Precise Microdetermination of Uranium-235 in Irradiated Graphite Fuel Systems via Short-lived Iodine Fission Isotopes GEORGE BUZZELLI General Atomic Division of General Dynamics Corp., John Jay Hopkins laboratory for Pure and Applied Science, Sun Diego, Calif. Microgram quaintities of U235 in irradiated fuel systems can be accurately determined via neutron activation analysis utilizing the 2 1 -hour P3 fission isotope. Standard iodide carrier containing excess hydrazine as a reductant is added to the sample. The samples, together with uranium reference standards, are irradiated simultaneously. Iodine is separated in a single oxidatiori extraction step into CCI,, which is subsequently counted for The chemical yield of both standards and samples is determined colorimetrically. The precision and accuracy of this method for U235 at the parts per million level based on This is a 1%. standards are withiri simple, rapid, and sensitive method. Subnanogram quantities of U235 are readily determined by modification of the method, employing the 53minute 1134 fission isotope.
*
A
of irradiated fuel bodies requires a variety of methods to determine all of the constituents. Radiochemical techniques are rather uell defined and are quite adaptable to a variety of systems for the examination of the fission elements. Conventional techniques for the determination of the fuel constituents are not always adequate because of their lack of specificity, sensitivity, and/or precision. Knowledge of the residual U235content in a fuel system is necessary for the calculation of burnup, or for the determination of postirradiation isotopic ratios of heavy elements if the fuel contains source material. Burnup can be conveniently derived by radiochemical techniques--viz., relating the activity of one or more suitable fission elements to the fissionable isotope. In reactors such as the High Temperature Gas-cooled Reactor (HTGR), the fuel bodies are essentially graphitic, being composed of pyrolytic-carbon coated spherules of UC2 and ThCz dispersed in a graphite matrix. Each particle may contain as little as gram of highly enriched uranium and varying amounts of thorium. Subsequent to dissolution, the samples used NALYSIS
for analysis (one or more particles each) are volumetrically diluted, yielding final uranium concentrations as small as micrograms, or even submicrograms, per milliliter. The precise microdetermination of uranium in these diluted fuel solutions is a formidable problem. Adequate sensitivity has been demonstrated by Smales (5) and Ledicotte (3, 4 ) utilizing neutron activation techniques with subsequent pursual of the fission product isotope, Bal4O, However, their method was not entirely satisfactory with r e spect to simplicity and accuracy. I n addition, their method was limited to older samples-Le., samples removed from the irradiation source for 6 to 8 months. The interference is due to the presence of the Ba-l" (tllz = 12.8 days) produced during the original irradiation of the fuel. Because time is an important factor in analytical work, the need for a reliable and appropriate method of analysis for numerous samples containing various amounts of uranium is enhanced. The method described here combines the sensitivity of neutron activation with the reliability of a carrier technique utilizing a conveniently determinable fission product isotope. The isotope is generated during a brief neutron irradiation, and isolated in a single oxidation extraction step. The selection of a suitable fission isotope was based on the following attributes: simple and reliable chemical manipulation (preferably amenable to solvent extraction), including yield determination; high fission yield; conveniently short half life; and emission
Table 1.
Isotope I131
I132 I135
(Xe133 I134 I136 (Xe-'"m (Xe13'
of one or more abundant and well defined photons suitable for gamma spectroscopy. Iodine-133 (tl/z = 20.8 hours) was selected for this investigation. The U 2 3 5 thermal neutron fission yield of this isotope is 6.65% (a). It decays by beta emission to 5.27-day Xe133 with an associated gamma photon of 0.53 m.e.v. (Figure 1). The gamma branching ratio is 0.93. The half life is short enough to offer adequate sensitivity, yet long enough to permit a reasonable cooling period subsequent to irradiation. Any necessary correction due to decay during the analysis period will be small. The extraction chemistry of iodine is rapid, clean, and amenable to a precise colorimetric yield determination. There is no interference from the main fission iodine-i.e., 8.0day Ilal-since all of the isotopes are relatively short-lived (see Table I). Otherwise, iodine can easily be expelled by volatilization during the solution process. Dissolution of the individual fuel specimens is effected by mechanical pulverization in a nitric acid medium, followed by a brief digestion period in a hot water bath. Specimens which include particles embedded in graphite can be wet ashed using a mixture of nitric and perchloric acids. This wet ashing will readily destroy the graphite but will not immediately affect the coated particles. At this point mechanical pulverization of the particles will greatly reduce the dissolution time. The metal carbides are quite reactive in an acid medium once the coatings have been cracked. The final acidity of the sample is approximately 3N with
Principal Gamma Activities
Half life 8.05d. 2.3 hr. 20.8 hr. 5.27d. 53 min. 6.7 hr: 15.5 mm. 9.3 hr.
Fission yield 2.93 4.38 6.65 8.06 6.4
of Fission Iodine Isotopes (6) Principal gamma rays, m.e.v. 0.364,0.64,0.284 0.67,0.78,0.53 0.96 0.53,0.85 0.081) 0.848,1.10 1.14, 1.24, 1.80,(0.53) 0.53) 0.250)
VOL 37, NO. 1 1 , OCTOBER 1965
1405
respect to nitric acid. An exact volume of the sample aliquot is pipetted into an irradiation vial containing standard iodide carrier in a reducing media. Standard reference solutions are prepared in the same manner. These vials are sealed and irradiated in a neutron flux of 2 X 10l2 neutrons per sq. cm. per second for 30 minutes. After a IO-hour cooling period to permit decay of the carrier activity and to maximize decay of the precursors, the samples are subjected to separation chemistry. Duplicate aliquots are withdrawn from each vial and pipetted into extraction tubes. The iodide is oxidized to iodine and extracted into carbon tetrachloride. An aliquot of this solution is taken for gamma counting and subsequent yield determination. Standard solutions containing uranium are processed in duplicate subsequent to the sample analysis. EXPERIMENTAL
Apparatus and Reagents. Analytical grade reagents were used throughout the work. A stock iodide carrier solution (10.00 mg. per ml.) was made up using potassium iodide (containing 1 ml. of hydrazine per liter of solution) and was standardized gravimetrically. The irradiation carrier standard solution (2.500 mg. per ml. and 10% with respect to hydrazine hydrate) was made up using the stock iodide solution. For construction of the colorimetric calibration curve, an iodide standard solution (1,000 mg. per ml. and 4% with respect to hydrazine concentration) was prepared using the stock solution. The calibration curve covered the range 0 to 0.400 mg. per ml. A Beckman Model DU spectrophotometer was used for the colorimetric yield determination using 1.00-cm. matched cells. The relative variance of each cell was determined using carbon tetrachloride reagent a t 520 mp. The General Atomic TRIGA Reactor facility was used for the irradiation. All samples were counted with a Nuclear Data Model 180 FM, 512 channel analyzer using a 3 X 3-inch sodium iodide solid detector within a 4-inch thick lead shield. This analyzer includes an integrator resolver unit. A Model K-500-4 vortex mixer, sold by Scientific Products, and disposable culture tubes (22 X 175 mm.), Scientific Products catalog No. T1300-101, were used for the extraction work. Cranium-235, NBS-930 was used to prepare the uranium standard. The u30, was ignited a t 850" C. overnight. The oxide was weighed out and dissolved in nitric acid. Dilutions of 1.347, 2.020, 2.694, 3.367, and 6.735 pg. per ml. were made up from this stock solution using calibrated pipets and flasks. The final acidity of these solutions was adjusted to 3N with respect to nitric acid to simulate sample composition. Procedure. With a calibrated pipet (or semimicro buret), 2.00 ml. of standard iodide carrier (2.500 mg. 1406
ANALYTICAL CHEMISTRY
I
0.530 (1-133)
j.67 (I-132) n 0.70
L CHANNEL NUMBER
Figure 1. Gamma ray spectrum of 20.8-hour isotope and activity due to Il3l1 11321 and Xe*35(energies in rn.e.v.1
per ml.) was dispensed into clean, dry 2-dram polyvials (polyethene vials). Then 3.00 ml. of the sample solution was added to the vial; the vial was heat-sealed and labeled, and the contents were mixed by inverting. A set of reference standards was prepared by pipetting 3.00 ml. of the uranium standard solution into the vials containing iodide carrier; for example, suitable reference standards were made up as one vial containing 1.347 pg. per ml. and a second vial containing 6.735 pg. per ml. The vials containing the standards were heatsealed, labeled, and mixed by inverting. The standards and samples were irradiated in a thermal neutron flux of 2 X loL2neutrons per sq. cm. per second for 30 minutes. It was convenient to perform the irradiation in the afternoon or evening to allow decay of the carrier and other short-lived fission product activities overnight. This time was also useful in permitting decay of the 50-minute isomer of Te133, thus optimizing the 1 1 3 3 activity. Labeled vortex tubes were prepared by pipetting 10.0 ml. of carbon tetrachloride into each and immersing them in an ice bath. The irradiated vials were carefully opened and duplicate 2.00-ml. aliquots were immediately withdrawn from each vial and pipetted into separate vortex tubes. The tube walls were rinsed with 1 ml. of water and the tubes were allowed to chill for 1 to 2 minutes. Fuming nitric acid was slowly added until free iodine was liberated as indicated by the brown color. The tubes were allowed to stand for another minute, then placed on the vortex mixer and agitated a t high speed. Several drops of saturated sodium nitrite were added, the iodine was extracted into the carbon tetrachloride by shaking for 30 seconds, and the phases were allowed to separate (a water-clear, transparent aqueous phase indicates completion of extraction). The entire mixture was then transferred to a clean 40-ml. centrifuge tube and centrifuged for 1 to 2 minutes. A 5.00-ml. aliquot of the carbon tetrachloride layer was withdrawn and pipetted into a labeled 2-dram screwcap glass vial for counting purposes. It was convenient to utilize this count-
ing fraction for yield measurement. The iodine-carbon tetrachloride mixture was stable for up to 20 hours. The iodine-carbon tetrachloride was read a t 520 mp relative to carbon tetrachloride as the blank, using a slit of 0.010 mm., a blue phototube, and a sensitivity setting of 1. The extracted solutions were counted with a multichannel analyzer over a period long enough to accumulate approximately 50,000 counts under the 0.53-m.e.v. peak. One half life after the irradiation, 1 pg. of UZ35will yield approximately 6000 counts per minute of 1 1 3 3 under these counting conditions. When using the solid detector (sodium iodide), it was desirable to use a sample positioner to precisely locate each vial in the identical position. A lucite cap approximately inch thick with a hole to accommodate the 2-dram glass vial in the geometric center was suitable for the positioner. The counting was repeated for one standard and one sample solution to obtain an experimental half life value for decay correction. Peak areas were integrated and the net counts above background were calculated. These values were corrected with respect to the chemical yield and decay. The amount of in the samples was obtained by quantitative comparison of these corrected peak areas to the standard uranium corrected peak areas. RESULTS AND DISCUSSION
The simplified method described above for the determination of P6 relies on the success of retaining the fission iodine in solution and equilibrated with the carrier iodide. To accomplish this, a strong reducing media was required. Hydrazine was employed as the reductant. When present in excess, hydrazine will reduce both iodine and iodate to the iodide. Acid solutions of hydrazine are stable and not affected by oxygen ( I ) . Thus, the fission iodine produced was born into a reducing environ which served to keep both iodine species (active and carrier) in the same solution-stable oxidation
Table 11.
CHANNEL. NUMBER
Figure 2. Gamma ray spectrum of 1134 plus Xe135m (energies in m.e.v.); 10 k.e.v. per channel, channel markers every 16 channels
state. With equilibl*ation exchange of the iodide species assured, the subsequent chemistry could be simplified to a single oxidation extraction step. No apparent interferences were observed in the graphic fuel system. Evidence supporting this concept is indicated by the consistency of the data. The results are shown in Tables 11, 111, and IV. Table 11 summarizes the results of analysis for U2S5employing standard solutions. The specific activity of each solution and the overall average value is indicated in the last column. Solution analyses of irradiated fuel particles are shown in Tables I11 and IV. These samples were irradiated to a P5 burnup of 20%. The fuel concentrations of the latter are a tenfold dilution of those in Table 111. The net counts above background which have been corrected for chemical yield are listed in each table. Each duplicate set of these corrected counts was averaged and normalized to the median sample counting times for the decay correction. It should be emphasized here that these decay corrections be derived from experiinental half lives. The actual half life of 1133 via the 0.53 is influenced by the presence of the decay product Xe135 from I 1 S 5 (see Table I). The uranium content of these sample solutions was calculated from these normalized values relative t o the standards and is listed in the last column. The high degree of precision and accuracy exhibited hLere is a result of using exact volumetric ratios and incorporating a precise yield corrective step. This method is capable of high yield, and therefore any necessary corrections are small. The ITCC14 solution is colorimetrically stable for up to about 20 hours. The results are shown in Table V. With samples containing two or more fig. of P 3 5 per ml., sufficient counts of I I 3 3 are obtained in 5 minutes counting time t o yield data with a statistical accuracy of +0.5%. Yield corrections are also accurate to within much less than ilyo. The
Sample 1.347A-1 -2 1.347B-1 -2 2.020 -1 -2 2.694 -1 -2 3.368 -1 -2
U236Method;
Yield,
%
93.6 92.2 78.4 94.2 93.6 93.1 93.8 91.8 92.4 91.8
Net counts, correctedb 35,566 33,280 33,724 33,376 50,294 50,330 67,079 65,668 84,815 83,798
via Standard Uranium Solutions"
Average 34,426 f 1143
Normalized t o (3) 33,851
25,131
33,460f 184
33,185
24,636
50,312f 18
50,312
24,907
66,373f 705
66,924
24,842
84,306f 508
85,739
25,455
Irradiated a t 250 kw.for 30 minutes. b Counting times are uniform for all samples, via., 8.00 minutes.
Counts/pg.
U*%
Av. 24,994 & 239
(1
Table 111.
Sample STD 1-1 -2 613 -1 -2 617 -1 -2 619 -1 -2 622 -1 -2 STD 11-1 -2 4
Yield,
70
97.5 97.8 100.0 100.0 99.0 98.8 99.3 98.9 97.5 100.0 97.4 97.5
U235Method; Analysis of Irradiated Fuel Particles-I
Net counts corrected" 42,334 42,940 29,995 30,074 94,000 92,408 75,876 78,090 20,244 20,862 6b,, 500 61,700
Ut%,fig. Normalized Average to ( 4 ) 42,637f 300 41,600
Per unit Per ml. count, 10-6 1.347 3.238
30,025 f 40
29,483
0.959
93,153f 745
92,221
2.98
76,983f 1107
76,983
2.49
20,553f 309
20,759
0.670
61,600f 100
62,733
2.020
The counting time is uniform for all samples, viz., 4.00minutes. Table IV.
Sample 613 -10 -20 617 -10 -20 619 -10 -20 622 -10 -20 STD 1-1 -2 STD 11-1 -2
3,220
Av. 3.229f 9
U235Method; Analysis of Irradiated Fuel Particles-11"
Net Yield, counts, Normalized % corrected Average to ( 3 ) 95.0 4,129 4,146f 19 4,097 95.8 4,164 96.6 12,858 12,870 13 12,562 97.5 12,883 99.2 10,197 10,151 f 46 10,151 98.3 10,105 97.5 2,731 2,775 f 44 2,779 98.3 2,819 97.5 27,892 27,956 f 66 28,448 98.3 28,020 100.0 110,080 110,450f 370 114,480 100.0 110,820
*
ua35,
pg.
Per unit Per ml. count, 10-6 0.0969 0.297 0.240 0.0663 0.6735
2.368
2.694
2.353
Av. 2.360& 7 These samples are 1 : 10 dilution of the previously analyzed solution (Fuel Particle Analysis I,see Table 111). * The counting time is uniform for all samples, viz., 8.00minutes.
iodine calibration plot is linear, with a standard error of estimate of 0.001849, which for the iodide concentration utilized here (0.200 mg. per ml. and a corresponding absorbency of 0.719) results in a relative standard deviation of 0.26a/0. Thus, with duplicate measurements of each sample and standard, an overall precision and accuracy of =t 1% is quite feasible and practical. It has also been demonstrated that
Table
12-CCla Solution Stability
12, Absorbency mg./ml. (1) (2) Remarks o,200 o,718 o.718 (1) 1700 hr.,
2 -,/FI-,/65
-1
0.300 1.07 0.400 1.44
1.07 1.43
(2) 1330hr.1 3/10/65
VOL. 37, NO. 1 1 , OCTOBER 1965
1407
Table VI.
U235Determination," Comparative Results
Set I Sample Analyst A Analyst B Set IIb 613 617 619 622
0 966 2 99 2.47 0.673
0.959 2.98 2 49 0.670
0.0969 0.297 0 240 0,0663
All results are in pg./ml. (p.p.m.). solutions of Set I1 are 1:10 dilution of Set I.
* Sample
this method is useful for the relatively precise determination of submicro(see Table IV gram quantities of U235 and VI). The yield correction is independent of the quantity of uranium present. Thus, we are able to retain a precise yield measurement. Only the counting statistics, which, of course, are proportional to the quantity of uranium, will vary. Longer counting
times are necessary to attain maximum precision. An alternative choice is available in cases of extremely low levels of U235. The sensitivity of this method can be extended to subnanogram quantities by utilizing the isotope 1 1 3 4 ( t l i z = 53 minutes). The irradiation is carried out with a sample aliquot plus hydrazine (10%) and without iodide carrier. The gamma spectrum is shown in Figure 2. Preliminary work a t this level indicates that, with caution and recognition of background interference a t these levels, subnanogram quantities of L235 are readily determined. For a single sample, a double extraction of the iodide followed by precipitation as silver iodide and subsequent beta counting yielded a beta decay plot of the 53minute The gamma purity of this specimen was verified by gamma spectrometry.
ACKNOWLEDGMENT
The author expresses his appreciation to Jack Douglass for his assistance in performing the analyses reported here. LITERATURE CITED
(1) Audrieth, L. E., Ogg, B. A., "The Chemistry of Hydrazine," Chap. 6, U'iley, New York, 1951. ( 2 ) Katcoff, S., Nucleonics 18, No. 11, 201 11960). (3) Ledicotte, G. W., Brooksbank, W. A,, USAEC Report TID-7531 (Pt. l ) , pp. 71-8, 1957. (4) Alaklman, H. A., Ledicotte, G. W., ANAL.CHEN.27,823 (1955). (5) Seyfang, A. P., Smales, A. A., Analyst 78.394 (1953). (6) -Stehn,' J.- F., Nucleonics 18, Xo. 11, 186 (1960).
RECEIVEDfor review June 10, 1965. Accepted July 30, 1965. Research s u p ported in part by the U.S. Atomic Energy Commission under Contract AT(043)167, Project Agreement No. 17.
. .
Chronopotentiometric Study of Pheny mercuric Ion Adsorption on a Mercury Electrode R.
F. BROMANI
and ROYCE W. MURRAY
Department of Chemistry, University o f North Carolina, Chapel Hill, N. C. Both phenylmercuric ion and the produd of its one-electron reduction are strongly adsorbed on a mercury electrode surface. The adsorbed reactant gives rise to polarographic, potential-sweep chronoamperometric, and chronopotentiometric prewaves during which the adsorbed reactant is coulometrically electrolyzed before the diffusing solution reactant is reduced. Adsorption and desorption rates of both reactant and product are slow. A spike on the chronopotentiometric prewave has been interpreted as arising from a relatively slow reorientation process within the adsorbed layer leading to a transient reaction overpotential.
T
of adsorption in the electrochemical reduction of phenylmercuric ion was first studied by Benesch and Benesch (1, 2 ) and Vojir ( 2 4 ) . A prewave preceding the first diff usion-controlled, polarographic reduction wave was interpreted ( 2 ) as arising in the classical Brdicka manner (4) from adsorption of the reduction product (presumably the phenylmercury radical) on the D.M.E. surface. Because it has HE ROLE
1 Present address, Department of Chemistry, University of Nebraska, Lincoln, Neb. 68508
1408
ANALYTICAL CHEMISTRY
275J5
been noted (1, 2,9, 24) that both of the one-electron polarographic waves are irreversible and because the Brdicka (4) interpretation of polarographic prewaves and postwaves is strictly valid only for reversible electron transfers, it seemed appropriate to consider the electrochemical behavior of phenylmercuric ion more thoroughly. Adsorption of electroactive species a t electrode surfaces can be studied by chronopotentiometry. Various theoretical features of this technique with reference to adsorption have been described (8, 11, 15, 16, 21), and several applications have been reported (8, 10, 12-15, 17, 18, 25). Information on the order of reaction of the adsorbed and diffusing reactants, their kinetic interplay, and surface excess data can, in favorable cases, be derived from the morphology of the chronopotentiometric wave and the appropriate theoretical relationships (11, 15, 16, 21). The detection and measurement of product adsorption can be accomplished by using current reversal (8, 18). Application of the chronopotentiometric technique to the phenylmercuric system has produced evidence for the adsorption of both reactant and product and for the direct connection of the observed prewave to reactant adsorption rather than product adsorption. POtential-sweep chronoamperometry was
employed for the qualitative verification of several aspects of the chronopotentiometric results. EXPERIMENTAL
Reagents. Phenylmercuric hydroxide (Columbia Organic Chemicals Co., Inc.) was used either as purchased or after a single recrystallization from water. The two forms displayed identical electrochemical properties. A determination of total mercury was performed by refluxing 500 mg. of the recrystallized substance in 8.11 nitric acid for 1 hour and then titrating the free mercuric ion potentiometrically with EDTA (20). The phenylmercuric hydroxide solutions were prepared by dissolving weighed nortions of the substance without further -standardization. ANALYSIS. Calculated for C6H5HgOH: C, 24.45%; H, 2.05%; Hg, 68.1%. Found. C, 24.92%; H, 1.87Q/,; Hg, 68.1%. The supporting electrolyte solution was 0.lM acetate buffer, p H 5.2, prepared from ordinary reagent grade chemicals and redistilled (from alkaline permanganate) water, Solutions were thoroughly deaerated with high purity nitrogen before use and a nitrogen atmosphere was maintained throughout each experimental series. Instrumentation. Experiments using the various techniques were all carried out with a modular instrument based on the DeFord design (6),utilizing
