Thus in analyzing samples, both extracts should be spotted. I n the basic extract of urine from two individuals who had each ingested 100 mg. of diphenhydramine and tripelennamine, respectively, a second Dragendorff positive compound, probably a metabolite, was found in addit'ion to the drug. Technique and Reproducibility of R,. Silica gel coated plates have previously been prepared with solutions of sodium hydroxide, oxalic :tcid, and buffers (3). I n this study potassium bisulfate was preferable to the use of other acidic materials either in the adsorbent or in the solvent. The potassium bisulfate coatings gave almost no tailing of spots nor significant flaking of the coating off the plates. The other acidic systems tried--e.g., either acetic or hydrochloric acid in the solvents, or oxalic or phosphoric acid in the coatings-produced either tailing of the spots, a coating which flaked off the 1)lates readily, or both. The developing t,anks cont'aining methanol were used satisfactorily for 25 plates; replenishment of t,he methanol lost during the development was all that was required to attain reproducible R , values. Development of additional plates in the same methanol was not attempted but would appear to be
feasible. Solvent I had to be changed after every five plates since Rj 1-alues were lower on subsequent plates. This change in R, is probably caused by selective evaporation of diethylamine. Because the R j values of many of the antihistamines are similar in each system, the use of an Rj value as a n aid in identification was dependent on its reproducibility. The most crucial factor in attaining reproducibility was securing a constant degree of saturation in the developing tanks prior to inserting the plates. The described procedure has proved satisfactory in this respect. All R j values reported in Table I were obtained on plates conditioned and developed as described in the procedure. Plates oven-dried a t 120" C. for 2 hours and stored over fresh calcium chloride until used have given the same R j values. Plates stored in the laboratory over a relative humidity range of 40 to 55y0 and developed a t temperatures from 22" to 28" C. have not produced Rr values outside of the range found using the more rigid conditions. Plates conditioned a t room temperature and 80% relative humidity did give R j values in both systems I and I1 averaging 0.05 to 0.08 unit higher than those reported in Table I. Ordinary laboratory conditions would thus appear
to suffice for the attainment of reproducible R, values. The ranges of R/ values obtained from the five plates used for the data in Table I were compared with those from five plates developed in unsaturated tanks. I n system I, 22 of the 25 drugs showed ranges of less than 0.07 Rj unit under standard conditions whereas when development was in unsaturated tanks only nine drugs had ranges of less than 0.07 unit. Similar patterns were noted for systems I1 and 111. Relative Rj values partially compensate for this loss of precision but the range distribution indicates that saturated tanks are also desirable when identification is made using relative R j values. LITERATURE CITED
(1) Cochin, J., Daly, J. W., J . Pharm.
Expl. Therap. 139, 160 (1963). (2) Goldbaum, L. K., Kazyak, L., ANAL. CHEM.28, 1289 (1956). ( 3 ) Stahl, E., Arch. Pharm. 292, 411 (1959). ( 4 ) Walker, K. C., Beroza, AI., J . Assoc. O$c. Ogr. Chemists 46, 250 (1963).
RECEIVEDfor review May 3, 1964. Accepted November 19, 1964. Study supported by funds from grant R. G. 9863 of the National Institutes of Health, U. S. Public Health Service.
Expe rimentu I Reactor Therm a I - Ne utro n Activation Analysis Sensitivities HERBERT P. YULE John Jay Hopkins laboratory for Pure and Applied Science, General Atomic DivisionlGeneral Dynamics Corp. P . 0. Box 608, San Diego, Calif. 927 72
b Gamma-ray photopeak yields have been experimentally determined for 1 18 reactor thermal-neutron products of all elements from oxygen through lead (except Ne, Kr, and Xe). limits of detection (sensitivities) have been estimated from these data; the median sensitivity is about p.p.m. for a 10-gram sample. The use of reactor thermal neutrons for these measurements has several advantages. Because the yield data have been experimentally determined, they are not subject to inaccuracies found in theoretical estimates caused by poorly known nuclear parameters. The high thermal-neutron flux available in the reactor has made possible the observation of a number of additional activation products not included in previously published lists containing experimentally determined thermalneutron product yields. These additional products offer alternative means of analysis which may b e useful in circumventing matrix activation problems. The present list contains data
on essentially all important thermalneutron products for those elements amenable to neutron activation analysis utilizing gamma-ray counting.
F
of the power of the activation analysis technique is dependent (among other aspects) on a knowledge of the limits of detection for the chemical elements. For most of the elements determinable by neutron activation, the most sensitive method is detection of the radioactivities generated by thermal neutron capture. This paper reports, for many elements, limits of detection by gamma ray counting which have been experimentally determined with reactor thermal neutrons. I n principle, these limits of detection are easily calculable, and a number of tabulations of such sensitivities have been published (3, 7 , 8,10, 11, 13, 1 5 ) . Such calculations depend on experimental parameters of individual nuclides, ULL REALIZATION
and because these pai.ameters are not always accurately known, calculated limits of detection may be in error by more than an order of magnitude as shown by the data in references ( 7 , 15). Experimentally determined limits of detection have also been published ( I , 7 , 9 , f 5 ) ; most of these sensitivities have been determined using thermalized 14-m.e.v. neutrons with fluxes of about lo8 neutrons/sq. cm.-second. I n the present work, the neutron source was the General Atomic Triga Mark I reactor, and the thermal neutron flux was 4.3 X lo1*neutrons/sq. cm.-second. Several advantages are obtained with this much more intense flux: limits of detection are much lower (more sensitive) ; many additional activation products have been observed, including long-lived products not readily available with irradiations with 14-m.e.v. neutron generators, and these additional products offer alternative means of analysis in the event that interferences from activation of the matrix prevent analysis with shorter-lived activities. VOL. 37, NO. 1, JANUARY 1965
129
~
~~~~~~
Table 1.
Triga Mark I Neutron Fluxes'
Neutron flux, neutrons/sq. cm.-second 4 3 x 10'2 3 5 x 10'2 7 5 x 10" 1 2 x 10" 1 9 x 10'0
Neutron energy Thermal >10 k.e.v. >1 35 m.e.v. >3 68 m.e.v. >6 1 m.e.v. Taken from reference ( 2 ) .
Table 11.
Product Element
0 F Na
2 For a number of elements, thermal neutron activation limits of detection are not so sensitive as those available by reactor fast neutron activation (6, 15). Studies of the utilization of reactor fast neutrons for activation analysis are nearing completion, and it is planned to publish these results in the near future. EXPERIMENTAL
Samples were irradiated in the F-ring (outermost ring) of fuel elements of the General Atomic Triga Mark I reactor; flux data are given in Table I. Samples were encapsulated in 1/2-dram polyvials (obtained from Olympic Plastics Corp., Los Angeles, Calif.), and transported in and out of the reactor in polyethylene "rabbits" by pneumatic tube. Spectroscopically pure materials were generally irradiated in the form of the element or its oxide. Selection of sample size, based on calculations of Lukens(l6), was aimed a t producing lo4to 105-photopeak c.p.m. in a standard irradiation of 15 seconds, and typical sample sizes were 1 to 10 mg. Following irradiation] sample vials were removed from the rabbit, and counted with a 3-inch X 3-inch solid NaI(T1) detector coupled with a commercial 400-channel multichannel analyzer (Radiation Instrument Development Laboratories, Models 34-12 and 34-1213). A '/2-inch thick polystyrene beta-particle absorber was placed directly on the crystal, and the sample vial was positioned in the center of the absorber. The geometry of this counting arrangement is 31y0 for small samples and slightly less for larger samples. It was possible, in most cases, to choose sample activities to be much larger than those from impurities in the polyvial and the argon content of the air in the polyvials (about 1600 c.p.m. in the 1.29-m.e.v. photopeak at the end of a 15-second irradiation). When interferences of this sort occurred, the sample was transferred to a "cold" polyvial, allowing Ar41 activity to escape in a hood. Isotope identification was obtained from a knowledge of the sample element, the gamma-ray spectra, decay data, and literature data for the isotopes (3, 5, 14). Calculation of Limits of Detection. T h e first step in the calculation is to compute, for a prominent gamma ray, the number of counts in the photopeak above t h e Compton "background." This computation is described in a recent article by Soltys
130
0
ANALYTICAL CHEMISTRY
Si P S
c1
Ar
K Ca
sc
Ti V Cr IMn Fe co
Ni cu Zn Ga Ge As Se
Br Rb Sr
Y Zr
Kb Mo Ru
Rh Pd Ag Cd
Half life* 29 s 11 8
15h 9.5 m 2.3m 2.62 h 5.1 m 37.3 m 1 . 8 3h 12.5h 8.8 m 20 s
84 d 5.8m 3.76 m 27.8 d 2.58 h 45.1 d 5.24 y 1 0 . 5m 2.56 h 1 2 . 8h 5.1 m 13.8 h 245 d 1 4 . 3h 1.4h 48 8 LlOd 17 s 61 m 18.2 m 120 d 1.50 d 4.4 h 17.8 m 1.0m 1 9 . 5d 2.8 h 70 m 3.19 h 17 h
Isotope 0 1 9
F% NaZ4 Mg2'
A128
Si31 537
c 1 3 8
Ar41 K42 Ca49 Sc4'm
Sc460 Ti5l v52
Cr5l &In56 Fe59 Co60 comm
Ni65 cue4 CU66 Zn6gm Znb5 Ga72
Ge75 Ge75m As76 Se77m Seslm Sea' Se75 Br82
65 d
Brso Rbs8 Rbsbm R bs6Q Sr87m Sr8Sm Y9om Zr97 Nb97 Zr 95
35 d 6.6 m 14.6 m 66 h 4.5h 39.8 d 2.88d 36 h 4.4 m
Nb95 Nb94 1Mo'O' Mog9 Ru105 Ru103 Rug7 RhlO5u Rh104m
4.8m 13.6 h 21 9 249 d 24 s 2.3 m 49 m 54 h
pdlO9m
+
pdlogu
pd107m Agllom Ag11OQ Agios
Cdlllm Cd115~
Gamma ray energy, m.e.v. 0.20 1.63 1.37 0.84 1.78 1.26 No y 3.1 1.64 1.29 1.53 3.1 0.1_40 0.89 0.32 1.44 0.32 0.84 1.09 1.17 0.059 1.49 0.51 1.04 0.44 1.12 0.834 0.264 0.14 0 555 0 160 0 104 0.28 0.265 o 55 0 63 0 62 1 8 0 56 1.08 0.388
+
0.. 2~~. 23
0,203 0.750 0.666
+
0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0
768 87 191 141 72 498 216 31 051 556 19
088 21 656 656 630 24 335
+\J
Experimental Reactor Thermal-Neutron
Yield," photopeak c.p.m. per gram of element
Limit of detection,
3.3 x 1.8x 3.4 x 7.1 X 2.0 x 3.3 x
3000 0.55 0.0029 0.14 0,0049 30
105 109 109 lo8
10'0 105
fig.
1 . 4 X 106 3 . 2 x 109 2 . 4 x 101oc 1.1 x 108 6 . 7 x 107 8 . 1 X 10" 8 . 7 X 108 2 . 5 x 109 2 . 6 X 10" 2 . 2 x 107 2 . 8 X 10'' 9 . 9 x 104 3 . 6 x 107 3 . 2 X 1O1O 5 . 6 x 107 9 . 7 x 109 1 . 5 X lo8 1 . 2 x 108 1 . 9 x 10' 8 . 8 X log 3 . 4 x 108 9 . 2 X 109 6 8 X lo9 7 8 x 1010 4 7 x 108 4 . 4 x 108 2 . 4 x 107 2 1 x 109
70 0.031 0 . 00042c 0.094 1.5 0,0012 0 011 0.040 0.00038 0.45 0,000036 100 0.27 0.0032 0.180 0.0010 0.020 0.083 5.4 0.0011 0.029
1 4 x 1010' 4 2 x 108 9 o x 109 5 . 1 X 10' 8 . 0 x 109 1 . 8 x 108
0. 0O07Oe 0.24 0.011 2.0 0,0013 0.054 0.24 0,081,
x 105' 3 . 6 X lo8 1 . 1 x 109 2 . 1 x 108 8 . 7 X lo8 2 . 2 x 107 4 . 2 x 107 9 . 9 x 107 9 . 7 x 10'0 2 . 0 x 10'0 6 . 3 x 109 5 , 4 x 108 6 . 7 x 109 1 . 9 x 107
83,
4 . 2 X io7 1 . 2 X lo8/
1.2
2.3 X 2.9 X 2.6 X 2.6 X
10"
10'0 10'0 lo8
Prominent interference
0.41 0,0048 Br82
0.29 0.088 0.047 0.011 0.46 0.24 3.10
0 . ooin
0.52 0.0043 0.0034 0.0039 0.038
AI data have been corrected to the following conditions: 1) thermal neutron flux = 4.3 x 10l2neutron/sq. cm.-second 2 ) irradiation Deriod = 1.00 hour 3 ) activity is tbat a t end of irradiation 4 ) detector is a 3-inch X 3-inch solid IiaI(T1) crystal 5 ) beta-particle absorber interposed between source and detector is '/n-inch thick, 3-inch diameter polystyrene disk (6) counting geometry -317, (7) analyzers are 400-channel type.
Product Yields and Limits of Detection Product Element In Sn Sb Te I Cs
Ba La Ce Pr Xd
Sm Eu Gd
Tb IIY Ho Er
Tm Yb
Lu
Hf Ta
Half life' 64 m 9 5m
41 m 60 d 2 8d 25 m 8 Od 25 m izh 83 m 2 6m 40 2 h 55s
137d 32 5 d 19 h 11 I d 18h 12 m 46 5 h 22 n 93h 3 7m 18 5 h 73 d 75 s 2 3h 27 3 h 7 5h 2 5s 127 d 19h 42d 32 d
6 8d 37h 5 5 h 44 6 d 19 s 115 d
Gamma ray energy, m.e.v
3.5 x 1.1 x 7.4 x 15 x 4.9 x 5.6 x 3.5 x
1 274
0 326 0 153
0 0 0 0 0 0 0 0 0
1
0 0 0 1 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 0 0 0
0
0 0 0 0
Yield," photopeak cp.m. per gram of element
603 566 147 364 455 605 127 163 662 60 74 294 142 57 091
x 1.1 x 2.0 x 3.3 x 9.4 x 2.8 x 1.5 x 1.0 x 3 0 x 4 5 x 1.7 x 1.6 x 1.2
211
110 102 105 961 102 364 299 108 094 080 36 301 208 084 147 396 177 0 198 208 088 216 482
+
OS
Ir
Pt Au Hg
1 Od 90 h
16 7 h 10 m 19 O h 74 4 d 14 s 30 m 3 15d 2 70d 5 5m 42 m
TI Pb Bi
c
24 h 65 h 47 d 1s
108 108 109
109 107 10'0 108
0.090
108 109
10'2
0.049 0.0030 0.00019 0.0092 0.045 0.00046
x
108
0.048
2.1
108 109 107
108 107 108
107 109 1Olo
10'0 10" 10" lo8# 1090
10'0
109 108
SblZ4
0.28
2.0 x 3.3 x 5.2 x 1.1x 2.2 x 2.2 x
109
Prominent interference
0.000082
lo8 10l2 10" 10'0 108 1 . 0 x 1010 1 . 2 x l01L 3 . 2 x 107 2 . 0 x 109 9 . 2 X lo8
3.6 X 3.2 X 1.5 x 3.0 X 2.5 X 1.7 x 3.6 X 1.7 X 5.6 X 1. 5 X 4.7 x
1010
0,000028 0.0093 0.013 0.068 0.0020 0.018
0,00049 0.0030 0.11 0 0036 0.65 0.097 0.33 0.022 0.60 0.0063 0,0028 0.00030 0.00068 0.000033 0.400 0.0060g 0.028 0,000058 0.000018 0,00067 0.21 0.00098 0.0086 0.32 0.0051 0.011
16 m W Re
10'2 10s
Limit of detection, A%.
YbI7o K x-ray
0,041 0.482 0.137 0 . 155 0 61 0 328 0 317 0 39 0 318 0 158 0 411 0 203
6.7 X 4.7 x 1.6 X 4 8X 4 1x
0.368 0.157 0.133 0.19 0.278 s o y 0.57
1 . 3 X lo8 x 108 1 . 2 x 108 1 . 3 x 107 4 . 0 x 107
No y
log 109 10'O lo7 1010
1 0 x 10'0 8 8 X lo7 1 2 x 109 9 6 X 10'
1 4 x 1011 2 5 x 108
9.0
X lo6
-6
s = seconds; m = minutes; h = hours; d Assumes air contains 0.9347, Ar.
=
0.0015 0.0021 0,00060 2.1 0.00024 0.0010 11
0,086 0.10 0.000070 0.41
where A is the number of counts in the photopeak; A t , the length of the count in minutes; A, the decay constant; ti, the time between the end of irradiation and the midpoint of the count; W , the element weight; tz, tJhe "standard" irradiation of 1 hour; and tS, the actual irradiation time. Suitable corrections to At and tl are applied for multichannel analyzer dead time. The fraction (1 - e - X t * ) , (1 - e - h f ' ) converts the result to a standard irradiation time of 1 hour. The choice of a 1-hour irradiation, unnecessarily long for many isotopes, was made for the sake of convenience. The correction factor, F , was applied when counting times were longer than 1 half life, for, under this condition, the average activity (counts/ length of count) is larger than, rather than equal to, the activity a t the midpoint of count. The correction converts the average activity to that a t the midpoint of the count, and is given by
0.77 0.11
0.083 0.79 0.25 -200
days; y = years.
Se75 plus unidentified gamma ray. Equilibrium established between parent 4.4-hour Braomand its 0.62-m.e.v. gamma-rayemitting daughter, 18-minute Br80.. f Parent and daughter in equilibrium. a Inrludes large self-shielding corrections. d
e
and Morrison (19). Selection of the ganima ray to be computed was governed by three rules: (1) the photopeak had the highest counting rate for that particular isotope; (2) t'he photopeak was free of serious interference by other gamma rays produced from irradiation of the element under study; and (3) higher energy photopeaks had priority over those of about 100 k.e.v. or less. Rule 2 was not adhered to in those cases when it meant selection of a gamma ray with a much greater (poorer) detection limit, or where no interference-free gamma ray was available. The next step is to calculate the phot,opeak yield for 1 gram of element irradiated for 1 hour. This yield is the maximum photopeak count rate available in a convenient irradiation, and hence is used to calculate the minimum amount of a n element which could be detected by neutron activation a t the flux used in these irradiations. The yield is calculated from Equation 1
where .io is the activity a t the start
of the count and the count lasts from t = 0 until t = At. Self-shielding corrections were applied where necessary according to the scheme of Gilat and Gurfinkel (4). Sample size was generally chosen so that these corrections were small or negligible; only for gadolinium were severe corrections necessary. Limits of detection were then calculated from these results according to the scheme of Buclianan ( 2 ) . =
k y- 10+R
VOL. 37, NO. 1, JANUARY 1965
(3)
1 31
Table 111.
Summary of Experimental Detection Limits Determined with Reactor Thermal Neutrons
Minimum detection limit ranges, pg.
Atomic S o . range 8 to 9 t o 18 19 to 35
10 -6- 10- 4
10-4- 10 - a
Ar
Sc, v, Br
Mn
37 to 53
In, I
55 t o 56;
Au
72 to 82 57 to 71
Eu, Dy
Cs, Hf, Re, Ir Sm, Ho, Er,
where S is the limit of detection. The constant, k , is chosen according to half life ( 2 ) :
Half life _ _
~ lo
5
-before the daughter activity was recorded. These data were corrected to the end of the irradiation, using the daughter half life, as though t’he
132
ANALYTICAL CHEMISTRY
10 -3-1 0 - 2 Na, A1 co, c u , Ga, As Se Sr, Rh, Ag, Cd, Sn, Sb Ba, W La, ?id, Gd, Yb, Lu
10-2-1 0 -1
c1
K, Ti, Zn. Ge Rb, Zr, Mo, Ru, Pd, Te Ta, Pt, Hg Ce, Pr, Tb
10 -’-I
00
100- 10’
~ O J - I O ~ 102-103
B
Mg Cr, Ni
103-104 0
Si, S Fe
Ca
Y, Xb Pb
os
Tm
daughter were formed during the irradiation. When the parent half life was longer than the daughter half life, data were recorded after the establishment of equilibrium, and decay corrections were made with the parent half life. I n those cases where the daughter activity was produced both by radioactive decay and by neutron capture, proper account’ was taken of contributions from both sources, and again yields are quoted as of the end of a 1hour irradiation. Table I1 lists well over one hundred gamma-ray emit,ting thermal-neutron products for most elements from oxygen through bismuth. For many elements, several products with sensitive detection limits are listed. I t is felt that this compilation may be of considerable aid t,o t’hose engaged in nonroutine activation analysis with reactor thermalneutron sources, for this list is more comprehensive than similar ones based on accelerator-produced thermalneut’ron data. Table I11 summarizes the data of Table 11. The data are grouped according to atomic number as indicated along the left side of the table. Ranges of detection limits in micrograms are indicated at the top of the t,able. I t may be seen from t,he data of Table 111 that the median det,ection limit’ is about 10-3 p.p.m. for a 10-gram sample. Obviously, with higher neutron fluxes t,he limits of detection will be reduced. Scale-up is not entirely straight-forward for a number of element,s, such as As, Cd, and Sb, because the epithermal neutron-capture contributions to the observed activities of these elements are appreciable. Thus, the validity of a scale-up calculation depends on both t,hermal- and epithermal-neutron fluxes (Table I). For the elements oxygen, silicon, phosphorus, iron, yttrium, thallium, and lead, react,or-fast, neutron-induced reaction product’s offer lower detection limits than thermal neutron reaction products. For further details, see reference ( 1 5 ) .
ACKNOWLEDGMENT
The author thanks H. R. Lukens, Jr., for making available his theoretical yield calculations, and T’. P. Guinn, for many helpful suggestions. LITERATURE CITED
(1) Anders, 0.U.,,Yucleonics 18, No. 11,
178 (1960). (2) Buchanan, J. D., dlomprazis 8, 272 (1962). (3) Croutharnel> C. E., Ed., “Applied Gamma-Ray Spectrometry,” Pergamon Press, S e w York, 1960. (4) Gilat, J., Gurfinkel, Y., Nucleonics 21, Ko. 8, 143 (1963). (,5) Goldman, I). T., Stehn, J. R., “Chart of the Suclides” (rev. to December 1961), General Electric Co., Knolls Atomic Power Laboratory, Srhenectady, \T
.Y.
1-
I.
(6) Guinn, V. P., Sucleonics 22, S o . 3, TO (1964). (, 7.) Guinn, V. P.. Wagner, C. D.. .4xa~. CHEM.32, 317 (196f). (8) Jenkins, E. N., Smales, A. A., Quart. Rev. 10, 83 (19.56). (9) Koch, R;, C., “Activation Analysis Handbook, Academic Press, Sew York, 1960 (10) Leddicotte, W., in “Radioactivation Analysis, proceedings of the Radioactivation Analvsis Svnmosium held in Vienna, Ausiria, J h e ‘ 1959, Butterxorths, London, 1960. (11) Meinke, W. W.,A r a ~ .CHEY.31, 793 (1939). (12) Soltys, S. S . , llorrison, G. H., Ibid., 3 6 , 293 (1964). (13) Steele, E. L., in Proceedings of Jnstrument Society of America, 1962, National Analysis Instrumentation Symposium, Charleston, W. Va., April 30May 2, 1961. (14) Strominger, D., Hollander, J. AI., Seaborg, G. T., Revs. M o d . Phys. 30, 585 (1958). (15) Yule, H. P., Lukens, H. R., Guinn, Y.P., Xuclear Instruments and Methods, in press; Rept. GA-5073, General Atomic Division/General Dynamics Corp., San Diego, Calif.
G.
RECEIVEDfor review August, 24, 1964. Accepted October 9, 1964. Work supported in part by a research contract with the Division of Isotopes Development,j IT. S. Atomic Energy Commission [Contract AT(04-3)-16T, Project Agreement, T o . 181.
