
one embodiment, a filtering circuit is advantageously included between the PMT
801 and the
input to the ADC 805 to act as a Nyquist filter to prevent unwanted aliasing
in the sampled data.
In one embodiment, an LCR multi-pole filter is implemented using the AC
coupling transformer
804 as the inductive component.
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In an alternate configuration, the PMT 801 may be d.c. coupled to the input of
the ADC
805 using a high bandwidth analogue amplifier. A variety of other circuit
configurations will be
apparent to one skilled in the art.
The digital data produced by the ADC is advantageously passed directly to a
digital
processing circuit, such as a field programmable gate array (FPGA) 806. The
FPGA provides
high speed digital pulse shape processing and is configured to (1) record the
time of arrival of a
pulse, (2) determine the magnitude of the pulse and (3) determine the fall
time of the pulse in
order to discriminate between neutron and gamma interactions. This pulse-by-
pulse data is
histogrammed to a random access memory 807 and can subsequently be analysed by
a software
program running on a computer 808 to resolve detected count rates relative to
a dynamically
adjusted baseline. The result may be indicated to an operator through a visual
display screen 809,
a visual indicator, an audible sounder or any other suitable device in order
to signal when a
radioactive substance has been detected.
A variety of other methods to provide pulse-shape discrimination will be
apparent to
those of ordinary skill in the art.
FIG. 9 shows an application of a composite gamma-neutron detector in a mobile
system,
in a drive-by scanning configuration. Referring to FIG. 9, the gamma-neutron
detector 901 is
positioned in a vehicle 902. This configuration allows rapid re-location of
the detector 901 from
one site to another, and is also useful for covert scanning of vehicles as
they pass along a road. In
this embodiment, the vehicle 902 is driven to a location, such as a roadside,
and the detection
system 901 is activated. In one embodiment, one or more sensors (not shown)
that are located on
the vehicle 902 determine the presence of a passing object to be scanned, such
as a passing
vehicle, and the detection system 901 is turned on automatically. Once the
vehicle has been
scanned, the the gamma-neutron detector 901 is turned off automatically. Once
scanning at a
given location is completed, the vehicle 902 can simply be driven to a new
location and scanning
can recommence as required. This feature provides the capability for random
location scanning
in a reasonably covert manner.
When not actively scanning a vehicle at the scanning site, the gamma-neutron
detector in
its off state is used to record the natural background radiation and this
natural background rate is
used to set an appropriate alarm threshold for when additional activity is
detected in a passing
vehicle during the on state of the scanner.
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In another application, the composite gamma-neutron detector 901 is installed
in a
vehicle 902 that can be driven past stationary targets at a known velocity. As
the vehicle 902
drives by, radiation emission data is collected in order to determine the
presence of radioactive
materials in the stationary object.
FIG. 10 shows another application of one or more composite gamma-neutron
detectors in
a drive-through scanning configuration. Referring to FIG. 10, a plurality of
composite gamma
neutron detectors 1001, 1002 and 1003 are arranged as a fixed drive through
system, in a portal
configuration having a right, left, and top side, through which cargo vehicles
such as 1004 can be
driven. The signals from the detectors 1001, 1002 and 1003 are processed and
the result can be
seen on a display 1005. The display is also coupled to audible 1006 and visual
1007 alarms
which are automatically generated, when radioactive material is suspected on
the vehicle 1004
being scanned. The result on display 1005 and the alarms 1006 and 1007 may be
used to
determine if the vehicle 1004 needs further search, and the vehicle may be
diverted to a holding
area, for example, for a manual search. The drive through scanning system of
FIG. 10 also
employs a traffic control system 1008, which operates a barrier 1009 for
stopping the vehicles
for inspection. The barrier is lifted automatically once the scan results
appear on the display 1005.
In an alternative configuration, one or more gamma-neutron detectors of the
present
invention are installed with a baggage handling system employed at airports.
In this manner, the
system of present invention may also be used for detection of radioactive
materials in baggage
passing through an airport terminal. In another alternative configuration, one
or more gamma
detectors of the present invention can be installed in air cargo facilities
and at the entrance of
scrap metal facilities.
In a further embodiment of the present invention, a gamma-neutron detector is
combined
with a mobile X-ray scanner for generating composite gamma-neutron X-ray
images. This is
illustrated in FIG. 11. Referring to FIG. 11, a gamma-neutron detector 1101 is
installed on a
mobile X-ray scanner 1100. The mobile X-ray scanner 1100 further comprises an
X-ray scanning
system 1102 mounted on a vehicle 1103. In this case, the radioactive signal
from the gamma-
neutron detector 1101 is acquired simultaneously with a transmission X-ray
image from the X-
ray scanning system 1102. This allows signals from the gamma-neutron detector
1101 to be
correlated with the X-ray image data to help the operator locate the presence
of a radioactive
material within the load under inspection. Any of the mobile systems disclosed
in U.S. Patent
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Application Nos. 10/201,503; 10/600,629; 10/915,687; 10/939,986; 11/198,919;
11/622,560;
11/744,411; 12/051,910; 12/263,160; 12/339,481; 12/339,591; 12/349,534;
12/395,760; and
12/404,913 can be used.
In yet another embodiment, the gamma-neutron detector of the present invention
is
combined with an X-ray imaging system, in a portal or gantry configuration.
Referring to FIG.
12, a plurality of gamma-neutron detectors 1201 are co-located with a
transmission X-ray system
1202 arranged in a portal configuration. Objects or vehicles under inspection
can be passed
through this portal or gantry. This mode of operation again allows the
radioactive signals to be
correlated with an X-ray image of the object under inspection thereby
increasing detection
efficiency. For example, the occurrence of a high-attenuation area observed in
the X-ray image
and a small increase in gamma-ray and/or neutron signal below the threshold
could indicate the
presence of a shielded radioactive source.
FIG. 13 shows another embodiment of a gamma-neutron detector in a portable,
hand-held
configuration. Referring to FIG. 13, a gamma-neutron detection instrument 1300
is shown. The
instrument comprises a main unit 1301 and a handle 1302. In one embodiment,
the scintillation
panels of the composite gamma-neutron detector (not shown) are located in the
main unit 1301,
while the electronics and battery are advantageously located in the handle
1302 of the
instrument. An embedded indicator 1303 provides feedback to the operator on
the amount of
radiation present in the vicinity of the instrument 1300. This configuration
is very useful for
random searching, especially small objects and in searching nooks and corners
within a vehicle.
The novel approach of the present invention combines a neutron scintillation
detector
with a gamma detector to form a hybrid gamma-neutron detector. This approach
provides the
advantage of detecting dual signatures, thereby increasing detection
efficiency. Further, by using
the method of pulse shape discrimination, the system of present invention also
provides an
excellent separation of the neutron signal from the gamma signal. The system
of present
invention may be used in various configurations, depending upon the
application, including but
not limited to, fixed, drive-through portal, gantry, portable and hand-held.
The combined
detector can be used for sea cargo inspection, and vehicle inspection in land
crossings and scrap-
metal facilities, in baggage and air cargo scanning, and other applications.
The combined
neutron-gamma detector of the present invention and/or the neutron detector
portion and/or the
gamma detector portion is further designed to meet ANSI standards for
radiation detection.
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Compared to He-3 based systems, which face a problem due to short supply of He-
3, the
present invention does not limit the use of the system with a particular
nucleus. As mentioned
previously, any suitable material with high neutron thermal capture cross-
section with emission
of particles, such as Lithium (Li-6), Boron (B-10), Cadmium (Cd), Gadolinium
(Gd), and
Helium (3-He) may be used for radioactive material detection with the system
of present
invention. This feature helps to keep cost and supply under control. Further,
the combined
gamma-neutron detector of the present invention is more compact and lighter as
compared to He-
3 based systems, as the detector of present invention only uses, in one
embodiment, one set of
electronics whereas He-3 based systems multiple sets of electronics are
employed. It should be
noted herein that in other embodiments, the present invention may be used with
a plurality of
electronic sets.
Most Radiation Portal Monitors (RPM) deployed around the world employ plastic
scintillators to detect gamma rays and moderated 3He detectors to measure
neutrons. It is
important to note that in typical RPMs, only one or two 3He tubes are used per
module with a
suboptimal moderating configuration to reduce cost. This results in a neutron
detection
efficiency of few percent.
The proposed neutron detector can replace 3He detectors in Radiation Portal
Monitors
(RPMs) as its neutron detection and gamma-ray rejection capabilities are
similar to that of 3He.
Further, the detectors of present invention do not contain hazardous
materials, are commercially
available, do not require special transport permits, are very rugged -
mechanically as well as
environmentally, and are easy to manufacture at a reasonable cost. The
detectors are also suitable
for handheld and backpack detectors, where efficiencies exceed that of 3He.
Finally, the present
approach is suitable for integrated neutron and gamma-ray detectors, as it
employs a single PMT
with relatively simple and compact electronics.
As mentioned above, 10B, like 3He, has a high thermal neutron capture cross-
section and
emits two detectable high energy charged particles, but unlike 3He, is
naturally abundant. On the
other hand, the supply of He is rapidly dwindling and as a result, 3He gas has
become extremely
expensive and difficult to obtain. Although boron coated detectors have been
available in the
past and for example, utilized as reactor neutron flux monitors, they were
inefficient, limiting
their usage.
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The present specification, therefore, describes in one embodiment, a thinly-
coated 1013
flat-panel ionization chamber neutron detector, which can be deployed as a
direct drop-in
replacement for current Radiation Portal Monitor (RPM) 3He detectors. In
various embodiments,
the 1013 coating has a thickness range of 0.1 to 2.0 micron. In one
embodiment, the 1013 coating is
1.0 micron thick. A thicker coating means the energy losses are greater from
the charge particle
traversing through the coating into the gas chamber. This results in a
detriment to the signal.
However, a thicker coating can increase detection efficiency lowering the
number of layers
required to reach a certain efficiency.
In one embodiment, the detector of the present specification comprises an
argon gas cell
sandwiched between boron-coated anode and cathode electrode plates.
In one embodiment, parallel plate geometry is employed, which allows for
integration of
neutron moderating sheets, such as polyethylene, on the back of the electrode
plates to
thermalize the neutrons and then detect them with high efficiency. Optionally,
the moderator
can be replaced with plastic scintillator sheets that can be viewed with a
large area
photomultiplier tube to detect gamma-rays in addition to neutrons, as is the
case with existing
RPMs.
The present specification further describes a large-area detector that is
simple in its
construction and manufacture, easily scalable with respect to the unit cell
detector, easily
adaptable to a variety of applications, and low cost.
In one embodiment, as mentioned above, the approach in developing a large-area
1013-
based 3He replacement detector focuses on utilizing a parallel plate
ionization chamber concept,
which is illustrated in FIG. 14. Referring to FIG. 14, the basic geometry of
one unit cell detector
consists of a first boron layer 1401 and a second boron layer 1402, which are
high-voltage biased,
sandwiching a gas cell 1403. The two layers of boron capture thermal neutrons.
When a neutron
is captured, two charged particles, 'Li and alpha are emitted and ionize the
gas, thereby creating
free ions and electrons. The voltage applied 1405 sweeps the charges creating
a signal.
The following equation shows the neutron capture reaction of 10B:-
94% ,7 Li + ,4a 1.47 MeV
5
6%,Li + cf. 1.78 MeV
As seen in the reaction, a 'Li and alpha particle are emitted in opposite
directions. One
particle ionizes the gas in the gas cell 1403 creating free electron and ion
pairs. The high-voltage
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bias sweeps the ions creating a signal pulse proportional to the number of
electron/ion pairs
created. Because the chamber does not rely on multiplication of electrons,
which proportional
counters utilize to increase signal, lower voltages can be applied. In 94% of
the reactions, an
alpha particle receives 1.47 MeV, while it receives 1.78 MeV in about 6% of
the reactions.
1 B has the second highest thermal neutron capture cross-section for a low-Z
material.
The cross-section is 3837 barns, while 3He has a cross-section of 5333 barns.
Because 10B has
such a high thermal neutron capture cross-section, 10B-based detectors can
achieve 3He
equivalent efficiencies. The large-area parallel plate ionization chamber can
not only be designed
to be a pure thermal neutron detector, it can be designed and optimized to
detect fast neutrons as
well.
Fast neutron detection is in many cases more relevant to the inspection arena
than pure
thermal neutron detection efficiencies, as all neutrons, when produced, are
"fast" (with energies
above 0.1MeV). Indeed fast fission neutrons are one of the most important
signatures of a fission
event. In one embodiment, multiple unit cell detectors of FIG. 14 are stacked
together to
increase the intrinsic efficiency of the detector. In one embodiment, the
detector is multi-layered
and includes greater than 20 layers.
FIGS. 15A and 15B illustrate a first and second embodiment of fast neutron
detector
geometries, respectively, that can replace large-area Radiation Portal
Monitors. Referring to FIG.
15a, a unit cell detector comprises a first polyethylene layer 1501, a first
boron-coated metallic
layer 1503, a gas cell layer 1505, a second boron-coated layer 1507, and a
second polyethylene
layer 1509. In one embodiment, gas cell layer 1505 is comprised of argon. In
operation, a fast
neutron gets moderated by the polyethylene layer, thermalizes, and gets
captured by the boron.
The polyethylene layers thus serve to moderate fast neutrons.
As shown in FIG. 15B, a photon detector is integrated with the neutron
detector. Here,
instead of polyethylene sheets, a plastic scintillator is integrated into the
detector in the form of
two layers 1510 and 1520. The plastic scintillator serves a dual purpose; it
can moderate fast
neutrons and can detect gamma rays as well, since it is a gamma ray
scintillation detector. While
the design of FIG. 15A can replace the 3He module in the current RPMs; the
design of FIG. 15B
in a single module, can replace the entire gamma ray and neutron detection
modules of current
RPMs.
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As mentioned above, scalability of the detector to cover large areas is
achievable through
the parallel plate ionization chamber concept. FIGS. 16A through 16C
illustrate three exemplary
steps via which scalability can be achieved. FIG. 16A shows two stacked unit
cell detectors
1601 described in detail with respect to FIGS. 15A and 15B. In one embodiment,
the stacked
detector has dimensions in the range of 10cm x 10 cm x 1 cm. The stacked
detector, which
comprises two unit cell detectors 1601 comprise a total of four boron layers
1605, two argon gas
cells 1607 and three kapton layers 1609. The kapton layers 1609 are used to
provide rigidity to
the thin boron coatings. One of ordinary skill in the art would appreciate
that other suitable
materials may also be used for the purpose.
By adding more boron, or stated differently, by adding more layers of boron,
by stacking
more than one unit cell detector, the amount of neutron absorbing material
within the detector
stack is increased. With more boron, there is a greater likelihood of
detecting a neutron because
as the neutron passes through the detector there is a greater chance that it
will interact with at
least one layer of boron. Thus, in one embodiment, multiple unit cell
detectors are stacked
together to increase the intrinsic efficiency of the detector. In one
embodiment, the detector is
multi-layered and includes greater than 20 layers.
FIG. 16B illustrates another embodiment of scaling the detector of the present
invention.
In one embodiment, unit cells are "tiled" to achieve areas of up to 1 m2. Each
square 1605 in the
detector matrix 1606 represents one unit cell detector and by having a 10 tile
x 10 tile detector,
large areas can be achieved. Each tile has a separate electrical line 1607
feeding into a data
acquisition system. Tiles are separated by grooves 1608 for electrical
insulation.
In yet another embodiment, FIG. 16C shows the detector 1610 in a foldable
geometry,
which allows reaching much larger areas by attaching lm x 1 m detectors, such
as those shown in
FIG. 16B, folded together into a package. Folding allows for greater
transportability of the
detectors, which, when unfolded, achieves much larger detection areas, thereby
increasing
detection efficiencies.
FIG. 17 illustrates the detection efficiency of the B-10 detector of the
present invention by
plotting the number of 10B layers 1701 required to achieve the same thermal
neutron detection
efficiency 1702, as that of a 2-inch diameter 3He tube having 4 atm pressure.
In the exemplary
simulation, the number of capture events for each layer of 1-pm thick 10B is
calculated. This
thickness was chosen because the 1.47 MeV alpha particle range in boron metal
is around 3.5 lam.
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If the layer of boron is too thick, the charged particles lose all their
energy inside the layer and
get lost without contributing to the signal. Referring to FIG. 17, it can be
seen that 40 1-pm thick
10B layers are necessary to achieve the same thermal neutron detection
efficiency as the 3He tube,
which is around 85%, as shown by the line 1703.
The large-area 10B thermal neutron detector can also be a good fast neutron
detector. In
many active interrogation techniques, it is the detection of fast neutrons
that indicate hidden
special nuclear materials. FIG. 18 compares the fast neutron detection
efficiency of the 10B
neutron detector of the present invention to a 3He-based Differential Die-Away
Analysis
(DDAA) detector. The DDAA technique can detect the thermal neutron induced
fission neutrons
after the thermalized interrogating source neutrons die-away within the
detector. The figure plots
the die-away time 1801 of the 10B neutron detector and the detection
efficiency 1802 of the
detector as a function of polyethylene thickness, since polyethylene is
layered inside 10B detector.
The DDAA detector achieves a die-away time of 40 [is with a detection
efficiency of
around 25%. That means, for the same die-away time as the DDAA detector, each
polyethylene
layer in the 10B neutron detector must be a thickness of 6 mm, as shown by the
curve 1801.
Subsequently, the intrinsic detection efficiency of the 10B neutron detector
at this point is around
20%, as shown by curve 1802, which is very similar to the DDAA detector.
FIGS. 19A through 19F illustrate, in a step-wise manner, one embodiment of a
fabrication
procedure for a large-area boron substrate layer as used in the manufacture of
the unit cell
detector of the present invention, having an area of about 1 m2, in one
embodiment. The methods
proposed follow established semiconductor techniques, which are economical and
scalable.
As shown in FIG. 19A, in step 1900, a very thin sheet of copper foil 1911 is
utilized as
the metallic base for good electrical conductivity. In one embodiment, the
thickness of copper
foil 1911 is in the range of 50 ¨ 100 lam. In one embodiment, the copper foil
sheet 1901 has an
area that is 100 cm2.
As shown in FIG. 19B, in step 1910, the copper foil 1911 is attached to a more
rigid layer
1912, such as a Kapton layer, which provides the large areal structural
strength. The
copper/Kapton layer is then immersed in a ferric-chloride solution for etching
of the 10 cm x 10
cm tile pattern and individual electrical lines.
Once the traces have been etched, the layer is mounted onto a drum 1921 for
vacuum
deposition, as shown in FIG. 19C, as step 1920.
CA 02863633 2016-02-03
As shown in FIG. 19D, step 1930 shows the deposition of boron 1931 onto the
copper
surface 1911. For deposition, the substrates attached to the drum 1933 are
rotated around in a
sputtering chamber (not shown). In one embodiment, the sputtering chamber
comprises a
magnetron 1934 for Bi0C/B4C sputtering. With the use of a linear sputtering
source 1934, the
target-to-substrate distance can be decreased and also the losses of boron in
one-dimension can
be constrained. Further, the rate of deposition can be increased through
maximizing magnetron
power densities and through scaling methods. In one embodiment, an extra
electron emitter
embedded within the boron target during sputtering. The use of extra electrons
increases the
stability and temperature of the depositions which leads to faster and more
stable boron films.
The method of using an extra electron emitter is described in United States
Patent Number
7,931,787, to Hilliard, entitled "Electron-Assisted Deposition Process and
Apparatus".
Because boron is electrically conductive, a mask 1935 is used to block the
deposition of
boron onto the etched electrical lines, thus keeping the lines from shorting.
As shown in FIG. 19E, at step 1940, after the boron has been deposited, the
large-area
boron layer 1941 is taken out of the vacuum and is ready for installation onto
the detector.
As shown in FIG. 19f, in optional step 1950, a fast neutron detector is
fabricated onto the
detector, wherein the boron/copper/kapton layer 1951 is laminated onto a sheet
of polyethylene
1952.
After each layer has been fabricated, each individual substrate layer, as
described with
respect to FIGS. 15a and 15b, are then stacked/layered into the detector,
thereby increasing the
amount of boron and maximizing the neutron detection efficiency.
Thus, the unit cell detector of the present invention comprises at least two
boron coated
metal layer sandwiching a gas cell. In one embodiment, the detector comprises
a plurality of unit
cell detectors, which may include a total of more than layers.
For fast neutrons (fission spectrum), most of the neutrons will need to be
moderated
before the boron capture occurs. It should be noted that the cross section for
capture increases as
the neutron energy decreases. Once moderated, a neutron is absorbed or
captured by the boron,
which emits charged particles. Since the particles are emitted in 180 degrees,
only one will
traverse through the gas cell, creating detectable electrons/ions. If the
first polyethylene or
scintillator layer does not moderate the fast neutron, the second layer can do
it, up to the nth
26
CA 02863633 2016-02-03
layer, thereby increasing detection efficiency. While it is noted that a
neutron can lose all of its
energy on the first collision, this is not usually the case, thus
necessitating the use of the entire
unit cell detector in each layer of the stack, including the additional
polyethylene or scintillating
sheets. Therefore, as more layers are added to the stack, the probability of
detecting more
neutrons is increased.
The above examples are merely illustrative of the many applications of the
system of
present specification. Although only a few embodiments of the present
invention have been
described herein, it should be understood that the present invention might be
embodied in many
other specific forms without departing from the scope of the present
specification. Therefore, the
present examples and embodiments are to be considered as illustrative and not
restrictive. The
scope of protection being sought is defined by the following claims rather
than the described
embodiments in the foregoing description. The scope of the claims should not
be limited by the
described embodiments set forth in the examples but should be given the
broadest interpretation
consistent with the description as a whole.
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