Table I.
Compound
a
Retention Time
Relative Retention Time
Retention Vol. per G. of Stationary Phase, Ml./G.
Partition Coefficient 18 5 50 3 77 5 103 135 227 256 367
Methanethiol Ethanethiol 2-Propanethiol 2-Methyl-2-propanethiol 1-Propanethiol 2-Butanethiol 2-Methyl- 1-propanet hi01 1-Butanethiol 1-Pentanethiol
0 0 0 0 0
094 257 396 528 688 146 305 875 14
19 6 53.3 82 1 109 143 238 271 389 1070
1000
2-Thiapropane 2-Thiabutane 3-Thiapentane 2-Methyl-3-thiapentane 3-Thiahexane 5-Methyl-3-thiahexane 3-Thiaheptane
0 306 0 590 1 500
2 3 6 9
63 2 122 311 494 772 1260 2000
59 8 116 294 466 729 1190 1890
Thiacyclopentane
5 42
1120
1060
n-Pentane %-Hexane n-Heptane n-Octane n-Nonane
0 139 0 368 1 OOOa
28.8 76 3 207 562 1530
27 2 72 0 196 530 1440
1 1 1
5
382 722 09 65
2 708 7 36
Standard.
mercaptan standards were relatively pure distillation fractions which mere identified by boiling points and refractive indexes. The ethyl sulfides were prepared from the mercaptans. Other compounds may be employed as gas odorants, but Table I covers only those encountered to date in the particular samples studied. Tests on known mixtures containing a few of the listed compounds indicate that the areas under the chromatograms are proportional to the weight per cent present with a n accuracy of about 5% of the amount present.
*
LITERATURE CITED
(1) Desty, D. H., 3-;ature 179, 241 (1957). (2) Destv. D. H.. IThvman, B. H. F., A X A L . QCHEM. 29,320 (“1957). (3) Hoare, RI. R., Purnell, J. .H, Trans. Faraday SOC.52, 222 (1956). (4) Liberti, A,, Cartoni, G . P., Chim. e ind. (Milan) 39, 821 (1957). ( 5 ) Poiter, P. E., Deal, C. H., Stross, F. H., J . Am. Chem. SOC. 78, 2999 (1956). (6) Ryce, S. A,, Bryce, R. 9.,ASAL. CHEX 29,925 (1957). (7) Sunner, S., Karrman, K. J., Sunden, V., Mikrochim. d c t a 1956, 1144. RECEIVED for review January 8, 1958. -1ccepted May 19, 1958.
Activation Analysis with an Antimony-Beryllium Neutron Source ANlL K. DE1 and W. WAYNE MEINKE Department o f Chemistry, University o f Michigan, Ann Arbor,
-
,The potentialities of using a 1 to 5-curie antimony-beryllium neutron source for activation analysis have been explored. Determinations were made without chemical separation by following the gross decay of the irradiated sample, as well as by gamma spectrum measurements with a manual sweep, and an automatic sweep spectrometer. More than 12 elements have been studied, alone and in mixtures. A simple irradiation chamber coupled with simple manipulations permits routine handling of the source with safety. Sensitivity values for many elements have been calculated for this method, using 5, 1 5 , and 60minute irradiation times. Some experimental sensitivity values are also reported.
A
analysis is a very useful analytical tool for many systems. I n the past decade it has been used by CTIVATION
1 Present address, Jadavpur University, Calcutta, India.
1474
ANALYTICAL CHEMISTRY
Mich.
many investigators with access to nuclear reactors (6, 7 ) . Three escellent recent reviews (8, 3, 17) discuss many of its applications. lleinke and Anderson (8, 9) esplored the possibility of using low-level neutron sources for activation analysis. They used a 25-mg. radium-beryllium neutron source for determination of rhodium, silver, or indium, as well as several rare earth elements in the presence of all other elements. Portable neutron sources have also been used for boron detection in tissues (j),for analysis of boron and iodine samples (C), and for determination of lithium-6, boron-10, and uranium-235 (18). This study deals with the use of a portable antimony-beryllium neutron source for activation analysis in conjunction with several different types of scintillation detectors. Recent nuclear data for the radioisotopes used in this work are given by Sullivan (16). One major advantage of this chart is that it has evaluated the decay schemes for the reader, and gives directly
the percentage of the different beta and gamma rays associated n-ith the decay. APPARATUS, REAGENTS, PROCEDURE
Apparatus. Antimony-beryllium neutron source (maximum activity about 5 curies) and irradiation chamber. Scintillation well counter, NuclearChicago Model DS-3 with 2-inch additional lead shielding, and scaler. Manual sweep scintillation spectrometer, Atoniic Instrument Co., Model 513. -4utomatic sweep scintillation spectrometer, Spectrogammeometer, Radiation Counter Laboratories, Inc., Mark 20, Model 1. K & E Model 4236, manual compensating polar planimeter. Reagents. Ferromanganese, National Bureau of Standards, No. 68-b containing: 79.97% manganese, 12.52% iron, 6.77% carbon, 0.293% phosphorus, 0.006% sulfur, and 0.44% silicon. Pyrolousite ore, Kew Mexico, 30.9% manganese by chemical analysis. Silver nitrate solution, about 0.LV;
For irradiation, the neutron source was transferred from a lead pig t o a hole in the center of half of a 34cm. paraffin cube (Figure 1). The top half of the cube can be slid into place t o complete the chamber. The irradiation assembly was properly shielded with about 2 feet of lead and concrete shielding. With a %curie antimonyberyllium source, the dosage rate just outside the shielding at the front was 5 milliroentgens per hour. Handling with %foot long hooks made it possible for an operator to make as many as 100 transfers a week of this hot gamma source while receiving lrss than tolerance dosage.
k-. e-
RESULTS AND DISCUSSION
e .
-___..
Figure l .
____...-
.. .. ._-. .-
-. _--.-
--
-
Close-up of irradiation chamber
6.inch ruler in ~ i d u r eindicates dimensions
170 grams of reagent grade silver nitrate per liter of mater. All other chemicals used in this work were of C.P. or analyzed reagent grade. Procedure. Individual elements and synthetic solid mixtures of a number of elements, or their appropriate compounds, were irradiated with the antimony-beryllium source for times varying from 8 minutes t o 2 hours, depending upon the nuclear characteristics of the radioisotopes produced. Each sample was then allowed to stand for 3 minutes to allow the decay of any short-lived isotopes. The total gamma radiations of the sample were then measured with a scintillation well counter; or the gamma spectrum was measured with a manual swecp gammaray spectrometer, or a n automatic sweep gamma-ray spectrometer. Known amounts of the pure clement to lic determined n-ere employed as standards for quantitative measurements. Mixtures were also analyzed by a rapid one-step chemical procedure rvhich enriched the desired radioisotope. IRRADIATION METHODS
Antimony-Beryllium Source. T h e neutron sources were obtained from t h e Isotopes Extension of the Oak Ridge National Laboratory. A typical source pictured in t h e Oak Ridge catalog (14) measures about 1 inch in diameter and 1.5 inches high, and consists of a 35-gram cylinder of antimony inserted in a beryllium cup and sealed in an aluminum jacket. High energy gamma rays from antimony-124 produce neutrons by a (7-n)reaction on the beryllium. The useful thermal neutron flux from a 1-cnrir antimony-beryllium source is betwren 10' and 10' nrutrons
cm.? see.-', sufficient to activate a number of elements. The antimony-beryllium capsule from Oak Ridge costs $44; the irradiation charge to produce the antimony-124 is $45 per month. The average strength of the sources used was 2 to 3 curirs, the strongest, ahout 5 curies. Because antimony-124 decays with a half life of 60 days,, the source needs reactivation from time t o time. It was convenient to have two sources, one in the laboratory and the other in the Oak Ridge reactor for recharging the antimony-124. Irradiation Chamber and Techniques. The energetic neutrons from the source must be slowed t o thermal energy before they can become useful for activation. T h e sample was sealed in a soft elass vial or nlastic tube. 11 X 38 m m . r i n prelimiAary experil ments soft glass tubes, 8 and 6 mm. in diameter, nere also used. The sample holder was placed in a slot in p a r a f i slabs about 2.5 em. from the source, where the flux of thermal neutrons was at a maximum. At a shorter diutance, many of the neutrons have not been slowed t o thermal energies; at greater distances the geometry effect of the point sonrce reduces the neutron flux. The source and sample must he completely surrounded by many inches of paraffin, or the neutron flux will not be maximized, Borosilicate glass is not recommended as a container, because many of the neutrons will be absorbed by the boron in the glass. A 5-curie neutron source induced some activity even in soft glass, presumably from sodium and potassium in the glass. For precise work, especially on weak samples, plastic tubes are recommended.
Activation with lon-level neutron sources, such as radium-beryllium, plutonium-beryllium, and antimony-beryllium, has one major advantage over activation with larger sources such as a reactor or accelerator--only isotopes with a high activation cross section and a short half life are detected after short irradiations. The profusion of neutron-induced activities arising in a high flux accelerator or reactor is eliminated. Measurement of Gross Decay of Sample. For these experiments, gamma rays were measured with a scintillation well counter equipped with a thallium-activated sodium iodide crystal, inches in diameter b y 2'/, inches thick; the well measured 2L/82 inch in diameter and 1'/%inches deep. The entire crystal was sealed in 0.20-inch aluminum foil. A l'/&nch Dumont 6292 photomultiplier tube was used with the crystal. The background counting rate for this detector with the extra lead shield was approximately 300 counts per minute. EXPERIMENT1. OPTIMUM SAMPLE SIZE. For orientation purposes, six elements were studied to determine o p timum conditions of sample size, sample container dimensions, and irradiation time for a d v a t i o n analysis: silver, cobalt, manganese, bromine as lead bromide, iodine as lead iodide, and europium as europium oxide. Three containers were used: a soft nlass vial. 38 mm. long X 11 mm. inhiamete; X 0.80 mm. thick; a soft glass tube, 40 mm. long X 8 mm. in diameter X 1.2 mm. thick; and a soft glass tube, 40 mm. long X 6 mm. in diameter X 1.3 mm. thick. Varying amounts of each sample werc used to obtain empirical information as t o the self-absorption of the radiations produced. Figures 2 and 3 are typical plots of thesc measurements that indicate the general pattern of selfabsorption as the glass container size was decreased from 11 t o 6 mm. and the order of magnitude of the activity induced by this source. Each sample, except silver, was irradiated for 30 minutes in the neutron source. Silver . .. was irradiated for f ' ' VOL. 30. NO. 9. SEPTEMBER
7000
I
-
6 mrn. TUGE A 8rnrn. TUBE I I rnm -UBC
I
V n 5 6 ( 2 6 riOCIRS) 6 m m. TUBE I 8 mm. W E E I I mm, V I A L
a
2 6,500cr W
a c
c
6,000-
I 1,500'
02
04 06 08 IO S A M P L E W E I G H T (GRAMS)
I
12
I
Figure 2. Self-absorption of 2.3-minute silver-1 0 8 after 8-minute irradiation and 3-minute decay
Activity of unknown Activity of standard
COBALTICOXIDE
AND
SILVER-LEAD
of these materials mere irradiated and the gross decay of the gamma rays was measured. The resulting two-component decay curves could be resolved into the 2.3-minute silver-108 and 10.7minute cobalt-60 or 25-minute iodine128 components (Figures 4 and 5 ) . B y extrapolating the resolved lines to the end of the irradiation, the initial activity of each component could be determined. The amount of each element could then be calculated by comparing with the initial activity (determined under exactly the same condiTable I,
Analysis of Silver-Cobaltic
- weight of unknown - weight of standard
Plastic containers were used in these experiments. Tubes for each determination and for the standards were filled to approximately the same height, so that self-absorption and counting geometry differences between samples would be small.
EXPERIMENT 2. ANALYSISOF SILVER-
This graphical method of analysis of mixtures gave results accurate to within 5 to 7y0 of the true values as shown in Table I. The total analysis time for the silver-cobaltic oxide mixture was about 11/* hours, while that of the silver-lead iodide was about 21/2 hours. EXPERIMENT 3. DETERhfIKATION OF COPPER OR ALUMINUhf IK PRESEKCE OF OTHER Low CROSS SECTIONELESynthetic mixtures of copper and other elements such as nickel, bismuth, zinc, iron, tin, and lead, often associated with copper, were analyzed for copper with corresponding amounts of pure copper as a standard. Aluminum was similarly estimated in a number of mixtures including iron, nickel, chromium, zinc, silver, magA
~
~
~
~
Oxide and Silver-Lead
.
Iodide Mixtures
(Resolution of gross decay in scintillation well counter) Ag in MixNo. 1 2
3
4
5 0
ture,
Taken, Grams 70 Found, Grams Ag,1.0143 cOz03,0.6780 59.9 Ag,0.969 C0,0.457 Ag, 1.0165 COzO, 1.1460 47.0 Ag,0.954 C0,0.864 24.4 Ag,0.434 CO, 1,067 Ag,0.4600 Coz03,1.4205 Ag,0.5642 PbIz, 1.2030 31.9 Ag,0.523 I,O.619 Ag,0.4700 PbIz, 1.9515 19.4 Ag,0.492 I, 1.007 column 4 - column 2 yo error = x 100. column 2
1476
ANALYTICAL CHEMISTRY
04 06 S A M P L E *EIGHT
I 1 08 10 (GRAMS)
I
I2
I
Figure 3. Self-absorption of 2.6-hour manganese-56 after 30-minute irradiation and 3-minute decay
tions) of a known pure standard of that element :
were then allowed t o decay 3 minutes before measurement. All counts were corrected to the end of irradiation. The curve for the 6-mm. tube in Figure 2 bent up at the 1.1-gram point because some of the sample was outside the scintillation well detector. The depth of the detector well was another important limit on the total sample size, in addition to the selfabsorption of the radiations. Although the smaller tubes gave a higher activity per gram than the larger ones, their limited capacity necessitated the use of 11-mm. tubes for all succeeding measurements.
IODIDE MIXTURES. Known mixtures
I
02
Errors, yo Ag, - 4 . 5
Ag, -6.0 Ag, -7.0 Ag, -7.3 Ag, $4.6
CO, -5.5 CO, + 5 . 0 CO, -6.0 I, -6.4 I, -6.2
nesium, and calcium. Table I1 gives the results of these runs in nhich the sample was irradiated for 30 minutes, cooled for 3 minutes, and counted for 5 minutes. EXPERIMEKT 4, DETERhfINATION O F IODIDE IN MIXTURESAFTER A ONESTEP CHEMICAL COKCENTRATION. Iodide was determined in several mixtures containing sodium iodide and various other salts such as manganous sulfate, cobaltous nitrate, nickel nitrate, barium nitrate, zinc nitrate, yttrium nitrate, and cerous nitrate. After a 1-hour irradiation the sample was transferred to a 250-ml. beaker and dissolved in 100 ml. of water. The iodide was precipitated in the cold by stirring in a slight excess (-5 ml.) of -0.W silver nitrate solution. The suspension was heated to boiling for 1 to 2 minutes, and then filtered on a small Buchner funnel. The precipitate n as first washed with water, then with a few drops of acetone, transferred to a plastic test tube, and counted in the scintillation well counter. Corresponding amounts of pure sodium iodide were employed as standards. The chemical operations required about hour; the total operation including the 1-hour irradiation and counting of the 25-minute 1-128 required 31/2hours. Results are shown in Table 111. Gamma Spectrum Measurement with M a n u a l Sweep Spectrometer.
This spectrometer consisted of a detector, stable high voltage supply, linear amplifier, single-channel pulse height analyzer, and scaler. The detector was a scintillation rrell similar in size t o the well counter used previously for gross decay determinations. Counts were taken at differe n t energies of the gamma-ray spectrum, and t h e spectra drawn manually. The photopeak for each gamma ray occurs a t a pulse height proportional to the energy of the incident gamma ray. Qualitative identification of gamma
IO?
I
-I
I 1 20
1
I
1
I
I
I
40 60 80 DECA" TIME [ W N )
I
I
I
I
IO0
120
Figure 4. Graphical analysis of typical silver and cobaltic oxide mixture after 15-minute irradiation and 3-minute decay
Figure 5. Graphical analysis of typical silver and lead iodide mixture after 15-minute irradiation and 3-minute decay
Scintillation well counter
Scintillation well counter
emitters was made by comparison of the voltage of the photopeak with a calibration curve for the instrument, previously made with standard radioisotopes of known gamma energies, Gamma Spectrum Measurement with Automatic Sweep Spectrometer.
This spectrometer consisted of a detector probe, stable high voltage power supply, linear amplifier, singlechannel pulse height analyzer, linear count-rate meter, strip-chart recorder, timer, and scaler. The detector was a thallium-activated sodium iodide crystal, 13/4 inches in diameter by 11/~ inches long (no well), coupled with a Dumont 6292 photomultiplier tube. An extra shield of lead and concrete bricks minimized the background. The gamma ray spectrum of a sample was automatically scanned and recorded on a chart. Counts under a photopeak could also be totaled on a register. I n quantitative determinations, counts in the photopeaks of the unknown were compared with a known, after appropriate decay corrections had been made. Calibration curves of gamma energy were prepared a t several channel widths (1, 2, and 3 volts), using the same standard radioisotopes described above. EXPERIMENT 5 . DETERMIXATION OF ~IANGAKESE IX MIXTURES.Manganese was determined in mixtures containing manganese or manganese dioxide, and potassium iodide, cobalt, nickel nitrate, zinc oxide, or ferric oxide. The activities due to iodine, cobalt, and the other elements were not large enough to permit accurate measurement. The manganese spectra were measurable, however, after 1- or 2hour irradiations and a 3-minute decay.
Typical spectra for two of the mixtures are given in Figures 6 and 7 ; the manganese-cobalt mixture shows a similar distinct peak for manganese on both gain settings. With a strong neutron source and a sample containing about 1 gram of manganese, a n irradiation period of 1 hour was sufficient. A similar amount
Table 11.
of pure manganese or ferromanganese alloy was used as a standard. After the decay period, portions of the spectra were scanned a t gain 4 and gain 8, with a channel width of 2 volts. The number of disintegrations in the photopeak was recorded, by turning on the count switch when the pen recorder started tracing the peak, and turning
Determination of Copper and Aluminum in Synthetic Mixtures
(Gross gamma count in scintillation well counter) Found, Grams Taken, Grams Weight 70 No. Cu, 1.044 1 Cu, 59.8 Cu, 1,0574 P\'i(N0&.4HzO, 0.7156 Cu, 0.717 Cu, 48.3 2 Cu, 0.6866 Ni(lu'0&.4H~O,0.7378 cu, 1.252 Cu, 51.4 Cu, 1.2485 BizOs, 1.1822 3 4 Cui 0.6420 Bii03; 1.4486 Cu, 0.673 Cu, 30.4 Cu. 1.0036 ZnO. 0,5453 Cu, 1.015 Cu, 64.8 5 6 Cui 0.6345 ZnOi 0 5785 Cu, 52.3 Cu, 0.650 7 Cu, 1.1778 Fe203,0 6130 Cu, 1.151 Cu, 65.8 Cu, 59.8 Cu, 0.859 Cu, 0.8486 Fe203,0.5691 8 9 Cu, 93.6 Cu, 1.044 Cu, 1 0020 Sn02, 0.0676 Cu. 0.6104 SnO?. 0.0815 Cu, 88.2 Cu, 0.637 10 11 Cu, 52.1 Cu, 1.082 Cu: 1.0305 PbO.' 0.9495 12 Cu; 0.5824 PbO; 0 9800 cu, 37.3 Cu, 0.596 Al, 0.3285 13 Fe2Os10 2696 Al, 54.9 Al, 0.333 A41,0.4785 Fe20s, 0 2025 14 Al, 0.488 Al, 70.2 Al, 0.3070 Xi(KOs)z*4HzO, 0.3032 15 Al, 43.2 81, 0.294 Al, 0.3253 Ni(NOs)v4HzO, 0.4443 16 Al, 0.332 AI, 42.2 Al, 0.3705 Cr20a,0 2653 17 AI, 58.2 81, 0.376 41, 0.3645 Cr203,0 2375 18 '41, 54,6 AI, 0 371 Al, 54. 4 ill, 0.3399 ZnO, 0 2846 Al, 0.355 19 Al, 0.2625 ZnO, 0 1975 Al, 0.268 Al, 57.1 20 21 Al, 0.3144 Si02,0 3244 .41, 0.300 AI, 4 9 . 1 22 Al, 0.3754 SiOz, 0 3494 All 0.368 Al, 5 1 , 8 23 Al, 0.348 Al, 72.6 Al, 0.3560 MgO, 0 1341 24 Al, 0.2951 MgO, 0 1778 Al, 0.291 Al, 62.4 Al, 0.3590 CaHP04.2H20, 0.5376 25 -41, 0.353 Al, 40.0 Al, 0.312 41, 46. 1 Al, 0.2984 CaHPOa.2H20, 0.3489 26 column 4 - column 2 Error, /o x 100. column 2 ~
Errorla
%
-1.3 $4.5
+0.3 +4.8 +1.1 +2.1 -2.3 +1.2 +4.0 +4.4 +5.0
+2.2
+1.3
+1.9 -4.3 +2.0 +l.5 $1.6 +4.4 $2.5 -4.5
-1.9 -2.2 -1.4 -1.7 +4.4
~
5
VOL. 30, NO. 9 SEPTEMBER 1958
1477
Table 111.
KO. 1 2 3
NaI, 0 SaI, 0 NaI, 0
4
KaI, 1 NaI, 0
5
6
k
SaI, 0
NaI, 0 SaI, 0
Determination of Iodide in Synthetic Mixtures after Chemical Concentration (Gross gamma count in scintillation n ell counter)
Taken, Grams 81.00 MnS04.H20,1 0999 4206 Co(203), 6H20,0 8315 8530 co(\003)~ 6H20,0 6001 IYI(NOS)~ 6H20,O 3624 0195 Z ~ I ( N O ~0) ~2155 , 7000 Y(so3)3, 0 5079 7987 Y(NO3)3, 0 3795 Ba(N03)?, 0 8795 7834 Ce(S03)2 6H20,0 8396 7774 Ce(NO& 6Hi0, 1 2101
it off when the pen recorder reached the tail of the peak. After background and decay corrections had been made, this number was compared with the corrected number of disintegrations in the photopeak of the standard sample, and the amount of manganese in the mixture calculated. Results of these analyses are shown in Table IT’. These analyses could be completed in ll/zto 2 hours. Also shown in Table IV are the results obtained using the manual m e e p spectrometer. A few synthetic mixtures containing manganese and potassium iodide or cobalt in various ratios were analyzed for manganese ITith this spectrometer. Samples were contained in 11-mm. glass screw-cap vials. They were irradiated for 30 minutes and allowed to stand for 3 minutes before spectral measurements began. Manganese metal (1.1070 grams), potassium iodide (1.7950 grams), and cobalt metal (1.3043 grams) were used as comparative standards. After irradiation, the samples were counted a t different pulse height settings, with a channel width of 1 volt, and relative gains of 4 and 8 for the manganese-potassium iodide mixtures, and 8 and 16 for the manganese-cobalt mixtures. All points of the spectra mere corrected for decay of the isotope concerned, before the best smooth curves were plotted. Figure 8 shows the type of spectra obtained. On this same scale the photopeak for the 10.7-minute cobalt-60 isotope falls a t pulse height values of about 40 and 80 for gains of 8 and 16, respectively. The areas under the photopeaks were measured with a planimeter, and compared with those obtained from a knon-n standard. Weight of element in unknown Weight of element in standard area under photopeak of unknown area under photopeak of standard Where a photopeak lies as a bump 1478
ANALYTICAL CHEMISTRY
Iodine,
%
38 2 30 0 38 1
Activitv, C./M., Corr. to End of Irradiation 2270 1050 2260
IODIDE M I X X S E (GAIN 4 )
Yields of Iodine,
yo
100 3 94 4
103 1.
95 1
57 0 58 0 31 2
2560 1770
2120
10d 3
38 9 31 5
LO50 1890
105 0 98 0
96 1
on another slowly varying curve, it was resolved for this determination. The over-all time for these analyses was 2 t o 211’~ hours, including the time for planimeter measurements. E X P E R I M E K T 6. DETERRIIKATION OF
MANGANESE I N FERROMAKGANESE ALLOY AKD PYROLOUSITE ORE. Five different amounts of the ferromanganese alloy were analyzed, using similar amounts of pure manganese dioxide as standards. Samples w r e irradiated for 1 hour and cooled for 3 minutes, and the spectrum n-as scanned a t two different gain settings as in Experiment 5 , Results are listed in Table T’. Also listed in Table T’ are the results obtained when five different-sized samples of the SBS standard alloy were analyzed by gross gamma count in the scintillation n ell counter. Corresponding amounts of pure manganese dioxide were used under identical conditions as standards. The iron produced no measurable activities under such conditions. A pyrolousite ore sample was similarly analyzed using the two methods. I n a determination using the dutomatic Slveep Spectrometer, trials with 2.0798 and 4.5800 grams of ore gave values of 34.0, 33.2, and 31.5, 32.47,, respectively, for a mean of 32.8% manganese content (us. 30.9 by chemical methods). The ferromanganese alloy served as a standard. Using the scintillation !vel1 counter, two samples of a finply pondered pyrolousite ore were analyzed; the ferromanganese alloy served as a standard. The 2.58-hour manganese-% n-as the only activity found in these samples. Two trials gave values of 30.9 and 32.97,, ivith a mean of 31.97, for the manganese content of the ore. Samples were irradiated for 15 minutes. cooled for 3 minutes, and counted for 5 minutes in this nondestructive analysis, after it had been proved that no activity other than the manganese was detectable. Sensitivity of Activation Analysis w i t h an Antimony-Beryllium Source.
io0 PULSE ? E ; G H T
0
Figure 6. Spectra of manganesepotassium iodide mixture after 60minute irradiation and 3-minute decay Automatic sweep spectrometer
l
MANGANESE-POTASSIUM IOCIDE M!XTURE (GAIN 8 )
100
200
PULSE HEIGHT
Figure 7. Spectra of manganese and potassium iodide mixture after 60minute irradiation and 3-minute decay Automatic sweep spectrometer
SPECTRA OF AND POTASSIUM IODIDE e
A
4
Goin 4 GoNn 8
.-
1-1
Figure 8. Spectra of manganese and potassium iodide mixture after 3 0 minute irradiation and 3-minute decay
_-
+
Lu 4c- v
oc
KL9ii.lRTYS
,
7
iu
-~
Figure 9. Elements calculated to have a thermal neutron activation sensitivity in a well counter of 10 c./min.-mg. or more with 60-minute irradiation in a 1.5-curie Sb-Be source 230
6GC
400 sJ,SE
b E SHT
Sensitivity values (counts mine-’ mg.-’), for isotopes of elements in t h e periodic table t h a t would be activated, have been calculated for irradiation times of 5, 15, and 60 minutes (Table VI). The revised cross section Talues of Hughes and Harvey ( I ) , recently published as a plot of atomic cross section against half life of daughter isotope ( I O ) , were used for these calculations. The laxer sensitivity limit for the table mas arbitrarily taken as -40 counts per minute for a 1-gram sample or about 12% of the 300 counts per minute background of the scintillation detector. The underline and wavy lines in Table V I are used to indicate isomeric states in the radioactive products ( I O ) . Table IV.
Table V.
(National Bureau of Standards 68-b: Mn, 79.97%) Gamma Spectrum Measuremrsiit n ith Automatic Sweep Scintillation Gross Gamma Count in Scintillation Well Counter Spectrometer FerroFerro70 Mn, manganese lnanganese Found, Errorja % taken, & Found. Error,(i taken, Gain Gain 10. grams Grams yo % grams 4 8 4 8 1 0.2321 0.180 77.3 -3.3 1.5837 83.5 77.1 +4.4 - 3 . 6 78.9 82.0 -1.3 $ 2 . 5 2 0.4038 2,1032 0.327 80.9 $ 1 . 0 3.1167 82.4 78.2 $ 3 . 0 -2.3 3 0.5730 0.446 77.8 -2.8 4.7291 77.1 76.6 - 3 . 6 -4.2 4 1.0040 0.777 77.3 -3.4 5 1.5837 4.7291 76.6 77.2 - 1 . 2 $ 3 . 5 1 274 8 0 . 5 + 0 . 8 yoM n found - 79.97 x 100. a Error, 70= 79.97
Determination of Manganese in Mixtures with Gamma Spectrum Measurement
Taken, Grams
7 0
Error,a %
Found, Grams
Xln in Mixture, No.
Determination of Manganese in Ferromanganese Alloy
Gain 4 Mn
Gain 8 I
;\In
Gain 4 I
RIn
--.L
I
(XI,
8-- __
I
hLIn
M A K U A L S W E E P SCliYTILLATIOS SPECTROMETER
1
2 3 4
Mn, 1.3320KI,1.6520 Mn, 0.5220K1, 1,3500
Mn, Mn; Mn, Mn,
0.3720KI,2.7510 0.0948KI; 1.6455 1.0248 Co, 1,0267 0.1132C0, 1.4139
44.6
1.294
27.9 11.3
0.526 0.360
5.4 50.
1:oio
1.188
1.370 2.060 1.220
1 270 0.405
0.350
1.310 1.280
2.060
-2.9 $1.0 -4.1 ...
-3.6
+l.5 -1.9 -1.9
-4.9 -5 1
$4.8
-5.2 -1.9 -4.8 CO, + 2 . 9 CO, t2.8
-58
1.320 .., c o , 1.060 - 2 . 0 -3.8 7.4 co, 1.430 , . , c o , I ,450 , , , co, + i : 4 ... SUTOMAT~C SWEEPSCISTILLATION SPECTROMETERC Mn Mn in Found, Grams Error,= 56 1-0. Taken, Grams mixture, Gain 4 Gain 8 Gain 4 Gain 8 1 lln,0.5507 KI, 2,0268 21.4 0.585 0,520 $6.2 -5 8 2 Mn, 1,3320 XI, 1.6520 44.4 1.393 1.380 +4.5 +3 5 3 M n , 2 0968 III, 1 0291 67.1 2.160 2,084 $2.9 -0.5 4 >In, 4.0332 KI, 1.1512 -3.3 +2.4 81.8 4 769 5,049 5 M n ,1,2230 Co, 0,9995 55.0 1,276 1.189 +4.2 -2.8 6 A h , 1.0238 Co, 2.9830 25.6 0,968 0,982 -5 5 -4.0 7 ;\In02, 1.1410 hTi(n’O&.6H20,2.2915 21.0 0 761 0.749 +5.5 t8.8 8 ?vIn02,0 . 5795 Ni(NO&.6HsO, 2.1399 13.5 0.350 0,384 $4.5 t4.9 9 R‘InOs,2,0455 ZnO, 0.8400 44.5 1 268 1 ,248 -2 0 - 3 . .i 10 MnOp,1.3931 ZnO, 0.9764 37.2 0 019 0.922 +44 +.18 11 Mn02, 1.8270 Fe203,1.1468 38 8 1 15.5 1,109 +3 2 -4 0 12 Mn02, 1.0545Fe20s,1.5185 25 8 0 697 ... +4.6 column 4 - column 2 a Error, 70= x 100. column 2 Data for cobalt obtained a t gain 8 and gain 16 instead of gain 4 and gain 8 as in other runs (Sos. 1-5). The areas under photopeaks due to Mn-56 in runs 4 and 6 were too small to permit any measurement. Comparative standards used for h’os. 1-3, 5, and 6 were 2.1351 grams of M n > for N o . 4, 5,0011 gram of %In, and for Kos. 7-12, 2.1032 grams of NBS ferromanganese alloy (68-b). Times of irradiation were 1 hour for Nos. 4,7-11, and 2 hours for Kos. 1-3, 5, 6, and 5 6b
co,
..,
o:&o
I . .
12.
VOL. 30, NO. 9, SEPTEMBER 1958
1479
Sensitivity of Activation Analysis with -1.5-Curie
Antimony-Beryllium Neutron Source and Scintillation Well Counter"
Atomic Cross Section, Barns 0.009 0.56 0.0056 0.21 0.138 0.56 0.069 0.0020 10 22 0.0075 4.5 0.0088 13.4 16 0.030 2.69 0.56 0.19 0.84 1.35 0.165 4.2 0.63 0.015 0.3 0,0046 1.47 4.3 1.73 0.057 0.0104 0.033 0.13 1.2 1.0 0,019 0.13 12 140 3.2
Emtl. Sensiiivity, Calculated Sensitivity, C./Min.-Mg. 60 Min. h a d . , 5-Min. irrad. 15-Min. irrad. 60-Min. irrad. C./Min.-Mg. 0.085 0.086 0.094 ' 0,021 0.25 0.062 0.267 0.016 0.036 0,059 2.21 1.99 1.56 0.322 0.119 0.97 0.99 0.483 0.169 1.87 0,009 0.003 0.030 0.040 0.0136 0.006 0.021 94.9 105.2 95.8 0.006 0.018 0.079 0.060 0.032 0.079 25.6 40.4 47.2 0.080 0.052 0.093 8,445 2.815 33.78 33.78 3.76 43.2 97.0 169.1 0,019 0.006 0.075 0.349 0.115 1.50 4.66 2.61 5.89 0.328 0.118 1.095 0.93 1.255 3.31 7.68 0.158 0.680 0.052 82 0.189 0.693 0.065 1620 1.123 0.258 0.97 0.085 0.3 6.03 5.95 57 0.024 0.0086 0.320 17 1.35 0.528 1.11 0.044 0.042 0.048 276 0.531 0.175 18 18 31 1.575 7.08 2150 0.026 0.080 0.349 264 0,090 0.007 0.022 77 0.004 0.012 0.0456 17.8 0.055 0.142 0.348 0.075 0.292 168 0.223 0.026 3780 0,010 0.032 0.139 6.6 3.86 7.55 10.53 14.3 0.189 0.039 0.094 270 0.201 0.192 0.016 0.048 1126.3 61.3 4.5 103.8 1472 0.73 1314 1336 0.290 816 0.129 0.391 22 0.056 0.142 0.04 2.3 22.6 214 238 167.4 0.40 563 53.5 513 506 49 0,025 0,017 0,046 0.172 3180 0,010 0.32 0.003 174 0.238) 0.062 0.11 0.021 0,893 0.811 0.085 0.787 1.2 Single underline. Neutron activation produces a metastable daughter activity. Two underlines. Two metastable states. Wavy line. Activation to the ground state of an isomeric pair. Underline and wavy line. Total cross-section values for formation of the ground state both directly and by decay of the shortlived metastable state. Daughter Half Life, Min. 0.18 900 9.5 2.3 37.5 109 744 8.5 0.33 1.22 x 10' 5.8 3.76 3.6 155 10.4 154.2 768 5.14 52 20.2 852
The calculated sensitivity values for each irradiation time were normalized to the experimentally determined values for manganese. Each sensitivity value is given in terms of counts per minute in the xell counter. For each element two normalizing factors were considered: a factor comparing the half life and atomic cross section of the isotope to that of manganese, and a factor for the decay during the irradiation of some of the isotope formed. Manganese was taken as the reference standard for the calculations because it has a relatively large atomic cross section with small error (13.4 f 0.3 barns), a reasonable half life (2.6 hours), and a good counting efficiency for its gamma rays 1480
ANALYTICAL CHEMISTRY
(assumed to be 100% for these calculations). For the 60-minute period many of the calculated sensitivities have also been compared with experimentally determined sensitivities. About 1 gram of an element in a 40 b y 12 mm. plastic tube was irradiated for 1hour and cooled for 3 minutes, and a gross gamma count in the scintillation well counter taken for a t least 10 minutes. These experimental values are also given in Table VI. There are several reasons why the calculated sensitivity values are always higher than the experimental values. When the product isotopes emit primarily beta particles, or gamma rays
which are highly converted, there is considerable absorption of the radiations in the sample. Similarly, lowenergy gamma rays may be appreciably absorbed in the sample and sample container. The counting efficiency of energetic gamma rays, however, is almost 100% as shown in the case of sodium, chlorine, and arsenic. These calculated sensitivity values can be summarized in graphs similar to Figure 9 for a given irradiation time. Table VI or these graphs show all elements vhich might interfere with gross gamma-ray measurements in the determination of one element t o a particular sensitivity without chemical separation. For example, alumi-
Sensitivity of Activation Analysis with -1.5-Curie Antimony-Bervllium Neutron Source and Scintillation Well Counter" (Continued!
Daughter Half Life, Min, 54.1 0.22 40 10 4030 1.3 21
558 72 25 25 3.9 192 85 2400 2040 1152 108 2820 24 552 1080 3.6 1.05 x 105 1.3 139 1638 450 1.85 X lo6 4 . 6 x 104 6060 108 222 9800 16.4 1 . 6 X lo5
1.4 1 . 1 x 105 1140 6190 1080 31 1620 5.5
42 23 5
num-magnesium-silicon alloys containing as little as 5 to 10% aluminum can be analyzed for aluminum with a 5-minute irradiation with little interference, while the same type of alloys containing manganese or copper require resolution of decay curves. Similarly, one could expect in 5-minute irradiations to measure lY0 amounts of vanadium in stainless steels containing iron, chromium, and nickel, with manganese causing interferences. These sensitivity values may be increased or decreased as the strength of the neutron source is varied, but their relative position will always remain the same. It has not yet been possible to evaluate the decay schemes of each
Atomic Cross Section, Barns 139 50 0.0075 0.012 3.89 0.013 0.013 0.15 0.041 0.076 5.5 0.013 0.017 0.36 8.4 0.11 10 0.21 37 1.24 670 1.0 0.18 >22 733 1000 60 1.3 130 15.4 19 0.8 34 104 0.03 19 9.7 37. I 47.2 2.1 0.66 100 370 80 0.7 0.28 0.28 96 0.029 0.07 2.78
Calculated Sensitivity, C./Min.-Mg. 5-min. irrad. 15-min. irrad. 60-min. irrad. 82.8 476 0.006 0.033 0.032 0.107 0,019 0.009 0,019 0.094 6.82 0.073 0.003 0.137 0.115 0.002 0.285 0.064 0.431 1.603 39.9 0.030 1.054 0.007 6390 236.6 1.205 0.095 0.023 0.011
0.103 0,244 5.03 0.349 0.055 0.004 0.221 0.221 1.437 0.003 0.012 870 0.111 2.31 0,004 0,009 0,297 1,949 0.129 0.373 3.67
227.0 480 0.016 0.074 0.096 0.115 0.049 0.027 0.053 0.266 19.22 0.116 0.009 0.380 0.350 0.005 0.865 0.186 1.307 4.52 120.9 0.092 1.655 0.021 7020 704.2 3.65 0.288 0.070 0.033 0.312 0.710 15.1 1.058 0.135 0.012 0.671 0.670 4.35 0.009 0.035 959 0.335 6.99 0.011 0.026 0.783 5.90 0.237 0.615 9.67
isotope as to the actual sensitivity to be expected. Table VI shows the maximum sensitivities obtainable, and the maximum number of interfering elements in a determination. It should be emphasized, however, that factors such as self-absorption, etc., may reduce the number of interfering elements in any determination. Short half-lived radioisotopes have been included in Table VI for completeness, although they tvould require special techniques for full utilization. They can be removed as interferences by allowing the sample to decay for 8 to 10 half lives. Gamma-ray energy discrimination, such as utilized in the spectrom-
ExDtl. Sensiiivity, 60 Min. Irrad., C./Min.-Mg.
786 526
::E}
0.422 0.126 0.118 0.115 0.189 0.646 46.7 0.137 0.035 1.471 1.52 0.023 3.76 0.709 5.68 I 10.97 1 521 0.401 1.886 0.092 7710 2774 15.87 1.239 0,307 0.146 1.372 2.51 60.9 4.646 0.290
1
0.020
0.394 7.00 0.290 1.080 1.060 0.600 2.85 93.8
0.052
2.92 2.941 18 93 0.040 0.154 1052 1.472 0,049 0.112 2.13 25.9 0,305 0.737 24.3
0.460
eter experiments, might also eliminate many interferences. Reproducibility. The precision and accuracy of measurements made with the automatic-srveep equipment were determined by analyzing the same sample of ferromanganese alloy ten times a t gain 4 and a t gain 8. The replicate analyses were made a t intervals of a t least 24 hours, to allow the activity of manganese-56 to decay to background. A value of 77.6% with a standard deviation of d = l . O % rcsulted from the 10 runs, compared with a value of 79.97 i 0.02% reported by the National Bureau of Standards. To check the reproducibility of the gross gamma count technique, six repVOL. 30, NO. 9, SEPTEMBER 1958
* 1481
licate measurements were made on another sample of the standard alloy. A mean of 78.5 5 0.7% was obtained. General Evaluation. This nondestructive method of analysis gives results accurate t o within a t least 5 t o 6% for all three techniques. T h e sample size was limited t o about 2 to 3 grams in the gross-decay and manualsweep-spectrometer methods, but could be increased to about 6 grams in the automatic scan technique. Because the precision of the measurement is a direct function of the total number of counts accumulated, it is advantageous to use the maximum sample size for analysis. Of the t\Vo typical spectrometer assemblies, the automatic recording instrument has better resolving power (about loyo, since a well crystal was not used), shorter scan time, and has the additional advantage of automatic recording and measurement, but it is more costly ( ~ $ 5 5 0 0 )than the other spectronieter ( ~ $ 3 1 0 0 ) . The 60-day half life and the highenergy ganiina rays of the antimony are definite disadvantages in using this type of neutron source for activation analysis, although its relative
cheapness and unrestricted availability tend t o compensate. Other neutron sources without these disadvantages such as polonium-beryllium ( l l ) , actinium-beryllium ( I S ) , plutonium-beryllium (15), and californium-252 spontaneous fission ( l a ) , may soon help to answer the demand for portable sources of higher neutron flux for use in routine analysis. ACKNOWLEDGMENT
The authors wish to thank the Michigan Memorial Phoenix Project for its generous support of this work. They also acknowledge the help of I. B. Ackermann in designing the irradiation chamber, and in making preliminary measurements. LITERATURE CITED
(1) Hughes, D. J., Harvey, J. A., U. S.
Atomic Energy Commission, Rept. BNL-325 (July 1955). (2) Jenkins, E. I., Smales, A. A., Quart. Rw. (London) 10, 83-107 (1956). (3) Loveridge, B. A,, Smales, A. .4., “Activation Analysis and Its Application to Biochemistry,” Methods of Biochemical Analysis, Vol. 5 , pp. 225-72, Interscience, Sew York, 1957.
(4) Mayr, G., Nucleonics 12, No. 5, 58-60 (1934).
(5)’ Miyr, G., Brunner, H. D Brucer, M., Ibid., 11, No. 10, 21-5 (1953). (6) Meinke, W. W.,AKAL. CHEY. A ,
736-56 (1956). (7) Ibid., 30, 686 (1958). (8) M$nke, W.K., ilnderson, R. E., Ibid., 25, 178-83 (1953). (9) Ibtd., 26, 907-9 (1954). (10) Meinke, W. K.,Maddock, R. S., Ibid., 29, 1171-4 (1957). (11) Mound Laboratory, U. S. Atomic Energy Commission, Rept. TID-5087 (July 1952). (12) Sucleonzcs 14, NO. 6 , 105 (1956). (13) Ibzd., 15, NO. 9, 192-3 (1957). (14) Oak Ridge Katiorfd Laboratory, Oak Ridge, Tenn., Radio-isotopes, Special Xaterials and Services,” Cataldg, 1957. (15) Sheldon, J., Williams, J., Brit. -4tomic Energy Research Establishment, Rept. AERE-M/M-80 (August 1954). (16) Sullivan, \I-. H., “Trilinear Chart of Kuclides,” U.S. Government Printing Offire, TTashington 25> D. C., January 1957. (1;) Tavlor, T. I., Havens, K. R.,Jr., “Berl’s Phvsical Methods in Chemical .4nalysis,” Vol. 3, -pp. 539-601, Academic Press, Sew E ork, 1956. (18) Wanke, H Monse, E. U., 2. Naturforsch. loa, 6$7-9 (1955). RECEIVED for review January 28, 1958. Accepted April 7, 1958.
Chemical Oxygen Demand of Petrochemical Wastes Modification of the Standard Catalytic Reflux Pro’cedure F. W.
BERTRAM,
0.T.
CARLISLE, J. E. MURRAY, and G. W. WARREN‘
Union Carbide Chemicals Co., Texas City, Tex.
C. H. CONNELL University of Texas Medical Branch, Galveston, Tex.
,A modification eliminating chloride interference in the determination of chemical oxygen demand by the standard dichromate reflux procedure with silver sulfate catalyst is presented. The conventional procedure is altered, so that chlorides are first quantitatively oxidized by refluxing the sample in the absence of silver sulfate catalyst. The catalyst is then added, and refluxing is continued to complete the oxidation. An inorganic chloride correction, based on an independent determination, can then be accurately applied to the total chemical oxygen demand. Applications of the modified procedure are evaluated, showing that all the advantages offered b y the standard dichromate catalytic reflux method plus an elimination of interferences due to inorganic chlorides may be expected. 1482
ANALYTICAL CHEMISTRY
A
of industrial wastes have far-reaching significance in monitoring water pollution and efficiency of operations. I n working with waste water streams from petrochemical industries, the test results of most general interest are usually the chemical oxygen demand (C.O.D.) and the biochemical oxygen demand (B.O.D.) ( 2 ) . A determination of the waste’s C.O.D. is made when a quick estimate of the strength of industrial wastes is desired. Such a determination does not differentiate between biochemically stable and unstable components, and may not be directly correlated with B.O.D. but instead represents a more nearly accurate measurement of the contaminants present. Most of the procedures originally proposed for C.O.D. used permanganate, chromate, or iodate as the NALYSES
oxidizing agent, none of which gave complete oxidation or consistent results except with a few specific wastes. A dichromate reflux procedure (4) introduced in 1949 gave a high degree of oxidation of certain types of compounds, but was ineffective in oxidizing completely the organic aliphatic aoids. and numerous other compounds. The use of silver catalyst with the dichromate reflux method Ivas investigated and in 1951 results were published ( 5 ) , which showed that nearly complete oxidation was accomplished with most compounds including aliphatic acids. The presence of chlorides in appreciable quantities caused lorn and unpredictable results, and no remedy was proposed. This procedure was, 1 Present address, Union Carbide Chemicals Co , South Charleston, W. Va.
