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Anal. Chem. 1084, 56,696-700
Mass Spectrometric Determination of Burnup of Thorium-Uranium Dioxide Fuel L. W. Green,* C. H. Knight, T. H. Longhurst, and R. M. Cassidy General Chemistry Branch, Atomic Energy of Canada Limited Research Company, Chalk River Nuclear Laboratories, Chalk River, Ontario, Canada KOJ 1JO
The isotopes 14'Nd and ls+lMNd were investigated for use as fission monitors. A two-column anion-exchange procedure was used to separate these and U and Th from the fuel matrix, and the purified fractions were analyzed by thermal ionization mass spectrometry. Relative standard deviations of Nd, U, and Th determinations by isotope dilution were ~ 0 . %. 7 A computer-generated simulation of the irradiation was used to estimate the effective flsslon yields for '&Nd and 's+lMNd. Burnup results with 's+luNd as the flsslon monitor showed excellent agreement with results obtained by a high-performance liquid chromatographic method that used lagLaas the fission monitor; the average difference between the two methods was 0.02%. The '"Nd results were biased high by up to 4%; this was attributed to a 14'Nd neutron capture effect. Resuits obtained with the initial heavy element content estimated from the weight and initial composition of the fuel, instead of from analyses for the actinides, showed excellent agreement (average difference = 0.2 % ) with the conventional method.
In thermal neutron reactors thorium fuel cycles use fertile material much more efficiently than do conventional uranium fuel cycles and are potentially less costly and longer lasting (1). Thorium fuels are being developed for the natural uranium fueled CANDU reactors, which are sufficiently flexible to be compatible with thorium fuel cycles. A significant part of the fuel development work is burnup determination, the topic of this paper. The dominant fissile nuclides in thorium-based fuels are the or 239Pumixed with the starting material and the formed during irradiation. The uranium isotopic composition and net fission product yields are substantially different from those of more conventional fuels, and conventional burnup techniques such as 235Udepletion, 148Nd,and neodymium isotope ratio methods (2-4) cannot be applied directly. Consequently, the options for burnup procedures were reviewed with special attention to the following: fission monitors, separation methods, and mass spectrometric techniques. Neodymium-148 has long been the fission monitor of choice for U and (U, Pu) fuels, because of its nearly equal yields from 235Uand 239Pu. However, this attractive feature does not extend to Th-based fuels; the 148Ndfission yield for 233Uis 34% lower than that for 235U( 5 ) . A more serious drawback with 148Ndis its reported susceptibility to a 147Ndneutron capture effect (4, 6-8). Indirect measurements by y spectrometry of the thermal neutron capture cross section of '"Nd yielded an estimate of 440 X cm2 (7), which is -9 times greater than the previously accepted value of 50 X cm2 (9). The flux dependence of 148Ndlevels in highly enriched UAl fuels is consistent with the higher value (4). The ASTM 148Ndprocedure (2) includes a correction for 148Ndformed from 147Nd,but Wilson et al. have questioned the validity of this correction (10). More work is required to accurately determine the 147Ndcross section and its effect on 148Nd burnups. 0003-2700/84/0356-0898$01.50/0
The sum of '&Nd plus I@Ndis not susceptible to significant neutron capture effects and has been recommended as a fission monitor for U fuels (4,8,11). However, for use with Th-based fuels, the effective fission yields of 145Ndand 146Ndmust be estimated because their respective yields from 233Uand 235U are substantially different (5). Lanthanum-139 and l@Ceare potential fission monitors, which, because of their nearly equal yields (respectively) from 233Uand 235U,do not require elaborate fission yield estimates. However, because of difficulties with mass spectrometric determination of La and Ce, described later, their use as fission monitors for mass spectrometry was not pursued. The sum of 146+146Nd was used, and a computer-generated simulation of the irradiation was used to estimate the effective fission yields. Results were also obtained with 148Ndfor comparison. For separation of Th, U, and Nd fractions, a review of the literature revealed classical anion-exchange methods for separation of Pu and U from the fission products (12),Nd from the other fission products (13),and T h from the lanthanides (14). The separation of Nd from other fission products appeared directly applicable to (Th,U)02fission product fractions, but modification of the other two procedures was required to separate T h and U and the bulk fission products from each other. The separation of these by anion exchange in HC1-acetone was investigated. Mass spectrometric determination of Nd did not present any new difficulties, but that of T h and U did. The radioactivity associated with 228Th,232U,and 233U,produced by irradiation, and 23"Th,used as a spike, limited the quantities of T h and U used for mass spectrometry to the nanogram level. Furthermore, T h is much more difficult to ionize than U, due to its refractory nature and higher ionization potential. Consequently mass spectrometry of T h and U was investigated in detail and is the topic of a separate paper (15). Finally, in an attempt to shorten the burnup procedure, a weight-based method for estimating the initial heavy element content was investigated.
EXPERIMENTAL SECTION Reagents and Materials. Nitric, hydrochloric, and hydrofluoric acid solutions were prepared from double-distilled nitric and hydrochloric acids and reagent grade hydrofluoric acid, diluted with deionized water. Reagent grade acetone and methanol were used for preparation of HC1-acetone and HNOgmethanol solutions, respectively. AGMP-1 anion resin (Bio-Rad Laboratories, Richmond, CAI, 100-200 mesh, was used for chromatographic separations. For the separations in an HC1-acetone eluent, fresh resin was equilibrated in freshly prepared eluent each day and used on the eluent, large same day. For separations in an "Os-methanol batches of resin were washed with 8 M HN03 and stored in 80% methanol/l.6 M "OB. These resins were then packed to a height of 6 cm in Pyrex Econo-Columns (Bio-Rad Laboratories). Fractions were collected in Pyrex beakers and aliquots for mass spectrometry were placed in polyethylene microvials. Acropor (Gelman Instrument Co., Ann Arbor, MI) anion-exchange membrane disks (1 mm diameter), covered with a starch-rhenium deposit, were used for mass spectrometry of U and Th. The disks were equilibrated with 2 M "OB prior to use. The starch-rhenium solution was 42 g/L in starch and 0.50 0 1984 American Chemical Society
ANALYTICAL CHEMISTRY, VOL. 56, NO. 4, APRIL 1984
g/L in Re; the Re solution was prepared from zone refined Re metal (Rhenium Alloys, Inc., Elyoria, OH) dissolved in H202-HNO% Primary Standards and Spikes. Isotopically natural thorium metal, -1 g measured to the fifth decimal place, was dissolved in 10 mL of 0.05 M HF/13 M HN03and diluted with 8 M HN03 to yield a 232Thprimary standard that had a concentration of 0,19117 mol/kg. For the 23@Thspike solution, 91.54% enriched 23@hI' as Thoz powder (Oak Ridge National Laboratory, Oak Ridge, TN) was dissolved as above, diluted with 0.5 M "OB, and standardized by isotope dilution mass spectrometry with the 232Thprimary standard as a spike. The average of three determol/kg. minations was (3.06 f 0.01) X Isotopically natural Ndz03powder (Spex Industries,Metuchen, NJ) was heated at 970 OC to constant weight and stoichiometry (16) and then dissolved in 0.5 M HN03 and diluted to give a primary standard that had a total Nd concentration of 5.7917 X W3 mol/kg. The solution was stored in a 100-mL Nalgene volumetric flask. For the IwNd spike solution, 96.17% enriched IwNd as Nd203 powder (Oak Ridge National Laboratory) was dissolved and diluted in 0.5 M HN03 and standardized by isotope dilution maas spectrometry with '&Nd of the Nd primary standard as the spike. The average of five determinations was (2.046 f 0.02) X mol/kg. This solution was stored in sealed glass ampules until needed. National Bureau of Standards SRM 993 was used as a U spike solution; the 23sUisotopic abundance and concentration were mol/kg, respectively. This isotope 99.8195% and 2.8219 X was chosen for the U spike because of the availability of the assay and isotopic standard solution and because U contamination in the rhenium filaments occasionally caused interferences at mass 238. An accurately weighed aliquot of this was mixed with accurately weighed aliquots of the 230Thand '%Nd spike solutions to yield a triple spike solution for the determination of the concentrations of Th, U, and Nd by isotope dilution mass spectrometry. Apparatus. Fuel dissolutions and separations were performed in a hot cell; a glovebox was attached to the cell to facilitate sample removal and other work with low levels of radioactivity. A remote-controlled Mettler (Mettler Instruments, AG, Switzerland) AK160 electronic balance was used for weighing fuel samples and solutions in the hot cell. A Vibrograver (Burgess Vibrocraft, Inc.) mechanical vibrator was used to remove the fuel from the fuel cladding. Cathodeon (Cathodeon La., Cambridge, England) type 553 triple filament assemblies equipped with tantalum side filaments and a rhenium center filament were used for mass spectrometry. The original rhenium filaments were replaced with zone refined rhenium ribbon (Rhenium Alloys, Inc.), 0.025 mm thick X 1mm wide, to avoid interference from trace natural U. The mass spectrometer was a Nuclide (Nuclide Corp., State College, PA) S.U.2.2 instrument equipped with a 16-stage electron multiplier detector. Data were acquired and processed with a General Automation (General Automation, Inc., Anaheim, CA) SPC-16/45 computer. Procedure. (a) Fuel Sampling and Dissolution. The 36 elements of the fuel bundle (BDL 417 AAX: 232Th,97.3%; YJ, 0.016%; 23sU,2.45%; 23BU,0.0093%; 238U,0.225%; DzOcoolant) had been arranged in the usual CANDU geometry of three concentric rings. Four of the elements from each ring were selected for analysis and allowed to decay for 2 years following removal from the reactor. A shorter decay time could have been used. A section containing -20 g of fuel was cut from the midsection of each fuel element; care was taken to avoid contamination from other fuels and to avoid loss of small fragments. The latter was important because of the high neutron absorptivity of Th and the resultant large burnup gradient across each fuel element. A mechanical vibrator was used to separate the fuel sample from the zircaloy sheath and all of the fuel particles were collected and accurately weighed. The sample was then transferred to a m-mL round bottom flask and refluxed in -300 g of 0.05 M HF/13 M HN03 until dissolved (17); between 12 and 40 h were required for dissolution. The fuel solution was weighed and an aliquot was diluted with water to yield a Th concentration of 1mg/g. Two 1-g aliquots of this solution were taken for separations; one of these was spiked with sufficient triple spike solution to yield a sample to spike ratio in the range 0.05 to 20. N
697
(b) Column Separation. Thorium and uranium were separated from each other and the fission products as follows. (1)Each aliquot of the dilute fuel solution was evaporated to dryness, dissolved in 1 mL of 1 M NaN02/12 M HCl, and reevaporated to dryness; then the last two steps were repeated. During the evaporations, the solutions were stirred and heated to near boiling to ensure equilibration of sample and spike. (2) The residue was dissolved in 0.25 mL of 75% (v/v) acetone/25% HC1 and transferred to an anion-exchange column that had been preconditioned with the same solution. (3) The fission products were eluted with 5 mL of 75% acetone/25% HC1 and the eluate was stored. (4) Th(1V) was eluted with 5 mL of 40% water/35% acetone/25% HC1; the eluate was evaporated to (5) Americium dryness and redissolved in 0.5 mL of 8 M "OB. was eluted with 5 mL of 12 M HCl and the eluate was discarded. (6) U(V1) was eluted with 10 mL of 8 M HN03 and the eluate was treated as in step 4. (7) The fission product fraction (step 3) was converted to a nitric acid solution by successive evaporations in H202-HN03,and Nd(II1) was separated as described by Marsh et al. (13). This procedure yielded optimal separations for one batch of resin, but adjustment of the eluent volumes and/or compositions was required for each new batch. The thorium fraction (step 4) was diluted 10-fold with 8 M "OB; then 10 r L of the dilute solution was mixed with 100 r L of the uranium fraction (step 6). Three Acropor disks were added and allowed to equilibrate with the solution for 1h; each disk adsorbed 15 ng of U and 40 ng of Th. One was transferred to the center filament of a filament assembly, covered with 3 WLof the starch-rhenium solution, and heated to dryness. (c) Separation with a Tributyl Phosphate Impregnated Resin Bead. For some of the fuel samples an alternative method (15) was used to extract U and Th. The dissolver solution was diluted to 16 mg/mL Th in 0.08 M HF/6 M "OB and an aliquot was equilibrated with an XAD-2 resin bead that had been impregnated with tributyl phosphate. After the equilibration, the bead was removed from the fuel solution and equilibrated with a 0.025 M HF/8 M HN03 solution. Several Acropor disks were added to this solution; each disk adsorbed approximately 14 ng of U and 45 ng of Th. Subsequently, a disk was transferred to a rhenium filament for mass spectrometry. (d) Mass Spectrometry. The mass spectrometry of U and Th has been described elsewhere (15);briefly, the filament assembly was inserted in the source and the center filament was heated to 1700 "C for U isotopic analysis and then to 1900 "C for Th. Twenty to thirty replicate isotopic analysis were completed for each element. The conventional side filament technique was used for Nd; a center filament temperature of 1950 "C was used. Approximately 80 ng of Nd was deposited on each of the side filaments. Observed isotope ratios were corrected for an instrumental bias factor of 0.21% per mass unit for Th and U and 0.58% per mass unit for Nd; the bias favored lighter isotopes.
-
RESULTS AND DISCUSSION Separations. Experiments with the HC1-acetone system showed that Th(1V) and the bulk fission products could be readily separated from each other by proper choice of HC1 and acetone concentrations. Uranium, however, was strongly retained in this system and was best eluted with HN03. Consequently, the elution scheme outlined earlier was selected, and it yielded satisfactory separations (Figure 1). The Nd fractions usually contained some Pr and Pm, but these did not interfere with Nd mass peaks. Occasionally the separation from Sm was poor; these separations were repeated until a satisfactory separation was obtained, because Sm interfered with Nd a t masses 148 and 150. Mass Spectrometry. Figure 2 shows a mass spectrum of the unspiked Nd fraction separated from element 4. In addition to the peaks for Nd, the spectrum shows peaks for Ba (135-138), Ce (140, 142), Pr (141), Pm (147), and BaO (151-154). The Ba and Ce signals were mostly due to Ba and Ce contamination in the filaments and prevented the use of '39La and as fission monitors. A tantalum center filament technique (18)was attempted for Ce but the signals were very
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ANALYTICAL CHEMISTRY, VOL. 56, NO. 4, APRIL 1984
FI SSI 01.1 PRODUCTS
Table I. Precision of Th,U, and Nd Determinations concentration, mmol/kg of fuel analysis no. 145t146Nd 148Nd U Th
fin4’
5.5739 5.6347 5.7012 5.6082 5.6366 5.6553
1
2 3 4
5 6 7
5.635 0.043 0.76
av S
%RSD
1.3560 1.3778 1.3855 1.3811 1.3603 1.3708 1.3827 1.373 0.011 0.84
86.893 87.820 86.547 86.988 87.596
3651.9 3686.2 3708.0 3729.4 3696.9 3677.4
87.17 0.52
3692 27 0.72
0.60
Table 11. Thermal Fission Yields from Reference 5 atom % yielda fission product ELUENT VOLUME ( m i )
I4’Nd I M Nd 145t146 Nd I4*Nd
Figure 1. Separation of fission products, thorium and uranium, by anion exchange in HCI-acetone; AGMP-1 100-200 mesh resln. a
233
u
YSU
3.39 i 0.02 2.54 + 0.02 5.93 * 0.04 1.27 i. 0.01
3.92 i 0.01 2.98 * 0.01 6.90 i. 0.02 1.67 i. 0.02
Uncertainties are 1standard deviation. 0 02 arnuls
0 028 amu I s
___j
lOfA
P
l a Ba
r?
.o I ILL
Nd
IOfA:
230
232
233
23L 235
236
238
m/ 15L
Figure 3. Mass spectra of Th and U from irradiated (Th,U)02 fuel: element 21 spiked with 230Th;fllament temperature for U, 1700 O C ; for Th, 1900 OC.
145
r
1LO
Ce
112
Ce A -
m/z Figure 2. Mass spectrum of Nd from irradiated (Th,U)O, fuel; element 4 unspiked.
noisy and the ionization efficiency was low. The Nd signals were sufficiently strong and stable for precise isotope ratio measurement; a relative standard deviation of 0.5% was obtained for a typical unspiked sample analyzed through the entire procedure three times. Table I shows results for Nd concentration determinations in a fuel analyzed repeatedly through the entire procedure; these data
show that the relative standard deviations of 148Ndand 145+146Nd concentrations were 0.8%. Figure 3 shows mass spectra of Th and U from element 21 spiked with 230r’h;the peaks are well-defined and interference free. A large 233Upeak was present in all of these samples. Plutonium appeared in the U spectrum during the initial heating period, but it evaporated quickly and thus was not an interference. The relative standard deviation of a Th or U isotope ratio was 0.296, determined from six replicate analyses through the entire procedure. Table I includes results for Th and U concentration determinations in a typical fuel analyzed repeatedly through the entire procedure. These data indicate that the relative standard deviations of Th and U concentrations were 0.7 % and 0.6 % , respectively. Effective Fission Yields. Effective yields (Y) for the fuel samples were estimated from the following relationship: Y = x3Y3 x5Y5 (1)
+
where Y3and Y5= la+16Nd (or other fission monitor isotope) fission yield from 2a3Uand 236U,respectively (Table 111, and x 3 and 3c5 = fraction of fissions in 233Uand 235u, respectively. The reactor physics computer code LATREP (19) was used and 236U. From to estimate the fraction of fissions in 233U
ANALYTICAL CHEMISTRY, VOL. 56, NO. 4, APRIL 1984
6Q9
Table 111. Estimated Fission Fractions and Effective Yields for BDL 417 AAX effective av flux
fission fraction fuel ring
outer middle inner
u
Z35U
u3
145+146
1.548 1.581 1.589
6.605 6.684 6.704
0.696 0.777 0.798
0.304 0.223 0.202
effective yield Nd 14*Nd
lo1* neutrons cm-2 s-l 13.4 7.88 6.01
Table IV. Atom Percent Fission in Fuel Bundle BDL 417 AAX atom percent fission mass spectrometric
ring
element
outer
4 9 14
middle
21 24 28 30 32 33 35 36
18
inner
% RSD
145t146Nd
2.32 2.17 2.28 2 -27 1.47 1.37 1.40 1.42 1.07 1.10 1.11 1.11 1.1
145t146Nda
2.29 2.18 2.27 2.27 1.45 1.37 1.43 1.40 1.07 1.10 1.12 1.11 0.8
148Nd
HPLC 139La
LATREP
2.38 2.25 2.35 2.36 1.48 1.41 1.43 1.49 1.08 1.12 1.13 1.13
2.32 2.21 2.28 2.31 1.41 1.37 1.40 1.45 1.08 1.12 1.12 1.07 1.2
2.17 2.17 2.17 2.17 1.42 1.42 1.42 1.42 1.15 1.15 1.15 1.15
1.1
Initial heavy element content estimated from weight and initial composition of fuel. input parameters such as the initial composition of the fuel, the physical arrangement of the fuel elements, and the fluxtime history, the code calculated the following: the energy and spatial distribution of the flux throughout the bundle, the effective cross sections and reaction rates of the actinide and fission product nuclides, the total number of fissions and captures in each nuclide summed over the entire irradiation, and the burnup. An important feature of the code was the ability to simulate the detailed flux-time history, because for this fuel the branching of 233Pa,and hence the production of was strongly dependent on the flux level and on reactor shutdowns. Optimum fit of the simulation to the irradiation was obtained by scaling the fluxes by a factor that yielded the best match to the U and Th isotopic data. The fission fractions were estimated from the initial and final abundances given for the heavy nuclides and are listed in Table I11 along with the corresponding effective fission yields. The combined fission fractions in nuclides other than 233U and 23sU(e.g., 23BPu)were
