Table VI. Gas Stream Flows (Basis: 100 lb/hr Coal Feed)]
Gas composition,% Stream
Total flow, scfh
H2
CHI
C2Hs
1 2 3 4
2700 2400 2090 610
67 64.5 74 43
2.5
1.0
CO
CO2
33 22.5 26 57
9.5
of 2Hz:lCO. Using an oil yield of 3 barrels per ton of coal, the synthesis gas (0.75Hz:lCO) consumption is calculated to be 4000 scf per barrel of oil. In comparison, based on the similar data for the hydrogen run at 450 OC and 15 min (Table V), the hydrogen consumption would be 4700 scf per barrel of oil when the feed gas is hydrogen. It should be noted, however, that the consumption of synthesis gas depends a great deal on the H2/CO ratio, the amount of water, residence time, temperature, and other variables affecting the water-gas shift reaction. Thus, depending on the use of off-gas, one can optimize the consumption of hydrogen and carbon monoxide accordingly. The use of synthesis gas as the feed gas may be applied to Synthoil, H-Coal, Consol, and other catalytic hydrogenation processes. The advantages are obvious: (1)the high cost of hydrogen production is saved, and the thermal efficiency increased; (2) the synthesis gas usage can be optimized, and the supply can be obtained a t less cost than hydrogen from the gasification of the char produced in the process; and (3) the off-gas, after scrubbing out carbon dioxide and hydrogen sulfide, is a low-Btu gas. With the H2/CO ratio of about 3, it is suitable for use in methane production or methanol synthesis. If one recycles the gas, only a small bleed stream is
necessary to keep CHI level down. The bleed gas could be burned for process heat. Conclusions High sulfur bituminous coal can be liquefied and desulfurized by hydrotreating with synthesis gas a t 3000 psi and 425-450 “C in the presence of cobalt molybdate-sodium carbonate catalyst, steam, and recycle oil. The sulfur content and the viscosity of the oil product both decrease with the amount of hydrogen consumed whether synthesis gas or hydrogen is used as the reactant gas, but less total hydrogen is required for the same oil product quality when synthesis gas is used. The synthesis gas consumption can be conveniently optimized, the cost of hydrogen production is saved, and the off-gas can be optionally burned as a low-Btu gas or utilized for methane production or methanol synthesis. Acknowledgment The authors thank Dr. F. W. Steffgen for his valuable discussions. Literature Cited Akhtar, S..Friedman, S., Yavorsky, P. M., AlChESymp. Ser. No. 737, 70, 106 (1974). Appell, H. R., Miller, R. D.,Wender, I., presented before the Division of Fuel Chemistry, American Chemical Society, April 10-14, 1972. Fu, Y. C..Illig. E. G.. Metlin, S.J., Environ. Sci. Tech., 8, 737 (1974). Pitchford, A. C., U S . Patent 3 728 252 (1973). Vernon, L. W., Pennington, R. E., U S . Patent 3 719 588 (1973).
Receiued for review July 25,1975 Accepted March 6,1976
Presented before the Division of Fuel Chemistry, 169th National Meeting of the American Chemical Society, Philadelphia, Pa., April 6-11,1975. Mention of commercial products is for identification only and does not constitute endorsement by the U S . Energy Research and Development Administration.
Improvement of Fission Products Decontamination through Dibutyl Phosphate Masking in a Purex Process Takeshi Tsujino,* Tadaya Hoshino,’ and Tetsuo Aochi ReprocessingDevelopment Laboratory, Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken. Japan
A modified method adding inactive zirconium or hafnium ion is proposed for the purpose of improving decontamination factors (DF) of fission products (FPs) in a primary separation process in the reprocessing of highly irradiated nuclear fuels. The feasibility of this concept was experimentally proven by both batchwise extraction and process studies with miniature mixer-settlers. By the addition of inactive zirconium, actual improvement factors of DF of about 4 and 2 were obtained for Zr-Nb-95 and Ru-Rh-106, respectively, in the extraction process with the solvent added DBP, simulating solvent exposure to an FBR fuel processing (-2 W-h/l. of solvent). Inactive hafnium ion would be an interesting element which could be usable not only for DF improvement but also for the criticality control as a soluble neutron poison.
Introduction Solvent extraction with tri-n-butyl phosphate (TBP) is now the only practical method for the reprocessing of irradiated fuels. I t is expected a t present that the modified Purex
Research and Development Section, PNC, Japan. 396
Ind. Eng. Chem., Process Des. Dev., Vol. 15, No. 3, 1976
process could be adaptable to the processing of highly irradiated fuel as from an FBR. In this case, one of the process problems to be solved is that there will be radiation damage to the extractant and a decrease in the decontamination factor (DF) of fission products (FPs) is feared, due principally to the formation of dibutyl phosphate (DBP) (Tsujino, 1968).
Although an effective solution against solvent exposure is being given practically by use of a rapid contactor (ORNL staff, 1970), some chemical solutions for the T B P system have also been proposed such as (1)separation of fission products before extraction (Lopez-menchero et al., 1967), (2) the addition of fluorine ion in the scrubbing, especially for zirconium decontamination (Lefort et al., 1968), and (3) inactive zirconium addition in the dissolution or extraction step. The outline with respect to item (3) was presented briefly by the present authors (Tsujino, 1970a), and comparisons of the latter two proposals then followed (Swanson, 1971). It is considered that process concept (2) is more effective for DF improvement than concept (3), but the problem of corrosion with the fluorine ion has to be solved in the extraction and also in further waste treatment. In the present paper, the effect of inactive zirconium addition is shown experime~tsllyin detail on the simulated solvent exposure to an FBR fuel processing. The addition of hafnium ion is also proposed, not only for the improvement of DF for FPs, but also for controlling criticality as a soluble neutron poison.
Principle It is well known that the reprocessing solvent T B P decomposes by radiation and/or hydrolysis to form principally DBP and monobutyl phosphate (MBP) which forms a stable complex with metal cations Mn+ as
-
+
[n-DBP or (n/2)MBP] M n + [(DBP),M or (MBP),/2M']
+ nH+
(1)
The stability of the complex increases in the following sequences (Hahn et al., 1964) Zr4+ > Pu4+ > U022+ > RuN03+ > Nb5+
(2)
If the M"+ is a radioactive cation such as 95Zr4+or 95Zr02+, it is clear that DF of 95Zrfrom U or Pu decreases by the above complex formation in the solvent. When an inactive metal cation such as Zr4+, Zr02+, or H F + is added with a higher concentration than that of 95Zrin the extraction or dissolution step, it forms a stable complex with DBP or MBP followed by eq 1 and an improvement of DF for FPs is expected by the "DBP masking effects", theoretically shown by the competition of these equilibrium reactions. Another effect of soluble neutron poison is also predicted when the added inactive cation such as HF+possesses a large absorption cross section for the neutron. Experimental Section Based on the principle mentioned above, simulated flow sheet studies of FBR fuel processing were performed by miniature mixer-settler banks in order to clarify the effect of inactive Zr addition. Batchwise distribution experiments were also undertaken to obtain fundamental data. Reagents a n d Apparatus. TBP, the purified product for reprocessing usage, from Daihachi Chemicals Inc., Japan, was used without further purification. DBP was obtained from Kanto Kagaku Chemicals Inc., Japan, and the ratio of DBP/MBP was 95/5. Dodecane was imported from Progil, France, and meets the specifications of the French AEC (CEA). Plutonium nitrate was supplied by NUMEC (U.S.A.), and was purified by an ion-exchange method. Uranyl nitrate was the product of the Allied Chemical and Dye Corp. (U.S.A.) and it passed the specification of the American Chemical Society. The oxalate of Zr-Nb-95 was imported from the ORNL (U.S.A.) and was converted to nitrate form by the treatment with "03 and H202. Ru-Rh-106 in the form of nitrosyl trinitrate was imported from the Radiochemical Center of Am-
Figure 1. Actual chemical flowsheet obtained for improvement of DF's by DBP masking effect.
ersham (U.K.) and was used as received. Gross fission products were obtained from the 1AF solution of the 3rd hot test (HR-3) in the JAERI Reprocessing Test Plant (Ishihara et al., 1972). Other reagents than those mentioned above were of extra pure quality produced in this laboratory. The apparatus was the same as that used in the previous work (Tsujino, 1970a). One miniature mixer-settle unit (stage volume 25 ml, volume ratio of mixerhettler = 1/3; 16 stages) made by "SONAL" (France), was installed in a glove box which also utilized home-made chemical pumps. Experimental a n d Analytical Methods. An ordinary batchwise extraction was performed at room temperature (-20 "C) in a 5-ml polyethylene tube. The mixing and settling times were 2 and 6 min, respectively. They correspond to the residence time in the mixer-settler bank. According to the chemical flowsheet shown in Figure 1, extraction runs were also carried out in the same way as in the previous work (Aochi, 1971). Based on the primary separation process in the JAERI Reprocessing Test Plant (Ishihara, 1972),DBP is added to the fresh solvent with a concentration of M, which corresponds to a solvent exposure of about 2 W-h/l. of solvent. The concentration of the added inactive ZrO(NO& was selected to be low in order to maintain a phase stability. Plutonium was ordinarily analyzed by @-countingand by cerimetry when necessary. The concentration of uranium was determined by K2Cr207 titration or by spectrophotometry with dibenzoil methane or with NH4SCN. Fission products (FPs) were analyzed by y-counting and the species of FPs were determined by y-ray pulse height analysis when necessary. Nitric acid was analyzed by NaOH titration masking U with NazC204. The concentrations of TBP and DBP were determined by the acid equilibrium method and by the Zr-95 extraction method, respectively. Results a n d Discussion Fundamental Batchwise Extractions. As shown in Figure 2, the distribution ratios (Kd) of radioactive zirconium (Zr-Nb-95, Zr*) decrease with increasing concentration of inactive zirconium (ZrO(N03)2).This tendency is noticeable a t high concentrations of DBP as expected from the DBP masking effect. At low concentrations of DBP, the dilution effect of Zr* is predominant and the decreasing tendency of Kd is not so large. The y-spectrum change of organic phase after extraction shows that the addition of inactive zirconium does not influence the behavior of Ce-Pr-144, but it is effective for the improvement of the decontamination factor (DF) of Ru-Rh-106 as well as of Zr-Nb-95, especially at high concentrations of DBP. The salting-out effect due to inactive zirconium theoretically increases the Kd of extracted metal. No noticeable Kd Ind. Eng. Chern., Process Des. Dev., Vol. 15,No. 3, 1976
397
20.30 00
0
[ZrO ( NO,),]
mol/(
Figure 2. Effect of inactive Zr addition on Kd of active Zr* [Zr*]i; -0.92 pCi/l. at 20 "C, 30 vol% TBP--dodecane.
increase of FPs, U, and P u due to the salting-out effect was observed below a concentration of 0.3 M ZrO(NO3)z. A decreasing tendency of Kd for P u and U is observed by the radiolysis of T B P (Tsujino, 1966) and its degree of change would be negligible below a concentration of M DBP in the actual process. I t is estimated that the addition of inactive zirconium promotes emulsion formation, especially a t high concentrations of DBP. By the preliminary examination using a feed and ZrO(NOa)z, solution containing 250 g/l. of U, 3N "03, it was found that the phase is stable below 0.02 mol/l. of ZrO(NO& under DBP concentrations between 0 and 0.13 M. Countercurrent Extraction. Through the above preliminary test was shown the feasibility for improving the DF of FPs, especially of Zr-Nb-95, by the addition of inactive zirconium. This process concept has actually been proven in the JAERI Reprocessing Test Plant with low burn-up fuel (400-600 MWD/T). From the results obtained above, it is estimated that the addition of inactive zirconium is expected to increase the DF more noticeably in the reprocessing of high burn-up fuels. In order to prove this prediction experimentally, the simulated solvent to FBR fuel processing is prepared by adding DBP to a concentration of 10-3 M and its flowsheet study was carried out thereby. Chemical Flowsheets and Concentration Profiles. An actual chemical flowsheet observed a t the steady state is shown in Figure 1.In the Pu 10 run, which is the base line run, neither DBP nor inactive Zr was added. Effects of DBP and inactive Zr addition are revealed by the P u 20 and the Pu 30 runs, respectively. By the present flowsheets, 5-6 h operation was required to attain the stationary state. No phase trouble as emulsion or third phase formation was observed during the process run. As shown in Figure 3, concentration profiles for Pu, U, and "03 at the stationary state are normal ones as expected and no large differences were observed among those three runs. As shown in Figures 4 and 5, remarkable differences were found in the concentration profiles for Zr-Nb-95 and RuRh-106. In the case of DBP addition ( P u 20 Run), fission products, especially Zr-Nb-95, are extracted into the solvent by the DBP complex formation and the tailing behavior, which appears in a general scrubbing step, is not clear in the scrubbing. By the addition of inactive zirconium (Pu 30 Run), abnormal profiles, obtained in the P u 20 Run, could be converted to normal ones and lower organic concentrations (1AP) of fission products than that of the P u 20 Run were observed. These results clearly reveal the DF improvement of fission products (FPs). This improvement could be elucidated by the "DBP masking effect" which is shown above. As estimated from the 398
Ind. Eng. Chem., Process Des. Dev.. Vol. 15,No. 3, 1976
I
Figure 3. Concentration profiles of plutonium, uranium, and nitric acid (Pu 10, 20, 30 runs). I
I
I
I
E
1 -
l
IAF 1
1
,
1
1
/
1
1
,
stability of the DBP complex, the effect on the behavior of Zr-Nb-95 is larger than that of Ru-Rh-106. Behavior changes of other FPs are not expected from the results which were obtained during the above preliminary test. Material Balance and Decontamination Factors. As summarized in Table I, material balance is generally satis-
10-2-
10
lo-'
I1./
0
t
I
-
[Zr-95), : 092 ,uc,/me o t Room Temp I-ZG'CI
I
1 2 3 4 5 6 7 E 9 1 O i l ! 2 ! 3 1 4 1 5 1 6 ~
%ge
nos.
IAP
Figure 5. Concentration profiles of Ru-Rh-106 in organic phases. Table I. Material Balance, 70
(
1AP) x 100 1AF +
Run no. Species U Pu "03
Zr-Nb-95 Ru-Rh- 106
Pu 10
Pu 20
Pu 30
98 99 99 109 105
99 99 106 120 107
99 97 102 130 118
Table 11. Decontamination Factors of FPs from U and P u (lAF/lAP) __
~~
~~
~
~~
Run no.
Zr-Nb-95
Ru-Rh- 106
Pu 10
-41
20 30
-5
-53 -25 -43
-19
1o - ~
IO+
IO-'
[DBP]
lA P
factory in all the runs. Some unbalance is observed for FPs, which may be due to the plus error in their analysis of low concentrations with large quantities of U and Pu. Recovery of U and P u is observed to be above 99%. This indicates that any troublesome influence for the U and P u recovery is not apparently found by the addition of DBP and inactive zirconium. No difference in the loss of U and P u was found among three runs (0.05%in U and 0.2% in Pu). As shown in Table 11, improvement of DF, which is lowered by the DBP addition, was actually observed upon the addition of inactive zirconium for Zr-Nb-95 and Ru-Rh-106, respectively. In the base line run ( P u 10 Run), lower values of D F are observed than the ordinary ones in a Purex process. This may be due to low activity of FPs which is used initially. With the addition of inactive zirconium, the concentration of unnecessary impurity increases in the system. I t is estimated that the impurity is easily separated from the product in the further second and third purification cycles.
Wott-hr/
