ML19317H401
| ML19317H401 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/31/1980 |
| From: | Barletta R, Robert Davis, Gangwer T BROOKHAVEN NATIONAL LABORATORY |
| To: | |
| Shared Package | |
| ML19317H398 | List: |
| References | |
| RTR-NUREG-0591, RTR-NUREG-591 NUDOCS 8006040119 | |
| Download: ML19317H401 (24) | |
Text
_.
N_.I STATUS REPORT ON LEACHABILITY, STRUCTURAL INTEGRITY, -
AND RADIATION STABILITY.0F ORGANIC ION EXCHN!GE RESINS SOLIDIFIED IN CEMENT AND CEMENT WITH ADDITIVES May 1980 R. E. Barletta, R.' E. Davis, T. E. Gangwer, N. Morcos, D. G. Schweitzer and A. J. Weiss The first stage Epicor-II resins from THI-2 will have Cs-137 activities 3
3 (approximately 40 Ci/ft ) that are about 10 times greater than the con-
~
centration guides (.02 to.04 Ci/ft ) for shallow land burial sites proposed 3
in 10 CFR Part 61 and NUREG/CR-1005.
The composition of waste from the TMI-2 auxiliary building cleanup differs in radionuclide content from the waste nor-mally collected on spent reactor resins and is considerably more concentrated in isotopes of cesium.
The total inventory is about fifteen times the nomal al-location for three months operation of TMI-2 and is equivalent to about 20 years production of waste by a PWR and about five years of nuclear waste production by a BWR.
L e
The waste considered in this report is considerably higher in inventory and specific activity than any spent reactor resins previously placed in shallow land burial sites.
Therefore, we have assumed that if this material is going to be considered for shallow land burial in a fom solidified with cement, leachability properties of this type of matrix should be understood, predict-able, and of a sufficiently low value so that the radionuclide release rate and total release over the burial lifetime should be comparable to other materials put into the shallow land burial sites. Once transportation and shipping requirements have been met and the waste foms have been buried, we believe that
' a measure of the adequacy of ixtobilization of the radionuclides is directly related to the leachability from the vaste fom and container.
11 Cf I
006040 tt 3
r
--. nee s w-sw-
,o:s
. + -
,.4
l-Sunnary 'of _ Conclusions The preliminary conclusions that were arrived at in the course of assem-l
' bling this report are summarized below.
i 1.
We.were not able to find any study or set of studies which was system-atic enough ~and extensive enough to answer or address the questions we believe i
i are pertinent to shallow. land burial for waste of the specific activity and f
quantity of the first stage TMI-2 resins.
l -
2.
We were not able to infer from published material that specialists in
.the field are aware that cesium and perhaps other radionuclides ' appear to be rapidly removed from the ion exchange resin during the initial stages of mixing with cement'and water.
The cesium displaced from the resin appears to distrib-ute itself'approximately equally between the water and the cement.
i-3.
We were' not able to find leaching data on samples larger than about a liter.
(The two types of TMI-2 containers are about 1400 liters and 4800 liters.)
4.
We do not believe adequate evidence exists that justifies predicting leach rates of large samples 'from data on small samples.
l S.
Along with known differences between anion and' cation resins, we be-l lieve the swelling of the' organic resias with frequent concomitant cracking of the resin-cement matrix in laboratory experiments may be associated with (a) improper resin-cement-water ratios, (b) improper curing conditions for the matrix and (c) inadequate' saturation of the resins with H 0.
.If so, this 2
problem may be avoided.
l 6.
We do not believe that there is a unique stability or integrity region l
i for relative compositions of resin, cement, and H 0.. Existing evidence shows 2
i that for similar conditions-the structural integrity nay depend on the type of
' 2'
~-
j ve n.e.,,.
,msm-wmm ew a --
-s
-w-~
resin rather than its weight fraction.
Since rest of the available data deal with mixes that tend to maximize resin proportions, this problem may be avoided in mixes with greater cement and/or H O proportions.
2 7.
The literature on venniculite and zeolite additions to cement-organic resin mixes and cement-liquid waste mixes is pronising but sparse.
Ball milled venniculite gives improved leach resistance in both cement-resin mixtures and cement-liquid waste forms.
European studies indicate no adverse mechanical ef-fects on cement-resin additive mixtures.
Studies on cement-liquid wastes at Savannah River Laboratory showed the resulting concrete was weak.'
8.
The fundamental processes causing radiation damage in resins are not unders tood. However, resins loaded at 40 Ci/ft3 will undergo significant de-composition. The extent of decomposition is specific to resin type.
Prelimi-nary calculations indicate a moderate level of gas generation as a result of radiolysis.
9.
Since data on.small samples indicate that large fractions of the cesium can be leached in hundreds of days and since there do not appear to be suffi-cient studies on impregnation of these systems with asphalt or polymers, the first stage TMI-2 organic resin waste may not be adequately immobilized by sim-ply mixing with cement and solidifying. We believe this type of waste should be treated by thoroughly understood more stringent waste management procedures.
3
~
"4 4%,
-,r.-
,-4 w---
.--i
1.
Mechanical properties of Organic Ion Exchange Resin / Cement Haste forms The available literature was reviewed to determine whether there were any correlations between the preparation characteristics (i.e., water, cement ra-tios, resin loading, and curing procedure) and the mechanical integrity of resin / cement conposites.
These canposites have been known to exhibit poor me-chanical strength during both preparation (1.2) and upon subsequent immersion f a seemingly good composite in water (1,3) as evidenced by swelling, o
cracking, and in extreme cases, complete desintegration of the composite.
.he papers reviewed cover a range of 11 anion, cation, and mixed bed' granular resins (seeTable1). The results of these experiments, as well as the resin loadings and water content for experiments involving granular ion exchange resin / cement composites are shown in Table 2.
Given the state of the art, it is pos.;ible to draw several conclusions from these data.
1.
The lack of mechanical integrity is not a problem generic to either all ion exchange resins or to any particular resin type.- Acceptable com - -
posites have been produced using anion exchange resins,(2) cation exchange resins, and mixed bed resins.(3) Conversely, composites made from all thrae types of resins have been known to exhibit poor mechanical properties.(1) 2.
Resin loading per se is not a particularly sensitive variable for loadings less than 50 wt.% (17 wt.% dry resin).(3) 3.
Fast setting cements may be somewhat superior matrix materials for granulated resins than ordinary cement.(3)
In other experiments
-using powdered resins,(3) it was concluded that low heat qual-ity cement was more suitable.
i 4
,-._m
m Table 1 Resin Types Used for Preparation of Resin / Cement Composites 1.
Amberlite IRN-150 (Mixture of IRN-77 and IRN-78).
2.
Amberlite IRN-77. Strong acid cation exchange resin loaded with Na+ and containing 55 ut.5 water.
3.
Amberlite IRN-78.
Strong base anion exchange resin loaded with Cl and containing 60 wt.% water.
4.
Dowex 1 x 8.
Strong base anion exchange resin.
5.
Dowex SAR. _ Strong base anion exchange resin, Cl-form.
Resin was de-watered and contains -approximately 45 wt.% water.
6.
Amberlite-200. Strong acid cation exchange resin, Na+ form resins was dewatered and contains approximately 45 wt.% water.
7.
Dowex 50.
Cation exchange resin, H+ form, saturated with water.
8.
Dowex 21K. ' Anton exchange resin, OH-form, saturated with water.
9.
Dowex 50 W-21K. Mixed bed ion exchange resin, H+-OH-form, saturated with water.
10.
Dowex 50 W-21K. liixed bed ion exchange resin, Ha+-0H form 2.+.urated with water.
- 11. Diaion SKIB. Cation exchange resin, H+ form.
5
..e
, Q %,w:
.v==
w-e = *-
Table 2 Mechanical Properties, Resin Loadings, and Mater Content of Resin-Cement Composite:
Weight Ratio Resin Good Cement : Resin : Added Loading Resin Type..
Mechanical Ref.
Water w/cb (wt.%)
(see Table 1) Integrity c go, la 0.9 0.9 0.9(1.4) 31 1
no 2
la 0.9 0.9 0.9(1.4) 31 2
no 2
la 0.9 0.9 0.9(1.4) 31 3
yqs 2
la u
u u
14.3 4
u 1
a-d i
la u
u u
23.5 4
u d
la 0.25 0.4 0.4 14 4
u i
la 1
0.28 0.28 44 4
ud 1
la 2
1.2 1.2(2.1) 48 5
no 1
la 2
0.85 0.85(1.8) 52 5
no 1
la 2
1 1(1.9) 50 6
no 1
If u
u (0.65) 14e 7
yes 3
If u
u (0.65) 14e 8
no 3
If u
u (0.55) 14e 9
no 3
If u
u (0.65)14-17e 10 no 3
19 u
u (0.56) 4.3e 10 yes 3
19 u
u (0.45) 4.3e 10 yes 3
19 u
u (0.62) 5.88 10 no 3
19 u
u (0.51) 5.8e 10 yes 3
19 u
u (0.64) 7.38 10 no 3
19 u
u (0.52) 7.38 10 yes 3
19 u
u (0.67) 8.60 10 no 3
19 u
u (0.60) 8.60 10 no 3
19 u-u (0.70 10e 10 no 3
19 u
u (0.62 10e 10 no 3
19 u
u (0.65 16e 10 no 3
19 u
u (0.52 100 10 yes 3
1h 0.7 0.4 0.4 17.6 11 u
4 d
aPortland Type II.
b ater/ cement wt. ratio, value in parentheses is ratio with total water.
W cNo cracking or swelling during incorporation curing or subsequent water immer-sien.
d sed in leach tests, condition of specimen not given.
l U
eDry resin.
f ortland Type 300 (NS 3050).
P 9 Portland RHC (rapid hardening).
h ortland blast furnace slag.
P uunspecified.
6
.~
s
4.
For a particular resin (or resin nix) there exists a window of water cement ratios within which an acceptable co:nposite may be produced.
Water content either greater or less than this range produces compos-ites exhibiting poor mechanical properties.
This window is not the same for all resins or for all resins of a given type.
For the gran-ular. resins.which they studied, Bonnevie-Svendsen, et al.(3) indicate that the region of the water / cement ratio within which an acceptable product is produced is small. The window was somewhat larger for the powdered resins tested.
Thus, in order for valid concl'usions to be drawn concerning the mech nical properties of a given resin / cement composite, scoping experir.ents must be perfonned using resin or resin mix identical to that of interest.
(
5.
Since, in a water deficient environment, one might expect a competition for available water between ion exchange resins and cement, curing con-i ditions and most particularly those relating to water availability (i.e., humidity, amount of free standing water during curing, etc.)
might be expected to affect the mechanical integrity of the product.
Indeed, improved strength has been noted for cement products when curing is done in the presence of a large excess of water.
In the papers reviewed however, no systematic study of this variable has been undertaken, and it appears that, at best, composites were allowed to cure.'in water-saturated air.(3) Thus, the effect of curing conditions upon the mechanical ~ strength and integrity of ion exchange resin / cement composities is unknown.
l 7
- 2.. Leaching of Organic lon Exchange Resin / Cement Waste Form Historically, the reasons for solidification of low level wastes arose from consideration of interim storage and transportation concerns.
Furthermore, under existing low level waste regulations, ion exchange resin wastes are nor-mally acceptable 'for disposal if shipped in a deuatered form.(5) For these reasons, we believe there has been limited interest in detailed characterization of the leaching of solidified resin wastes.
No systematic studies of resin /
cement waste forms have been identified.
While the present review has identi-fied several studies of resin / cement waste form, they are of li'aited scope, of-ten focused on a single resin or cement type or a very limite'd subset thereof.
There are approximately 20 generic types of synthetic organic ion exchange res-ins, and literally hundreds of specifics brands, exhibiting property variation by manufacture and even production batch.(12) Leach behavior of a resin /
cement waste form is not a generic property. Leaching varies with the resin and cement types.
The leach behavior of cesium is stressed since cesium represents the most mobile fission product of immediate concern.
As a baseline for comparison of resin / cement waste foms, the results of experiments to determine the leach behavior of unbound ion exchange resins 'are shown in Figures 1, 2, 3, and 4.(6) Figures 1 and 3 indicate that ion ex-137 s and 85 r when leached with change resins exhibit some retention of C
S distilled water. When ion exchange resins are leached with saltwater, the total activity of Cs and Sr are leached within five days.
The leaching of resins in groundwater (figures unavailable) is intemediate to that in distilled water and sal twater. (6) However, the Cs release was essentially complete after ten days.
8 t
if ionic spe-Recent sccping experiments were ptrfomed at BNL to detemine uld elute activity from an ion exchange resin.
.cies present in unset cement co oximately Rohm and _lfaas IRH-77, a strong acid cation resin, was loaded wi 3 expected for the 137 s (substantially less than the 40 Ci/ft 1 mCf/ft3 of C
Resin, water, and cement mixtures, having waste first stage Epicor-II resins).
f 0.8, were prepared to cement wt. ratio of 1.8 and water to cement wt. ratio o For Portland type II cement, resin / cement samples we with two cement types.
i fraction and a contacted for 5 and 60 minutes, then seperated into a res n tilnes were S and cement / water fraction. Similarily for Lumnite cement, contact 137 s activity before and i
C Ion exchange resins were counted for i
After 5 minute contact, for both Portland II and Lumn 45 minutes.
after cement contact.
While, for Portland 11 ce-cements, approximately 10% Cs activity was eluted.
For Lumnite w *t and 60 minutes contact, approximately 30% Cs was released.
l Cs was released.
cement arid 45 minutes contact, approximately 35% of the initia i
sent in l
The maximum extent to which Cs could be replaced by the ion However, these preliminary experiments cement / water mixture was not determined.
h e resin indicate a substantial amount of Cs was liberated from the ion ex j
Since both the average ion exchange resin upon mixing with cement and water.
i f
rable at the site binding energy and the kinetics of elution would be more un a be 1rading expected for first stage Epicor-II resins, these resu j
.t viewed as conservative.
d to Earlier studies (6) on Portland II cement waste fom had be i
and cobalt.
Cement
,s L
measure the decontamination factors for cesium, stront um, f 1 to insure inten-samples were prepared with a water to cement wt. ratio o No resins were incorporate' tional free standing water above cement samples.
l 9
i 3
these experiments. After a 3-day curing, samples of the free standing water were analyzed,for the radioisotopes of interest.
Decontamination factors (DF) were calculated as the ratio of radionuclide in the initial waste (mci /mL) to the radionuclide concentration in the free standing water (mci /mL).
The values reported were 137Cs, DF = 1, 85Sr, DF = 11; and 60 o, DF = 200.
These C
results indicate scant incorporation (interaction) of the Cs with the cement matrix in contrast to Sr and Co.
It was concluded that a mixture of resins in cement liberates a substantial fraction of cesium with an uncertain degree of incorporation within the cement matrix.
The use of additives to' enhance radio-nuclide retention is discussed in the following section.
The earliest examination of leaching of ion exchange resin /cc sent waste-fann was reported by Holcomb.(7) This work predates IAEA leach testing and reporting conventions.
Samples were prepared with 50% Portland cement, 25%
mixed resins, and 25% water by volume. The total volume, surface to volume ratio, and concentration of radionuclides in the sample were not reported.
Static leach tests in 3 liter of seawater were performed.
The fraction of 137Cs activity leached was approximately 3-4% in one year. Another leaching study, which also predates IAEA recommendation, on powder resin / cement waste form has been reported by the Hanford Engineering Development Laboratory.(1)
Samples, having volumes of 131 cm3 and volume-to-surface ratio of 0.85 cm, were leached in 250 mL of deionized water with rechanical agitation. After 14 days, the cumulative fractions of released reported were 0.8cm for cesium and 0.1 cm. for strontium.
More recently, several studies on ion exchange resins / cement composite have been reported.(3,4,8) Results representative of studies are sho' a in Figures 5, 6, and 7.
Leaching experiments carried out in these studies conform to the 10
q IAEA' recommendation.(9) Fran these studies of small samples, ger qlly indi-cate that 30 to 100 percent of the initial Cs is leached within 100 days. Vari-ations can be attributed to many parameters; type of resin, type of cement, res-in to cement 'and water to cement ratios, length of curing period before leaching was initiated, and possibly the surface to volume ratio of the sample.
The mechanisms 'of leaching are not well understood, and hence an unambig-uous interpretation of leach test results is not currently possible. However, the-following diffusion model has been proposed.
In the model, the fraction, f, release at time, t, is given by f=h (1) where D is the diffusion ccefficient, and S and V are the geometric surface and volume of the sample.
If the model was applicable, then a plot of f versus t1/2 would yield a line passing through the origin.
However, see Figure 6 for example, the model gpnerally is not followed. During an initial period of approximately 25 days, the leaching is significantly more rapid than the latter period, 'which is linear. Beyond the initial period, the leach process has been described by the following equation,
+a=mK+2 (2) f=f where a is simply the f axis intercept.
The rationale often given far this ex-pression -is that after the initial-period of rapid leaching dominated by a not well understood mechanism, the leaching is diffusion controlled.
A primary goal of leach testing is to characterize what is to be expected after disposal of a candidate waste fom.
Of interest here is the extension of 11
data obtained on small laboratory samples to predict the behavior of realis-tically sized waste forms.
To date, no study of a resin cement composite has been reported which explores the functionality of fraction release and the sur-face to volume ratio expressed in Eqs. I and 2.
Matsuzuru(10) has reported a study on the effect of sample dimension on the leaching of sodium sulfate waste /
cement composites. Samples having surface to volume ratios over the range 2.47 to 0.587 cm-1 were leached in accordance with IAEA recommendations.
The re-suits could be represented by Eq. 2.
Linear dependances on S/V were observed for both the m and b; see Figure 8.
Extrapolation of this data can provide an estimate of the release from a realistic sized waste form.
Estimates for a 200 L drum were given in reference 10 and are reproduced here in Table 3.
Whil e these results are not directly applicable to resin / cement composites, specimen number 8 is spectulated to be most applicable.
Clearly, significant gains in cesium retention are predicted for large monoliths.
The largest sample tested in the study was 10 cm (3.9 in.) diameter and 12 cm (4.7 in.) high.
Given the empirical nature of interpretion of leach results, it would seem important that this type of measurement be extended to larger if not actual sized resin / cement waste form.
Table 3 Es:I=sted frac: ion leached 'from a cement com7osite of 2001 drum size gg;g, I.cach!c; p:ra: eters Leaching frae: ion No.
m(day-8 8) 1 yr 10 yr 50 yr 100 yr 200 yr e
"'Cs 1
- 1. 3 x 10 * * -
- 4. 7 x 10-s 7,2xjo-
- 1. 3 x 10-8
- 2. 2 x 10-'
- 3. 0 x 10-'
- 4. O x 10-'
2
- 7. J x 10.a
- 8. I x 10-8
?. 2 x 10-8
- 5. 3 x 10-3,gx jo-:
j,5xjo-a
- 2. I x 10-8 3
- 1. 4 x 10"
- 4. 7 x 10-8
- 7. J x 10-8
- 1. 3 x 10-8
- 2. 4 x lo-'
- 3. I x 10-8
- 4. 3 x 10-'
4 1.1x10-8
- 1. 3 x 10-8
- 3. 4 x 10-8
- 7. 9 x 10-8
- 1. 6 x 10-8 12 x 10-8 3.1 x lu-8 s
- 2. 4 x 10-8
- 4. 6 x 10-8
- 1. 4 x 10-8
- 3. 2 x 10-8
- 4. 6 x 10-8
- 6. s x 10 '
6
- 8. 6 x 10-8
- 1. 6 x 10-8
- 5. 2 x 10-'
- 1. 2 x 10-r
- 1. 6 x 10-8
- 2. 3 x 10 8 7
- 1. 2 x 10-s
- 5. 7 x 10-8
- 1. J x 10-8 3.1 x 10-8
- 6. 2 x 10-8 S. 6 x 10-8
- 1. 2 x 10-8 8
- 1. 9 x 10-8 1.1 x 10-8 J.7 x 10-8
- 1. 3 x 10-8
- 2. 6 x 10-8
- 3. 7 x 10-8
- s. 2 x 10 8
- sr 13
- 2. 5 x 10-3 9.1 x 10-8 5.7 x 10-8 1.GxID-8
- 3. 4 x 10-8 4.8x108 6.Sx108 14
- 3. 5 x 10-8
- 2. 3 x 10-8 9.0x10-8
- 2. 3 x 10.s
- 5. 0 x 10-8
- 6. 9 x 10-8
- 9. 7 x 10 '
17
- 3. S x 10-8
- 1. 3 x 10-8
- s. 6 x 10-8
- 2. 4 x 10-'
- 5. 3 x 10-8
- 7. 4 x 10-8
- 1. 0 x 10-'
13
- 5. 5 x 10-s
- 3. I x 10-8 1.4 x 10-8
- 3. E x 10-8
- 7. 7 x 10* 8
- 1. I x 10 *
- 1.5x10-8 e 'Co
~il 1.1 x 10 8
- s. l y.10" '
2.S x 10.
- 7. 5 x 13-8
- 1. c x 10"
- 2. 2 x 10-8 3.07.10
- 1.6/10 *
- 1. 2 x 10.-'
J. 7 x 10" 1.1 x 10-8
- 2. 3 x 10-8
- 3. 2 x 10-5
- 4. 4 y.10-*
+
E3
- 3. 4 x 10-8
- 1. 2 x 10-5 7.7x10-'
- 2. 0 x 10 8
- 4. 7 x 10-8
- 6. C x 10"
- 9. 3 x 10" 26 5.5e.104
- 1. 5 x 10-*
- 1. 2 x 10 '
- 3. 5 x 10 8
- 7. 4 x 10" 1.1 >: 10-*
- 1. s y 10" 12
Inherent -in scaling to a realistically sized waste form, is the assump-tion that the sample is a structurally sound monolith (a waste form, absent of any cracking which would increase the effective surface to volume ratio).
In the case of resin / cement composites, large monoliths have not been demonstrated.
In Summary D
1.
The leach behavior of resin and cement waste forms depends upon resin type, j
cement type, loading (waste to cement ratio), curing, type of leachant, and temperature.
2.
Preliminary experimental evidence suggests substantial liberation of cesium
_from resin immediately after mixture with cement and probably little in-corporation within the ce. ment matrix.
3.
Recent leach testing of small samples indicates little retention of cesium:
l 30-100% loss within 100 days.
4.
Some experimental evidence exists to suggest substantial improvement of radionuclide retention with increased size of waste form.
However, this has not been verified for resin / cement waste forms of realistic size.
Reduced leachability may be realized if the waste form exhibits good me-chanical integrity.
3.
Effect of Additives Incorporated In Concrete Waste Forms A variety:of additives have been used in inc solidification of radioactive wastes in cement to alter properties of the resultant waste form, such as set tire, strength, swelling, and leachability. Spent ion-exchange resins consti-tute a major fraction of radioactive waste from nuclear reactors.
Cement waste forms containing ion-excnonge resins often swell, crack, and tend to disinte-grate in water. Addition of vermiculite to the resin / cement mixture is reported to reduce this tendency of the waste form to swell and crack, however,'there is 13
a paucity of information available in the literature on the effects of additives incorporated in ion exchange resin / cement waste forms.
Bonnevie-Svendsen, et al.,(3) discusses the leachability of cesium from
- ion exchange resin / cement products and the effects of additives, including vermiculite, on the quality of the resin / cement products.
High leaching of cesium is a main objection against incorporation of fission product wastes in
- cement, and the leaching of cesium fran resin / cement products is sometimes higher than from pure cement.
With 2.5% by wt. vermiculite (particle size 30-70 mesh) cesium. leach coefficients in cement were reduced 2 to 3 or'ders of magni-tude without impairing the mechanical properties.
No resins were incorporated in these waste forms.
Similar effects were achieved for cement products with approximately 18% (dry wt.%) r.ew granular resins (H+.S0 -2),10% new pow-3 dered resin (H+-0H-), and 9% spent powdered resin.
The reduction of cesium leach rates with increasing vermiculite content from 2.5% to 10% cement weight is shown in Figures 9, 10, and 11. Since vermiculite is known to have cationic adsorptive properties, it is reasonable to expect that vermiculite will aid in recapturing the cesium that is released from the ion exchange resins.
A Savannah River _ laboratory study (ll) tested a series of sorbents for.
cesium' sorption kinetics, compressive stre'ngth' of concrete containing the sor-bent, and leachability.
The best cesium sorbents, in terms of sorption kin-etics, were Linde AW-500 and Zeolon Z-900 out of the following materials:
Z-200, Z-500, Z-900, Linde 13-X, Linde AW-300, Linde SK-40, Linde AW-500, clinoptilolite, and vermiculite.
The effect of the sorbents on the compressive strength'of concretes was measured with and without simulated sludge waste.
In the absence of sludge, AW-500 and Z-900 had little effect on 'b3 strength of concrete, whereas. Z-200,- AW-300, and vermiculite severely weakened the concrete.
14
. I 8.
l
In the presence of sludge, Z-900 and AW-500 actually strengthened the concrete while vermucilite weakened it badly.
No data are available for resin / cement waste forms.
Comparative leachability of cesium from two concretes containing 2 wt.% of the sorbents are shown in Table 4.
Table 4(11)
Leaching of Concretes
% Cesium Leached After 28 days Sorbent High Aluminum Cement Pozzolanic Cement AW-500 5.7 4.6 Z-900 6.0 3.9 Vermiculite 6.3 3.8 Z-500 6.5 4.5 Z-200 9.1 6.5-1 Clinoptilelite 13.1 9.2 AW-300 17.9 9.3 On the basis of leachability, AW-500, Z-900, and vermiculite were the best sorbents for cesium.
It has been shown that addition of silicates to cements reduces free calcium in the final product.
It was postulated that free calcium in concrete might displace cesium from ion exchange sites in the solidified ma-terial. However, addition of silica gel to cement containing sorbents had no additional effect on cesium leachability.-
Conclusion These data suggest that if ion exchange resins are solidified in cement, additives may reduce the potential for swelling and cracking and at the same time decrease the cesium leachability resulting from its desorption by free cations in the ceme:.;.
15
4.
Radiation. Stability of Organic Ion Exchange Resin Systems s.
3 level wj11 A loading of the organic ion exchange r_esins to the 40 Ci/ft 9 rads.
Based on the literature (12) such result in total dose exposure of 10 a dose level will result in significant N'ecomposition of the resins. The loss of the ability to retain radionuclides and generation of organic products,'which
. would be free to migrate away from the resin sites, will occur in these systems.
Resin degradation processes may continue even after the majority of the radia-tion has decayed due to long lived reactive species formed by various chemical reactions. For example, radi'ation induced oxidation can generate peroxides which will subsequently decompose and react with the resin.
Production of such strong oxidizing species could also result in conditions favoring exothennic chemical reactions.
Fcr more than three decades there have been extensive applications of ion exchange materials in the nuclear process industry.
Process related problems of a serious nature caused by radiation chemical changes in the ion exchange medium have been mentioned occasionally in the open literature as well as in safety evaluation studies of radioactive mr.terial processing.
The present literature survey (12) identified a number of reports on experimental studies relevant to evaluating the process related properties of organic ion exchangers. Still, there are numerous problems associated with the radiation chemical changes of
~
ion exchange materials due to radiation damage which are not properly under-stood.
Thus, in spite of the apparent abundance of experimental studies, thc e is no reliable information to allow one to predict the chemical effects on lorj term storage of ion exchange materials with loading comparable to the first stage TIM-2 materials.
16 LL
Understanding the effects of radiation on ion' exchange materials may be im-portant in deciding how disposal of spent resins should be carried ~ out. From the present survey of current knowledge on the radiation effects on both organic resin and inorganic zeolite ion exchange materials, it can be-concluded that:
1.
The fundamental processes involved in radiation damages to ion exchange materials are not understood.
2.
Structural and chemical constituent modifications can significantly change the radiation resistance of ion exchange materials.
3.
There have not been systematic investigations to devel'op standards for use of and loadings of ion exchange resins.
4.
There is a need to extend the radiation effect studies to investiga-tions of secondary reactions of the radiolysis products with the materials to be used in waste disposal.
Gas Evolution as a Result of Irradiation Irradiation of ion exchange resins results in evolution of gaseous prod-ects. There are only a few reported experimental studies which have attempted to study the nature of gas evolution and make some quantitative estimat es of the gaseous product formation during the radiolysis of ion exchange resins (12),
The main gaseous products fonned during the irradiation of ion exchange resins are hydrogen and carbon dioxide. Other gases such as carbon monoxide, oxygen, methane, nitrogen, and oxides of nitrogen and sulfur have been identified. The radiation chemical yields (G values) for the gaseous products, taken from Table VII of reference 12, are reproduced here in Table 5.
In cation exchange resins, she evolution of hydrogen is a linear function of absorbed dose.
Hydrogen evolution is also shown to increase with the water content (swelling) of the resin. Sulfur dioxide is formed by the direct action 17
.of. ionizing radiation on dry forms of sulfonic acid resins.
In the presence of water, the S02 undergoes secondary reactions to fonn acidic solutions and for this reason, S02 is not identified during the irradiation of moist and wet resins.- Carbon dioxide and carbon monoxide are products of irradiated resins.
Like hydrogen, the evolution of CO2 and C0 is quite linear with absorbed dose.
In general, the radiation chemical yield of carbon monoxide is significantly less than that of carbon dioxide.
Based on the G values in Table 5, which range for the various gases from 0.001 to 3.0, the yield of gas from an organic resin subject to 109 rad will correspondingly range from.02 to 70 cm3 of gas per gram of resin.
An esti-mate of the corresponding gas yield from cement is 1.3 cm3 per gram of ce-ment.(6)
If one assumes a G = 0.01 for gas evolution from resins subjected to 109 rad, then pressures in the range,of 1.5 to_15 atmospheres would be pro-duced in a sealed 4800 L (4 ft. dia. by 4 ft. high) cylindrical canister con-taining solidified waste with 60 volume % resin and a void volume of from 1 to If the gas were CO, for example, the container would be subjected to 10%.
2 the pressure.
If on the other hand, the gas were S02 and the void volume was filled with water, then the 9.2 moles of S02 produced would result in an acidic solution whose stre. if b sould be in the range of 0.07 to 0.7 M.
' These preliminary calculations serve to illustrate that the quantities of gases'which could be-c:-"c.
ted are not trivial, and therefore, gas generation warrant's further conside.ation.
Organic Ion Exchange Failures Ion exchange resin failures have been reported for systems involving ni-
.tric acid. Critical ignition temperatures as low as 135 C have been reported 18
for resins in the nitrate form with the potential of this going below.100 C for radiation damaged resin systems.
Self ignition temperatures for anion resins depend on resin form, heating rate, resin brand, cross linkage, loading, resin bed volume, aging and pretreatment. Of particular concern to the TMI resin sys-tems is the resin bed volume.
Increasing the geometrical dimensions of the resin bed was reported.to lower the ignition temperature.
As pointed out in the Miles report (13) although only nitric acid / resin system have been observed to be unstable to date, one cannot assume any resin system is safe:
"Therefore, experimental work to establish safe limits in these areas before'using a given ion exchange system should be considered mandatory." As a first step, the pre-cautionary measures such as those listed in Table 3 of Miles' paper should be verified for the specific resin systems under consideration.
Conclusion We do not know if every one of the many uncertainties discussed in this 5
report is directly or indirectly related to safety aspects of shallow land buri- !
I al. However, we believe that the uncertainties do war. ant that more stringent waste management procedures be applied to the TIM-2 first stage Epicor II 7
resins.
19
<r..
o.
9 o
I s.
r U
- c. ist g..
s -e$
3 a
imm 7?Prasatsies y
- 00's,
9.& t 2 Ol5fitLED *atte LE ACHANT 2
v/$.8 75 *io'8 cm
/
o j.
/
h
- 0007 s
p-4 4
gf2 Vmnwww a
O to 20 30 40 SO 60 70 60 90 LE ACM Tsut. C&TS 137 85 Figure 1.
Cs and Sr release from
) static leaching IRN-77 bead resin f in distilled water.
w 7
E
.4 i
e i
0 5
E y.
,,to 9 g.
,.. : : : :- -- : : : : : --- 1.o 9y U 'o 8
{'
j@6 e
. 5,.85 6
w y, C
SALT WATER LEACHANT -05 Y yT 1RN-77 BEAD AESIN 4
E-
}l WS
- 8.75 = 163 cm 2
~'
. O S
to 15 20 LE ACH TIME.0ATS 137Cs and 85 Figure 2.
Sr release from saltwater.(6)esin for static leaching in IRN-77 bead r 30
.k*.es
- 032 d O
PCarca %*t" At$1N d
DistsLLCD *atEm Lt Atmant d*
v/5 9 25. so'* sm
'I?*20
.,..e-*"=#-*,
C22 *I Va b,. k.
~.
80
- Osl g g
A 1
5u
- 3 O
.0 20 30 - a0 $0 60 70 80 90 LtacM Tipt.Dav$
137 85 r release from S
' Figure 3.
Cs and PHCpowderedresinfgrstaticleaching in distilled water.t 1-20
, U
m 4
w 7
s s
8 n.
E U
go gi sog 8Y 8 e Cs*IST G
o C ~.
- Sr - 85 N
$[y6 PCH PQwCRED RESIN sat.T %ATER LE ACHANT
- 0.S w v/S. 9.2S m oo'* cm
- l
> =
E 2
N1 S
E 2
0 3
3 S
30 is 20 o
~
Lf ACH TIME, CAYS 137 85 r release from Figure 4.
Cs and S
PCH powdered resin for static leaching in saltwater.t6) 6.
~
i.0->>.
i.--.+,.
.s r. > e. s e OS N,e.e.ew C.- W
-CS
,04
-04 l
e
[c3
- 03 I n
g3 P4**"/"*
. y o'
Ol
'33 rn"12 =
m a
a 6
.0 s....,..~
p.;r Figure 5.
Leaching of 137 s from ion C
exchange resin incorporgtgd in polyethylene or cement.(8; t
l l
es-3 3 e.
_ 3 r
f.**
_ 33 0.3 -
?
- - * ~ ~ ' _ _ -. "
E 5c f',,.de./
i e
- ~
o.- -I,,f
_3, 2
a r.ne, { ear Figug7 Plots of f vs t for leaching 6
of.
Cs from ion-exchange resin in-corporated in polyethylen r cement, o polyethylene, a cement.
21
- 9. ~
w g
=~* ~ posGe'eC res.n.OC.
..grc f*
e
.s Ii
- ls
..... --- -~
g*&f,.
-e-- - - w. -
^
Sf 0VCr reso.0C e
. t.c i
. - + - -
..anc O
C' 20 30 to 50 60 70 43 SY"~~C I f. C0y t ~
Figure 7.
Effect of cement qualities on the cesium leach rate of resin / cement products.
OC-Portland cement, Type 300; LH -)
portland cement, rapid hardening.(3 5
< 14
. a
. iG 3
' *~
"g v
2
. g E
=
. c 1
j;
. g O.~
0 0
8 2
2 Siv Figure 8.
Typical plots of m and a(10) vs S/V-(Specimen No. 1) 22 e
k,1
- OC to'g-8 F.
tL eOC.25 ee me c OC.5 */. veru&,.
=
e
- OC.CY. wme.
so'-
6 lo "
' ' 2 o [,,' ',, j 1
o
- Figure 9.
Accumulated Cs leach rates for powdered resin R
Ia
=
n n
cement vermiculite products.
OC-Portland cement, type 300.(3)
S*t..
Ih-
- og b,
OC. 2 5 */..ermc.
OC. 5 '/. werer.c.
y OC.e me.
c.
y c.
g
,,.,?
5.,',,,,,,,.,,,,,,,,,,
' ri. e.,. - i I
e Figure 10. Accu ulated Cs leach rates R
for granular resin Ia
=
n n
cementvermic01teproducts.
OC-Portland cement, Type 300.(3) f
-x to'{
i :
t
<t d
c'-
p ec. 5 v.. mc.
c'r
.I t i. e i,o r
,o 2o
, ao Figure 11. Acc mulated Cs leach rates R. = Ia -
for.'Oscarshamn spent n
n resin-cement vermiculite products.
(Cs tracer added). 0C-Portland cemenc,
-Type 300.(3) 23
- 98 L-_.
References 1.-
R. E. Lerch, Division of Waste Management, Production and Reprocessing Programs Progress Report, January to June 1977, HEDL-TM-77-74 (July 1977).
2.
P. Colombo and R. M. Neilson, Jr., " Properties of Radioactive Wastes and Waste Containers,"BNL-NUREG/CR-51101 (1979).
3.
M. Bonnevie-Svendsen, K. Tallberg, P. Aittola, and H. Tollback, " Studies on the Incorporation of Spent Ion Exchange Resins From Nuclear Power Plants in Bitumen and Cement," IAEA-SM-207/78 (1976).
4.
N. Moriyama, S. Dojiri, and H. Matsuzuru, Health Physics, 32,549-52 (1977).
5.
Code of Federal Regulation, Title 10 Part 71 and Title 49 Part 173 and 177,
.(1978).
6.
P.' Colombo and R. M. Neilson, Jr., " Properties of Radioactive Waste and Waste Containers," First Topical Report, NUREG/CR-0619, (1979).
7.
R. R. Holcomb, "Martime Reactor Waste Disposal Studies: Solidification of Ion-Exchange Resin With Portland Cement for Radioactive Waste Disposal,"
0RNL-2899 (1960).
8.
N. Moriyama, et al., Journal of Nuc. Sci. and Tech., 12(6) (1975).
9.
E. D. Hespe, Atomic Energy Review., 9,(1), 195 (1971).
10.
H. Matsuzuru and A. Ito, Journal of Nuc.. Sci. and Tech. 15(4),(1978).
137 s Sorbents for Fixation in Concrete,"
- 11. M J. Plodinec, " Evaluation of C
DP-1444 (1977).
12.
T. E. Gangwer, M. Goldstein, and K.K.S. Pillay, " Radiation Effects on Ion Exchange Materials," BNL-50781 (1977).
13.
F. W. Miles, " Ion Exchange Resin System Failures in Reprocessing Activi-ties," Nuclear Safety, 9,(5) (1968).
24 t