Rapid scanning system for fuel drawers

A nondestructive method for uniqely distinguishing among and quantifying the mass of individual fuel plates in situ in fuel drawers utilized in nuclear reactors is described. The method is both rapid and passive, eliminating the personnel hazard of the commonly used irradiation techniques which require that the analysis be performed in proximity to an intense neutron source such as a reactor. In the present technique, only normally decaying nuclei are observed. This allows the analysis to be performed anywhere. This feature, combined with rapid scanning of a given fuel drawer (in approximately 30 s), and the computer data analysis allows the processing of large numbers of fuel drawers efficiently in the event of a loss alert.

BACKGROUND OF THE INVENTION 
1. Field of the Invention 
A nondestructive method and apparatus for uniquely distinguishing among and 
quantifying the mass of individual nuclear fuel plates in situ in fuel 
drawers utilized in nuclear reactors is described. The method is both 
rapid and passive, reducing the personnel hazard of the commonly used 
irradiation techniques. It comprises simultaneous collimated neutron and 
collimated, energy resolved gamma ray analyses. The spatial resolution of 
both is sufficient to identify the smallest expected unit plate which is 
2.5 cm long. 
Said simultaneous neutron and gamma ray analysis is crucial to avoid 
intentional attempts to confound the procedure for the purpose of theft of 
fissile materials. The several pieces of information obtained for each 
fuel element make successful illegal substitution of materials in an 
attempt to deceive the system, virtually impossible, whereas, for example, 
simple neutron monitoring could be quite readily overcome. High spatial 
resolution of the present method also prevents the averaging of neutron or 
gamma events essential in the deception by substitution. 
The invention is a result of a contract with the Department of Energy. 
2. Description of Prior Art 
The novelty search has produced several patents which differ from the 
instant invention in one or more of several ways. The most important and 
prevalent difference is that many of these patents teach neutron 
irradiation of the sample to be investigated with subsequent analysis of 
the resulting neutrons and gamma rays which arise from nuclear 
decomposition (U.S. Pat. Nos. 3,496,357 and 3,786,256 exemplify this 
approach.) The instant invention provides sufficient sensitivity to simply 
monitor the spontaneous decomposition of normally radioactive nuclei, 
thereby eliminating the hazard and inconvenience of neutron irradiation. 
The second salient difference is that those patents which do not actually 
mention neutron irradiation but suggest simultaneous neutron and gamma ray 
monitoring of samples do not teach energy resolved gamma analysis (See, 
e.g., U.S. Pat. No. 3,786,257.) A third group mentions either resolved 
gamma or resolved neutron analysis but not a combination of the two (See 
U.S. Pat. No. 3,717,765 for a description of the use of gamma 
spectroscopy, for example). A maximum of non-destructively and rapidly 
obtained sample information is crucial to the method of the instant 
invention which uses such analysis to quantitatively identify all 
radioactive elements present in nuclear fuel samples to reduce the 
possibility of willful deception by potential thieves. 
The simultaneous, rapid use of both spatial resolving power along with 
isotope specific gamma and neutron materials identification of an 
unirradiated fuel sample is therefore unique. Further, the use of .sup.3 
He proportional counters in a collimated neutron scanning system is novel 
by itself. 
STATEMENT OF THE OBJECTS 
An object of the present invention is to rapidly, non-destructively, 
uniquely and quantitatively identify the mass of individual nuclear fuel 
plates according to their radioisotope composition. 
Another object of the invention is to reduce the hazard and inconvenience 
of such analysis which are usually performed by simultaneous neutron 
irradiation of the sample and investigation of the resulting neutrons and 
gamma rays produced from nuclear disintegration. 
Another object is the prevention of repeated, premeditated theft of small 
quantities of fissile materials by the alteration of the isotopic 
composition of the fuel elements. 
Another object is the examination of said nuclear fuel plates in situ in 
the fuel drawers currently utilized in nuclear reactors. 
Another object is the rapid characterization of irradiation products of 
non-fuel materials with regard to transuranic isotopes and other 
radioactive species. 
Other objects, advantages and novel features of the invention will become 
apparent to those skilled in the art upon examination of the following 
detailed description of a preferred embodiment of the invention and the 
accompanying drawing. 
SUMMARY OF THE INVENTION 
The method of this invention teaches simultaneous neutron and energy 
resolved gamma scans of nuclear fuel elements with spatial resolution 
sufficient to identify the smallest expected unit plate which is presently 
2.5 cm long, at a rate of 2.5 cm/s. The neutron scan senses the 
spontaneous fission neutrons from .sup.240 Pu and serves to characterize 
the isotopic content of this element in the fuel plate. Collimation of the 
neutrons essential for the spatial resolution is achieved by a 
polyethelene slot arrangement. The reduced neutron counts resulting from 
this collimation are compensated for by the use of four .sup.3 He 
proportional counters arranged at 90.degree. intervals around a circle 
perpendicular to the direction of travel of the drawers. An energy 
resolved gamma ray scan simultaneously characterizes the .sup.239 Pu, 
.sup.241 Pu, .sup.241 Am, and .sup.235 U in particular as well as any 
other isotopic content for species which emit spontaneous gamma radiation. 
Appropriate energy resolution of these specific gamma rays is provided by 
a Ge(Li) detector. An on-line least-squares analysis comparison of the 
observed scan data to that expected for the given drawer is achieved using 
a microprocessor control system. The apparatus of this invention further 
includes a conveyor system whereby fuel drawers are carried past the gamma 
and neutron detectors. The method and apparatus can be used to analyze 
more general radioactive products from irradiation of materials other than 
fuel elements. It is both rapid and passive reducing the inconvenience and 
personnel hazard of the commonly used irradiation techniques.

DESCRIPTION OF THE PREFERRED EMBODIMENTS 
In a preferred embodiment, the system is designed to scan typical fuel 
drawers at a rate of one or two per minute, with a sensitivity adequate to 
detect the presence or absence of a typically 1-in.-long plutonium fuel 
plate in a drawer containing as much as 36 in. (total) of plutonium fuel 
and a variety of coolant, structural, and fertile mockup materials also 
present in the drawer. The technique assumes that the plutonium fuel 
plates in the facility inventory are well characterized but does not 
require that all fuel plates have the same .sup.240 Pu, .sup.241 Am, etc., 
isotopic content, these radioisotopes being in sufficient abundance to be 
easily monitored and sufficient in number to ensure against clandestine 
theft of nuclear materials by simple alteration of the fuel element 
composition in a given set of drawers. An example of possible fuel 
inventory variation is shown in Table I. Table II shows the measured or 
calculated neutron and isotope-specific gamma intensities for the 
plutonium fuel plates listed in Table I. The neutron and gamma line 
intensities are tabulated as relative values per inch of fuel plate, with 
the Pu/U/Mo fuel of 11.56% .sup.240 Pu content arbitrarily designated as 
the unity response. 
TABLE I 
______________________________________ 
TYPICAL ISOTOPIC VALUES FOR A VARIETY 
OF PLUTONIUM FUELS IN A CRITICAL FACILITY 
Plutonium 
.sup.239 Pu 
.sup.240 Pu 
.sup.241 Pu 
.sup.241 Am 
Per Inch 
Fuel Type (%) (%) (%) (%) (g) 
______________________________________ 
Pu/Al 95.25 4.50 0.20 0.24 34.1 
Pu/U/Mo 90.80 8.66 0.51 0.46 20.0 
Pu/U/Mo 87.00 11.56 1.20 0.59 31.1 
Pu/Al 74.20 22.33 2.86 1.80 35.3 
Pu/U/Mo 68.70 26.40 3.39 2.19 37.7 
______________________________________ 
TABLE II 
______________________________________ 
MEASURED (OR CALCULATED) NEUTRON AND 
GAMMA RESPONSES FOR A VARIETY OF 
PLUTONIUM FUELS 
Relative Relative Relative 
Total Total .sup.239 Pu 
Total .sup.241 Am 
.sup.240 Pu 
Neutrons (414 keV .gamma./ 
(662 keV .gamma./ 
Fuel Type 
(%) per Inch inch) inch) 
______________________________________ 
Pu/Al 4.50 0.73 .+-. 0.07 
(1.26 .+-. 0.13) 
(0.42 .+-. 0.04) 
Pu/U/Mo 8.66 0.47 .+-. 0.05 
(0.67 .+-. 0.07) 
(0.49 .+-. 0.05) 
Pu/U/Mo 11.56 1.00 .+-. 0.10 
1.00 .+-. 0.10 
1.00 .+-. 0.10 
Pu/Al 22.33 3.17 .+-. 0.32 
0.96 .+-. 0.10 
3.44 .+-. 0.30 
Pu/U/Mo 26.40 (2.75 .+-. 0.28) 
(0.95 .+-. 0.10) 
(4.47 .+-. 0.45) 
______________________________________ 
As can be seen in Table II, each of the five classes of plutonium fuel has 
a characteristic signature when the three independent quantities (total 
neutron, .sup.239 Pu .UPSILON., .sup.241 Am .UPSILON.) are considered. 
Furthermore, based on experimental measurements with 25 separate fuel 
plates taken from among the first four classes listed in Table II, the 
uniformity in the signature from plate to plate within a class appears to 
be quite good. This is probably due to the excellent quality control 
required in the manufacture of such plates. The Pu/U/Mo plates have a 
neutron output consistent with 100% spontaneous fission, but the Pu/Al 
plates have an additional Al (.alpha.,n) component (.about.70% for the 
4.50% .sup.240 Pu plates and about 50% for the 22.33% .sup.240 Pu plates). 
A design of the scanning apparatus required to perform a fuel-drawer 
inventory measurement is shown in the FIGURE. Fuel drawers are loaded on 
to a "conveyor belt" that transports them, first through a collimated 
fast-neutron detector, and subsequently past a collimated intrinsic Ge or 
Ge(Li) gamma detector. For ease and accuracy of drawer identification, it 
is recommended that each drawer be tagged with a "grocery store"-type 
laser-scan identification label. This label could also contain information 
on the drawer's current plutonium content. A photocell-laser sensing 
system would read the label. The drawer identification could be routed to 
a microprocessor- or minicomputer-based data acquisition system. The 
observed "signature" for each inch of fuel plate in the drawer (as 
determined by the neutron output and at least two isotopic gamma lines) 
could then be compared to the expected values. Passing would require that 
each inch of the drawer checks against expected values. Count rates for 
the conceptual design of the FIGURE (both neutron and gamma) are such that 
a statistically reliable signature for an inch of any plutonium fuel plate 
considered in Table 2 can be obtained in &lt;1 s. The signature for the 
combination of any two fuel types (two rows of fuel plates in a drawer) is 
also obtained statistically in less than 1 s/in. Table 2 also shows that 
accuracies of each attribute measurement need only be .+-.10% in order to 
verify a signature. For most cases, an even poorer accuracy would suffice. 
The apparatus and method has been tested in separate neutron and specific 
gamma scanning prototypes and an integral unit is under construction. The 
method can be considered successfully proven, based on these prototype 
data with actual Pu fuel plates simulating typical Pu containing fuel 
drawers. 
The neutron scan (.gtoreq.10% solid angle and overall 4% direction 
efficiency) senses primarily spontaneous fission neutrons from .sup.240 Pu 
and thus, serves to characterize .sup.240 Pu isotopic content. Collimation 
of the neutrons essential for the spatial resolution is achieved by a 
polyethylene slot arrangement. The reduced neutron counts resulting from 
this collimation are compensated for by the use of four .sup.3 He 
detectors arranged at 90.degree. intervals around a circle perpendicular 
to the direction of travel of the drawers. The gamma scan in principle 
could be used to additionally characterize .sup.239 Pu, .sup.241 Pu, 
.sup.241 Am, .sup.235 U, and .sup.238 U (as well as any other) isotopic 
content, by utilizing various specific gamma lines. In the present 
version, however, a more limited selection of strongly emitted .sup.241 Am 
and .sup.239 Pu gamma lines will be used. (661 keV for .sup.241 Am 
characterization and 414 keV for .sup.239 Pu characterization.) A Ge(Li) 
or HPGe detector will be used to assure proper energy resolution of these 
specific gamma rays. Thus, with the present invention, a quantitative 
characterization of three separate transuranic isotopes is achieved in the 
scan. Since fuel plates of differing types (a typical U.S. zero power 
reactor may have five separate fuel plate types in its inventory) will 
have different mixes of the .sup.240 Pu, .sup.239 Pu, and .sup.241 Am 
isotopes, the method serves to uniquely distinguish among the five types 
and to quantify the mass of fuel plates. A key feature of this system is 
the microprocessor control of the entire operation. This is necessary 
since the sheer volume of data generated in a short period of time is 
immense. In addition, the identification process requires an on-line 
least-squares analysis comparison of the observed scan data to that 
expected for the given drawer. The amount and rate of data analysis 
required is easily handled with a micro-NOVA-type unit. Microprocessor 
control will also facilitate transfer of the scan results to a larger CPU 
wherein detailed inventory can be kept up to date. 
At present, this system will probably be most useful in rapid Pu fuel 
inventory measurements at facilities having large Pu inventories. With 
suitable modifications (primarily in the acceptance cross section of the 
scanner), the system might also be useful in verifying a wide variety of 
well-characterized Pu and U containing materials, for example, a unit in 
which a continuous active neutron interrogation is performed during the 
scan process is under investigation. With this feature, .sup.235 U may be 
identifiable through .sup.235 U(n,f) reactions and detection of prompt or 
delayed neutrons. Additional specific gamma ray lines would be selectable 
for this purpose as well (186 keV for .sup.235 U, 1001 for .sup.238 U, 
etc.). The principle would be essentially the same, a simultaneous 
collimated neutron and specific gamma ray characterization of the 
materials. It is clear that such a system is considerably more difficult 
to subvert than verifications based on either neutron or gamma signatures 
alone. In addition, by requiring a signature check on each inch of fuel 
plate (as opposed to a signature check of an integral drawer), a 
successful diverter would have to be quite resourceful. Finally, the 
instant invention, unlike conventional techniques, does not require sample 
irradiation with its required proximity to an intense neutron source and 
inherent hazards. Therefore, the analysis can be performed virtually 
anywhere. 
The foregoing description of a preferred embodiment of the invention has 
been presented for purposes of illustration and description and is not 
intended to be exhaustive or to limit the invention to the precise form 
disclosed. It was chosen and described in order to best explain the 
principles of the invention and their practical application to thereby 
enable others skilled in the art to best utilize the invention in various 
embodiments and with various modifications as are suited to the particular 
use contemplated. It is intended that the scope of the invention be 
defined by the claims appended thereto.