Cation Exc ange Removal of adioaetivitv from astes J
H. GLADYS SWOPE AND ELAINE ANDERSON1 .4rgorane National Laboratory, Lemont, I l l .
Laboratory studies on the use of a cation exchanger for the removal of mixed fission produce activity from tap water are discussed. A sulfonic acid-typc cation exchanger can remove 75 to 80yo beta-gamma activity over long periods of time. The resin capacity for removing radioactivity t o this extent is >260,000 gallons per cubic foot whereas the theoretical capacity is 6000 gallons per cubic foot for tap water containing 300 p.p.m. total solids and 85 p.p.m. of total hardness as calcium carbonate. The 6000 gallons per cubic foot is the point of break-through of total hardness-i.e. leakage of calcium and magnesium ions. This means that low level wastes requiring a removal factor of no more than 80% can be economically processed by a convenient method. The resin can be regenerated or burned. Experiments have been conducted to determine the best operating conditions. A pH of 2.5 provides the best decontamination without filtration. A pH above 8 provides better decontamination but the waste must be filtered to remove precipitated hydroxides or the ion exchange column clogged. A flow- rate as high as 10 gaIlons per cubic foot per minute does not reduce the decontamination obtained. Fission product analysis indicate that ruthenium and cesium are the main radioactive species present in the effluent.
T
HIS paper deals solely R ith the use of a high capac
exchanger of the styrene base sulfonic acid type f moval of niixed fission product radioactivity added to laboratory t a p water. The results of routine processing of laboratory wastes through a cation exchanger are reported for a 6-month period. Other work of inteiest in this connection has been reported by Ayres ( 1 ) . No attempt has been made to survey the literature dealing Tj-ith the removal of radioactivity by inn exchange since reviews of the literature of ion exchange appear each January in Analytical Chemistry ( 2 ) and in INDUSTRIALa m ENQINEERIA-i~ CHEMISTRY (3). To investigate all the variables in so heterogeneous a solution as a general laboratory waste Jr-as obviously impossible since the waste is never the same from day to day or even from minute to minute. Therefore, a standard feed was used which was laborator3 t a p water "spiked" with mixed fission product radioactivity If a cation exchanger is used for water softening, regeneration of the exchanger is required when the presence of calcium arid magnesium ions indicate the end of a cyrIe-i.e., hardness breakthrough. This point is defined in this paper as break-through. B fair removal of gross beta-gamma radioactivity ('i5yO) from mixed fission products was obtained after this point for long periods of time. After adjusting t a p water to p H 2 5 and spiking i t with mixed fission product radioactivity, each cubic foot of the resin removed 76y0of the activity from 260,000 gallons of water. TYPE A N D K I N D O F RESIN
The stabilities of 12 different resins toward various chemical agents and organic solvents R-ere investigated in the first experiments (1949). Not all the resins investigated are still on the market, but Nalcite HCR (Dowex 50) had the highest capacity 1
Ill.
78
Present address, International Minerals and Chemical Corp., Chicago,
for rcmoval of radioactivity and appeared to be one of t,he more stable of the cation resins investigated. These preliminary experiments, however, showed t,hat a cation resin of the polystyrcnedivinylbenzenesulfonic acid t,ype was necessary for good removal of mixed fission product radinact,ivity. The capacities, p H range, and fundamental propertics of various resins are given in boolrs on ion exchange by Kunin and Myers (4)and by Nachod ( 6 ) . CAPACITY OF RESIN FOR 4CTIVITY REMOVAL
For the laboratory experimental work, t a p water was used l o which mixed fiseion product radioactivity was added. The laboratory water is obtained from deep wells and is softened by the lime-soda process in a water treatment plant. Hardness averaged 85 p.p.m. as calcium carbonate, total solids 300 p.p.m., and p H 9.0. The softening capacity of Kalcite HCR-Pl;a+ resin ia 6000 gallons per cubic foot for the laboratory t a p water. I n normal water softening practice the resin would be regenerated a t this point. However, since these exchange resins were to be used for the removal of radioactivity the prevnce of nonradioartive calcium and magnesium in the effluent \vas immaterial. Therefore, from the standpoint of economy the ability to remove activity for a long time without regeneration is an asset. For low level laboratory wastes requiring a decontamination factor of only 2 t o 5 (50 to 80% removal), the use of cation exchange is much less expensive than evaporation. Decontamination factor (D.F.) is defined as counts/min./mI. in feed ______ cnunts/min./ml. in effluent and per cent removal as counts/min./ml. in feed - counts 'min./ml. in effluent x 100 counta/min./ml. in feed
INDUSTRIAL AND ENGINEERING CHEMISTRY
)
Vol. 47, No. 1
Exchange-
- T o n
TABLE I. REMOVAL OF RADIOACTIVITY BY HYDROGEN AND SODIUMCATIONRESIKS Feed, Counts/ Min./M1. HCR-Na
A Volume, Gal./Cu. Ft.
5,s00a 7,300 2 13,100
HCR-H
Weighted Average
5,800a 7,300 Z: 13,100 a
Weighted
Average Hardness break-through.
.
+
+'
Effluent Counts/ Mi;./Ml. + HCR-Na
HCR-H
Alpha Activity 14 25
250 326
328
8 23
292
293
17
20
x
3.5 x 3.8 x
248
105
106
3.7 X 108
3.5 x 10s 3.8 x 105
2.4
6.2 x 104
Beta Activity 2.2 x 104 6.7 x 104
3.6 X 106
4.5 X IO'
4.7 X 104
104
REWOVAL OB
Before Hardness Break-through -~ Throughput Over-all Activity volume, gal./cu. ft. 5900 6200 6100
2.5 4.0 5.5 8.0 8.0 9.5
4900 4800 5500 6100
6100
5900
integrated
D.F. 10.4 13.6 13.0
14.5 10.7 2.8 7.0 6.0 18.7
removed,
%
90.4 92.6 92.3 93.1 90.6 64.8 85.7 83.4 94.7
~
+
Removed, % HCR-H+ HCR-Xa
31 14
18 13
96.8 92.9
17
16
94.2
15 6.1
16 5.7
93.1 83.7
92.0 82.4
8.2
7.7
87.8
89.7
+'
94.4 92.4 ,
93.2
The wastes vary in p H from 2 to 10 with the majority between 2 and 4. Therefore, to find out what effect p H has on the capacity of the resin for the removal of radioactivity, p H 1.8, 2.5, 4.0, 5.5, 8.0, and 9.5 were investigated I n order to have a standard feed in all of the ion exchange experiments, laboratory t a p water was used. The t a p water has a p H of about 9, a hardness of about 85 p.p.m. as calcium carbonate, and a total solids content of 300 p.p.m. The calcium content as C a t * is I 1 p.p.m., and the magnesium content as Mg++ is 16 p.p.m. The t a p water was adjusted to the p H desired by the addition of nitric acid or sodium hydroxide and was then spiked with about 2-year old mixed fission products from dissolver solution. As the removal of low level radioactivity was the prime objective, about 1000 counts per minute per ml. were added as the spike. The feed was made in 15-gallon stainless steel drums that delivered the spiked feed to 4-liter glass feed bottles. The total activity was determined on these smaller batches daily. Flow rate was 2 gallons per minute per cubic foot (13 ml. per minute). Experiments were continued a t all pH's after hardness breakthrough in order to determine the capacity of the resin for removal of radioactivity. Table I1 shows the volume of feed treated to hardness break-through, the total amount passed through the columns, and the removal of activity. Contrary to expectations, no plateaus were observed in the decontamination obtained as the volume of liquid passing through the columns increased. Instead, fairly constant decontamination factors were obtained after hardness exhaustion. This was true even when 260,000 gallons per cubic foot of resin were passed through the column. At a p H of 2 5 decontamination factors of 3 (66%) were still being obtained when the column was shut down.
As cation resins are usually marketed in the sodium form, it would be more economical t o use this form for waste processing, provided as good removal of activity could be obtained with the sodium as with the hydrogen form. The sodium form gave only slightly less decontamination than the hydrogen form. For these experiments and in all subsequent ones, unless otherwise noted, Xalcite HCR resin was added to a 6/g-inch diameter borodicate glass column, 24 inches long. The resin is measured by adding i t t o distilled water contained in a 100-ml. graduated cylinder. The resin was allowed t o settle in the water until i t reached a volume of 49 ml. ( 3 cubic inches). It was then washed into the glass column. The resin was held in place by a 200-mesh stainless steel screen. The resin was backwashed with distilled water for 1/2 hour at such a rate as to expand it 50% of its volume. This was done for classifying the resin and removing any air pockets. The feed used was t a p water adjusted t o p H 2.5 with nitric acid and spiked with about 3 X lo5 counts per minute per milliliter of 2-year old mixed fission beta activity. The flow through the column was maintained a t 13 ml. per minute equivalent t o 2 gallons per minute per cubic foot. The effluent was collected in 13 fractions before hardness break-through, equivalent to 38 liters, and in 27 fractions after break-through, equivalent to 48 liters. Samples were analyzed for total hardness using the Versene method and for alpha and beta radioactivity. The activity was counted on a Nuclear Measurements Corp. proportional counter, giving a 50% yield for alpha and a 62% yield for beta.
ON
+
EFFECT O F pH
COMPARISON O F SODIUM A N D HYDROGEN FORMS OF RESIN
TABLE 11. EFFECT OF p H
Integrated D.F. HCR-H HCR-Na
Table I gives the results for alpha and beta radioactivity both before and after hardness break-through. The hydrogen form was superior to the sodium form for the removal of alpha activity before hardness break-through.
Laboratory experiments showed that an equivalent of 260,000 gallons of t a p water could be processed per cubic foot of H C R resin before the operation was stopped, because the unit was getting so radioactive that it required shielding. A decontamination factor of 3 (66% removal) or a greater one had been maintained for several months using the unit daily. When a removal of only 65 to 80% mixed fission products was requirrd, the resin could be used without regeneration for the equivalent of 44 cycles.
Feed pH 1.8 2.5 2.5
+
GROSS BETAACTIVITY BY CATIONEXCHANGE
After Hardness Break-through Total throughput Over-all Activity volume, integrated removed, Remarks gal./cu. ft. D.F. % 29,000 6.1 80.2 Average of 2 runs 21,400 5.0 80.0 Average of 2 runs 260,000 4.1 75.6 Flow rate, 10 gal./min./ cu. ft. 20,800 5.6 82.2 With Alsop filter 26 500 3.8 73.5 20 400 2.3 57.4 Average of 2 runs 28,200 2.4 58.4 24,300 3.0 67.2 With Alsop filter 12,300 6.3 84.2 With micruinetallic filter
:
~
January 1955
INDUSTRIAL AND ENGINEERING CHEMISTRY
79
'
pH
i
1.5
u
T O T A L p IN FEED
pfi
,
5.5
,
r
10,ow
low 100
10
more effectively and thereby t o increase the decontamination obtained. Some effluent fractions from the second runs made a t p1-l 1.8, 2.5, and 5.5 were analyzed for ruthenium, cesium, strontium, total rare earths, and zirconium. Since the mixed fission product activity used in these experiments was about two yesre old, about 70% of the gross bet,a activity was derived from these elements. Break-throug!i of a specific species is the point where the amount of the species exceeds t,he tolerance lev'el in water or meets the feed line-whichever point conies first'. Eardness break-through is the point where calcium and magnesium ions are first found in an effluent fraction. For example, the tolerance level for ruthenium-106 in mater is 0.1 microcurie per milliliter (pc/ml.) whereas t'he tolerance for strontlium-90 is 8 x 1 0 - 7 pc/ml. In no case does the effluent fraction exceed the tolerance level for ruthenium so that break-through is determined when the ruthenium in the effiuent is equal to the ruthenium in the feed. For etrontiuiii-DO, however, the amount of stront,ium iii t,he effluent fraction equals the tolerance level long before it equals the amount of strontium in the feed. Table 111 was prepared on this basis.
1
10,000
TABLE 111. BREAK-THKOUGH o r FISSIOX PRODUCTS O N CATIOX COLUlfN
lMxi K
e
Feed pI-1
100
Hardness Cesium Ruthenium Strontium Rare Earths
10
I
a
Break-through Volumes, Gal./Cu. Ft. 1.8 2.5 5.5
6,000 3,000 5,600 4,300 > > > 2 4 , OOOQ
6,300 5,800 9,100 4,000 > >>22,oooa
5,100 7,000 4,300 4,300 > > >21,000"
Last fraction taken, t o t d rare earths were retained on the ion exchange
column. 100
IO
1 C."'
TPLERANCE IN WATE
1W.WO
10,oM
1000
1W
10
1 LITER I lWOGAL./FT?
10 30 50 70 PO 110 130 10 30 50 70 W 110 1 3 0 10 30 50 70 90 li0 130 1.5 3.07.6l0.613.716719.8 1 . 5 3 . 0 7.610.613.716.719.81.5 IO 7.610.613.7lb719.8
THROUGHPUT VOLUME
Figure 1.
Distribution of Fission Products in Effluents from Cation Exchange Columns
Feed I
Tap water adjusted to pH shown and spiked with -xed fission product activity 5/8-inch diameter borosilioate glass cantaining 3 cubic inches Nalcite HCR-Na+ Resin 2 gallons per m i n u t e per cubic foot of resin
Columns:
Flow rate:
REMOVAL O F SPECIFIC FISSION PRODUCTS
If 65 to 80% gross alpha-beta activity can be removed from large volumes following hardness exhaustion in a cation exchange column, these columns have practical value in industries and other organizations concerned with the disposal of small quantities of low level radioactive wastes. Also, if it were found that certain radioactive species were eluting more than the others perhaps new resins could be developed to remove these species 80
The extremely low level of activity used made the analyses of the single fission products very difficult; therefore, effluent saniples mere concentrated 20: 1 when enough sample was available, otherwise 10: 1. Figure 1 plots the results obtained. The conclusions t o be drawn from these data are
1. Cesium can be retained on a cation column for :I period prior to hardness break-through. As the p H is increased from 1.8 to 5.5 this period increases. However, cesium completely breaks through at the same time a s the calcium and magnesium. The normal capacity of Kalcite HCR is 30,000 grains-for cesium it i F about 15,000grains. 2. Although ruthenium is not retained satisfactorily on a cation exchanger, more is retained a t p H 2.5 than a t 1.8 or 5.5. 3. A cation exchanger a!one, 11hen using a high capacity styrene base exchanger, removes strontium-90 satisfactorily to hardness exhaustion of the resin. Subsequent work using a much higher level of activity has confirmed these results. A considerable amount of strontium is still removed past hardness exhaustion of the resin since the amount of strontium in the effluent mas just starting to approach the feed line as the experiment was shut down The tolerance level for strontium-90 is so low that 1.1 counts per minute per nil. (8 X 10-7 pc/ml.) exceeds the tolerance. Since the amount of strontium was so low in these effluents, the counting errors were sometimes greater than 2007, and the analytical errors greater than 100%. I n the case of the total rare earths not enough effluent fractions were analyzed to be sure of the slope of the curve, but in no caw did the amount of rare earths in the effluent fractions even approach the feed line which was considerably below the tolerance level. A few zirconium analyses were made, but results were too erratic to make even an educated guess as t o the trend.
INDUSTRIAL AND E N GINEERING CHEMISTRY
Vol. 41, No. I
-Ion TABLlC
Feed
11'.
EFFECTOB' FILTRATION O N RADIOACTIVITY REMOVAL BY CATION VARIOUSpH's A Volume Throughput, Gal./Cu. Ft.
2 . .5
filter Element Alsop
2.5b
None
8.0
8.0
None Alsop
9.5
Micrometallic 200 mesh
PH
6,100 14,700 6,400 15,100 6,100 6 , 100
8,400 9,900 5,900 ? ,300 a . 000
Average 8-7 Bctivity, Counts /Min./Ml. Final Feed Filtrate effluent 1 . 2 x 108 83 950 169 89 1 . 2 x 108 1 . 1 x 10' 227 262 ... 1 . 3 X 10s 233 1 . 4 X IO' 408 1 . 2 x 103 ... 266 830 493 920 308 49 5 . 0 X 10' 569 274 289 880 268
Exchange EXCHANGE .4T
Integrated D.F. Ion exchange
Filter
Over-all 14.5' 5.6
...
...
14.1a 5.2
...
...
5.0n
...
,..
...
1.7
3.0 8.8 3.0
...
6 Oa 2.9
...
1 .9 6.3 2.1 1.1
3.3 18.6O 18.2
3.3
Hardnesx break-through. 6 Average of two runs.
(1
The tolerance levels (6) for the radioactive species considered axe rc/ml. Rutheniuni-I 00 Cesium137 Strontium-90 Cerium-144
0 1 1 5 x 108 X 10-7
EFFECT O F CROSS LINKAGE
0.04
EFFECT OF FILTRATION PRIOR TO CATION EXCHANGE
As this experimental work was in progress the question v-as often raised whether or not filtration alone would have been as effective as ion exchange; perhaps the resin was acting solely as a filter. At p H 2.5 most of the ions should be soluble and filtration should not be beneficial. At the other pH's filtration should help remove any of the ions precipitated as the hydroxides. These assumptions were confirmed by the experimental work. At p H 8.0 very little was gained by filtration alone but at 9.5 there was enough precipitate to filter before passing through the ion exchange column. Filtration alone accounted for most of the decontamination obtained after hardness break-through. The results of the fission product analyses showed that the activity removed after hardness break-through was the total rare earths precipitated as their hydroxides (Table IV). EFFECT OF FLOW RATE
Preliminary experiments had indicated that there was no appreciable difference in the decontamination obtained after hardness exhaustion when operating in the range of 1 to 14 gallons per minute per cubic foot. Processwise it would be expedient to operate a t a faster flow rate than that usually recommended for Kater softening, namely 2 gallons per minute per cubic foot. Therefore, two columns containing HCR-Na+ resin were operated a t 2 and 10 gallons per minute per cubic foot. The same feed was used and 21,400 gallons per cubic foot of tap water adjusted to pH 2.5 with nitric acid and spiked with mixed fission products t o about 1000 counts per minute per ml. was passed through the columns. (Table V). Run 2 was another experiment made at the regular flow rate, namely 2 gallons per minute per cubic foot, and using the same type feed. Both before and after hardness break-through there was almost no difference in
TABLE v. FLOW RATE EFFECT Tiow Rate, Gal./Cu. Ft. 2 10 2
Capacity, Gal./Cn. F t .
the decontamination obtained. Therefore, a flow rate as high aa 10 gallons per minute per cubic foot is satisfactory for removal of beta radioactivity when the feed is a t pH 2.5.
REMOVAL O F GROWBETARADIOACTIVITY T l l H O U G B CATION EXCHANQE: RESINCOLUMN
ON
Decontamination Factors Before hardness After hardness break-through break-through 13.9 18.8 12.2 6,200
Nalcite HCR is polymerized styrene with divinylbenzene a8 the cross-linking agent, the nuclear sulfonic groups supplying the only ionic group. B y changing the amount of cross linking the porosity of an ion exchange resin is varied-the less cross linking the greater porosity. The results indicate that a better removal of cesium and ruthenium is desirable; perhaps a resin with a different cross linkage would help. Hence cross linkages of 1, 4, 8, and 16%, all 50- to 100-mesh size were tried. Standard Nalcite H C R resin is 8% cross linked and 20 to 50 mesh. The resins were all measured as in the previous experiments; 3 cubic inches were used. I n this experiment, however, the resins were supplied and used in the hydrogen form. Since a pH of 2.5 gave the best over-all removal of activity in earlier experiments, the t a p water for these experiments was adjusted t o this p H and then spiked with mixed fission product activity with about 1000 counts per minute per ml. (62% yield). An attempt mas made t o maintain a flow rate of 2 gallons per minute per cubic foot but this was impossible with gravity flow because of the small mesh size (50 to 100). Flow rates were, therefore, approximately 1.5 gallons per minute per cubic foot or less. In addition t o analysis for total beta activity, an analysis wm made for the amounts of cesium and ruthenium past the hardness break-through point in the effluents. With the standard resin, 8% cross linked and 20- t o SO-mesh, neither of these active ions was retained after hardness break-through. With the 50 to 100 mesh, only the 16% showed complete leakage a t these points and the 8% produced the least leakage. The individual results are shown in Table VI. The resin with the highest capacity before hardness breakthrough was the 16% cross linked; t h e capacity was equivalent to 9700 gallons per cubic foot. However, the highest decontamination factor (16.1 or 94%) to hardness break-through waa obtained with the 8% cross linkage. After hardness break-
5.3 5.0 4.5 15,200
Over-all
Before hardness break-through
Removed, yo After hardness break-through
6.7 6.2 5.5
92.8 91.6 91.0
81.0 80.1 77.9
21,400
6,200
15,200 ~~~
January 1955
INDUSTRIAL AND ENGINEERING CHEMISTRY
Over-all 85.0 83.8 81.8
21,400 ~~~
81
TABLE VI.
ANALYSIS O F
EFFLUEXTS AFTER
PASSAGE THROUGH
VARIOUSCROSS-LIXKED CATIOSRESIXS
Feed: Tap water adjusted t o pH 2.5 a n d spiked with mixed fission product activity. Beta 82 count/min./ml.", Cesium 11 oount/min./ml.Q, Ruthenium 9.9 counts/min./ml.a Column: S/,-inch diameter containing 3 cubic inches Nalcite HCR-H resin. Flow rate: About 1.5 gal./min.,/cu. ft. Beta Activity G a l /Cu. €t. in Effluents, Cross Counts/hIin./hIl.a Linkage, Throughput Hardness 70 volume, break-through Gross Cs Ru 1 5,200 1820 19.2 6.7 3.3 4 6,700 5200 17.3 7 8 4.0 8 7,:OO 7300 2.7 1.3 2.5 9700 26.8 10 10,300 20.3 11.0 a End window counter, 12% yield. +
obtained; the maximum was 98%, the minimum 17%. results are shown in Table VI11 CONCLUSIONS
A high capacity cation resin of the styrene base sulfonic acid type removed from 17 to 9870 mived fission product radioactivity from laboratory retention tank wastes. For water, an average removal of GO t o 75% iadioactivity is possible with a resin with a capacity of greater than 260,000 gallons per cubic foot of resin when the Tmter has a hardness of 85 p.p.m. as calcium carbonate and total solids of 300 p.p.m. The most satisfactory pH for the removal of radioactivity by cation exchange is 2.5.
TABLEL'II. through, both the 1 and 470 croes-linked resins were superior; the decontamination factor was taice that of the 8% cross-linked resin (Table VII). At the time these experiments 1% ere made the cost of the special cross-linked resins was five times the cost of the regular grade so that no further experiments were made. Today, the price of these crowlinked resins compares favorably with that of the standard grade and further studies may be in order.
RADIO-ACTIVITY REMOVAL BY NALCITE HCR-H +. AT VARIOUS CROSSLINKAGES
Feed: T a p a a t e r adjusted t o pH 2.5 with nitric acid and spiked with mixed fission product activity ea. 10s counts/min./ml. beta (62% yield). Column: &/s-inch diameter borosilicate glass, containing 3 cubic inched Salcite H C R - H " 50 to 100 mesh; bed height 12 inches. BV.
Cross Linkage,
Resin Shrinkage, c7 /U
Flow R a t e Gal. Lib. Cu. Ft. 1.5
Total Throughput, Gal./Cu. Ft. 1,800
38.3
1.5
1,s
0,100 5,200
1.5 1.3
8,000 7,300
1.4 0.82
4.2 0 83 Hardness break-through.
%
1 4
REXIOVAL OF RADIOACTIVITY FROM GENERAL LABORATORY WASTES
16.7 Q
At t h k laborat,ory there are two sewer systems-one for sanitary sewage, t,he other for suspect wastes. The latter contains liquid wast,ee whose radioactivity is below tolerance and any other laboratory wastes. The drains from all laboratory sinks discharge to 1500-gallon retention tanks. These are monitored, and if the activity is below tolerance the v-astes go directly to the drain discharging to the laboratory vaste t'reatment plant, otherwise the wast,es are transported to the so-called effluent processing building. Waste treatment at' the Argonne Satiorial Laboratory is discussed more fully in t>woother papers ( 7 , 8). Over tolerance low level, low solids wastes Irere processed by passage through an Alsop filter and then through a cat,ion column,
6.3 18 a
82
Integrated
D.F.
Removal, %
9.2a
89
5.0 9.6"
90
6.6 16,l a
85 94
8,700 9,700
3.0 12.2"
92
10,600
1.8
44
80
e7
Strontium-90, the most toaic of fission product activities, can be removed by a cation exchanger to hardness break-through. The reduction in decontamination as the ion exchanger is used prior to hardness break-through is due to the leakage of cesium and ruthenium. Total rare earths are removed both before and after hardness exhaustion of the resin.
4 Activityb, Filtered Feed Analysis__ HCR-XaBatch T o t a l Volume Hardness Total 6 Activity, Effluent. Volume, Through Column, as CaC03, Jolids, oounts/min.; Counts/hIin./ Gal. Gal./Cu. Ft. pH p.p.m. p.p.m.0 mi. hI1. 300 920 2 3 0 2420 329 8 038 2 5 22 1000 55 1660 216 1310 Pti 1073 2.7 13.5 20 2120 1405 3 2 1000 111 185 86 1100 4 1c 800 21 1071 12 214 880 1320 102 2109 112 3 1c 29 1000 7.4c 1540 102 2621 49 327 860 1200 144 2.3 3020 1000 281 108 1120 Ill 1.9 3392 330 3680 281 920 150 2.1 3697 170 62 1610 1180 141 4089 1070 106 2.3 134 1420 131 2.8 4501 127 1125 59 4920 1.9c 36 1080 150" 155 2090 5279 1080 l5Od 128 1.9: 28 1880 3.2 5843 1700 148 324 59 1020 3.6 6225 1240 1150 195 497 65 Z: 18,740 6226 Average 3.0 100 1550 230 66 Planchet method rt 10%. b Radioactivity determined o n Nuclear Measurements Corp. proportional counter beta yield p H adjusted t o -2.5 before filtration. d Assumed.
12 inches in diameter containing 3 cubic feet of Salcite HCRK a + resin. During the period May I, 1953 to January 18, 1954 a total of 18,740 gallons of liquid wastes was processed in this manner, The average p~ was 3, dissolved solids 1550 P,p.m,, and the beta-gamma radioactivity 240 counts Per minute Per ml. ((32% yield). An average removal of 72% radioactivity was
The
n.r. 41
Activity Removed,
70 98
75
4 0 6.7 2.2 17.8 3.5 6.6 2.6 1.2 3.3 1.3 2.2 1 3 4.6 5 4 7.7
77 78 82 87
3.6
72
=
85 64
94 73 85 62 17 69 21 64
62%.
ACKh OWLEDGM EN TS
The authors mould like to thank Ruth Juvinall and Arden Schilb who assisted in some laboratory work. Fission product analyses were made by the analytical group under the direction of D. Krause t o whom \Te are very grateful We also appreciate the suggestions of Elton Turk regarding the interpretations of the fission product analyses.
INDUSTRIAL AND ENGINEERING CHEMISTRY
Vol. 47, No. 1
d LITERATURE CITED
Ayres, J. A, IND.ENG.Cmnf., 43, 1526 (1951). Kunin, Robert, and McGarvey, F., Anal. Chem., 26, 104 (1954). Eunin, Robert, and McGarvey, F., IND.ENG.CHEM-,46, 118 (1954). ,
I
Eunin, Robert, and Myers, Robert J., “Ion Exchange Resins,” John Wiley & Sons, New York, 1950. (6) Nachod, Frederick C., “Ion Exchange,” dcadeinio Press, New York, 1949.
o
n Exchange
(6) National Bureau of Standards Handbook 52 (Supt. Documents Washington 25, D. C.), March 20, 1953. (7) Rodger, UT.A , , and Fineman, P., A-ucleonics, 9, 50 (December,
---*,.
1a‘;i 1
(8)
Rodger, W. A , , Fineman, P., and Swope, H. Gladys, “Disposal of Radioactive Wastes at Arnonne National Laboratorv.” Proceedings of t h e Eighth 1nd;strial Waste Conference,”Purdue University, May 4-6, 1953; Purdue Engineering Bull. 38, 474 (January 1954).
RECBIVEDfor review September 2 , 1954.
ACCEPTEDNovember 3, 1954.
Treatment of Chromic Acid Wastes EVALUATION OF METHODS R. F. LEDFORD iiND J. C. HESLER Industrial Filter & Pump M f g . Co., Chicago, 111.
Chromium plating lines and other related operations discharge appreciable quantities of chromic acid to the subsequent water rinses that remove plating solution film from the treated part. The amounts of chromium discharged are frequently large enough to warrant recovery or to require treatment o f the rinse waters to comply with antipollution laws. Chromic acid-copper stripping solutions must be treated to keep copper content low in order to maintain effcctive production rates, but the high cost of solution replacement often requires that the bath be kept in service after the stripping rate has decreased appreciably. Cation exchange treatment of these impure chromic acid solutions to remove metallic impurities, followed by reconcentration to the desired strength, makes recovery procedures financially attractive. Actual case histories illustrate how economic evaluation of ion exchange led to the selection of chromic acid recovery over conventional methods of waste disposal.
T
HE metal plating industry utilizes many chemical baths
that require frequent replenishment to replace “drag-out” losses and to maintain required purity of the bath itself. Drag out is removed from the plated parts by still or flowing rinses that then become contaminated with the plating solutions. Where these solutions contain compounds toxic to fish and plant life or to sewage plant organisms, treatment to destroy or convert them to a harmless form often must be practiced. Chromic acid and cyanide plating baths are typical t,oxic baths. Obviously, the cost of rendering these wastes innocuous is extremely important; the recent development of stable ion exchange resins capable of withstanding chemical attack has provided a means of turning former waste disposal problems to profitable recovery operations. This paper illustrates with actual case histories how evaluation of ion exchange treatment for recovery of chromic acid from several wastes led to its selection over conventional chemical treatment methods. The method permits re-use of recovered acid and eliminates its disposal cost as a waste material. CHROMIUM PLATE STILL-DIP RECLAMATION
A manufacturer expanding plating facilities was advised by local authorities that rinse water from chromium plating operations would require correction before approval of the expanded plant would be granted; the plating baths in use were the conventional type containing chromic and sulfuric acids in the amounts reported in Table I. A number of these plating baths, ranging from 800 gallons to 1500 gallons in size, were to be installed for chromium plating on both die cast and steel parts; January 1955
all plating was to be done over previous copper and nickel base deposits.
TABLE I. DATAO N CHRONIUM PLATIKG OPERATIOXS Chromic acid d a t i n g baths Number of baths CrOa concn., oz./gal. HzSOa concn., oz./gal. Rinse waste water estimated Total flow, gal./min. Daily volume, gal. Average CrOs concn., p.p.m.
5 40.0 0.40 400
500 000 ~
150
The expected total rinse water wastr flow based on data obtained from existing equipment was estimated a t 400 gallons per minute; rinse water analysis sampled during 20-hour operation showed chromic acid contamination ranging from 50 p p.m to more than 200 p.p m., mith a daily average of about 150 p.p.m. Since this degree of sewer contamination would not be permitted in the new plant, a program to reduce chromic acid in the rinses was initiated. Prior t o installation of additional equipment, lLsave”rinse or reclaim rinse tanks mere employed 011 some of the existing lines where drag out of chromium was excessive. These solutions were allowed to build to approximately 4 ounces CrOa per gallon and were returned to the plating tanks whenever possible, as permitted by evaporation and drag-out losses. However, it was estimated that return of all 6ave rinses to the proposed and existing tanks would not he feasible since the nature of the work required that there be sufficient drag out to maintain heavy metal contaminants below a level that would not adversely affect plating.
INDUSTRIAL AND ENGINEERING CHEMISTRY
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