Patent Number: 043615341
Section: description

Before describing the apparatus and technique in detail it will be helpful to consider the mathematics associated with the present invention. In a sample containing both aluminium and silicon, the grade of aluminium, Al, is related to the number of counts, G, recorded of gamma rays emitted by .sup.27 Mg at 0.844 MeV and/or 1.015 MeV, and to the sample weight, W, by the equation: EQU Al=a.sub.0 +a.sub.1 G+a.sub.2 W. (1) The constant coefficients a.sub.0, a.sub.1 and a.sub.2 are determined from linear regression analysis by calibrating the responses G and W of the apparatus against the aluminium content of known samples using linear regression analysis. The number of counts G is determined from the equation EQU G=G.sub.T -kJ, (2) where G.sub.T is the total number of gamma rays recorded in an energy window encompassing the 0.844 MeV and/or 1.015 peaks, J is the total number of counts recorded of 1.78 MeV gamma rays, and k is a constant. The term kJ is used to subtract the spectral continuum due to both the Compton scattered 1.78 MeV gamma radiation and background due to the neutron source. Thus EQU Al=a.sub.0 +a.sub.1 G.sub.T +a.sub.2 W+a.sub.3 J, (3) where a.sub.3 =-a.sub.1 K. Similarly, the chemical concentration of silicon, Si, in the sample can be related to the number of counts per unit time, H, of 1.78 MeV gamma rays due to the .sup.28 Si(n,p).sup.28 Al reaction and the sample weight, W, by: EQU Si=b.sub.0 +b.sub.1 H+b.sub.2 W, (4) where b.sub.0, b.sub.1 and b.sub.2 are constant coefficients obtainable from regression analysis by calibrating the responses H and W against the silicon content of known samples using linear regression analysis. In practice, with bulk bauxite samples, however, the total number of counts per unit time due to 1.78 MeV gamma rays from .sup.28 Al, J, contains two indistinguishable components H and I, where I is the component contributed by thermal neutron activation of .sup.27 Al. The number of counts, I, is proportional to the product of the number of thermal neutrons, N.sub.t, measured during irradiation and the aluminium concentration of the sample. Because the number of counts, G.sub.T, previously referred to, are related to the aluminium concentration, and since N.sub.t is proportional to the thermal neutron flux within the sample, the actual silicon concentration in the sample is given by: EQU Si=b.sub.0 +b.sub.3 J+b.sub.4 G.sub.T N.sub.t +b.sub.2 W, (5) where b.sub.3 and b.sub.4 are regression coefficients, and where b.sub.0, b.sub.2, b.sub.3 and b.sub.4 are calibration coefficients determined by linear regression analysis as described above. Equations (3) and (5) are used in analysis of materials in accordance with the present invention. The experimental arrangement devised to test the present invention, and illustrated in FIG. 1, comprises a neutron source 10 and thermal neutron detector 11 located close to a railway track 12 on which a small sample of material 13 can be moved. Also close to the track 12, but remote from the neutron source 10 and detector 11, is a gamma ray detector 14, suitably shielded by a lead screen 15 and a masonry shield 16. In the experimental facility shown in FIG. 1, the neutron source was 20 Ci of Am-Be, giving an estimated output of 4.4.times.10.sup.7 n/s. The samples of material 13 were contained in a rectangular brass box 17 (25.times.25.times.4 cm deep) and were irradiated by fast neutrons from underneath. The neutron source 10 was enclosed within a cyclindrical shell 21 of cadmium (see FIG. 2) which prevented the thermal neutrons emitted by the source from reaching the sample 13. The thermal neutron flux generated within the sample 13 was monitored by thermal neutron detector 11, which was a high efficiency neutron detector (filled to a pressure of 4 atmospheres with a mixture of .sup.3 He and Kr) which was also located beneath the sample container and adjacent to the neutron source. After irradiation by source 10 for 6 minutes, the sample 13 was transferred within 15 seconds to a position immediately above the gamma detector 14, which comprised a 127.times.127 mm NaI(T1) scintillation detector. The distance between the neutron source 10 and gamma detector 14 was about 7 meters. This separation between source and detector added a considerable component of distance shielding to the already appreciable shielding against source radiation provided by concrete brick structures 16 and 26 built around both the NaI(T1) detector 14 and the neutron source assemblies. The lead shield 15, of thickness 3 cm, which was built around the body of the scintillation detector 14 so as to leave only the upper plane surface exposed for measurements, provided further reduction of background radiation. Spectrum stabilization was obtained using 0.662 MeV gamma rays from a .sup.137 Cs source (not shown in the drawings) which provided a reference peak for a Canberra Industries Model 1520 analogue spectrum stabilizer. The energy spectra of gamma rays detected by the scintillation detector 14 were analysed in the initial phases of the development of the method by a Tracor Northern 4096-channel pulse height analyser (model TN-1700). At a later stage, when energy-pulse height calibrations had been fully established, Ortec single channel analysers, digital counters and a timer were used, as shown in FIG. 2, for their greater suitability to plant or mine site operating requirements. The amplifiers used were a Tennelec linear amplifier for the scintillation detector 14 and an Ortec spectroscopy amplifier for the neutron detector 11. Output from the digital counters was obtained with a strip printer. The partly schematic and partly block form diagram of FIG. 2 is essentially a more comprehensive illustration of the apparatus shown in FIG. 1. In particular, the high energy neutron source 10 and the thermal neutron detector 11 are shown more explicitly, with the neutron source 10 encased in cadmium shielding 21 and the source 10 and detector 11 positioned within masonry shielding 26. A single high voltage power supply 22 services both the thermal neutron detector 11 and the gamma ray detector 14. The output signal from the thermal neutron detector 11 is first amplified in gain control unit 23, and if the signal, when amplified, exceeds a required threshold value (determined by threshold device 24), it is supplied to the input of a digital counter 25. The output from counter 25 is fed into processor 32, which is usually a microprocessor or small computer, programmed to effect the required analysis from its three input signals. The other two input signals to the processor 32 are signals indicative of the values G and J (see the above description). These signals are derived from the gamma rays received by the gamma detector 14 as a result of the decay of, respectively, the .sup.28 Al and .sup.27 Mg isotopes formed during the irradiation of the sample. These inputs are obtained after the output of detector 14 has been processed by amplifier 26, a gamma ray discriminator, and digital counters 28 and 28A. The gamma ray discriminator has been shown in the drawing as two gamma ray single channel analysers 27 and 27A. In practice, these devices 27 and 27A may be a single unit comprising a multi-channel analyser with outputs from channels which have energy windows, typically about 0.35 MeV wide, centred on 0.844 MeV, 1.015 MeV and 1.78 MeV. Presettable timer 31 controls the operation of the digital counters 28 and 28A. Timer 31 will be synchronized with, but operating sequentially to, pre-settable timer 30 which controls the operation of the digital counter 25. The output from the processor 32 may be recorded. For example, it may be stored on magnetic tape, magnetic disc, magnetic card, punched tape, punched card, or on any other suitable medium. Alternatively, or additionally, the output from the processor 32 may be presented as a digital display, a paper print-out, or on a chart recorder. Those skilled in this field will appreciate that the actual form of the presentation of the output from processor 32 may be chosen to suit the requirements of the owner or operator of the equipment. Accordingly, a single, unspecified display unit 33 has been included in FIG. 2. In one example of the experimental testing of the present invention, bauxite samples were dried to less than 5 percent (by weight) free moisture and crushed to -6 mm particle size. It should be noted, however, that this amount of pre-treatment is not essential. The bauxite samples contained aluminium in the range from 26 wt. percent to 32 wt. percent, whilst the silicon concentrations ranged from 0.9 wt. percent to 4.5 wt. percent. The mass of the sample used for irradiation was about 4 kg. As expected, when the bauxite was irradiated with fast neutrons, the gamma ray spectra were dominated by the 1.78 MeV gamma ray peak due to .sup.28 Al, and by a spectral continuum of gamma rays which had undergone Compton scattering both within the detector and within the bulk sample. This continuum underlies the spectral peaks at 1.015 and 0.844 MeV due to .sup.27 Mg. The Comptom scattering processes in this example were dominated by gamma rays which initially had energies of 1.78 MeV, 1.014 MeV and 0.844 MeV originating from the sample, and 0.662 MeV due to the .sup.137 Cs stabilization reference source. In the case of bauxite, again as expected, the interferences from other constituents, such as the natural radioactive nuclides, was minimal, and those from .sup.56 Mn at 0.846 MeV, 1.811 MeV and 2.113 MeV were very small. (This nuclide, with a 2.57 hr. half life, can arise from the .sup.55 Mn(n,.gamma.).sup.56 Mn reaction, or from a .sup.56 Fe(n,p).sup.56 Mn reaction; the first of these reactions will contribute negligible .sup.56 Mn owing to the extremely low concentration of manganese in Australian bauxites, despite the relatively large cross section of 13.3 barns for that reaction; the second reaction, involving iron, contributes more .sup.56 Mn than the first, but constitutes a constant level of about 2 percent interference, the variation of which is only about 1 percent of the gamma ray signal from .sup.27 Mg.) Another source of spectral interference which occurred in this example arose from the neutron activation of the copper constituent of the brass sample container, which contributed a small peak at 1.05 MeV. It was necessary, therefore, to exclude from calculations all count data that would have been recorded in a narrow energy window, about 0.1 MeV wide, centred at 1.05 MeV. Apart from interferences to the spectral peaks due to .sup.28 Al and .sup.27 Mg from monoenergetic gamma rays emitted by minor constituents of the sample and sample container, there was also substantial interference from the continuum of scattered gamma radiation. The extent of interference with the 1.78 MeV spectral peak appeared to be insignificant owing to negligible gamma radiation apparent at higher energies. However, the 0.844 MeV gamma ray peak due to .sup.27 Mg received considerable interference from the substantial underlying continuum caused by Comptom scattering of the 1.78 MeV gamma radiation from the decay of .sup.28 Al and background from the neutron source. One technique that could have been used to overcome the interference problem when using multichannel pulse height analysers for neutron activation analysis is that which is described in the specification of Australian Pat. No. 468,970. That method entails an estimation of the underlying continuum which is based on the number of counts in an energy channel close to the relevant spectral peak. However, in the present experimental arrangement, an alternative method was effectively implemented with the use of single channel analysers for the activation analysis of bauxite. The method simply entailed the establishment of two particular energy windows. One window, centred at 0.844 MeV, is approximately 0.1 MeV wide. The other window, about 0.35 MeV wide, encompasses the 1.78 MeV peak. Implementation of these two energy-window conditions alone worked well because the counts accumulated within the spectral continuum occurring within the first narrow window are proportional, with good approximation, to the number of counts due to .sup.28 Al, 1.78 MeV gamma radiation. The counts recorded in these two windows were respectively denoted by G.sub.T and J in equations (3) and (5) for purposes of either determining the calibration coefficients, a.sub.i and b.sub.j, or for determining the chemical concentrations of silicon and aluminium in samples when calibrations, and hence coefficients, were already known. After performing a number of experiments with well-blended, effectively homogeneous, ore samples of accurately known composition, the data from the activation analysis were fitted against the known chemical assays for aluminium and silicon by linear regression analysis, in order to determine the constant coefficients in equations (3) and (5). The respective precisions for silicon and aluminium determinations in bulk samples were obtained in terms of the sample standard deviations (s) as shown below: (a) When using equations (3) and (5), and the method of the present invention, PA0 (b) When the contribution by gamma rays from .sup.27 Mg at 0.844 MeV is omitted from equation 5, PA0 (c) When the contributions both by the gamma rays from .sup.27 Mg at 0.844 MeV and thermal neutrons measured below the sample container are omitted from equation 5, for Al: s=0.43 percent Al PA1 for Si: s=0.14 percent Si PA1 for Si: s=0.19 percent Si PA1 for Si: s=0.82 percent Si As shown by the smaller standard deviations for the results obtained using the present invention, the present invention compares most favourably with alternatives (b) and (c). Comparisons between neutron activation determinations for aluminium and silicon, expressed as alumina (Al.sub.2 O.sub.3) and silica (SiO.sub.2) respectively, and determinations by conventional analysis are shown in FIGS. 3 and 4. The calibration equations used to calculate the neutron activation determinations of alumina and silica in FIGS. 3 and 4 were as follows: EQU Al.sub.2 O.sub.3 =71.04-0.946G.sub.T -9.636W-0.242 J (6) EQU SiO.sub.2 =12.93+0.665J-0.0477G.sub.T N.sub.t -2.61W (7) where J and G.sub.T are expressed in thousands of counts, N.sub.t in millions of counts, and W in kilograms. It will be clear to those skilled in this art that (a) the container 17 need not be of brass and thus need not generate a significant component of the gamma spectra being studied, (b) the rail and bulk sample of the experimental arrangement described above can be substituted by a conveyor belt carrying ore (or other material) between a neutron irradiation station and a downstream gamma monitoring station, to enable on-stream analysis for silicon and aluminium of the material being carried by the belt, and (c) the rail and bulk sample of the experimental arrangement described above can be substituted by the stationary walls and surrounding rock of a borehole, and both the source and detector can be simultaneously moved in the borehole to enable borehole logging for silicon and aluminium. For such an arrangement, the high energy neutron source, the thermal neutron detector and the gamma ray detector will be mounted on a borehole probe, which can then be lowered into a borehole to any required position to analyse the rock surrounding the borehole. Normally the signal processing equipment will not be included on the probe, but will be connected to the source and detectors by long cables.