|Publication number||USRE38910 E1|
|Application number||US 10/654,855|
|Publication date||Dec 6, 2005|
|Filing date||Sep 4, 2003|
|Priority date||Jan 10, 2002|
|Also published as||CA2416056A1, CA2416056C, CN1312470C, CN1431488A, DE60323891D1, EP1327878A1, EP1327878B1, EP2042859A2, EP2042859A3, US6567498|
|Publication number||10654855, 654855, US RE38910 E1, US RE38910E1, US-E1-RE38910, USRE38910 E1, USRE38910E1|
|Inventors||Robert E. Troxler, W. Linus Dep|
|Original Assignee||Troxler Electronic Laboratories, Inc.|
|Export Citation||BiBTeX, EndNote, RefMan|
|Patent Citations (9), Non-Patent Citations (5), Referenced by (20), Classifications (7), Legal Events (4)|
|External Links: USPTO, USPTO Assignment, Espacenet|
This invention relates to the measurement of density, and more particularly to a test instrument and method for measuring the density of a sample using gamma radiation. The invention is especially suited for measuring the density in a relatively thin zone below the surface of a sample.
In the asphalt pavement construction industry, portable nuclear gauges are frequently used for measuring the density of the asphalt pavement. Often, the asphalt paving material is applied in relatively thin layers, e.g. on the order of about one to two inches in thickness, over a prepared roadbed foundation or an existing paved roadway. Consequently, there is a need to measure density of the pavement sample in a relatively thin zone, e.g., one to three inches in depth, below the pavement surface. To this end, nuclear density gauges have been developed for directly measuring the density of a thin layer of paving material. For example, nuclear “thin layer” gauges of this type are described in commonly owned U.S. Pat. Nos. 4,641,030; 4,701,868 and 6,310,936. The gauges described in these patents use a Cesium-137 (137CS) source of gamma radiation containing approximately eight millicuries of Cesium-137. Gamma radiation that is Compton scattered from the underlying sample is detected by Geiger-Mueller tubes positioned to form two geometrically differing source-to-detector relationships, and the density of the material is calculated based upon the gamma radiation counts detected by the respective detectors.
Although the activity of the gamma radiation source in these gauges is quite small, in the millicurie range, and can be safely used by an operator with ordinary precautions and care, regulatory agencies impose restrictions on the handling, transport, storage and use of such gauges, and on persons qualified to operate such gauges. Consequently, there exists a need for a gauge which uses a radiation source of a much lower activity level which is not subject to the regulatory requirements of existing gauges.
It is therefore an object of the present invention to provide a nuclear gauge suited for measuring the density in a relatively thin zone below the surface of a sample, and which uses a low activity radiation source.
It is a more specific object of the present invention to provide a gauge which can operate using a gamma radiation source having an activity in the microcurie range, and more specifically with an activity of no more than 100 microcurie, and more desirably an activity of no more than 50 microcurie. Gauges employing these low activity nuclear sources are subject to fewer and less stringent restrictions and regulations, if any.
Prior attempts to produce nuclear gauges using low activity (microcurie) radiation sources have had limited success, primarily because of their limited levels of accuracy. By way of example, one prior nuclear gauge using a low activity nuclear source is described in commonly owned U.S. Pat. No. 4,766,319. The main difficulty in developing a gauge based on a low activity gamma radiation source is that the signal to noise ratio of the gamma radiation detection is low because of the relatively low gamma radiation flux from a low activity source. Background radiation from certain naturally occurring radioactive elements (e.g. K-40, U and Th) present in the material to be tested generate noise which cannot be ignored without sacrificing the accuracy of measurement. With conventional gauges using higher activity gamma radiation sources (e.g. a 8000 microcurie Cs-137 source), the signal to noise ratio is high and the background radiation does not contribute significant error.
The present invention provides a nuclear density gauge and method which is suited for measuring the density in a relatively thin zone beneath the surface of a sample of paving material. The gauge may be designed to measure the density in a zone up to a specific depth of, for example, up to 1 or as much as 3 inches beneath the surface of the material sample. The gauge uses one or more gamma radiation sources having a total activity of no more than 100 microcurie. The gauge includes a gauge housing having a surface adapted to be positioned on a surface of the material sample. The microcurie gamma radiation source is mounted within the housing for emitting gamma radiation through the base and into an underlying material sample. At least one energy selective gamma radiation detector is mounted within the gauge housing in spaced apart relation with respect to the gamma radiation source, with the detector being operable for producing signals representing the energy level of the detected gamma radiation. Density calculating means is connected to the detector and is operable for calculating a value for the density of the material based upon detected signals having an energy level within a predetermined portion of the energy spectrum of the gamma radiation detected by the detector. In one embodiment, the density calculating means includes an analyzer which is connected to the detector and is operable for classifying and accumulating signals from the detector into one or more channels corresponding to said predetermined portion of the energy spectrum. The analyzer may, for example, comprise a multichannel analyzer which classifies and accumulates signals in a plurality of discrete channels over the energy spectrum of the gamma radiation detected by the detector, and wherein at least one of these discrete channels defines said predetermined portion of the energy spectrum.
In one specific embodiment, the predetermined portion of the energy spectrum which is used for density calculation has a lower limit of 0.1 MeV or greater and an upper limit which is less than the characteristic primary energy of the source. The gamma radiation source may comprise at least one Cesium-137 gamma radiation source with a 0.662 MeV primary energy. Preferably, the detector is a scintillation detector, and the system may include an analyzer connected to the scintillation detector which is capable of identifying the counts which have an energy within the specified energy spectrum.
Some of the features and advantages of the invention having been described, others will become apparent from the detailed description which follows, and from the accompanying drawings, in which:
The present invention now will be described more fully hereinafter with reference to the accompanying drawings, in which preferred embodiments of the invention are shown. This invention may, however, be embodied in many different forms and should not be construed as limited to the embodiments set forth herein; rather, these embodiments are provided so that this disclosure will be thorough and complete, and will fully convey the scope of the invention to those skilled in the art. Like numbers refer to like elements throughout.
The present invention is based on the scattering and absorption properties of gamma radiation with matter. For gamma radiation with energies less than 2 MeV, there are two dominant interacting mechanisms with matter. In the 0.1 to 2 MeV energy range, the dominant mechanism is inelastic scattering (Compton scattering). For energies less than 0.1 MeV, the dominant mechanism is photoelectric absorption. In the 0.1 to 2 MeV energy range, the amount of gamma radiation scattering (energy degradation) is a function of electron density of the material and therefore, density is a fundamental measurement property. This results in a nuclear attenuation per unit-length mass-density that is less influenced by the material composition. At energies below 0.1 MeV, the photoelectric absorption of gamma radiation is sensitive to the atomic number of the material and hence to the chemical (elemental) composition of the material. Therefore, when a gamma radiation source of sufficient energy is placed near a material, and an energy selective gamma radiation detector is used for gamma radiation detection, gamma radiation mainly undergoing Compton scattering can be counted exclusively. With proper calibration, the gamma radiation count can be converted to an absolute density.
According to one specific embodiment of the invention, a 137CS gamma radiation source with a 0.662 MeV primary energy is used. However, other gamma radiation sources with different primary energy levels could be employed, such as 60Co for example. Gamma radiation interacting with the sample is measured with a detector, which is preferably an energy selective detector configured to detect gamma radiation in a predetermined energy spectrum. Gamma radiation detectors may be configured in various ways to be selective to a desired energy spectrum. For example, in the embodiment shown and described herein, an energy selective scintillation detector is used, specifically a sodium iodide (NaI) crystal mounted on a photomultiplier tube (PMT). When using a 137Cs source, gamma radiation interacting with the sample with energies in the predetermined range 0.1 to 0.4 MeV are counted. In a further specific embodiment, gamma radiation with energies in the predetermined range of 0.1 to 0.25 MeV are counted. The gamma radiation within this energy spectrum is that which has interacted with the underlying material and has been back-scattered to the detector. Because of Compton scattering, the radiation posses a lower energy level than the 0.662 MeV primary energy of the 137Cs source. For gamma radiation sources other than 137Cs, the upper limit would be selected in a similar manner based upon the energy distribution for the particular source selected.
One embodiment of a gauge in accordance with the present invention is shown in FIG. 1. The gauge is indicated generally by the reference character 10. The gauge includes a base 12 having a substantially planar lower surface and a gauge housing 14 which cooperates with the base 12 to protectively enclose the various components of the gauge. A handle 16 extends upwardly from the gauge housing 14 to facilitate transporting the gauge. On the upper side of the gauge housing 14 suitable input-output devices are provided, such as the keypad 18 and display 19 shown in the drawing.
Additional components of the gauge are mounted to the upper surface of the base 12. As shown, located adjacent one longitudinal end of the base 12 is a source plate 20. Source plate 20 is in the form of an elongate bar. In the illustrated embodiment, a series of three discrete radiation point sources 22 are mounted at spaced-apart locations to one side of the source plate 20. It will be understood that more than three discrete point sources could be used. In an alternative embodiment, not illustrated, the radiation source may be continuous and distributed along the entire length of the source plate. Alternatively, the sources may be arranged in a pattern, such as a circular pattern, surrounding the detector. In any event, the total activity of the gamma radiation sources does not exceed 100 microcurie. In the particular embodiment illustrated, the gamma radiation source is Cesium-137 and each individual point source of Cesium-137 has an activity of no more than 10 microcurie.
The source plate 20 is preferably mounted so that it can be readily removed from the base plate 12. In the embodiment shown, the source plate 20 has two vertically extending holes adjacent each end which are adapted to receive threaded fasteners, such as bolts 24, and which threadably engage suitably tapped holes 25 formed in the base plate 12. This arrangement makes it possible to remove the radiation source, either for replacement or for taking background radiation counts, as explained more fully below. It also ensures that the source plate 20 is reliably and consistently located at the same position when installed on the base 12, since the distance and geometrical relationship between the source plate 20 and the radiation detector must be consistently maintained for accurate and reproducible results. For radiation safety, the source plate 20 may be tethered to the gauge to prevent loss while removed from the gauge.
An energy selective detector system is mounted to the base 12 adjacent the opposite end from the source plate 20. In the particular embodiment illustrated in
Radiation shielding 36 is also mounted on the base plate 12. The shielding 36 is located directly between the source plate 20 and the radiation detector assembly to inhibit gamma radiation emanating from the gamma sources 22 from passing directly from the sources to the detector. Consequently, the only gamma radiation from the sources 22 that is received by the detector is radiation which has passed through the base 12 into the underlying material sample, and which has interacted with the material sample before being scattered back upwardly through the base 12 to the sodium iodide crystal 26. Thus, the gauge operates in the “back-scatter” mode. Any suitable material capable of blocking gamma radiation can be used as the shielding 36, with lead or other dense metals being typical.
The functional components of the circuit board 34 are shown schematically in FIG. 2. An analog-to-digital converter 38 transforms the amplified analog signals from amplifier 30 into digital signals quantifying the energy level of the gamma radiation (photon) count. The output of the analog-to-digital converter 38 is directed to an analyzer device, which in the illustrated embodiment is a multi-channel analyzer (MCA) 39 which accumulates the number of gamma radiation (photon) counts of different energy levels into a plurality of channels, each channel corresponding to portion of the energy level spectrum. For purposes of density calculation, only a predetermined portion of the overall energy spectrum detected by the detectors is considered. Thus, only the accumulated counts from one or more of the channels corresponding to this predetermined portion are considered for the density calculation. For example, in one specific embodiment, this energy spectrum has a lower limit of 0.1 MeV and an upper limit of 0.4 MeV when a 137Cs gamma radiation source is used. In a more specific embodiment, the lower limit is 0.1 MeV and the upper limit is 0.25 MeV. Other channels of the analyzer representing other slices of the energy spectrum may be considered for taking standard counts or in compensating for background radiation. The output of the MCA 39 is directed to a processor 40 containing a set of stored instructions suitable for converting the accumulated gamma radiation (photon) counts from the MCA into a density value. The processor 40 is operatively connected to the keypad input device 18 and to the output display 19.
Preferably, the source or sources of gamma radiation are configured so that gamma radiation emanates from a laterally extending area or zone so as to provide for a number of individual of pathways along which the gamma radiation may travel downwardly into the underlying sample. The resulting backscattered radiation also travels along a number of pathways back up to the detector system. In the embodiment illustrated, there are three discrete 10 microcurie point sources of 137Cs mounted on the source plate 20, and the gamma sources are oriented along a line generally perpendicular to a line passing directly from the source plate to the detector. Since the detector is capable of receiving radiation over its entire area, there are numerous paths of travel for the gamma radiation passing downwardly into the underlying sample and being backscattered to the detector system. It will be appreciated that similar results would be achieved from a source which extends along the entire length of the source plate 20. To make more efficient use of the detector area, the detector system may include a plurality of smaller sodium iodide crystals and associated photomultiplier tubes arranged side-by-side, instead of the single crystal 26 and photomultiplier tube 28 shown in FIG. 1. By providing multiple paths of travel in this manner from the source to the detector, the gauge is able to see a larger volume of the sample and the error caused by the surface roughness of the sample is thereby reduced.
Scintillation detectors are sensitive to temperature fluctuations. In the digital spectrum produced by the MCA, the energy level of the gamma radiation detected by the scintillation detector is correlated into one of many (e.g. 512) channels representing the counts corresponding to a particular gamma radiation energy level or range. This spectrum may be represented graphically as extending in the x-direction, with the total number of counts in each channel extending in the y-direction. When the temperature fluctuates, the spectrum fluctuates non linearly in the x-direction. Therefore, a peak once centered on one channel may end up centered on a different channel. If one wants to find the gamma radiation (photon) counts in channels between Clower, representing the energy Elower, and Cupper, representing the energy Eupper, because of these fluctuations, the counts obtained from using the “raw” spectrum will have uncertainties due to the temperature sensitivity. An analog or digital spectrum stabilizer is used to stabilize the spectral drifts resulting from temperature fluctuations in the NaI detector. For purposes of spectrum stabilization, the gauge is provided with an additional 1 microcurie 241 Am gamma radiation reference source 45 mounted near the detector 26 in the embodiment shown in FIG. 1. The 0.056 MeV peak from the source 45 is used as a reference point by the MCA for stabilization of the spectrum.
During a 4 minute counting time, the MCA collects counts, which are then corrected for signal amplitude fluctuations and stored in a buffer. At the end of counting, the MCA gives the stabilized spectrum.
In an alternative approach, spectrum stabilization could be carried out without requiring an additional radiation source for reference. A tiny “leak” hole could be provided in the shielding 36 so that a small fraction of the gamma radiation can pass directly from the source 22 to the detector 26. In this instance, the 0.662 MeV peak of the gamma radiation source itself can be used as a reference point for spectrum stabilization.
In order to obtain an accurate density measurement, it is necessary to quantify the background gamma radiation from the sample and its surroundings. Conventional nuclear density gauges avoid this issue by using a stronger gamma radiation source (e.g. about 8000 micro Curie) resulting in such a large signal to noise ratio that the effect of background radiation can be ignored. With the present invention, there are several possible approaches to compensating for background gamma radiation. According to one approach, for example, the source plate 20 can be physically removed from the gauge and placed in a location shielded from the detector. Then, the sourceless gauge can be operated to obtain a gamma radiation count representing the background spectrum. According to another approach, the gauge can be constructed with a source which can be moved from an unshielded active position when operated for density measurement, to an internally shielded location within the gauge when operated for background calibration. One exemplary embodiment using this approach is illustrated in FIG. 3. To avoid repetition, like reference numbers with prime notation (′) added are used to identify elements in this embodiment which correspond to elements previously described. In this embodiment, the gamma radiation source 22′ is located on a disk 52 which is mounted for rotation within a shielded enclosure 54. Both the disk 52 (
Gamma radiation background may also be estimated “on-the-run” based upon a measurement of the gamma radiation counts having an energy level at or about 1.460 MeV. The element potassium has a long-lived radioisotope, K-40, that emits 1.460 MeV gamma radiation. Since potassium is present in the minerals typically used as the aggregate for an asphalt paving mix, Compton scattering of the 1.460 MeV gamma radiation produces background radiation in the energy spectrum which is of interest for density measurement. Another approach involves mathematical fitting of the straight-line part of the 0.662 MeV gamma radiation peak. The slope of this line can be used to estimate the background. Still another approach involves having a separate smaller detector system for background measurement. This detector may be connected to gauge electronics with a cable and placed in the side of the detector that is away from the sources, or may be placed outside the gauge enclosure.
Nuclear density gauges use radioactive sources having a finite half-life. The source activity decreases with time due to disintegration of nuclei. To compensate for the varying source activity, the measured gamma radiation count is normalized to the count on a standard. This count ratio is then independent of time. In conventional gauges, this standard is a polyethylene block. The present invention can employ any of several methods for acquiring a standard calibration count. For example, in one approach, the gauge can be placed on a standard plates two to three inches thick and of a surface area one or two times the footprint size of the gauge. These standard plates can be magnesium, aluminum, or a combination of magnesium and aluminum, and backscatter counts are acquired on each plate. The gamma radiation streaming from source to detector is completely stopped by the shielding, so that only a backscatter reading is acquired, and counts are taken in a particular energy window, for example 0.1-0.25 MeV (for a 137Cs source with 0.662 MeV primary energy).
In another approach, a small bore hole is formed in the shield to provide a direct path for the gamma radiation from the source to the detector so that the detector could see a direct beam of gamma radiation of 0.662 MeV energy. The net counts in the 0.662 MeV 137Cs peak can be used as the standard count, when the gauge is placed on the standard plate as well as on-the-run. Here, on-the-run means when the gauge is placed on a testing material. When the gauge is placed on the testing material, the standard count (the net counts in the 0.662 MeV primary energy) is taken simultaneously with the backscatter density count.
In still another approach, a small Geiger Muller tube is incorporated in the gauge housing near the primary source and is used to ascertain the standard count. This tube is inside the gauge and is not affected by the density of the underlying material.
As with other nuclear gauges, the gauge has to be calibrated to convert gamma radiation counts to material bulk densities. Preliminary calibration was performed using three solid metal calibration plates: a magnesium plate with soil equivalent density of 109.8 pcf (pounds per cubic foot), a composite magnesium/aluminum plate with soil equivalent density of 133.3 pcf, and an aluminum plate with soil equivalent density of 161.2 pcf. The gauge was operated in the backscatter mode. Counts in a 0.1 to 0.25 MeV energy window were used to estimate the density. The background radiation from the sample and its surrounding was measured by obtaining counts when the 137Cs gamma radiation source was removed from the gauge. The gauge was placed on the magnesium plate and three 4-minute counts were obtained. The average of these counts was calculated as Cbgd1. The gauge was then placed on the magnesium/aluminum plate and three 4 minute counts were taken. The average count was calculated as Cbgd2. The source was reinstalled in its operative unshielded position in the gauge and the gauge was placed on the magnesium plate and three 4 minute counts were collected. The average count was calculated as CMg. The gauge was then placed on the composite magnesium/aluminum plate and three 4 minute counts were obtained. The average count was calculated as CMgA1. The gauge was then placed on the aluminum plate and three 4 minute counts were collected. The average count was calculated as CA1.
The 4 minute background count Cbgd is given by Cbgd=(Cbgd1+Cbgd2)/2. The background corrected counts on the magnesium plate was used as the standard count (Cstd) where Cstd=CMg−Cbgd.
The count ratio (CR) for each sample was then calculated using the following equation: CRplate=(Cplate−Cbgd)/Cstd where Cplate is the count on a particular calibration plate. Table 1 shows the data.
The calibration counts are used to determine the calibration constants by fitting to a standard equation of the form
Where A, B, and C are the fitting coefficients or calibration constants and D is density.
The best fit gave the following values for the three calibration constants.
A portable calibration unit can be produced with sandwiched 1-inch thick Mg and 1-inch thick A1minimum plates. The 1-inch plate of Mg itself is formed by two 0.5-inch plates. The plates preferably have a surface area about one to two times the footprint of the gauge.
Background Count: Place the plates flat on the ground with the 1-inch Mg plate facing up. Place the gauge, with the source removed or in the shielded position, on the plate. Acquire counts for 4 minutes (Cbgd).
Standard Count: Place the plates flat on the ground with the 1-inch Mg plate facing up. Place the gauge, with the source install or in the unshielded operative position, on the plate. Acquire counts for 4 minutes (Cstd,raw). The standard count Cstd=Cstd,raw−Cbgd,
Mg Count for Calibration: CMg=Cstd,
MgA1 Count for Calibration: Now remove the top 0.5-inch Mg plate. Place the gauge, with source installed and active, on the plate and acquire counts for 4 minutes (CMgA1,raw) MgA1 Count CMgA1=CMgA1, raw−Cbgd.
A1 Count for Calibration: Now turn the plates so that the 1-inch A1 plate is facing up. Place the gauge, with source installed and active, on the plate and acquire counts for 4 minutes (CA1, raw). A1 Count CA1=CA1,raw−Cbgd.
The counts as acquired above may now be used as described in Calibration Example 1 to obtain calibration constants.
The calculation of the density of a material sample is preferably carried out by a suitably programmed microprocessor or by any other functionally equivalent device, such as an application specific integrated circuit or a general purpose computer. The gauge is placed on the sample to be measured and a count is obtained for a suitable period of time, such as 2 to 4 minutes. From the MCA, stabilized counts for the particular portion of the energy spectrum of interest are obtained. Then using the density equation and calibration constants obtained as described in the Calibration Examples above, a value for the density of the sample may be obtained. This value is displayed to the user on the display 19 of the gauge.
In a preferred implementation of this method, the calculations are carried out on the accumulated gamma radiation (photon) counts repeatedly at a frequent intervals as the counting proceeds, such as everyone to two seconds, treating each as a frequency packet, and a digital filtering algorithm is utilized to decrease the statistical variation of the packet. Instead of waiting until the end of a 2 to 4 minute count to display the density value, this approach makes it possible to provide to the user an almost real-time display of the calculated density value while the count is still proceeding. The density values may be displayed to the user graphically as a function of time, as shown in FIG. 1. As the digitally filtered density value settles down to a steady state, the user may decide to accept the calculated density value as being sufficiently accurate, and to discontinue the measurement procedure without waiting until the end of the full two or four minute count. Thus, this calculation method can reduce the time required for taking density measurements and can thereby increase efficiency and productivity.
According to a further modified embodiment of the invention, it is possible for the user to adjust or to set the depth of field of the gauge so that density measurements can be obtained from a specific depth into the underlying material, such as a depth of up to one inch or up to three inches. This is achieved by adjusting the source to detector geometry. In particular, in this modified embodiment, the source can be adjustably positioned at one of several different distances from the gauge. In the embodiment shown in
Many modifications and other embodiments of the invention will come to mind to one skilled in the art to which this invention pertains having the benefit of the teachings presented in the foregoing descriptions and the associated drawings. Therefore, it is to be understood that the invention is not to be limited to the specific embodiments disclosed and that modifications and other embodiments are intended to be included within the scope of the appended claims. Although specific terms are employed herein, they are used in a generic and descriptive sense only and not for purposes of limitation.
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|U.S. Classification||378/89, 378/86|
|International Classification||G01N23/06, G01N23/203, G01T1/36|
|Apr 4, 2006||CC||Certificate of correction|
|Oct 27, 2006||FPAY||Fee payment|
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|Oct 20, 2010||FPAY||Fee payment|
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|Oct 22, 2014||FPAY||Fee payment|
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