Table 11. Accuracy of the Ag/PHEN/TBF Method in Aqueous Solution Absorbance against reagent Sample blank at no. %Om@
Silver(1) content Found Certified Difference 1.23% 1.220% +O.Ol% 1 0.282” 7.19% 1.220z -0.03% 2 0.28W a Each absorbance reading in columns 2 and 3 represents the mean of two absorbance readings obtained from two separate solutions. = 550 mp, and all subsequent absorbance measurements of the extracts were made at this wavelength. Calibration Curves and Beer’s Law. By following the recommended procedures, the validity of Beer’s law was examined in both aqueous and organic media. In aqueous medium, it was found that Beer’s law was obeyed over the range 5.39-107.87 micrograms of silver(1) per 50 ml, or approximately 0.1-2.0 ppm of silver(1) in the final concentration with molar absorptivity of 35,000 at 550 mp. I n the case of solvent extraction, it was found that Beer’s law was valid over the range 2.69-53.93 micrograms of silver(1) per 25 ml or approximately 0.1-2.0 ppm of silver(1)
in the final concentration with molar absorptivity of 55,000 at 550 mM. Application of Method. Spectrophotometric analyses were performed o n four solutions obtained from a silver alloy (Analyzed Silver Alloy Let. No. 10, Thorne Smith Co., Michigan). This alloy was analyzed as recommended (23). Two stock solutions were prepared, each containing 50.00 mg of the alloy. The solutions were diluted with doubly distilled water to 250 ml. From each solution a 3-ml aliquot was pipetted into a 50-ml volumetric flask and neutralized with 1M ammonium hydroxide. The sample was then treated according to the recommended procedure. The absorbance measurements were made at 550 mp against a reagent blank. Duplicate solutions were prepared from each stock solution of the alloy. The results obtained are summarized in Table 11.
RECEIVED for review March 29, 1968. Accepted July 18, 1968. Work supported by a grant of the National Research Council of Canada. M.T.E. is grateful for financial support as a Postdoctoral-Fellow from the same grant and a special research grant from Daihousie University. Presented at the 51st conference of the Canadian Institute of Chemistry, Vancouver, June 1968. (23) A. I. Vogel, “Quantitative Inorganic Analysis,” 3rd ed.,
Longmans, London, 1961.
Neutron Capture Gamma-RayActivation Analysis Using Lithium Drifted Germanium Semiconductor Detectors S. M. Lombard and T. L. Isenhour Department of Chemistry, Unicersity of Washington, Seattle, Wash. 98105 The performances of a planar and a coaxial Ge(Li) semiconductor gamma-ray detector are compared with that of a Nal(TI) scintillation detector for neutron capture gamma-ray activation analysis. The superior resolution of these large volume Ge(Li) diodes more than compensates for their lower efficiency because post-irradiation chemical or half-life resolution is not possible in the capture gamma-ray method. Data are given for the efficiency, resolution, and observed background of the detectors; several sample containers are compared; and detection limits for 18 elements in aqueous solution are presented.
CONVENTIONAL activation analysis is a method of elemental analysis in which the sample is irradiated by neutrons, charged particles, or gamma rays produced in nuclear reactors or accelerators, and the resultant delayed radiation is analyzed, often after chemical isolation of the desired products. Neutron capture gamma-ray activation analysis is based on the instantaneous decay by gamma-ray emission of nuclear excited states produced by the capture of thermal neutrons. I n this method, the irradiation of the sample and the measurement of its activity must be done simultaneously. The radiative capture of thermal neutrons was first observed in 1934 by Lea (1) and by Amaldi and his associates (1) D. E. Lea, Nature, 133, 24 (1934).
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ANALYTICAL CHEMISTRY
( 2 ) . It has since been one of the principal sources of information on nuclear energy levels. Initially thermal neutrons were produced by radium-beryllium sources [using the Be(a,n) reaction] imbedded in a moderator such as paraffin. Samples were positioned in the area of highest thermal neutron flux within the moderating material. These sources produced very low thermal neutron fluxes and were of little value for capture gamma-ray studies. The advent, in the early 1940’s, of nuclear reactors made high thermal neutron fluxes available for study of capture gamma reactions. Most capture gamma-ray spectra have been obtained from samples placed in a collimated thermal neutron beam external to the reactor shielding. Thermal neutron fluxes available at these external beam ports are typically of the order of lo6 n/cm* sec. Particle accelerators are also available which produce thermal neutron fluxes of about the same magnitude. The energy of neutron capture gamma radiation ranges from about 75 keV to about 10 MeV and the spectra of most nuclides are fairly complex. The relatively low neutron fluxes available for thermal neutron capture result in rather low sample activity. For these reasons the most important
(2) E. Amaldi, 0. d’Agostino, E. Fermi, B. Pontecorvo, and E. SegrC, Recerca Sc., 2, 461 (1934).
properties of detectors used for capture gamma radiation are efficiency, resolution, and applicable energy range. Table I gives approximate values of these properties for the six types of detectors most commonly used for capture gamma-ray studies. Magnetic pair spectrometers, which transfer the gamma photon’s energy t o electron-positron pairs whose trajectories in a magnetic field are related t o their energy, were the first high resolution detectors used for capture gamma radiation. The efficiency of the magnetic pair spectrometer increases with gamma-ray energy and its practical lower limit is about 3 MeV. The magnetic compton spectrometer analyzes the energy of electrons produced by Compton scattering of the gamma ray in a target. Its useful energy range extends t o well below 1 MeV. Crystal diffraction spectrometers achieve high resolution through the use of quartz crystal monochrometers coupled to a scintillation counter. The upper limit of these detectors is 3-5 MeV and thus they, like the magnetic spectrometers, are not applicable to the entire range of capture gamma-ray energies. All magnetic and diffraction spectrometers have low counting efficiency and therefore require the use of large samples, u p t o about 1 kg, and long counting periods, from several hours to several days, t o obtain the capture gamma-ray spectra of most nuclides. An atlas of capture gamma-ray spectra obtained with high resolution spectrometers has been compiled by Groshev et al. (3) and is currently being revised and updated ( 4 ) . Thallium-doped sodium iodide [NaI(Tl)] scintillation detectors have very high photopeak efficiency for gamma radiation but relatively low resolution. These detectors may be used over the entire range of capture gamma-ray energies. However, because many capture gamma-ray spectra are quite complex, NaI(T1) detectors have not been extensively used for the analysis of capture gamma radiation. Greenwood and Reed ( 5 ) have compiled a n atlas of thermal neutron capture gamma-ray spectra of 74 elements using NaI(T1) detectors. I n 1962 the first lithium drifted germanium [Ge(Li)] semiconductor detectors were fabricated (6). Large volume Ge(Li) diodes presently available combine resolution comparable t o most magnetic and diffraction spectrometers with much greater photopeak efficiency and are free of intrinsic limitations o n their applicable energy range. This paper discusses the advantages of the use of a Ge(Li) diode over a NaI(T1) detector in neutron capture gamma-ray activation analysis and presents the detection limits of the method for 18 elements in aqueous solution. EXPERIMENTAL
An apparatus designed for neutron capture gamma-ray activation analysis has been described in detail (7). A collimator-filter located within the reactor beam port yields a thermal neutron beam with a flux of 1.1 X 10’ n/cm*.sec (3) L. V. Groshev, A. M. Demidov, U. N. Lutsenko, and V. 1. Pelekhov, “Atlas of y-Ray Spectra From Radiative Capture of Thermal Neutrons,” Pergamon Press, London, 1961. (4) G. A. Bartholomew, A. Doveika, K. M. Eastwood, S. Moraro, L. V . Groshev, A. M. Demidov, V. I. Pelekhov, and L. L. Sokolsvski, Nirclear Data, Sectioti A , 3, 367 (1967). ( 5 ) R. C. Greenwood and J. H. Reed, U. S. At. Energy Comm. Rept. IITRI-1193-53 (1965). (6) D. V. Freck and J. Wakefield, Nature, 193, 669 (1962). (7) S. M. Lombard, T. L. Isenhour, P. H. Heintz, G. L. Woodruff, and W. E. Wilson, Znt. J . Appl. Radiation Isotopes, 19, 15 (1968).
Table I. Characteristics of Detectors Used for Capture Gamma Radiation Energy range Detector type (MeV) Efficiency Resolution Magnetic compton 0.3-12 10-5-10-7 2-3 % Magnetic pair 3-11 10-5-10-7 -1 Double flat crystal 0.05-5 10-5-10-’ 0.5% Bent crystal 0-3 10-5-10-7 o.001-1 .o% Scintillation No limits 10-1 5-10 Ge(Li) No limits 10-2-10-4 0.1-1
z
z
Table 11. Characteristics of Gamma-Ray Detectors Used in This Work Ge(Li) NaI(T1) Planar Coaxial Size 3 X 3 inch 7.5 cm2 .X. 11 mm 20cm3 Operating +700 +loo0 - 1800 voltage Pre-amp Current Vac. tube charge FET charge sensitive sensitive sensitive Resolution at 8% 5 keV 3 . 2 keV 662 keV Peak/compton 8/l 3.5/1 17.5/1 at 662 keV ~
~~
and a minimum of fast neutrons and gamma rays. The beam is modulated by a cadmium chopper and the amplified signal from the detector is analyzed by a multichannel analyzer which adds o r subtracts the incoming signal from the spectrum according t o the magnitude of the routing pulses from the chopper control unit. Table I1 lists the characteristics of the three gamma-ray detectors used in this work. The first Ge(Li) semiconductor detector is an Isotopes, Inc., planar diode with a n active volume of 8.3 cm3. Its dewar-cryostat is designed t o place the detector -4 inches above the floor at the end of a horizontal cold finger 12 inches long. The Simtec, Ltd., Model P-10 pre-amplifier is attached to the side of the cryostat by a BNC connector. An Ortec true coaxial Ge(Li) detector was made available on a temporary loan. Its horizontal cryostat is mounted on top of a 25-liter dewar, placing the detector 25 inches above the floor. An Ortec Model l l 8 A pre-amplifier mounted o n the cryostat uses a microdot connector. Both detectors are maintained at liquid nitrogen temperature and -10-6 mm Hg pressure. The main amplifier in this system is an Ortec Model 440 selectable active-filter amplifier. All amplifiers are equipped with pole-zero cancellation circuits (8) for optimum gain stability at high count rates. A high-resolution pulse height logic unit, Model 217B, is used with the T M C 1024 channel analyzer t o accommodate the Ge(Li) detectors. System performance with each detector is shown in Table 11. The memory contents of the multichannel analyzer are transferred instantaneously to a Data Machines, Inc., Model Data 620 computer. Mathematical manipulations and readout of one spectrum are performed while the next spectrum is being recorded. Spectra are plotted in either linear or logarithmic format by a Calcomp x-y plotter and/or printed by a Teletype printer. This arrangement allows maximum use of reactor operating time. The planar Ge(Li) detector is shielded o n four sides by 4 inches of lead. A 2-inch-thick lead brick in front of the de(8) C. H. Nowlin and J. L. Blankenship, Rec. Sci. Instrum., 36,
1830 (1965). VOL. 40, NO. 13, NOVEMBER 1968
1991
.
x = Polyvial
e
e 'e
%
e = Puortz
0
0
a
e = Glass
o = Background
e
e . 6 . 0
CHANNEL NO. Figure 1. Background spectra of three types of sample containers tector contains a tapered hole to allow maximum counting geometry. The lead shield is lined on the inside with a graded shield consisting of 0.050 inch tin and 0.015 inch copper. Thie shield is, of necessity, open at the back to accommodate the dewar-cryostat. The shield for the coaxial Ge(Li) detector is essentially the same configuration except that it is raised 20 inches on a stack of lead bricks to allow for the height of the dewar-cryostat. Polyethylene, glass, and pure quartz have heen considered for use as sample containers. Figure 1 compares the background spectra in the energy range 0-1 MeV produced by sample vials made of these materials. Both borosilicate and soda-lime glasses contain sufficient boron to produce a serious
Figure 2. Experimental apparatus at A. 6. C. D. E. F.
1992
reactor beam
Dewar-cryostat of planar Ge(Li) detector Beam catcher Detector shield Sample vial Chopper Chopper control unit
ANALYTICAL CHEMISTRY
port
interference at 471 keV. The scattering of thermal neuti by polyethylene results in an undesirably high backgro level and reduces the neutron flux at the sample. The I trihution of pure quartz (99.8% Si03 to the backgroun very small and thus it was chosen for use in sample contair The vials are supported in a n aluminum holder boltec the front of the lead shield. Figure 2 shows the experimental apparatus at the rea beam port. The cable from the amplifier extends I feet to the multichannel analyzer located on the ground le
RESULTS AND DISCUSSION Comparison of NaIiTI) and GeWi) Detectors. The supe resolution of the Ge(Li) detectors results in a higher ratis peak height to background and also enables closely ne boring photopeaks to be distinguished. Figure 3 shows combined spectra of Gd(Ey = 79 keV and 88 keV) and (E? = 95 keV) obtained with the planar Ge(Li) detector. In spectra taken with the NaI(Tl) detector the elect] positron annihilation peak at 511 keV is spread over a n en, range from about 450-570 keV. Three of the 18 elem considered produce prominent capture gamma-ray peak this range. The Ge(Li) detector not only reduces the rela intensity of the annihilation peak but also resolves it to keV (FWHM) facilitating analysis of other peaks in energy region. Figure 4 compares a portion of the spectra of 1 ml of ai ous solution containing 1 ppt boron obtained with the I (TI) and the planar Ge(Li) detectors. In the NaI(TI) s trum, the boron peak at 471 keV is not resolved from annihilation peak, making analysis for boron difficult, u in the Ge(Li) spectrum the two peaks are completely resol The Cd(n,y) peak at 558 keV is due to C d used in the grs shield. At the normal operating power of the University' of W ington Nuclear Reactor (100 kW), there exists a radiation
r
Eu
8
BORON
‘io
80 90 C H A N N E L NO.
~
Figure 3. Spectrum of Gd
100
+ Eu
of -20 mr/hr within 1 foot of the sample in all directions. Because of its larger volume, the NaI(T1) detector records a much higher background level than do the Ge(Li) detectors which offsets to some extent the lower photopeak efficiency of the latter. Figure 5 compares the background spectra obtained with the three detectors. Because the background count rate is substantially lower with the Ge(Li) detectors, the “dead time” in the analyzer is reduced, In addition, the statistical error associated with the subtraction of this background from the total spectrum is smaller with the Ge(Li) detectors. When the spectrum was taken with the coaxial Ge(Li) detector, tin had been substituted for cadmium in the graded shield, eliminating the Cd(n,y) peak but resulting in a peak at 595 keV
C H A N N E L NO. Figure 4. Portion of boron spectrum with NaI(T1) and planar Ge(Li) detectors due to thermal neutron interaction in the G e itself. This effect has been substantially reduced by surrounding the lead shield with 0.040-inch sheet Cd. Some detection efficiency is sacrificed with the Ge(Li) detectors, particularly at higher energy. However, of the 20 elements which have the lowest detection limits for neutron capture gamma-ray activation analysis, 19 have intense capture gamma peaks at less than 1 MeV, and 15 of them have peaks below 0.5 MeV. Figure 6 compares the total peak
r I-
ANNIHILATION
cu
Y
0.10
U W
a 1
U
-
I-
O I-
PLANAR I
20 0
600
1000
CHANNEL N O . Figure 5. Background (nothing in sample position) of the three detectors
O.Ol0
‘
’
0.2
I
I
0.4
I
I
0.6
0.8
r.0
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Figure 6. Efficiencies of NaI(T1) and planar Ge(Li) detectors VOL. 40, NO. 13, NOVEMBER 1968
1993
Table 111. Detection Limits of Neutron Capture Gamma-Ray Activation Analysis Detection limit 1 ml H20) Planar Coaxial (pg in
Element
E-t(keV)
1.0 0.5 95 30 20 101 Gd 79 & 180 0.3 0.2 Ho 120 & 140 180 95 320 I 136 160 Tm 155 Ce 184 130 b 500 Se 148, 166, 242 187 17 11 DY a 60 151 & 230 sc Er 186 & 200 44 32 co 279 250 190 369 140 25 Hg 0.9 0.6 Sm 334 & 440 477 1.5 0.9 B a 250 518 Cl 558 4.5 2 Cd 320 Nd 698 a Not available with planar detector because of irreparable damage to detector caused by loss of vacuum in cryostat. * Not available with coaxial detector because of analyzer malfunction during loan of detector.
Eu In
efficiency in this system (product of intrinsic and geometric efficiencies) of the planar Ge(Li) detector and the NaI(T1) detector in the energy range 0-1 MeV. I t is important to note that the decrease in total peak efficiency does not directly relate to the decrease in analytical sensitivity. Greater resolution confines the counts to fewer channels, thereby decreasing the total background counts which interfere. Hence in the case of background interference, peak height ratios are more directly related to sensitivity than peak area ratios. Reference to Figure 4 shows that, because the entire boron peak is recorded in about five channels by the planar Ge(Li) detector us. 30 channels by the NaI(T1) detector, its peak channel count rate is only decreased by a factor of of -3:l (6000 cpm us. 19,000 cpm) instead of the factor of 37:l predicted by the efficiencies. The intrinsic efficiency of the 20 cm3 coaxial detector had not been determined at this writing but it would be substantially higher than that of the planar detector above about 0.3 MeV. Detection Limits. Table 111 lists 18 of the elements calculated by Isenhour and Morrison ( 9 ) to have the lowest detection limits by the capture gamma-ray method. Aqueous solutions of these elements were prepared and 1-ml samples were analyzed for 10 minutes live time with each of the Ge(Li) detectors. The net counts under the photopeak at the gamma-ray energies shown in Table I11 were determined and the detection limits based on the quantity of the element necessary to produce a number of counts equal to twice the standard deviation of the background were calculated. This is equivalent to a one-sided confidence level for detection (9) T. L. Isenhour and G. H. Morrison, ANAL.CHEM., 38, 162
(1966).
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ANALYTICAL CHEMISTRY
of 9 7 . 5 z . A blank sample of 1 ml of distilled water was used to determined the background. A calculation based on the assumption that the neutron beam originates as a point source at the top of the graphite stacked in the reactor beam port shows that the thermal neutron flux at the sample for the coaxial detector (with the top mount cryostat) is 0.66 of that at the sample for the planar detector. At the position of the coaxial detector the neutron beam is more divergent than it is at the lower position, resulting in a higher background due to neutron interactions in the material around the sample and detector. In addition, the background at the coaxial detector is substantially increased by its proximity to the beam catcher, The data shown in Table I11 for the coaxial detector has been corrected for these effects. Because the criterion for the detection limit is a function of the square root of the background, an increase in counting time results in a lower detection limit. The data in Table I11 are the detection limits computed for a counting time of 100 minutes. CONCLUSIONS
The data presented clearly demonstrate the superiority of large-volume Ge(Li) diodes over NaI(T1) detectors for analysis by the capture gamma-ray method of the elements considered. The high resolution of Ge(Li) detectors would, in principle, permit the simultaneous analysis of all of the elements in Table I11 except perhaps Ce and Dy. The detection limits shown in Table I11 are below 1 ppm for four elements, below 100 ppm for seven more, and 500 ppm or less for the remaining seven. The thermal neutron flux in the core of the UWNR is -1012 n/cmZ/sec. There are reactors currently in operation with in-core fluxes of 10’5 n/cm2/sec. When using one of these facilities, the detection limits for neutron capture gamma-ray activation analysis would be considerably lower. This method is both rapid and nondestructive and is adaptable to computer-controlled automated analysis. ACKNOWLEDGMENT
The authors thank W. E. Wilson for programming the on-line computer and W. P. Miller for his assistance in conducting the reactor experiments.
RECEIVED for review July 8, 1968. Accepted August 16, 1968. Research supported by the National Science Foundation. Presented in part at the 155th National Meeting, ACS, San Francisco, April 1968.
Correct ion Use of Powder Covered to Enhance Sensitivity in Detection of Aerosols via Internal Reflection Spectrometry IN THIS ARTICLE by N. J. Harrick [ANAL.CHEM.,40, 1755 (1968)] an error appears in the title. Thelatter should read “Use of Powder Covered Surface to Enhance Sensitivity in Detection of Aerosols via Internal Reflection Spectrometry.”
