ML20080D293
| ML20080D293 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/31/1983 |
| From: | Adams J, Dougherty D BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| CON-FIN-A-3162 BNL-NUREG-51689, NUREG-CR-3381, NUDOCS 8308300319 | |
| Download: ML20080D293 (67) | |
Text
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i NUREGICR-3381 l
BNL-NUREG-51689 1
I l Evaluation of the i Three Mile Island Unit 2 Reactor
! Building Decontamination Process 4
Prepared by D. Dougherty, J. W. Adams Brookhawn Nationci Laboratory Prap:: red for U.S. Nuclear Regulatory Commission l-l l
000!Oo!do!oO!3bo P
PDR l
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NUREGICR-3381 BNL-NUREG-51689 Evaluation of the Three Mile Island Unit 2 Reactor Building Decontamination Process Manuscript Completed: May 1983 Data Published: August 1983 Prepared by D. Dougherty, J. W. Adams Brookhaven National Laboratory Department of Nuclear Energy Upton, NY 11973 Prcpared for Division of Waste Management Office of Nuclear. Material Safety and Safeguards U.S. Nuclear Regulatory Commission a
Wrahington, D.C. 20555 NRC FIN A31S2 hi -
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-i,i,i
ABSTRACT Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams.
Solid wastes being disposed of in commercial sballow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings.
The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61.
It appears that much of the Epicor-Il ion-exchange resin being disposed of in commerical land burial will be Class B and require stabiliza-tion if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B.
Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation. Results indicated that both radionuclide con-tamination and chelating agents leach from strippable coating vaste. It
- swells and partially dissolves upon immersion in water and organic liquids, biodegrades readily in soll and, upon irradiation, generates gas, principally hydrogen. The coating is thermally stable up to at least 1000C and is not an unacceptable flame and smoke hazard. The strippable coating samples from the TMI-2 reactor building decontamination testing contained significent strontium-9G and cesium-137 contamination and would be Class B under 10 CFR Part 61.
A-111
CONTENTS 1
ABSTRACT.
iii j
CONTENTS.
v FIGURES.
vi TABLES.
vii ACKNOWLEDGMENTS.
ix
(
1.
INTRODUCTION.
1 2.
WASTE MANAGEMENT AT TMI-2 REACIOR BUILDING CLEANUP.
7 21 Reactor Building Decontamination Planning.
7 2.2 ' Decontamination Wasces.
7 3.
STRIPPABLE COATINGS - GENERAL.
13 3.1 Introduction.
13 13 3.2 Description.
4.
STRIPPABLE C0ATINGS AT TMI.
15 4.1 Introduction.
15 4.2 Testing ALARA 1146 Strippable Coating for Characterization as Radwaste.
15 16 4.2.1 Scoping Tests..
4 4.2.2 Testing of Strippable Coating Frcan TMI-2.
34 5.
REFERENCES.
47 APPEMDIX A - 10 CFR PART 61.35: RADI0 ACTIVE WASTE CLASSIFICATION FOR DISPOSAL IN SHALLOW LAND BURIAL.
51 APPENDIX B - TECHNICAL DATA SHZET FOR ALARA 1146 CECON STRIPPA33LE COATING...
55 APPENDIX C - CUMdLATIVE DOSE CALCUIATIONS FOR THE ALARA 1146 STRIPPABLE C3ATING SAMPLES FROM THE THI-2 REACTOR BUILDING GROSS DECONTAMINATION EXPERIMENT.
59 v
FIGURES 1.1 Vertical Cross-Sectional Schematic of the T41-2 Reactor Building Showing the Floor Levels in Comparison to the Major Components of the Nuclear Steam Supply System 3
4.1 Percent Activity Loss vs Time for Imperial 1146 Strippable Coating Leached in Deionized Water.
17 l
4.2 Percent Weight Gain of ALARA 1146 Strippable Coating Upon Immersion l
in Toluene, Kylene, Water, and a Solution of Water Saturated With Toluene and Xylene.
20 4.3 Weight Gain of ALARA 1146 Strippable Coating Upon Immersion in LSC Cocktail.
21 4.4 Drying Curves at Room Temperature in Air for ALARA 1146 Strippable Coating Following 34 Days Immersion in Toluene, Xylene, Water, and a Water Solution Saturated With Toluene and Xylene.
22 4.5 Drying Curve at Room Temperature in Air for ALARA 1146 Strippable Coating Following 34 Days Immersion in LSC Cocktail.
23 4.6 Biodegradative CO2 Cas Evoluation From Scoping Tests on Imperial 1146 Coating in Barnwell and Hanford Soils.
25 4.7 Plot of Gas Generation vs Co-60 Gamma Dose for ALARA 1146 Strippable Coating.
29 4.8 Effect of Radiclysis n t.LARA 1146 Coating vs Unirradiated Specimen. 31 4,9 Leach Test Results for TM1 Strippable Coating.
37 4.10 Siodegradative CO2 Gas Evolution From TMI Strippable Coating Material in barnwell and Hanford Soils.
41 4.11 Probable Range of Biodegradation of THI Strippable Coating in Barnwell Scil..
42 4.12 Probable Range of Biodegradation of TMI Strippable Coating in Hanford Soil.
43 vi
TABLES 4.1 Immersion and Drying Data for ALARA 1146 Strippable Coating.
23 4.2 Scoping Test Fesults on the Biodegradation of ALARA 1146 Strip-pahle Coating in Soils From the Barnwell, SC, and Hanford, WA, Shallow Land Burial Sites.
26 4.3 Cas Composition Analysis Fran Co-60 Gamma Irradiation Tests of ALARA 1146 Strippable Coating in Air and Under Vacuum.
30 4.4 Gas Analysis From the Thermal Testing of ALARA 1146 Strippable Coating.
33 4.5 Strippable Coating Samples Fran the THI-2 Reactor Building Gross Decontamination Ter. ting.
35 4.6 Activity Distribution Between Leachate and Coating for a Sample Coating From the TMI-2 Gross Decontamination Experiment.
38 4.7 The Estimated Activities of Cs-137 and Sr-90 in the Five TMI Coating Pieces Considered as a Batch.
39 4.8 Biodegradation Test Results for Strippable Coating Samples Fron the TMI-2 Reactor Building Cross Decontamination Experiment.
40 C.1 Nuclides aad Relevan! Eecay Data Used in Calculation of Dose to Strippable Coating.
59 C.2 Total Absorbed Dose Calculations fo: the AL/.RA lif.o Strippable C.:ating Samples Fraa the TMI-2 Reactor Building Glos 9 Decontamination Experfuent 61 vii
ACKNOWLEDGMENTS The authors thank Drs. Richard E. Davis, Robert E. Barletta, and Karl J.
Swyler for helpful discussions during the course of this work. We also thank Mr. James D. Smith and James Clinton for their technical assistance with ex-perimental portions of this work.
Special thanks are due to Ms. Colleen E.
Shea without whose assistance and advice the biodegradation portions of this work would have been sparse indeed.
1 We also acknowledge with sincere appreciation the efforts of Ms. Nancy Yerry and Ms. Kathy Becker for the preparation of this manuscript.
1 1
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I.
i f
ix 1.
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I EVALUATION OF THE THREE MILE ISLAND UNIT 2 i
REACTOR BUILDING DECONTAMINATION PROCESS 1.
INTRODUCTION It has been more than 3-1/2 years since the March 28, 1979 accident at the Three Mile Island Unit 2 Nuclear Generating Plant which lef t the reactor core damaged and the reactor building contaminated with fission products.
l j
During this time plans (1-7) have been made and ef fort (8-12) expended to cleanup and recover TMI-2.
These ef forts are directed toward decontaminating l
the reactor building, removing the nuclear fuel f rom the damaged core and de-contaminating the reactor coolant system and connected systems. Decontamina-l tion of the reactor building will initially focus on thase areas needed for l
defueling. These areas, as shown in Figure 1.1, include the 305-ft. level, the 347-ft. level (the operating floor) and the polar crane. The reactor coolant system water must also be cleaned up to reduce the concentration cf dissolved fission products prior to defueling.
Major milestones (13) in the cleanup and recovery ef fort have included the following events:
(1) Venting to atmosphere of the 44,000 Curies of the fission product gas krypton-85 which was released in the reactor building during the accident. This was done between June 28 and July 11, 1980.
l l
(2) Removal of dissolved fission product activity from the 565,224 gal j
cf tritiated accident water in the Auxiliary and Fuel-Handling l
Huilding (AFH3). The ficcion product activity wcs removed by ion ex-chan6e (Epicor II*) between Uceember 1979 and December 1980. The i
highly loaded ion-exchange resins are being transferred co the Department of Energy (DOE) for disposal. The tritiated accident 3
water (<1 pCi/cm ) is being stored for reuse at TMI.
(3) Removal of the dissolved fission product activity f rom nore than 500,000 gal of tritiated accident water which flooded the basement (282-ft elevation) cf the reactor building. The fission product i
{
activity was sorbed onto zeolite demineraliner beds in the Submerged Demineraliter System (SDS) between September 23, 1981 and March 5, 1982. The tritiated water ef fluent from the SDS was polished (i.e.,
the small amount of fission product activity not sorbed onto the zeolites. was removed) using Epicor-II ion-exchange resin. The highly loaded zeolites are being transferred to DOE for disposal.
Activity levels in the Epicor-II polishing resins were within the Hanford, Washington, commercial land burial site limit (<1 pCi/g) l i
- Proprietary ion-exchange system of Epicor, Incorporated.
1
for disposal in the dewatered form and without solidification. The 3
tritiated accident water (<1 tCi/cm ) is being stored for reuse at TMI.
(4) The Gross Decontamination Experiment was performed between October 29, 1981 and March 26, 1982. This effort was directed to-ward reducing radiation levels on the 305-f t. elevation and above and toward determining the relative decontamination ef fectiveness of the several techniques selected for testing.
(5) Gross Decontamination of the reactor building including the 305-f t.
elevation and above started September 17, 1982. The effort was de-signed to reduce smearable levels of contamination to the point that workers will be able to remove much of the bulky protective clothing and the respirators.
(6) Removal of fission product activity from the 88,000 gallons of water in the reactor coolant system (RCS) using the SDS has been attempted but has not been completely successful. Strontium-90 from the fuel debris apparently is dissolving into the cleaned water and restoring the strontium-90 activity to near its initial level. This phenome-non may influence the RCS cleanup plans which were not finalized as of this writing.
Major events (12-16) in the cleanup and recovery yet to occur include:
(7) Defueling of the reactor. Pla ns for removing the fuel f rom the re-actor await detailed i'nspection of the core. This inspection began with the " Quick Looks"(13,14) into the reactor vessel with a re-mote TV camera and will continue when the reactor vessel head is removed, which is scheduled for sccetime in mid 1983.
It is hoped that fuel removal from the core can be completed by late 1985.IIS)
(8) Decontamination of the reactor coolant syste's and connected systems which have been contaminated by fuel debris. The decontamination techalques aqa procedures to be used for this cleanup will not be decided upon until studies being perf ormed by the Electric Power Research Instituta (EPRI) recomaend appropriate techniuues.(12)
(9) Cleanup of the refueling canal during and af ter reactot defueling.
Transfer of the core debris from the reactor to a packaging f acility may result in substantial soluble and particulate contamination of the refueling canal.
(10) Hands-on decontamination of the 305-f t, elevation and abcve. This is intended to return the working environment in this part of the reactor building to near normal (i.e., pre-accident) conditions.
This effort will start sometime after Gross Decontamination, Item 5.
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Figure 1.1 Vertical cross-sectional schematic of the TMI-2 reactor building showing the floor levels in co: parison to the major components of the nuclear steam supply system.
(11) Decontamination of the 282-f t. evaluation. The 600,000 gallons of J
highly contaminated water, Item 3, filled this part of the reactor to a depth of about 8 f t.
This elevation has much higher radiation fields than the upper portions of the reactor building.
(12) Disposal of the more than 1-million gallons of tritiated accident water, Items 2 and 3.
Dil is under specific NRC order (16) not to release any of the accident water.
In addition to the considera-tions of 10 CFR Part 20 with regard to this water, the NRC and the State of Maryland are conducting studies on potential socioeconomic impacts of various alternatives for disposal of the accident water.
Flushing with both low and high pressure water spray has been the princi-pal decontamination method for removal of as much loose debris and smearable and soluble activity as possible. The tritiated accident water (<1 pCi H-3 per mL) is being used for the water flushing decontamination activity. It is routed through floor drains from the upper levels in the reactor building to the basement (282-f t. elevation) from where it is then sent to the SDS for cleanup and return to tank storage for reuse. The Gross Decontamination Experiment (GDE) also tested, in small scale, detergent (Turco 4324 NP-10%)
and phosphoric acid (Turco 4512-A 10% Normal) solutions in conjunction with a floor sc rubber.(10) These tests were conducted on two 128 f t2 areas, onc for each solution, on the floor of the 347-ft elevation. The liquid wastes, approximately 26 gallons for each solution, were wet vacummed into 55-gal drums and then transferred to storage. The storage container is an 8000-gal tank at TMI in which all liquid wastes are combined for storage. Strippable coctings were also tested during the GDE. These coatings are applied as liquid s. Af ter drying to a rubbery film, the coatings are then peeled of f for disposal. This resuits in only solid waste which is then drummed for dis-pos al. Problems associated with the procescing of radioactive waste produced by decontamination methods other than water flushing have provided a major inhibition to their use at TMI-2.
The cleanup activities listed above are generating a large quantity of radioactive vaste.
Some of this waste is unique - such as the damaged core.
Foac of the waste concains exceptionally large rmounts of activity and is not suitable for disposal in comercisi land Furial. Disposal of these wastes is provided for in the Nemorandum of Understanding (17) (MOU) between the NRC and the Dt.pattment of Energy (DOE). The MOU states that DOE may, on an R and D or reimbureable basis, take posession of the nuclear fuel and those radio-active wastes from TMI-2 which are not suitable for disposal in commercial I
I land burial. However, a large amount of waste will go to commercial land burial.(18,19)
The radionuclides of concern in the cicanup wastes are long-lived fission products principally Cs-137 and Sr-90, rather than neutron activation prod-ucts (e.g., Co-60) associated with reactor " crud".
Fuel debris from the reac-tor core will also contain transuranic (TRU) contamination.
In wastes from 4
i cleanup of the RCS and possibly the 282-f t elevation, the TRU contanination may be of most concern. The strippable coatings being used in the cleanup contain chelating agents.
Since the cleanup of TMI-2 will continue well into the latter half of the decade, much of the waste generated will fall under the regulation of 10 CFR Part 61, "Lic ensing Requirements for Land Disposal of Radioact ive Wast e.'
The activities of alpha-emitting TRU and the long-lived fission products Cs-137 and Sr-90 are used in 10 CFR Part 61 to classify radwaste for disposal pur-poses. The waste classification section of the regulation, 10 CFR part 61.55, is attached as Appendix A.
The NRC has cont racted Brookhaven National Labora-tory (BNL) under FIN A-3162, Task 7, " Evaluation of the Three Mile Island Unit 2 Reactor Building Decontamination Process,* for technical assistance in char-acterizing these decontamination wastes.
Evaluation of the strippable coat-ings being used in the decontanination ef fort is specifically included in this task.
Evaluation of these wastes for complianc e with the provisions of 10 CFP Part 61 is a part of this characterization.
The decontamination wastes being generated in the cleanup of the TMI-2 reactor building, as of this writing, are described and their performance under 10 CFR Part 61 is evaluated.
Strippable coating properties and uses are presented along with the results of testing for radionuclide and chelating agent leachability, thermal stability, biodegradability, and radiation stabil-ity.
The activity that strippable coating waste f rom TMI-2 may contain are estimat ed f rom samples of coating f rom the Cross Decontanination Experiment.
)
P P
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2.
WASTE MANAGEMENT FOR IMI-2 REACTOR BUILDING CLEANUP 2.1 Reactor Building Decontamination Planning Reactor building decontamination planning is based on the alternatives outlined in the Final Programmatic Environmental Impact Statement (PEIS).(3,16) The specific technologies chosen have been based on ef fec-tiveness, waste management, and other considerations. The approximately chro-nological listing in Section 1 of major events in the TMI-2 cleanup also pro-vides a basis for categorizing the various streaus of radioactive waste being generated in the reactor building cleanup.
Items 1-4 have been completed and item 5 is scheduled for completion no later than the end of January, 1983.*
2.2 Decontamination Wastes Waste stream characterization for various decontamination and waste pro-cessing alternatives has been estimated in some detail in the PEIS.(3) The wastes generated from the activities of Items 2 and 3, Section 1, consisted of SDS zeolites, Epicor-II ion-exchange resins, and general trash. The SDS zeo-lite waste and the Epicor-II vaste were highly radioactive and DOE has agreed to take possession of these under the MOU, An estimate of the Classification that these wastes would have had under 10 CFR Part 61 (draf t version) is available.(20) DOE is using some of the Epicor-II wastes, Item 2, and the zeolite wastes, Item 3, in various research programs related to radioactive waste and the rest is being disposed of as 'special waste" on a one-time disposal basis.(20,21) The entire reactor core from the defueling opera-tion, Item 7, will be taken by DOE.(17) yastes contaminated by TRU concen-trations in excess of the levels authorized for commercial burial may also be accepted by DOE.(17) Some wastes from (Section 1) Items 8, 9, and 11, may be transferred to DOE under the MOU provision cited in the preceeding sen-tence. The THI-2 licencue, General Public Utilities, is disposir.g of the low leve3 waste arising from the cleanup operations at the Hanford, Washington, commercial radioacitve waste Land burial site.(18) The waste being shipped to the Hanford commercial burial site le similar te wastes bring raatinely generated at other nuclear power facilitics.
Characterization of IM1-2 redwaste la accomplished usi gamma measure-ments from which radionuclide inventories are estimated.(22g3)
Two cate-g: ries of radwsste he.ve becn defined in the reactor building cleanup so far:
(1) Normal Unit-2 radweste and (2) Makeup and purification (MUP) system rad-waste. Normal Unit-2 radwaste is being generated by cleanup activitics on the 305-ft elevation and above. The activity in normal Unit 2 radwaste has a specified f ractional composition of 0.471 Cs-137 and 0.0234 Sr-90.
The re-mainder consists of the Ba-137m and Y-90 daughters of Cs-137 and Sr-90 plus some Cs-134. The MUP system, which is located on the 281-ft level of the AFHB, consists of ion exchange resin bed demineralizers and heat exchangers.
- Personal canmunication between D. Dougherty (BNL) and D. Geifer, Bechtel l
National Corporation, at TMI, December 14, 1982.
i 7
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This system is connected directly to the reactor coolant system.
It is sched-uled to be decontaminated as part of the RCS cleanup. As of this writing no MUP radwaste has been generated. MDP radwaste has been specified to have the following isotopic fractional distribution:
0.176 Cs-137, 0.234 Sr-90, and plus 2.74 x 10-4 TRU. The TRU content limits the fission product activity in a container to avoid exceeding the 10 nCi/g limit imposed by the Hanford burial site. The container activity limits necessary to avoid exceeding the TRU limit of 10 nCi/g are listed in Reference 22.
The activity inventory in a waste container is estimated for both categories of radwaste using the follow-ing equation, Container Activity (mci) = (C.F.) (Average Measured Dose Rate [ mrem /h])
in which C.F. is a converrion factor which depends on the container and whether the activity in the waste is normal or MUP. Three types of con-tainers, each with its own C.F., are used at TMI-2 for solid waste disposal.
3 LSA boxes.
t These are compacted and non-compacted 55-gal drums and 98 f t Volume reduction capability for solid radwaste at TMI-2 consists of a 30,000 psi compactor for 55-gal drums as of the writing of this report. A 5 gph evaporator is available for liquid wastes but, as of this writing, has never been used.(12) This small evaporator was to be tested on chemical decontamination liquids but there were no plans to place the evaporator in service in the foreseeable future.* There is no solidification facility at TMI-2.
A f acility for solidifying waste in cement is available at TMI-1, but it is not allowed to be used for TMI-2 wastes. A 30-gph evaporator was con-sidered for installation in 1979, but has not been purchased.
Incineration was also a considered option for volume reduction, but an incinerator has not been purchased.(12,24)
Solid waste is packaged in 55 gal drums and 98-cu. f t. L.S.A. boxes for storage, shipment, and disposal. Campactible wastes can be volume reduced by compact ion into 55 gal drums.
As a general rule, caapaction of waste is per-formed to the maximum extent possible. However, waste that would exceed the LSA classification under compaction and waste whose activity is such that the compactor operators would receive too larga a dose are not compacted. That waste suitable for laad burial is shipped cc Hanford, Washington, comnerical burial site for disposal (18) in placarded, exclusive use trucks.
Shipments are now averaging about two per month ** although they were more frequent when radwaste shipments from TMI-2 started in November 1979.
A fully loaded f
trailer holds 20 LSA boxes (1960 f t3) or approximately 150, 55-gal drums 3
(1100 ft ) or a mixture of the two.
TMI shipment number Rs-82-028-II, April 29, 1982, may be representative of typical radwaste shipments from
- Personal communication between D. Dougherty (BNL) and P. Carmel (Bechtel) at TMI, December 17, 1982.
- Personal communication between D. Dougherty (BNL) and T. Moslak (NRC) at THI, September 8,1982.
P 8
h
TMI-2.*
This shipment contained 6 LSA boxes and 68 drums. Five of these drums labelled " compacted trash" would be Class B under 10 CFR Part 61 for the Sr-90 content or the Sr-90 plus Cs-137 content.
Gross decontamination of the 347-f t elevation was performed with water spray and a mechanical floor scrubber using a non-abrasive pad and water. The contaminated water is cleaned up using the SDS and Epicor-II systems. The activity loading of the zeolites in the SDS from this cleanup is not known.
The activity loading of the Epicor-II resins is detennined by monitoring activity levels of the influent to and effluent from the Epicor-II system.
The loading is limited to <1 pCi/g. The Hanford license ** states that resins having activities less than 1 pCi/cm3 with half-lives greater than 5 years can be disposed of in the dewatered condition. Since dewatered Epicor-II resin has an apparent density of approximately 0.7 g/cm3, the 1 pCi/g load-
- 0. 7 g/ cm3 -
ing at TMI is below that allowed by Hanford (i.e.,1 pCi/cm3 i
1.4 pCi/g). Hence, spent resins have been dewatered and shipped to Hanford for disposal.
The Epicor-II system for polishing the SDS ef fluent has three sections:
the pre-filter stage (PF), the first stage (K filter) and the second stage (2K f il te r). Through 1982 42-PFs, 8-K and 7-2K filters had been used.***
Of these totals 23-PFs, 5-Ks and 6-2Ks were stored at TMI awaiting disposal. The PFs were loaded to approximately 1 pCi/g while the K and 2K filters were loaded to lesser activities. PFs are shipped in shielded casks for disposal, 2K filters do not require shielding for shipment and K filters are shielded or not on a case by case basis. Disposal costs have led to studies of the cost effectiveness of continuing present procedures. An alternative currently being considered is to increase the activity loading on the Epicor-II resins followed by solidification prior to disposal.****
By way of illustration, the 10 CFR 61 classifications that Epicor-II radwaste would have if loaded to 1 pCi/g with activity of Normal Uni t-2 and MUP radioisotopic fractional compositions are calculated. The fractional activities of Cs-137 and Sr-90 in Normal Unit-2 Radwaste are 0.471 and 0.0234, res pe ctively. The cumulative fraction under 10 CFR Part 61 from Table 2 of Appendix A, using the gravimetric loading limit of 1 pCi/g and the apparent
- Copies of the shipping papers for this shipment were foruerded to BNL l
from a request for information on a typical solid waste shipment for commercial land burial.
- The State of Washington, Radioactive Materials License ~WN-1019-2, Amendment No. 15 in accordance with renewal application dated October 28, 1981, expiration date November 30, 1985.
9 l
- - - ~ -.
- dcasity of approximately 0.7 g/cm3 for dewatered Epicor-II resin is
- o 0.471 x 1 pCi/g x 0.7 g/cm3+
0.0234 3
x 1 uCi/g x 0.7 g/cm = 0.74 4
3 3
1.0 pCi/cm 0.04 pCi/cm 1
Performing the same calculation using the MUP radwaste fractional contents for Sr-90 and Cs-137 leads to a value of 4.2.
These numbers mean that if the
_ radionuclide fractional compositions in normal and MUP radwaste are carried 1
through in the SDS effluent to the Epicor-II system, then a 1 pCi/g loading would be Class A for the normal Unit-2 radwaste radionuclide composition and Class B for the MUP fractional distribution.
In actuality, any radionuclide composition effluent from the SDS has a-larger strontium fraction than the i
influent.. This is because cesium is held up on the SDS demineralizer beds much more than is strontium, as shown by the SDS data presented in Table I and Figure 3, Reference 11.
Specifically, the decontamination factors listed in Table I, Reference 11, for processing the reactor building sump water are 1
140,000 for Cs-137 and 590 for Sr-90.
The ratio of Sr-90 to Cs-137 activities in the influent to and effluent from the SDS were 0.04 and 10.2, respectively.
i It is therefore likely that all of the Epicor-II resins loaded to even a modest fraction of 1 pCi/g will be Class B.
4 Gross decontamination of the 305-f t elevation is being performed with water spray and a mechanical floor scrubber, as was done on the 347-f t level.
Phosphoric acid solution (Turco 4512-A 10%) will also be used on parts of the 305-ft level. The phosphoric acid radwaste solution is to be neutralized and combined with other liquid waste in the holding tank at TMI.
Since TMI-2 has
. no evaporation and solidification facilities for liquid wastes, they are i
stored. These stored liquid wastes may be solidified by vendors with mobile
)
f acilities called in for specific jobs. This was done once as part of a dem-l onstration of the Dow solidification system but there are no specific plans for disposal of the stored liquid wactes as of this writing.
Strippable coatings are. being used on some decontaminated sections of the 305-and 347-ft elevations and the polar crane to provide a protective barrier against recontamination. Except for the GDE testing, these coatings have not i
been used for decontamination on the upper part of the reactor building.
These protective layers of coating are scheduled to remain in place until after the reactor head lift which is planned ~for Spring, 1983. The activity,
that these coatings contain will depend on both how much contamination they remove from the gross-decontaminated surface and how much they receive from contamination of the top surface. The classification of the protective coat-
.ing. waste under 10 CFR Part 61 cannot be. reasonably estimated until it is peeled f rom the surfaces and assayed. prior to disposal.
Cleanup of the RCS water, Item 6, is generating highly loaded SDS zeolite and pre-filter radwaste. The pre-filters' will also conet!.n TRU activity from
~
suspended fuel debris. These wastes will be taken by D00 10 w.
z.,
,, -, = -,,-
Items, 7, 8, and 9 are interrelated to the extent that DOE will take the nuclear fuel and may take other TRU contaminated radwaste (in excess of the 10 nCi/g limit of the Hanford commercial burial site).
It is uncertain if DOE will accept wastes having TRU activities in excess of 10 nCi/g.
It is also uncertain if the 10 CFR Part 61 TRU limit of 100 nCi/g (Table 1, Appendix A) will be applied. Since much of the core has been reduced to rubble, the amount of fuel particulate contamination in radwaste resulting from these cleanup activities may be large. Therefore, SDS pre-filters from these cleanup activities may contain TRU contamination. The SDS zeolites from water cleanup in Items 7, 8, and 9 will be highly loaded with fission products and DOE will take these. Epicor-II polishing resins are loaded so as not to ex-ceed 1 pC1/g in the dewatered form. This allows disposal of these wastes at the Hanford commercial disposal site without solidification. When 10 CFR Part 61 takes effect, thic resin waste will require solidification if loaded to near 1 uCi/g, as is the practice as of this writing.
The decontamination procedures to be used in the hands on decontamina-tion cf the TMI-2 reactor building have not been determined as of this writ-ing.
Therefore, the waste streams that will be generated f rom this activity have not yet been defined. The Gross Decontamination has involved removing smearable contamination without significant surface penetration. There is some evidence that paint removal by abrasion, grit blasting or similar methods may ultimately be needed for complete decontamination.(8,25) Less of coolant accident (LOCA) tests have shown that epoxy painted surfaces do not decontami-nate readily af ter a LOCA. (25) Results from the GDE indicate that about 10%
of the contamination remains in epoxy painted surfaces after gross decontami-na tio n.
The activity remaining in floor surf aces averaged approximately 0.5 uC1/cm2 for Cs isotopes after GDE.(8)
There are no firm plans for the decontamination of the 282-f t elevation, Item 11, as of this writing. The wastes resulting f rom clesnup of this area of the reactor building are anticipated to contain very high levels of fission product activity and may contain significant TRU activity.
Waste management at TMI-2 is evolving as the waste streams from the vari-ous cleanup activities are processed.
For wastes to be disposed of by commer-cial burial, efforts have been made to limit activity to LSA limits. Wastes which do not meet the LSA criterion are stored at TMI since, as of this writing, there are no procedures for packaging and shipping wastes which do not meet the LSA criteria.(26) These non-LSA wastes include some solid waste plus the chemical liquid wastes in holding storage at TMI-2.
The liquids may be solidified by contractors brought in for these specific jobs, but this has not been planned as of this writing.
11
(
3.
STRIPPABLE COATINGS - GENERAL 3.1 Introduction Strippable coatings are film-forming compositions which are applied as liquids and which, af ter drying, are peeled from the surface. The great ad-vantage of strippable coatings for radioactive decontamination is that only solid waste is produced. They have been found to be useful for removing loose debris and smearable contamination (9,27-33) and for significantly decreas-ing the level of particulates in the air.(9) The coatings are also useful as a barrier to protect cleaned surf aces from recontamination.
If the protec-tive layer is contaminated, it can be removed and replaced or, preferably, a second layer of coating can be applied trapping the contamination betwen two layera of film before stripping. They have also been used to immobilize loose contamination on equipment to allow moving it without spreading contamina-tion.(30) Strippable coatings are most effective on smooth surfaces.
Thicker applications, multiple applications, one on top of the other, and/or reinforcement with a matrix, such as cheesecloth, may be required to success-fully strip from porous, pitted or corroded surf aces. The coating may not ad-here to some surfaces, such as Teflon, or to oily surfaces. Decontamination f actors (DF) of 10 are routinely obtained with one application of strippable coating, which compares favorably with DF obtained using liquid decontamina-tion agents.(29) 3.2 Description Chemically, strippable coatings consist of a film forming agent, usually in colloidal suspension, in a solvent. Other chemicals are added to the mix-ture to improve the decontamination or physical characteristics.
Several One(g able coating formulations have been described in theAnother(28,32) uses stri literature.
) uses polyvinyl alcohol as the filming agent.
prevulcanized rubber latex. A fast drying composition (33) uses a solution of copolymers of polyvinyl chloride and vinyl acetate. Although the filming agent alone is effective in removing loose contamination, the addition of che-lating agents such as EDTA and NTA can significantly increase the decontamina-tion effectiveness of these fo rmula tions. (28,31,32 )
A typical formulation for polyvinyl alcohol based coatings (31) was given as follows:
2-10% polyvinyl alcohol in water, 1% EDTA, 15-20% ethanol, 0.02% sodium carbonate and 1-2% glycerine or ethylene glycol. This formula-tion consists of only 4-13% solids with the remainder being solvent which evaporates to lea.e the rubbery strippable film. The ethanol (optional) is included to aid the wetting of the surf ace.
The sodium carbonate raises the pH and helps fix ions such as strontium and barium in the set coating. Glyc-erin or ethylene glycol acts as a plasticiser to keep the texture of set coat-ing elastic and easily peelable from the surface.
The prevulcanized rubber latex strippable coating (28,32) comes in lig-uid form as an alkali stabilized emulsion with water. The formulation, which 13
i a
incorporates an alkaline complexing detergent, is called Detex (for Detergent
)
latex). Fine pumice can be added to Detex to obtain the added benefit of ab-rasion when the coating is applied with scrubbing.
A f ast drying plastic strippable coating of copolymeric polyvinyl chlor-
\\
ide (83-85%) and vinyl acetate (15-17%) is claimed to film on contact and to be removable af ter only 3 minutes drying time. The rapid filming allows this coating to be sprayed onto vertical surfaces without running. However, vola-tile organic solvents which allow the rapid drying can cause ventilation and flammability problems.
The characteristics that strippable coatings produce only solid waste in relatively small volumes and can be used for both decontamination and for pro-tecting clean surfaces make them useful for radioactive decontamination. At i
least three vendors
- in the United States of fer strippable coatings for this use.
i 1
1 i
l l
1 l
i
- Imperial Professional Coatings Incorporated, New Orleans, LA; Turco Prod-ucts, Carson, CA; RAD Services, Incorporated, Pittsburgh, PA.
14 s
4.
STRIPPABLE COATINGS AT IMI-2 4.1 Introduction The strippable coating being used in the cleanup of the TMI-2 reactor building is Imperial
- ALARA 1146 Decon.
(Two strippable coating formulations -Imperial 1146 and 1148-were tested in the Gross Decontamination Experiment but only the 1146 coating was recommended for use because of prob-less encountered in stripping the 1148 coating.)
Imperial ALARA 1146 Decon is a waterborne, vinyl composition incorporating chelating agents. A technical data sheet, put out by Imperial, on the properties of the 1146 coating is re-produced in Appendix B.
The identities and quantities of chelating materials in 1146 are considered proprietary by Imperial.(34) According to the data sheet, the coating decontaminates a surface both physically and chemically; it absorbs and chemically binds heavy metal isotopes while wet and, upon curing, mechanically locks these absorbed contaminants plus smearable and loose con-taminatiou into a polymer matrix.
It is also claimed to be useful for pro-tecting clean surfaces, for inhibiting airborne contamination and for shield-ing against beta emitters.
(A typical 25-mil thickness of 1146 coating would provide effective shielding for low energy beta emitters, up to approximately 0.3 MeV.(35) However, this 25-mil thickness would be almost completely in-effective for shielding the 2.2 MeV beta electron from the Y-90 daughter of S r-90. )
Imperial 1146 strippable coating was tested on the floor of 305-f t elevation of the TMI-2 reactor building as part of the Gross Decontamination Experiment (GDE). This surface (l) consists of a concrete substrate finished with K and L** 7107 epoxy primer and K and L 7475 epoxy paint.
(All of the concrete surf aces in the reactor building are finished with K and L 7107 and 7475 epoxy primer and paint; reference 1, Table 4.1.)
For testing (10) in the GDE, the 1146 coating was applied to an approximately 500-f t2 area of the 305-ft. elevation in the northwest quadrant north of the open stairwell.
This application followed low pressure water flushing of the area to remove loose debris and water soluble contamination. A description of this test of the 1146 strippable coating including the volume and an estimation of the ac-tivity of the radwaste produced is included in Reference 10.
(This informa-tion is detailed in Section 4.3 of this report along with other experimental results on TMI strippable coatings from the GDE.)
4.2 Testing ALARA 1146 Strippable Coating for Characterization as Radwaste Tests were performed on ALARA 1146 coating purchased from Imperial Pro-fessional Coatings, Incorporated, and on samples of ALARA 1146 from the GDE testing on the 305-ft. elevation floor of the TMI-II reactor building. Char-acterization of ALARA 1146 is based on test results for leachability of Co,
- Imperial Professional Coatings, Incorporated, P.O. Box 2977, New Orleans, LA, 70189.
- Keeler and Long, Incorporated, Watertown, CT.
I 15
Sr, and Cs radionuclides and chelating agents and on tests for stability to-ward irradiation, biodegradation, immersion and heat.
Scoping test results on
~
the purchased material are presented in Section 4.1 while test results on the samples of the actual TMI-II coating are presented in Section 4.2.
4.2.1 Scoping Tests 4.2.1.1 Sample Preparation Samples of ALARA 1146 coating were prepared according to directions in the technical data sheet, Appendix B.
The shelf lif e of liquid coating mix is 4 months according to Appendix 8.
The liquid coating mix was used within 1 month of its arrival in accordance with vendor recommendations. The liquid i
material was stirred before and during application and was spread evenly over a surface to cure for at least 2 days (full cure) before stripping.
The coating was spread with a groved Teflon sheet instead of being sprayed or applied with a roller as is suggested in the technical data sheet.
Sprayi ng and rolling were impractical on the laboratory scale of these tests, and as long as the coating is merely spread and not mechanically scrubbed, the method of application should not matter. The surfaces on which the coating was spread included Lucite sheets and small painted forms
- of concrete and steel similar to the epoxy painted surfaces in the TMI-II reactor building. The stripped coating used in all of the scoping tests was 18 + 2 mils thick.
For the radionuclide leach tests, 40 mL of coating mix was spiked with 0.2 mL of an aqueous solution containing 20 pCi Cs-137, 40 pCi Sr-85 and 40 pCi Co-60.
Spiked samples for leach testing were prepared this way rather than by con-taminating the surf aces with the spike, allowing it to dry and then applying the coating in order to avoid contamination of the painted samples and to prepare the contaminated coating such that the activity would be held in the coating as firmly as possible. This would then provide radionuclide leach data which should be a measure of the best perf o rma nc e (i. e., lowest leach rates) that could be expected of coating used for actual decontamination.
There was no detectable activity remaining on the epoxy painted forms af ter the coating was stripped indicating that the distribution coefficient between the coating and epoxy painted surfaces is large and in favor of the coating for the radionuclides as tested. The spiked coating mix was stirred for five minutes before being applied to the K and L epoxy painted specimens for cure.
4.f.1.2 Radionuclide Leach Testing for Sr, Cs and Co I
Scoping test samples of ALARA 1146 spiked with Cs-137, Sr-85 and Co-60 were prepared as described in Section 4.2.1.1.
Half-inch square test specimens - three each f rom the K and L painted steel and painted concrete l
- Specimen forms furnished by the Keeler and Long Corporation (K and L) i included 2x4x1/4-in. carbon steel coupons primed and painted with 6548/7107 epoxy primer and E-1-7475 epoxy enamel and 2x4x2-in. concrete blocks finished with 4129 epoxy concrete curing compound, 6548S epoxy surfacer
(
and E-1-7475 epoxy enamel.
I 16
l forms - were cut f rom the spiked coating which had been peeled f rom the paint ed surfac es. These specimens were then leached in 32.3 mL deionized water for a leachant volume to specimen surface area ratio of 10.0 cm.
Leachants were changed and the activity remaining in each of the test s p ec i-mens was counted at 1,2,3,4,8,9, 11, 15, and 18 days. A GeLi detector was used for gamma counting. Counts were integrated over 5 or 10 minutes, a decay correction applied for Sr-85, and the measured activity compared to that initially present in the sample. Figure 4.1 shows a plot of leached activity at each of the points of measurement stated above.
The activity eached was calculated according to Eq. (4.1) for each isotope.
I Activity Leached (%) =
[ Co - Ci x 100 (4.1) i CO C0 and Ci are the initial activity and the activity at day 1, r es p ec-tively, measured in a test s p ec im en.
The standard deviation of the Sr measure-ments (six test specimens) is indicated by the error bar attached to each date point. The standard deviation for the Co and Cs data were all less than the width of the symbols and are not shown.
o_ 4 _ __ _
o-e
(._._ _ _.. m._
r o.
u o _
o
=
I o
t o-~
/
5
(
o<
@ o__
4 60Co
- O--
85 137Cs a f
o I
l 0
5 10 15 Time (days)
Figure 4.1 Percent activity loss vs time for Imperial 1146 strippable coating leached in deionized water.
17
I The data in Figure 4.1 indicate that Cs and Co leached readily from I
the coating upon contact with water. All of the Cs activity leached from the coating within four days with over 95% being released in the first leach vol-ume.
The behavior of Co paralleled that of Cs except that the coating re-j tained about 4% of the initial Co activity. The behavior of Sr was more com-plex and varied considerably between test specimens, as indicated by the mag-nitudes of the error bars. About 5% of the Sr activity leached immediately.
However, the curve shows 'that a longer period of contact with water was needed to release the major part of the Sr activity from the coating. About half of Sr activity was released readily and then the rate of release slowed markedly.
The radionuclide leaching of the coating measures the combined result of at least two ef fects: (1) the ef fect of the chelating agents in the coating and, (ii) the properties of the coating material itself. Chelating agents tend to solubilize metal ions and would be expected to enhance leachability whereas radionuclides more strongly bound to the coating material leach less i
readily. The relative quantity of chelating agent to radionuclide metal ion generally is very large and not a limiting factor.* One may speculate from Figure 4.1 that the Cs and most of the Co may be held in the set coating in the chelated form whereas most of the Sr may be bound preferentially to the coating.
4.2.1.3 Chelate Leach Testing The ALARA 1146 strippable coating contains chelating agents the identities and quantities of which are considered confidential by Imperial.**
According to the definition in 10 CFR Part 61.2, " Chelating Agent" means amine polycarboxylic acids (e.g., EDTA, DTPA), hydroxy-carboxylic acids, and polycarboxylic acids (e.g., citric acid, carbolic acid, and glucinic acid).
10 CFR 20.311b requires that radioactive waste containing more than 0.1% by weight of chelating agents be identified on the shipping manifest and that the weight As part of an agreement (3gercentageofchelatesinthiswastebeestimated.) between Imperial Profession Coa Imperial disclosed the identities and quantities of chelating agents in the ALARA 1146 composition to BNL.
Based on this information from Imperial, BNL performed standard analyses for the chelating compounds in leachate samples f rom leaching ALARA 1146 coating in deionized water. The results from these tests on chelates leached from the coating were consistent with the quantities l
- The quantities, in gram atoms (or moles) per UCi of activity, of the radionuclide of concern are: Cs-137 = 8.4x10-ll, S r-90 = 8.1x10-ll, Co-60 = 1.5x10-ll. Chelating agents generally have molecular weights in the range of about one hundred to several hundred. Assuming, for purposes of conservative calculation, a molecular weight of 1000 and a concentration of 0.1% by weight results in lx10-6 moles of chelate per gram of material.
Therefore, f or an activity loading on the order of 1 pCi/g in our hypo-thetical example, there would be more than 10,000 chelate molecules for each radionuclide ion.
- Personal communication between R. Taylor (Imperial) and R. Barletta (BNL), June 7, 1982.
18
of chelates which were indicated by Imperial to be present in the ALARA 1146 composition. As such, the concentration of chelating agents in the coating l
wastes does not require identification on the shipping manifest.
4.2.1.4 Immersion Testing Immersion tests were conducted to determine changes which might be ex-pected when strippable coatings are contacted with five liquids that are be-lieved to bound the range of liquid compositions that may be found in a shal-low land burial environment. These included a commercially prepared liquid scintillation counting (LSC) cocktail, toluene, xylene, deionized water, and deionized water saturated with xylene and toluene. The LSC cocktail forms a two phase suspension with water so this mixture was not tes ted. The results for water and the water saturatea with oluene and xylene are the more rele-vant to conditions expected in a burial environment.
One-inch square coating specimens were weighed and immersed in 50 mL of the liquids. Four samples were exposed to each liquid. Weight gain was measured as a function of time at 1, 2, 6, 8, 9, 13, 17, and 34 days of immer-sion. Data were calculated as:
Weight Gain (%) = Ni~N x 100 (4.2)
Wo where Wi is the weight of the sample at time i and W is the weight of the o
sample before immersion. These data are shown in Figures 4.2 and 4.3.
Af ter 34 days of immersion, the samples were removed from the liquids to air dry in the laboratory.
Samples were weighed af ter 1, 2, 3, 4, and 7 days and percent weight changes calculated with respect to the pre-immersion weights. These data are shown in the plots in Figures 4.4 and 4.5.
(The points at day zero in Figures 4.4 and 4.5 are the same as those at day 34 in Figures 4.2 and 4.3, respectively.)
Dimensions of the initially 1-in. x l-in. samples were measured fol-lowing the immersion phase and again following the drying phase. The immer-sion liquids were also evaporged and the residues weighed. These data are listed in Table 4.1 along with c g ruted values for the volume gain based on the observed weight gain and the density of the immersion liquid.
These data indicate that ALARA 1146 coating is af fected by exposure to water and the organic liquids tested.
lhe coating swells upon immersion and shrinks upon drying. A part of the coating dissolves upon immersion, as evidenced by the presence of a residue upon evaporaticn of the immersion lig-uid.
For water and water saturated with toluene and xylene, the residue is a dry powder. For toluene and xylene the residue is a viscous oil.
In water and the water solution saturated with toluene and xylene, the sum of the final weight (af ter drying) and the residue weight is less than the initial weight, as shown in Table 4.1, whereas for toluene and xylene, the final weight plus the residue weight is greater than the initial weight by approximately 10%.
This may suggest that toluene and xylene react with some constiraent of the coating.
19
o
@~~
O TOLUENE T
O XYLENE i
6 WATER SATURATED WITH 4
.i.
o TOLUENE AND XYLENE 4.
[
-}- WATER
()
IJ ny*
(__
U 9.G v
)
o
~~
~~
06.
t$o g
C O
s-re r-A g
f_
O uw f
,c
,1,
-A U
J o.
T g o.
cg v 0
w l
o
' A I
d.
N O
0.0 5.0 10.0 15.0 20.0 25.0 NO.0 E
Time (days)
Figure 4.2 Percent weight gain of ALARA 1146 strippable coating upon immersion in toluene, xylene, water, and a solution of water l
saturated with toluene and xylene.
20
i
}
=
2 l
2 l
c-
~
)
\\
=
L L
T
+-
)
h3-(
E' l
-E 3-3:
=
E LSC Cocktail 20 50 10 0
- 50 20 0 25 0 63 0 25 0 Time (days) t I
Figure 4.3 Weight gain of ALARA 1146 strippable coating i
upon immersion in LSC cocktail.
21
AI5 O'
O TOLUENE O XYLENE A WATER SATURATED WITH Q,
o.
TOLUENE AND XYLENE
-}- WATER i
n M
s v
0 $.
Ona 1
C C
~
.C
\\
O
- O
.c o -
00 ~
ou
.x.w
~r o
~'
I U
G w
- 1 Oo 4
4 4
4 4
a 0.0 1.0 2.0 10 4.0 5.0 0.0 7.0 Time (days)
Figure 4.4 Drying curves at room temperature in air for ALARA 1146 stippable coating following 34 days immersion in toluene, xylene, water, and a water solution saturated with toluene and xylene.
22
E.
s P
LSC Cocktail 4
- c--
8c. -
A M
e
'9~
s 6~ 3 1
-i s
C
\\
-3
=, r s
N 4
x-Y.N 3:
s.
E 00-10 20 30 40
=a 60 73
]
Time (days) i Figure 4.5 Drying curve at. room temperature in air for ALARA 1146 strippable coating following 34 days immersion in LSC cocktail.
J Tahl e 4.1 j
Immersion and firying Data (nr AI. ARA 1146 St rippable Coat ley 1
Immernion Phase Drying Phane Weight Cain (2)
Volume Gain (%)D Post Immersion Weight Fi nal Residue Maximum Finale Maximum Fi nale Dimenston Loss Dimension Weight (% of 1ritial Liquid (in + 0.01 in.)
(%)
(in + 0.01 in.)
Sample Weight)c Vater 38 + 2 23 + 1 84 53 1.01 6.2 +.1 0.88 4.0 Water Saturated with toluene 4,
and sylene 88 + 6 37 159 78 1.08 14.6 +.2 0.8A 5.7 Toluene 76 ~ 2 72 146 146 1.20 18.R i.2 0.87
'2A.7 Iylene 57 + 2 56 113 113 1.15 19 %.4 0.98 28.2 i.5C Cocktait 751 + 32 534 1.9
~
not availabled aAf t er 34 days immerainn.
bCalculated from weight gain data using the formulas Volume increase (%) =
-1 x 100 where n and W are l'.24 g/cm3 4
and the initial sample weight of each lal-in. piece of coating, renpettively, and o
e p1 eM Vg are the density of an weight gain in each of the immersion liquida, except 1.SC cocktail.
cThe leach liquids of each type were combined and evaporated so the value listed is the average for the four samples in each 11guld.
I dThe test samples were fiung on wires to dry. These samples had lost so much mechanical strength that they dripped of f wires into. viscous puddles. The puddled residues could be weighed but their dimensions could not be measured.
23
4.2.1.5 Biodegradation Scoping Tests The blodegradability of ALARA 1146 strippable coating was determined two ways. The rate of biodegradation in soils from the Barnwell, SC, and Hanford, WA, shallow land burial sites was quantitatively monitored by measur-ing the CO2 Produced from microbial respiration. Microbial CO2 Production in soil was monitored using the special flasks and procedure described by Bartha and Pramer.(36) Biodegradation in soil (37-41) is controlled by the inherent biodegradability of the material plus related soil f actors including moisture and nutrient mineral content, temperature and microbe types and popu3ation. In addition an evaluation of the ability of microbes to utilize the coating as a food source was performed using the ASTM tests G21 (Determin-ing Resistance of Synthetic Polymeric Materials to Fungi) and G22 (Determining Resistance of Plastics to Bacteria). The Branch Technical Poeition on Waste Form (204.1.5/TCJ/1/5/83, February 14, 1983) specifies the use of G21 and G22 for testing resistance to biodegradation of waste forms which fall under 10 CFR Part 61 Class B or C.
In G21 and G22, an ideal environment of moisture, temperature, and nutrient minerals is provided; these tests measure the inherent biodegradability (i.e., ability to be used as a food source) of a material to the microbes specified in the tests.
For biodegradation of Imperial 1146 strippable coating in soil, the 1146 coating was cut into strips approximately 1/2 x 2 inches. Fifteen to twenty of these strips with a total weight of approximately 12 g were used for each sample. Each of these samples was mixed in a special flask with 100 g moist soil (the moisture content was adjusted until the soil was wet but still porous and not muddy) from either the Barnwell or Hanford shallow land burial site. Two sample-in-soil tests were run for both of the burial site soils to provide an estimation of the reproducibility of the results. One sample plus 15 mL water was loaded into a flask without soil and two more flasks contain-ing 100 g of the moist soil f rom the Barnwell and Hanford sites were used as controls. The difference between the CO2 Produced in the flasks containing soil p,lus sample and the control containing soil only indicated sample biodeg-radation (more CO2 from sample flasks), inertness (no difference in CO2 production) or toxicity (decreased CO2 Production). A standard potassium hydroxide (KOH) solution, 0.100 M, was used to absorb any CO2 Produced in a flask. When the K0H solution absorbed 00, less acid was required for 2
titration of the solution to the phenolphthalein indicator end point. To obtain the amount of CO2 generated, the KOH solution was removed f rom the flasks and titrated with 0.0500 M hcl solution. The amount of acid needed to titrate the remaining KOH is related to the CO2 absorbed by CO2 (mg) = (B-V)NE where B = mL hcl to titrate the soil control, V = mL hcl to titrate the sample in the soil, N = normality of the hcl and E = 22 (the equivalent weight of CO ).
The time interval between titrations is determined by the rate of 2
CO2 generation so as to not expend more than 2/3 of the KOH in absorbing CO -
2 24
k-I f
Results from the blodegradations in Barnwell soil, El and B2, and in Hanford soil, H1 and H2, are shown in Figure 4.6 and listed in Table 4.2.
(
Figure 4.6 shows the total quantity of QO2 generated by each of the four t
samples with time. Table 4.2 lists data for each including the quantity of 002 generated and sample weights before and af ter the 220 days of blodegra-dation. The amount of biodegradation that has occurred is estimated from these data. Carbon dioxide generation monitored af ter the samples were re-moved from the soils indicated that microbial activity continued at a level
- significantly above that in the soil controls but was decreasing steadily with time.
8 O B1 0
,..... * * *... +
~
O B2
+
~
l pl a H2 pr
.' [
e o
g{s#"#
U n
mUo
. s
. u$
'if / /
E
- ~f 9
,d [
.c 5
M
/
o
- e r
d/
a l
o N
la S
if E
5 o
8.-.b' 00 400 ft0 0 110 0 K00 400 0 J 800 030 0 Tune (days)
-Figure 4.6 Biodegradative CO2 gas evolution from scoping tests on Imperial 1146 coating in Barnwell and Hanford soils.
25
Scoping Test Results on the Biodegradation of ALARA 1146 Strippable Coating in Soils From the Barnwell, SC, and Hanford, WA, Shallow Land Burial Sites Initial Final Biogenically Carbon Losta Weight Weight Generated
%CD W
-Wf x 100 C
n Sample Wo (g)
Wf (g)
CO2 (mg)
%C 0.6 Wo B1 12.11 11.25 875.5 3.77 6.28 7.10 B2 11.97 11.05 961.0 4.19 6.98 7.69 H1 11.24 10.31 1094.5 5.08 8.47 8.27 H2 11.24 10.48 1189.5 5.52 9.20 6.76 Control 11.74 12.18 nil
-3.75d aFrom the biogenically generated CO2 datc and the measured 52.3% C content of the coating used in these tests.
bThis is the estimated upper limit of biodegradation based on this CO2 generation data and published observations, References 39 and 40, that as j
little as 60% of the available carbon may be evolved as CO2 during blodegradation.
cPercent biodegradation from the initial and final weight following 220 days of biodegradation in soil at 20-240C.
dThe sampla did not detectably biodegrade in the absence of soil. The weight increase should not be due to retained water since all samples were dried both
- mfore and af ter the tests in the same environment.
Carbon dioxide monitoring provides a lower limit to the amount of biodegradation that has occurred. This is so for two reasons. First, as little as 60% of the biodegradable carbon in a substrate may be evolved as CO ;(39,40) the remainder is incorporated into additional micrabe mass or 2
excreted as metabolic waste other than CO.
Second, only carbon metabolism 2
is measured by this technique; thus, neither the metabolism of hydrogen and oxygen to produce water nor other transformations (e.g., sulfate or nitrace l
metabolism) which may occur in the substrate are accounted for in this method.
However, CO2 monitoring does provide a quantitative measure of biodegrada-tion and allows measurement of changes in the rate of biodegradation with 4
time.
Comparison of initial and final sample weights should be a direct l
measure of the amount of biodegradation that has occurred. Difficulties in separatirg tightly bonded soil f rom the coating, the loss of very small amounts of coating with the soil removed from the coating surface and contri-butions from microbe mass retained on the coating complicated the interpre-tation of this data. The weight added by retained soil and microbe mass is 26
offset to some degree by coating removed with soil.
A greater quantity of soil was incorporated into the coating than coating was scraped off with the soil (with the probable exception of B2 which is the only sample for which the biodegradation measured by weight change did not lie within the range of prob-able biodegradation as determined by carbon dioxide monitoring). These f act-ors considered together led us to conclude that the final weights of the sam-ples in soil (i.e., B1, B2, H1 and H2) may be each slightly high, perhaps by as much as, but certainly not more than, a few tenths of a gram.
There are several prominent features in Figure 4.6.
Two of these f ea-tures, the 10-day initial part of the curves with very low rates of CO2 g en-eration and the " kinks" in the curves from day 84 to approximately day 94, re-sulted from unplanned occurrences during the tests.
The initial region of each of the curves is characterized by a low rate of CO2 evolution followed by the onset, at day 10, of much higher rates of blodegradation. This was probably caused by cold temperatures (45-570F daytime, colder at night) due to failure of a steam valve which supplied heat to the lab.
From day 10 on, lab temperature was maintained between 68 and 740F.
The region of reduc ed CO2 evolution, which is especially notic eable in samples B1 and H1, f rom day 84 to about day 94 was caused by an accident on day 84 in which some alkaline solution was spilled into some or all of the samples. The microbes recovered from this trauma and the rates of CO2 generation returned to pre-accident levels within 10 days.
A difference in the way the coating samples were mixed with the soils is probably the cause of relative dif ference in the B1, B2 and H1, H2 curves.
Sample B1 was mixed with soil such that it had approximately 50% more surface area in direct soil contact. Specifically, there was less overlapping of coating stripe in B1 than in B2, H1, and H2.
B1 initially biodegrades sig-nificantly faster than B2 whereas, the rates for H1 and H2 were very similar to each other.
It appears that biodegradation of this material proceeds faster in Hanford soil, all other things being equal. The rates of CO2 gen-eration from all four samples are essentially. equal from about day 80 to 120, at which point the rate of CO2 generation in B1 slows significantly. The reason for this slowing is not clear.
It should not have been caused by the accident at day 84 since there was a complete recovery in CO2 generation and no similar slowing occurred in H1 which was also severely aff ected by the accident. Although CO2 generation from all of the samples was gradually slowing, this effect in B1 was much more pronounced.
1 The samples were removed f rom the soils on day 220 of the experiment.
CO2 production from these soils minus samples decreased significantly but did not return to background. This may be due to the small quantities of coating adhering to the soil in direct contact with the samples and possibly due to a soluble fraction of the coating, as is indicated by the immersion testing (Section 4.1.4), remaining with the soil. The rate of CO2 genera-tion in the u ils af ter sample removal gradually decreased toward background with time.
27 m
The ASIM tests G21 and G22 were run on coating samples approximately 1 x 1 in. in size. G21 (fungi) showed no growth, i.e., a growth rating of zero on a scale of 0-4 in the test.
G22 (bacteria) was positive for bacterial growth on the coating specimen.
(There is no quantiative scale in G22, results are reported as growth or as no growth.)
4.2.1.6 Radiation Testing The stability of ALARA 1146 strippable coating to ionizing radiation was tested using coating specimens sealed into Pyrex tubes both in an air atmosphere and under vacuum. Gamma irradiation was carried out in the Co-60 irradiation facility at BNL.
Data included pressure measurements for the tubes both before and af ter irradiation and mass spectroscopic analysis of the gas in the tubes af ter irradiation. The coating generated gas when exposed to ioni::ing radiation, as was expected, since chemically, it is an organic sub-strate.(42,43) It also was changed physically by the irradiation, becoming much stiffer and stronger with increasing dose. This change could also be an-ticipated since the polymeric coatin posure to ionizing radiation.(43,43)g would be expected to cross-link upon ex-The measured data from the sealed tube irradiations in air and vacuum are shown in Figure 4.7.
Three experimental configurations, as differentiated by the symbols, were used in the 20 cm3 sealed tubes: squares - 0.5-g coat-ing in air; triangles - 5.0 g coating in air, and ciceles - 5.0 g coating in vacuum. The vertical bars through the center of each symbol represent the standard deviation of the four measurements at each data point. The 0.5-g sample irradiations were performed first to guide the selection of points for further study. A preliminary gas analysis indicated that the negative slope of the line defined by the square data points up to about 108 rad was caused by oxygen depletion. At doses greater than 108 rad, radiolytic gas genera-tion became dominant. The line defined by the square data points at 5 x 108 rad and 1 x 109 rad indicate a G value for gas generation of 1.3.
The quantity of gas generated per gram of coating was less in the sealed tubes containing 5.0 g of coating than in the preliminary tests using 0.5 g coating. This is probably due to the higher pressures, as listed in Table 4.3, produced in tubes with the larger amount of coating. The higher pressures would increase the extent of back reaction (i.e., the coating would react with the radiolysis gas and reduce the net amount of gas produced) and a lower apparent G value. The pressure in the tubes containing 0.5 g of sample irradiated to 1 x 109 rad was slightly less than 2 atm.
Table 4.3 lists the gas generation data for doses of 108 rad and greater, for the points indicated by circles and triangles in Figure 4.8, i.e., for the 5.0-g samples irradiated in air and under vacuum. The pressure in these tubes at the end of the irradiation is also listed.
For reference, the composition of dry normal air at sea level and a gas analysis for an unir-radiated 5.0 g sample sealed from air are also listed. The sensitivity of the i
gas analysis is in the range of 0.05% by volume for detection. The data show that oxygen is scavenged from air over the samples and that hydrogen is the 28 i
7, primary radiolysis product being generated. Carbon dioxide, hydrocarbons in-cluding methane, ethane, propane, butane plus others and a small amount of carbon monoxide account for the renainder of the gas produced by radiolysis of the coating. Radiolytic gas generation causes also the coating to blister, as shown in Figure 4.8 for a piece of coating irradiated to 1 x 109 rad.
The gas analysis of the sample which was sealed in a tube but not irradiated was surprising - the coating appears to scavenge oxygen and liter-ate carbon dioxide, carbon monoxide, and a small amount of hydrogen spontan-eously at room temperature. The unirradiated sample remained sealed for 7 weeks prior to analysis.
It seems unlikely that biodegradation could be re-sponsible for the oxygen uptake and CO2 generation in the sealed tube based on the CO2 monitoring results of Section 4.1.5 on the moistened coating not in soil. Also, there was no moisture in the sealed tube to support biodegra-dation, and carbon monoxide, which was produced in the scaled tube, is a prod-uct of decomposition, r.ot biodegradation.
1 C
P s
~
c 9
k 3.
d 7
if
. c-5' l
}
5 E
}L c.
g 4
o I
~ e 4
l' f Ik/
nf lif lif Iki t
I r radiat ion Dow (rads)
Figure 4.7 Plot of gas generation vs Co-60 gamma dose for ALARA 1146 stri -
pable coating. Three sample configurations were used in 20 cm sealed tubes: squares - 0.5 g coating in air; trianges - 5.0 g coating in air and; circles - 5.0 g coating in vacuum.
29
T~
Table 4.3 Cas Composition Analysis From Co-60 C=ama Irradiated Tests of a
AIJutA 1146 Strippabis Coating in Air and Under Vacuum Irradiation Ces (2 by Volume)
Dose initial Pressure Other 1
0 Hg CH4 10 Hydrocarbons
( rad)
Environment (ata)
N2 02 Ar CO2 Hg CO CH4 CH26 3
ob 1.0 78.1 20.9 0.93 0.034 99.964 Oc air 0.9 72.5 11.8 0.90 10.8 0.25 3.7 99.95 5 x 108 air 4.8 13.3 12.6 65.4 2.0 2.1 0.7 2.2 1.1 0.6 100.0 L x 109 air 7.3 6.6 0.2 14.8 70.3 1.8 2.3 1.0 1.8 0.8 0.5 100.1 1 x 108 vacuum 0.7 0.09 12.3 76.8 2.0 1.9 0.6 2.5 2.7 1.2 100.09 1 x 108 vacuum 4.0 0.4 11.7 78.3 2.0 2.1 0.9 2.3 1.7 0.4 99.8 5 x 106 vacuum 6.7 0.2 15.0 75.1 2.6 2.6 1.1 2.0 1.0 0.5 100.1 dFive grams of ALARA 1146 cuating were irradiated in sealed tubes with a total volume of 20 cm3-The plenus volume is 16 cm3 based as the 1.24 g/cm3 density of the coating se can be determined f rom specifications in the ALARA 1146 Technical Data sheet, Appendix 3.
harsal air at sea level exclusive of water vapor.
C*ata f run a sample tube that was not irradia ted. It was sealed at the same time as the tubes for irradiation and the seal was broken f ar analysis 7 weeks later.- when all of the analyses were performed.
(
}
pm- -
m., my~s - ~- w g.x Qp:
'~
f ; :4 CS:l g.:
)h n-r U
h w.-; - 4 0
$~O a
s
,e Ilr -;d j
.:$f a
?
hh
.;:r4 7
C ig Y
t y
Ery
- ~
L':.
-?
Figure 4.8 Effect of radiolysis on ALARA 1146 coating (lef t) vs unirradiated specimen (right).
4.2.1.7 Thermal Testing The thermal resistance of ALARA 1146 strippable coating has been tested accordin Materials).(44)g to ASTM E84-77 (Surface Burning Characteristics of Building This test was performed to document the compliance of the 1146 coating to NRC Draft Regulatory Guide 1.120, " Fire Protection Guidelines for Nuclear Power Plants" (Revision 1, November 1977). ASIM E-84 compares the burning characteristics of a test material to those of asbestos cement board, which is nonflammable, and to red oak wood paneling. Test results are stated numerically on a scale of 0-100 in which the asbestos cement board results define zero and the red oak wood paneling results define 100 on the scale. A 20-25 mil thickness of the 1146 coating had a flame spread rating of 20 and a 30-35 mil thickness was rated at 40.
The NRC Draft Regulatory Guide 1.120 limits the flame spread behavior to a value of 50 for use in nuclear plants.
To determine the effect of heat on the coating, without burning, 1 x 1 in. pieces of coating were sealed into glass tubes at atmospheric pressure and then heated for 24 h at 50,100 and 2000C.
there were no visible changes 31
I in the coating samples heated to 50 and 1000C whereas the 2000C sample became darker in color. Table 4.4 lists the results of gas analyses on the atmospheres in the tubes. The final pressure in all of the tubes was reduced to a small fraction of the initial pressure of approximately 0.9 atm, which is normal for this procedure.* The disappearance of most of the oxygen from the air over the samples provides additional evidence that the 1146 coating scavenges oxygen from air. Specifically, it was previously observed (c.f.
line 2, Table 4.3) that oxygen was disappearing from the air over a piece of 1146 coating in a sealed tube at room temperature. Raising the temperature, as has been done in the thermal tests, would be expected to increase the rate of reaction. Likewise, the generation of carbon dioxide, hydrogen, and carbon monoxide was anticipated plus some pyrolysis at higher temperatures to produce various organic decomposition products. The absence of the nitrogen and argon, as indicated by the low final pressure, as well as the gas analysis was surprising since these gases are inert under the thermal testing conditions used (argon is chemically inert under all conditions). A possible explanation may be that some of the organics generated by thermal decomposition have sig-nificant vapor pressure at elevated temperatures but not at room temperature coupled with a leakage of gas from the tube at elevated temperatures and high internal pressures. One could then also expect condensate in the sealed tube af ter cooling. Although none was observed, no ef fort was made to determine the possible presence of condensate at the time of the experiment and a colorless, transparent film could have escaped observation. The explanation for the low final pressures in the tubes following the thermal tests is based only on speculation and these results should remain suspect until independently verified by additional experiments.
- The heat from sealing the break-seal tubes used in these procedures, as well as in the irradiation tests, normally reduces the pressure in the sealed tubes to approximately 90% of ambient pressure.
32
Table 4.4 Cas Analysis From the Thermal Testing of a
ALARA 1146 Strippable Coating t
l Test Final Gas (% by Volune)
Temperature Pressure b Total (OC)
(atm)
N2 02 CO2 HO H2 C0 Hyd roca rbons 2
50 0.01
-------------- no t m ea s u r ed --------- ---------
100 0.01 37.3 1.4 21.7 36.6 0.9 2.0 99.9 200 0.07 11.3 0.5 74.5 6.9 4.3 2.8 100.3 Tests were performed on 1 x 1 in. pieces of coating weighing ~0.5 g which were sealed into 20 cm3 glass tubes at atmospheric pressure and heated to the indicated temperature for 24 h.
bThe hydrocarbons detected included methane plus higher alkanes and benzene.
Alcohols were also detected.
4.2.1.8 Scoping Tests Summary The results of the scoping tests-on ALARA 1146 strippable coating, as presented in Sections 4.1.2 through 4.1.7, are summarized below.
e Radionuclide leach testing in deionized water showed that Cs leaches readily and canpletely from the coating; Co leaches readily except for a small amount retained by the coating while Sr leaches from the coating much more slowly and with large varia-
- ions in leach rate between different coating pieces.
I Chelate leach testing indicated that the chelating agents con-e tained in the 1146 coating leach in deionized water. However, the quantity of chelating agent contained in the coating is less than 0.1% by weight.
e Immersion testing showed that the 1146 coating is attacked by water and the several organic liquids tested.
It absorbed liquid, as indicated by weight gain and swelling during immersion. Addi-tionally, part of the coating dissolved as indicated by weight loss upon drying and residues remaining upon evaporation of the immersion liquids.
33
t Biodegradation tests showed that the coating biodegrades readily e
in moist soil. It appeared to biodegrade faster in Hanford soil than in Barnwell soil. The amouat of biodegradation measured by monitoring CO2 Produced from microbial respiration was confirmed by weight loss measurements on the biodegraded coating samples.
The biodegradetion is probably caused by bacteria and not fungi as indicated by the results of ASIM tests G21 and G22.
Radiation stability testing showed that the coating generates gas, e
primarily hydrogen with lesser amounts of carbon dioxide, hydrc-carbons, and carbon monoxide, with a G-value for total gas of approximately 1.3.
It also consumes oxygen. Radiation improves the physical characteristics of strength and toughness of the coating and may enhance the resistance to chemical and blodegrada-In sum, the 1146 coating showed no meaningful radi-tive attack.
ation damage, except for gas generation, to doses of 10 rad.
Gas generation becomes significant (i.e., the volume of gas gen-8 erated is approximately equal to the coating volume) above 10 rad.
i Thecsal stability testing showed that ALARA 1146 is only moder-e ately flammable and is not greatly affected by heat up to 1000C l
for short periods of time.
It scavanges oxygen from the air even at room temperature.
Some pyrolysis, as indicated by the genera-tion of hydrocarbons, occurs at tenperatures as low as 1000C.
4.2.2 Testing the Strippable Coating From TMI-2 4.2.2.1 TMI-2 Coating Samples i
Samples of strippable coating f rom the reactor building Gross Decon-tamination Experimant were sent to BNL for characterization. The TMI coating samples obtained consisted of five pieces individually packaged in numbered plastic bags. A summary of data on the coating pieces received from TMI in-cluding the bag number, weight, dimensions, and a gamma-activity measurement at approximately 5 cm is listed in Table 4.5.
Table 4.5 agree with previous observations (1,8) gamma ectivities listed in The. large dif ferences in the measured that the contamination in the reactor building is distributed inhomogeneously. There were no obvious differences between coating specimens from different bags, such as dif fering amounts of debris incorporated into coating samples with grossly differing activities.
In fact, since the surf ace to which the ALARA 1146 coating was applied had been flushed with water prior to application, very little debris was observed on any of the coating pieces. Due to the inhomogeniety of the activities on the different coating pieces, there is no way of knowing whether this batch is representative of the average activity for the strippable coat-ing radwaste produced in the Gross Decontamination Experiment or whether it may provide an estimate for expected activities on strippable coating radwaste which may be produced in decontaminating other parts of the reactor building.
34
Table 4.5 l
Strippable Coating Samples from the TMI-2 Reactor Building Gross Decontamination Testing (The bag numbers were on the plastic bags containing the individual coating samples as received)
Bi g Sample Sample Sample Activity at 5 cm N<..
Weight (g)
Width (cm)
Length (cm)
(mrem /h) 0 7.77 9.0 17. 5 3.5 31 6.95 9.0 19.0 230 52 4.50 9.8 15.2 150 162 5.94 8.5 16.5 250 163 4.13 8.5 11.5 4.0 7otal 29.29 The TMI coatings were tested for radionuclide leaching and for bio-degradation. The radionuclide leach tests, in which Cs-134,137 vere moni-tored, were performed to compare results from these decontamination coatings in which the contamination was removed f rout a surface with the scoping test results in which the radionuclides were added to the liquid coating in an aqueous spike. Following this leach testing, the leachate and the leached coating specimen were sent to a commercial analysis laboratory (EAL Corpora-tion *) for analysis for Sr and Pu content. Biodegradation tests in soil were run on these coatings to clarify the differences in biodegradation that were
(
suggested by the scoping b* odegradation tests.
4.2.2.2 Leach Testing and Radionuclide Analysis on TM1-2 Coating The TMI coating sample from bag 162 (Table 4.5) was leached in deion-ized water for 43 days and then the leachate and the leached coating residue were cent to EAL Corp., a ccamercial analysis laboratory, for further radionu-clide analysis.
Figure 4.9 shows the time intervals at which the leachate was changed and the cumulative Cs-137,134 activity leached f rom the sample. Each leachate volume was counted for activity in a gamma spectrometer. The activ-l ity was quantified using identical volumes with known activities of the spe-cific radionuclides in aqueous solution. No gamma emitters other than Cs-134,137 and the Ba-137m daughter were unambiguously detected in the leach-ate.
No Cs remained in leached coating sample and no other gamma emitters were unambiguously detected in the leached coating sample. Sample 162 had 140 cm2 surface (one side); the leachaat volume used was 70 cm3 for a volume-to-surface ratio of 0.5 cm.
(The use of a ratio of 10 as the IAEA recommends was not practical because of the large volume of leachate that
- EAL Corporation, Richmond, CA, performed radiochemical analyses for Sr-90 and Pu-239,240 on both the leachate and the leached coating 'esidue.
r 35 l
would be produced. Since the total amount of leachate was to be sent out for analysis, the total volume was limited to less than 2 L.
After gamma count-ing, the leachate was acidified to about 2 N with nitric acid and combined with previous specimens in a polyethylene bottle. The acidified leachate volume plus the wash (performed at EAL Corp) from the polyethylene bottle in which the leach coating was shipped af ter the coating was removed for analysis totaled 1370 mL.
The total volume of deionized water leachant was 1050 mL for a cumulative volume-to-surface ratio of 7.5 cm.
Table 4.6 lists the data on the activities of Cs-134,137, Sr-90, and Pu-239,240 from the leachate and residue of the 'Dt1 coating sample from bag 162.
The activities listed in Table 4.6 are for the TMI sample with the highest activity, according to the gamma counting results listed in Table 4.5.
Assuming that the Cs and Sr activities in the five coating pieces are distrib-uted proportionally according to the totals in Table 4.6 and the gamma counts in Table 4.5, the total activity of the five coating pieces considered as a batch can be calculated.
Total Cs-137 activity = _52 uCi Cs-137 (3.5 + 230 + 250 + 150 + 4.0) arem 250 mrem Total Cs-137 activity = 130 pCi Total Sr-90 activity = 2.28 pCi Sr-90 (3.5 + 230 + 150 +250 + 4.0) mrem 250 mrem 4
Total Sr-90 activity = 5.8.pCi These estimated total activities along with the corresponding specific and volumetric activities are listed in Table 4.7.
The limits for Class A radwaste in 10 CFR Part 61, Table 2, are 1 pCi/cm3 Cs-137 and 0.04 pCi/cm3 Sr-90.
The estiimated volumetric activi-ties in Table 4.7 exceed these limits for both of these radionuclides.
If these activities approximate those of the strippable coating, radwaste from GDE, it would be Class B under 10 CFR Part 61 and hence would require
(
stabilization.
{
The data in Table 4.6 and 4.7 were used to calculate the expected dose that a 55-gal drum of the TMI coating considered as a batch could be ex-pected to receive. This calculation, which follows the procedures presented in Reference 45, is shown in Appendix C.
The cumulative dose is estimated to be approximately 2 Mrad in 300 yrs.
36
(__
r o'
8 Cs-137 o
N*2 ccc c
c o
7
.S o so Total Activity o
l eachate 0
o Cs-137 32.0 pCi S p, Cs-134 4.3 pCi Sr-90 1.23 pCi yo Pu-239,240 0.006 nCi
.coe oo
' 8-A i
W
.?
-um
< M-o>
Bc
~po g El-o O
f ut t-Cs-134 0o0 - - -0 000 0- - - 0o0 - - - - - - - - - o - - - - - - - - - o- - - - - - - - - - - - - -o et o-a s
a s
0.0 75 15.0 225 30.0 375 45.0 525 Time (days)
Figure 4.9 Leach test results for a TMI strippable coating sample.
Cunulative leached activity for Cs-137 and Cs-134 are plotted vs time. Data points indicate leachate change intervals. The total leached activities for Cs-137, Cs-134, Sr-90 and Pu-239,240 are listed.
l 37
Table 4.6 Activity Distributiona Between Leachate and Coating for b
a Sample Coating From the TMI-2 Gross Decontamination Experiment Activity Total Specific Volumetric Activity in Remaining in Activity Activity Activity c
d Radionuclide Leachate Leached Coating in Sample in Sample in Sample Cs-134 4.3' pCi nil 4.3 pCi 0.73 pCi/g 0.91 pCi/cm3 Cs-137 52.0 pCi nil 52.0 pCi 8.7 pCi/g 10.8 pCi/cm3 Sr-90 1.23 pCi 1.05 pCi 2.28 pCi 0.38 pCi/g 0.47 pC1/cm3 Pu-239,240 0.006 nCi 0.02 nCi 0.026 nCi 0.004 nCi/g 0.005 nCi/cm3
,,m aThe Cs-134,137 activities were determined by gamma spectroscopy at BNL.
, The Sr and Pu activities were determined by EAL Corp. Richmond, CA.
D he sample came from bag 162 as listed in Table 4.5.
The coating is ALARA 1146 Decon, product T
of Imperial Professional Coatings Corp., New Orleans, LA.
cTotal activity divided by the sample weight of 5.94 g.
3 dSpecific activity in units'of pCi/g multiplied by the density of the set coating in g/cm,
3 as calculated f rom the Physical Properties fact sheet for The density. used is 1.24 g/cm ALARA 1146.
~
p Table 4.7 The Estimated Activitiesa of Cs-137 and Sr-90 in the Five TMI Coating Pieces Considered as a Batch Total Activity Specific Volumetric b
c Radionuclide (Estimated)
Activity Activity Cs-137 130 pCi 4.5 pCi/g 5.6 pCi/cm3 Sr-90 5.8 pci 0.20 pCi/g 0.25 pCi/cm3 aEstimated assuming a proportional distribution in the five TMI samples listed in Table 4.5 to the measured activities in sample 162 E
.as listed in Table 4.6.
L bTotal activity divided by the batch weight of 29.29 g.
cSpecific activity in pCi/g times the density (1.24 g/cm3) of the ALARA 1146 coating as calculated from the Physical Properties fact sheet.
4.2.2.3 Biodegradation Testing of TMI Strippable Coating Biodegradation of TMI strippable coating samples in backfill soils from the Barnwell and Hanford land buriel sites was monitored by CO2 genera-tion. The experimental procedures using the special flasks (36) were the same as those in the scoping tests of Section 4.1.5.
The sample size in these tests was greatly reduced to assure uniform mixing of-samples with soil and equal sample surface areas exposed to direct soil contact in each test. Each TMI coating sample consisted of 10 pieces, each I cm2, cut from the specimen in bag 52 (Table 4.5).
Samples were mixed with 100 g of moist soil from the Barnwell and Hanford land burial sites and loaded into the special flasks.
Soil. controls of 100 g of each of the same soils and a sample control nois-tened with 1-mL water were aise prepared. The soil moisture contents were raised to 12.5% for the Hanford soil and to 9.8% for the Barnwell soil from as received values of 8.9% and 4.2%, respectively.(46)
(The lower moisture holding ^ ability of the Barnwell soil appeared to be a result of its coarse texture _ and sandy characteristics.)-
Results of the 1HI biodegradation tests are listed in Table 4.8 and shown in Figures'4.10-4.12. Figure 4.10 shows the CO2 generation (ag) vs time from the samples in Barnwell, B, and Hanford, H, soils. Figures 4.11 and 4.12 show the CO2 generation expressed as percent carbon in the sample.
(The value of 52.3% C measured for'the purchased 1146 coating used in the scoping tests was also.used in these tests.. The actual TMI coating samples
' were not analyzed for total carbon.) The shaded area in these figures-repre-sent the range.of probable biodegradation based on the data, which is the-
' lower boundary, and -the curve prodt.ced by dividing each data point value by 39
\\
0.6, which defines the upper bounda ry.
This probable range is based on obser-vations(39,40) that as little as 60% of the carbon in a material undergoing biodegradation may be evolved as CO, the remainder being incorporated into 2
increased microbe mass or excreted as metabolic waste products other than CO. Table 4.8 summarizes the data for the biodegradation experiment moni-2 tored by CO2 generation plus data on the weight loss that occurred during the 208 days of the test.
Biodegradation measured by weight loss falls in the range of expected blodegradation from the CO2 measurements.
Table 4.8 Biodegradation Test Results for Strippable Coating Samples From the TMI-2 Reactor Building Gross Decontamination Experiment Initial Final Biogenically Weight Weight Generated W
Wf CO2
%Cb W
-Wf x 100 l
o Sample (g)
(g)
(mg)
%Ca 0.6 Wo B
0.544 0.526 48.62 4.65 7.75 3.31 H
0.554 0.505 94.86 8.92 14.87 8.84 Control 0.554 0.546 None 0
0 1.44 aFrom the biogenically generated 002 data and the measured 52.3% C content of the coating used in the scoping tests, Section 4.1.5.
bThe estimated upper limit of biodegradation based on CO2 generation and obser-vations, References 39 and 40, that as little as 60% of the available carbon may be evolved as CO2 during biodegradation.
cPercent biodegradation from the initial and final weight following 208 days of biodegradation in soil at 20-24oC.
These results support the results obtained in the scoping biodegrada-tion tests that the ALARA 1146 strippable coating used at TMI biodegrades readily in soil. The rate of biodegradation is initially rapid and decreases with time. The biodegradation of this material proceeds faster in Hanford soil than in Barnwell soil under the conditions of these tests. The differ-ences may be due to the greater moisture holding ability of Hanford soil rela-tive to Barnwell soil, or it may be related to the fact that Hanford soil con-tains more soluble ions,(46) and therefore, presumably more of the required trace minerals needed for microbe growth, or it may be a combination of these plus other factors. The blodegradability of the ALARA 1146 coating as re-flected by these test res. fir 5 shows that, under 10 CFR Part 61, Class B or C strippable coating radwasts would have to be isolated from the burial trench environment to prevent microbe attack.
l 40
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II' 00 30 0 60 0 90 0 idoo 150 0 180 0 210 0 Time (days) r s
Figure 4.10 Biodegradative CO2 gas evolution from TMI strippable coating material in Barnwell and Hanford soils.
i 41
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/
a 8
i 9' 0 150 0 idoo 150 0 210 0 6' 0 3' 0 0
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00 Time (days) 1 Figure 4.11 Probable range of biodegradation of TMI strippable coating in Barnwell soil.
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- Oat,08 IO IIanf*#d g 1.
43
4.2.2.4 Summary of TMI Strippable Coating Sample Tests Leach testing of the sample of ALARA 1146 coating f rom the DtI-2 GDE showed that Cs leaches from the coating immediately upon contact with water.
The Sr activity was portioned more or less equally between the coating and the leachate whereas, only about a quarter of the Pu activity was '.eached.
These results are consistent with the scoping radionuclide leach testing results and added the knowledge that Pu tends to be retained on the coating, at least for the low activity levels measured here.
The TMI coating samples biodegraded readily in Barnwell and Hanford land burial backfill soils. This susceptibility to microbe attack dictates that coating radwaste classed as B or C under 10 CFR Part 61 would require stabilization.
Decontamination using the ALARA 1146 strippable coating was shown to be ef fective (DF about 100) on the epoxy painted floor of the TMI-2 reactor building.(9,10) Its ef fectiveness in removing debris following low pressure water flushing was the best of the decontamination methods tested. Use of the strippable coating generated about 0.64 ft3 of solid compactible waste per 100 ft2 of application. Compaction would allow coating waste from decon-taminating 2000-2500 ft2 to be disposed of in one 55-gal drum.
The contamination taken up by the coating application in the decon-tamination test produced waste which, if compacted to fill a 55-gal drum, would have a contact radiation reading of less than 5 rem /h.(10) Equation 2.1 and the appropriate C.F.(22,23) for a compacted 55-gal drum of Normal Unit 2 Radwaste allows calculation of the activity contained in the hypotheti-cal drum.
Activity = (0.557 mci h/ mrem) (5000 mrem /h)
Activity = 2800 mci This activity in a 55 gal drum (210,000 cm3) results in an actitity density 3
of 13.3 pCi/cm, or 10.7 pCi/g based on the 1.24 g/cm3 density for the coating. This activity is consistent with, though higher than, the measured activities in Tables 4.6 and 4.7.
s l
The measured activity on the strippable coating radwaste used for de-contamination in the GDE provides evidence that such decontamination applica-tions may be expected to result in Class B radwaste order 10 CFR Part 61.
As 1
of this writing, it is not known whether the coating will be used for decon-1 tamination activities beyond its evaluation in the GDE testing. It is being
)
used as a protective coating over cleaned surf aces in the upper part of the reactor building including the 305 and 347-ft elevations and the polar crane.
Class B strippable coating radwaste would require stabilization under 10 CFR l
44 1
. - ~... -._.
1
~
Part 61.
Imperial Professional Coating has experimental solidification sys-tems* for this purpose; however, these solidification systems have not yet been evaluated for compliance with the stability provisions of 10 CFR Part 61.
a i
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i 1
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Personal communication between J. Adams (BNL) -and H. Lomasney (Imperial),
l July 30,1982, " Solidification Systems for Strippable Coatings."
i i
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REFERENCES 1.
TP0/TMI-007, "Three Mile Island Unit 2 Recovery Project Technical Plan for Reactor Building Gross Decontamination," Bechtel Ns.tional, Incorporated, May 1981.
2.
TP0/IMI-009, "Three Mile Island Unit 2 Recovery Project Decontamination Techniques Test Program," Bechtel National, Incorporated, May 1981.
3.
" Final Programmatic Environmental Impact Statement Relat ed to Decontami-nation and Disposal of Radioactive Wastes Resulting Fron May 28,.1979, Accident Three Mile Island Nuclear Station, Unit 2," U.S. Nuclear Regula-tory Commission," Office of Nuclear Reactor Regulation, KUREG-0683, March 1981.
4.
D. L. Geifer, " Planning for the TMI-2 Reactor Building Decontamination Experiment," Decontamination of Nuclear Facilities 2, International Joint Topical Meeting ANS-CNA, Chapter 5, 1-10, Sept ember 1982.
5.
D. Leigh, L. J. Coe, and G. R. Eidam, " Data Acquisition and Characteriza-tion for LSDE at TMI-2," Ibid,11-26.
6.
J. V. Gilbert, "TMI-2 Containnent Entry Program," Ibid, 27-46.
7.
H. L. Hondorp, " Planning for the TMI-2 Reactor Coolant System Decontami-nation," Ibid, 69-80.
8.
M. Drackley, " Preliminary Analysis of Reactor Building Cross Decontamina-tion Experiment Core Boring Samples," Memorandum to L. H. Barrett, TMI Program Office, USNRC, June 17, 1982.
9.
D. R. Dougherty and R. E. Barletta, " Trip Report - Meeting at TMI on June 18, 1982," Memorandum to File, Brookhaven National Laboratory, Nuclear Waste Management Division, June 30, 1982.
10.
P. R. Bengel and R. L. Mason, "Results of the Gross Decontamination Experiment and Implications for Future Decontamination Activity,"
Decontamination of Nuclear Facilities 2, International Joint Topical Meeting ANS-CNA, Chapter 5, 47-68, September 1982.
11.
K. J. Hof stetter, K. J. Hitz, C. G. Lookabill, and S. J. Eichf eld,
" Submerged Demineralizer System Design, Operation and Results," Ibid, 81-90.
12.
R. S. Daniels and J. d. Rodabugh, " Waste Management Planning for the Recovery of TMI-2," Ibid,91-100.
13.
H. Dieckamp, " March 28,1979 Plus 42 Months or a Status Report on the l
TMI-2 Cleanup Program," Decontamination of Nuclear Facilities, Keynote Add ress es, International Joint Topical Meeting, ANS-CNA, September 1982.
l 47
I 14.
J. Raloff, "TV at TMI: Hard Core Rubble," Science News 122 (5), 68, 1982.
15.
M. Gold, "Inside Three Mile Island," Science 82 3 (8), 16, 1982.
16.
R. Lo, and R. Bellamy, " Regulatory Aspects of TMI Cleanup" Decontamination of Nuclear Facilities 1, International Joint Topical Meeting ANS-CNA, Chapter 1, 1-5, September 1982.
17.
Memorandum of Understanding Between USNRC and USDOE Concerning the Re-moval and Disposition of Solid Nuclear Wastes From Cleanup of the Three Mile Island Unit 2 Nuclear Plant, Federal Register 47, 16229-16230, April 15,1982.
18.
R. W. Sexton, "NRC Pre-Notification Rule in Ef fect:
TMI Officials Relate Time Consuming Procedure," Hazardous Materials Transportation 5 (7), July 1982.
19.
R. S. Daniels, in an answer to a question on radwaste volumes yet to be generated in the TMI-2 cleanup after his presentation of the paper in Reference 12.
He was quoted in a newspaper article:
W. Immen, "Three-Mile A Blast is a Risk in Cleanup," The Globe and Mail, 1, Sep t embe r 16, 1982.
20.
T. C. Johnson and H. Lowenberg, " Classification of TMI Wastes," Waste Management '82, in Proceedings of the Symposium on Waste Management at Tuscon, Arizona, March 8-11, 1982, Volume II, 121-131, 1982.
21.
H. Lowenberg, "TMI-II Waste Management - DOE Programs for Handling Some Special Wastes," Memorandum to File, U.S.N.R.C., Office of Nuclear Material Safety and Safeguards, September 1,1982.
22.
Memorandum, S. R. Frey to J. Winkel, " Mathematical Factors for Curie Estimations for Unit-2 Radwste," GPU Nuclear Inter-Office Memorandum 9240-956, May 28,1982.
- 23. Memorandum, S. R. Frey to R. Hahn, " Revision of Correction Factors for Unit II Radwaste," GPU Nuclear Inter-Office Memorandum 9240-1038, June 30, 1982.
24.
Bechtel Northern Corporation, " Controlled Air Incineration Conceptual De-sign Study," GEND-021, EG and G Idaho, Inc., January 1982.
25.
B. J. Newby and K. L. Rohde, " Decontamination of Protective Coatings Following a Loss of Coolant Accident," in Proceedings of the First International Symposium on Decontamination of Nuclear Installations, pp. 117-128, 19 67.
1 26.
Bechtel Northern Corp., " Decontamination Experiment Post Execution Analysis," June, 1982.
48 i
27.
J. A. Mock, "Strippable Coatings," Materials Engineering 90(3), 86-89 (1979).
28.
B. W. Ariss, and C. R. Thomas, "The Use of Coatings to Facilitate Dacon-tamination," Proceedings of the First International Symposium on Decon-tamination of Nuclear Installations, 55-64, 1967 29.
J. F. Remark, " Plant Decontamination Methods Review," Electric Power Research Institute, Technical Planning Study, TPS78-816, EPRI NP-ll68, May 1981.
30.
C. S. Lacy, " Decontamination of a Fueling Machine Contaminated by Irradiated Fuel Debris," Presented at Corrosion /78, Paper No. 36, Houston, TX, March 1978.
31.
S. K. Robev, T. B. Sumerska, S. L. Todorov, D. K. Krstanov, T. I.
Marinova, "A Method for Deactivating Contaminated Surfaces by Film," in Proceedings Series-Practices in Treatment of Low and Intermediate Level Radioactive Wastes," pp. 751-757, IAEA, Vienna, Austria, 1966.
32.
A. Catherall and B. W. Ariss, "The 'DETEX' Technique of Decontamination,"
U.K.A.E.A. Report AHSB(S)R57, 27-39, 1963.
33.
O. A. Bernaola and A. Filevich, " Fast Drying Strippable Protective Cover for Radioactive Decontamination," Health Physics, Pergammon Press, Vol.
19, 685-687, 1970.
34.
Personal Communication, letter, between R. Barletta and H. L. Lomasney of Imperial Prof essional Coatings, Inc., June 7, 1982.
35.
Radiological Health Handbook, Bureau of Radiological Health, U.S. Depart-ment of Health, Education, and Welfare, 122, January 1970.
36.
R. Bartha and D. Pramer, " Features of a Flask and Method for Measuring the Persistence and Biological Ef fects of Pesticides in Soils," Soil Science 100(1), 68-70 (1965).
37.
R. H. Brink, " Biodegradation of Organic Chemicals in the Enironment," pp.75-100, in Environmental Health Chemistry: Proceedings of a 1979 Symposium on Chemical and Environmental Agents and Potential Human Hazards, Ann Arbon Science Press, Ann Arbor, MI, 1981.
38.
H. G. Schlegel, " Production, Modification, and Consumption of Atmospheric Trace Cases by Microorganism," Tellus 24, 11-20 (1974).
39.
G. Stotsky, " Microbial Respiration," Methods of Soil Analysis: Part 2, Chemical and Microbiological Properties, pp. 1150-1572, American Society and Agromony, Inc. Madison, WI, 1965.
49
1 40.
P. A. Gilbert, " Biodegradation Tests: Use and Value," pp. 35-45 in Biotransformations and the Fate of Chemicals in an Aquatic Environment:
Procedures Workshop, 1980.
41.
D. R. MacKenzie, F. Vaslow, D. Dougherty, and S. Chan, " Technical Factors Affecting Low Level Waste Form Acceptance Requirements," BNL-NUREG-31536, Draft Report, 33-38, August 1982.
- 42. Radiation Effects on Organic Materials, R. O. Bolt and J. G. Carroll, Editors, Academic Press, New York, 1963.
43.
Irradiation of Polymers, R. F. Gould, Editor, American Chemical Society Advances in Chemistry Series, 66, Washington, D.C.,1967.
)
44.
Southwest Research Institute, 6220 Culebra Road, P.O. Drawer 28510, San Antonio, TX, 78284; Project No. 02-5424-002, June / July 1979; The report is also available from Imperial Professional Coatings, Inc., P.O. Box i
29077, New Orleans, LA, 70189, as Technical Report No. 311-1-79.
45.
K. J. Swyler, R. E. Barletta, and R. E. Davis, " Review of Recent Studies of the Radiation Induced Behavior of Ion Exchange Media,"
BNL-NUREG-28682, Draft Report, November 1980.
46.
P. L. Piciulo, C. E. Shea, and R. E. Barletta, " Analyses of Soils at Low Level Radioactive Waste Disposal Sites," BNL-NUREG-31388, Draf t Report, June 1982.
t I
4 a
i' 50 l
l l
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~,
APPENDIX A 10 CFR PART 61.55: RADI0 ACTIVE WASTE CLASSIFICATION FOR DISPOSAL IN SHALLOW LAND BURIAL 51
Fed:ril RIgistsr / Vol. 47. No. 248 / Monday. December 27. 1982 / Rules and Regulations 57173 dispos:1 site befora they leave the site (iv) Waste that is not generally (iv)If the concentration exceeds the boundary.
acceptable for near. surface disposalis value in Column 3. the waste is not waste for which waste form and generally acceptable for near. surface i 61.54 Alternative requirements for disposal methods must be different, and disposal.
design and operations.
in general more stringent. than those (v) For wastes containing mixtures of The Commission may, upon request or specified for Class C waste. In the the nuclides listed in Table 2. the total en its own initiative. authorize absence of specific requirements in this concentration shall be determined by pr: visions other than those set forth in part, proposals for disposal of this waste the sum of fractions rule described in il 61.51 through 61.53 for the may be submitted to the Commission for paragraph (a)(7) of this secticn.
segregation and disposal of waste and approval, pursuant to 9 61.58 of this for the design and operation of a land part.
TABLE 2 disposal facility on a specific basis. if it (3) Classification determined by long.
finds reasonable assurance of lived radionuclides. If radioactive waste c',"*'M compliance with the performance contains only radionuclides listed in a.ao==
g(g obj1ctives of Subpart C of this part.
Table 1. classification shall be cd '
,a 3
determined as follows:
l 61.55 Waste classitication.
(i)If the concentration does not tea' 8 ** =** =* *a *aa 5 k '"
(a) Classification of waste for near exceed 0.1 times the value in Table 1 J" "*"*
l}
l}
surface disposal.
the waste is Class A.
cuo 7m o
n (1) Considerations. Determination of (ii)If the concentration exceeds 0.1 Q,,,,,,,,,,,
3, 5
,$ n'"3 5
iso I rwo tha classification of radioactive waste times the value in Table 1 but does not s,.so c o4 involves two considerations. First, exceed the value in Table 1. the waste is c..ist i
a yl_
m consideration must be given to tne Class C.
1n,....,,,
concentration of long. lived (iii)If the concentration exceeds the gs,,e,ar,gni gP,
, o w u..
.a C, opt 9
a radionuclides (and their shorter. lived value in Table 1. the waste is not on. n.a n.no. ous e e. coaceau.-
prscursors) whose potential hazard will generally acceptable for near surface
- ""4,*,,f*,','n ".*ffon,'"*",,;f"Jf. 'll &*.", 3 persist long after such precautions as disposal.
a *a=
- ='*
- ia* cias c 'aap=ad*a' * *"
a,,,
institutional controls, improved waste (iv) For wastes conta, m, g m,xtures of m
i form. and deeper disposal have ceased radionuclides listed in Table 1. the total (5) Classification determined by both to be ef fective.These precautions delay concentration shall be determined by long. and short lived radionuclides. If tha time when long-lived radionuclides the aum of fractions rule described in radioactive waste contains a mixture of could cause exposures. In addition, the paragraph (all7) of this section.
radionuclides. some of which are listed magnitude of the potential dose is Tm1 in Table 1. and some of which are listed limited by the concentrution and in Table 2. classification sball be availability of the radionuclide at the conc.a.
determined as follows:
time of exposure. Second, consideration
,J, *1',",
(i)If the concentration of a nuclide must be given to the concentration of listed in Table 1 does not exceed 0.1 cee shorter.hved radionuclides for which times the value listed in Table 1 the requirements on institutional controls.,
c,i4 e
class shall be that determined by the waste form, and disposal methods are cgi,47.c
.i.a concentration of nuclides listed in Table effIctive.
%,,,,,,,ve.o ene.i o2 2.
(2) Classes of woste. (i) Class A waste Tc-n 3
(ii)If the concentration of a nuclide is waste that is usually segregated from Z
,,,,m,,,,n, y,,,,,,,,,,
listed in Table 1 exceeds 0.1 times the oth:r waste classes at the disposal site.
y.
v.= *
'im value listed in Table 1 but does not Tha physical form and characteristics of M*,',
,Qo"o, exceed the value in Table 1. the waste 5
Clrss A waste must meet the minimum shall be Class C. provided the requirements set forth in i 61.56(a). If
'una... n.nm,,n e-y*a' concentration of nuclides listed in Table Clzss A waste also meets the stability (4) Classification determined by short. 2 does not exceed the value shown in r1quirements set forth in i 61.56(b). It is lived radionuclides. If radioactive waste Column 3 of Table 2.
nit necessary to segregate the waste for does not contain any of the (6) Classification of wastes with disposal.
radionuclides listed in Table 1.
radionue: ides other than those listed in (ii) Class D waste is waste that must clasification shall be determined based Tables 1 and 2. If radioactive waste mict more rigorous requirements on on the concentrations shown a Table 2.
does not contain any nuclides listed in waste form to ensure stability after However, as specified in paragraph either Table 1 or 2. it is Class A.
disposal.The physical form.and (a)(6) of this section. if radioactive (7) The sum of the fractions rule for ch:racteristics of Class B waste must waste does not contain any nuclides mixtures of radionuclides. For meet both the minimum and stability listed in either Table 1 or 2. it is C! ass A.
detcrmin.ing classification for waste that
)
requirements set forth in i 61.56.
( )If the concentration does not contains a mixture of radionuclides, it is liii) Class C waste is waste that not exceed the value in Column 1. the waste necessary to determine the sum of cnly must meet more rigorous is Class A.
fractions by dividing each nuclide's requirements on waste form to ensure (ii)If the concentration exceeds the concentration by the appropriate limit st:bility but also requires additional value in Column 1. but does not exceed and adding the resulting values. The mrsures at the disposal facility to the value in Column 2. the waste is appropriate limits must all be taken protect against inadvertent intrusion.
Class B.
from the same column of the same table.
The physical form and characteristics of (iii)If the concentration exceeds the The sum of the fractiens for the coh.mn Class C waste must meet both the value in Column 2. but does not exceed must be less than 1.0 if the waste class l
minimum and stability requirements set. the value in Column 3. the waste is is to be determined by that column.
j I:rth in i 617.,0.
Class C.
Example: A waste contains Sr.90 in a l
52 t
57474 Federal Resister / Vol. 47. Ns. 248 / M:ndry. Dec mber 27. 1982 / Rules and Regulations conceatration of 50 Cl/m* cnd Co.137 in maximum extcot przeticxble the owned in fee by the Federal or a State o concentration of 22 Ci/mL Since the potential hazard from the non-government, concentrations both exceed the values radiological materials.
(b) Institutiono/ control. The land in Column t. Table 2. they must be (b) The requirements in this section owner or custodial egency shall carry compared to Column 2 values. For Sr-90 are intended to provide stability of the out an institutional control program to fraction 50/150 = 0.33; for Cs.137 waste. Stability is intended to ensure physically control access to the disposal fraction. 22/44 =0.5; the sum of the that the waste does not structurally site following transfer of control of the fractions = 0.83. Since the sum is less degrade and affect overall stability of disposal site from the disposal site than 1.0. the waste is Class D.
the site through slumping collapse.or oper.ator. The institutional control (8) Determination of concentrations in other failure of the disposal unit and program must also include, but not be wastes.The concentration of a thereby lead to water infiltration.
simited to. carrying out an red;onuclide may be determined by Stability is also a factor in limiting environmental monitoring program at indirect methods such as use of scaling exposure to an inndvertent intruder, the disposal site, periodic surveillance.
frctors which relate the inferred since it provides a recognizable and minor custodial care, und other concentration of one radionuclide to nondispersible waste.
requirements as determined by the cnother that is measured. or (1) Waste must have structural Commission; and administration of rx.dionuclide material accountability. If stability. A structurally stable waste funds to cover the costs for these there is reasonable assurance that the form will generally maintain its physical activities. The period of institutional indirect methods can be correlated with dimensions and its form. under the controls will be determined by the expected disposal conditions such as Commission, but institutional controls actual measurements. The concentration of a radionuclisle may be averaged over weight of overburden and compaction may not be relied upon for more than the volume of the waste, or weight of the equipment the presence of moisture.
100 years following transfer of control of waste if the units are expressed as and microbial activity. and internal the disposal site to the owner.
nanocuries per gram.
factors such as radiation effects and I 61.56 waste characteristles.
chemical changes. Structural stability Subpart E-Financial Assurances (a)The following requirements are can be provided by the waste form
$ 61.61 Applicant qualifications and minimum requirements for all classes of itself, processing the waste to a stab!c assurances.
waste and are intended to facilitate form or placing the waste in a disposal Each applicant shall show that it handl.ng at the disposal site and provide container or structure that provides either possesses the necessary funds or protection of health and safety of stability after disposal.
has reasonable assurance of obtaining personnel at the disposal site.
(2) Notwithstanding the provisions in the necessary funds. or by a (1) Waste must not be packaged for il 01.56(n)(2) and (3), liquid wastes or combination of the two. to cover the disposalin cardboard or fiberboard wastes containing liquid. must be estimated costs of conducting all boxes.
converted into a form that contains as licensed activities over the planned (2)1.iquid waste must be solidified or little free standing and noncorrosive operating life of the project, including packaged in sufficient absorbent liquid as is reasonably achievable, but costs of construction and disposal.
material to absorb twice the volume of in no case shall the liquid exceed 1% of the liquid.
the volume of the waste when the waste f 61.62 Funding for disposal site closure (3) Solid waste containing liquid shall is in a disposal container designed to and stabmzetion.
contain as little free standing and ensure stability, or 0.5% of the volume of (a) The applicant shall provide noncorrosive liquid an is reasonably the waste for waste processed to a assurance that sufficient funds will be achievable, but in no case shall the stable form.
available to carry out disposal site liquid exceed 1% of the volume.
(3) Void spaces within the waste and closure and stabilization including:(1)
(4) Waste must not be readdy capable between the waste and its package must Decontamination or dismantlement of of detonation or of explosive be reduced to the extent practicable.
land disposal facility structures; and (2) decomposition or reaction at norma!
um and aWhaen d h @osM 9 61.57 t.abeting.
site so that following transfer of the pressures and temperatures. or of h pac e o was e must b disposal site to the site owner. the need explosive reaction with water.
(5) Waste must not contain, or be Y
Y for ongoing active maintenance is Class A waste. Class B waste, or class C eliminated to the extent practicable and capable of generating, quantities of toxic waste m accordance with 6 61.55.
only minor custodial care, surveillance, gases. vapors, or fumes harmful to persons transporting, handimg. or 1 61.58 Alternative requirements for waste and monitoring are required.These assurances shall be based on disposing of the waste.This does not classification and characteristics.
Commission-approved cost estimates apply to radioactive gaseous waste
.lhe Commission may, upon request or reflecting the Commission. approved
(
packaged in accordance with paragraph on its own initiative authorize other provisions for the classification and plan for disposal site closure and (4)(7) of thls section.
stabilization. The applicant's cost (6) Waste must not be pyrophortc.
characteristics of waste on a specific estimates must take mto account total Pyrophoric materials contained m waste basis. if. after evaluation of the specific capital costs that would be incurred if shall be treated, prepared, and packaged characteristics of the waste, disposal an mdependent contractor were hired to to be nonflammable.
site, and method of disposal. it fmds perform the closure and stabilization (7) Waste in a gaseous form must be reasonable assurance of compliance w rk.
packaged at a pressure that does not with the performance objectives in (b)In order to avoid unnecessary cxceed 1.5 atmospheres at 20*C. Total Subpart C of this part.
duphcation and expense, the activity must not exceed 100 curies per containct.
g 61.59 Insttv tlonal requirements.
Commission will accept financial (a) Waste containing hazardous.
(a) Land ownership. Disposal of suretics that have been consolidated rad oactive waste received from other with carmarked financial or surety biological, pathogenic, or infectious i
material must be treated to reduce to the persons may be permitted only on land arrangements estab!!shed to meet 53
APPENDIX B TECHNICAL DATA SHEET FOR ALARA 1146 DECON STRIPPABLE COATING 6
55
/N
_o
^"" "
l im,=r-rgi nLnKn
'W 1146 PROFESSIONAL COATINGS
((
l 9,, k#dENuVMM"MN N M5iFJ" 7 -.ry.w.-
p..
py,-- m -. r.,s
. www..7,
- p tv -.r-am--
l a
.A NP '-124""""""N d b%*IA*hEl L dye S b*1 NFN-DESCRIPTION: One package, water borne, strippable coating.
RECOMMENDED USES: Decontaminating radioactive polluted areas. Apply over contaminated steel, con-crete, wood, aluminum, or painted surfaces via spray or roll. Can be used for sealing (fixing) potential airborne contaminants, for protection of personnel from "smearable" contaminants. Also useful with beta emitters for
" shielding" purposes. Because this product produces an easily compacted (solid) waste, the workload on rad-waste processing facilities can be substantially reduced.
~
PRODUCT DESIGN FUNCTION:
- 1. Apply material over contaminated surface. While material is wet it attracts, absorbs and enemically binds heavy metal isotopes. During application the coating migrates into micro voids of surface to contact con-taminants. Upon cure, the product mechanically locks the contaminants into a polymer matrix. Stripping the film effectively cleans the substrate and produces a solid waste.
- 2. Apply material over clean surface to provide protective layer against future contamination.
- 3. Apply material over previously contaminated surfaces while still wet to inhibit potential airborne contami-nation during dry-out.
CAUTION: DO NOT FREEZE. Consult Imperial for specific instructions concerning high humidity applica-tions. Refer to chart on reverse.
PilYSICAL PROPERTIES:
TYPE Vinyl COLOR Yellow FLASli POINT N/A s
NUMBER OF COMPONENTS One POT LIFE @ 75*F N/A E-\\ -)
DRYING TIME @ 75'F AND 75Cc Ril C
%~.
A. TO TOUCll (FOOT TRAFFIC) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
\\
3 m
B. TO FULL CURE 2 days t
SIIELF LIFE 4 months 1
NET WEIGilT - l*s 9.0 pounds 5's 45.0 pounds 7 f PERCENT VOLUME SOLIDS 42%
HECOMMENDED DRY FILM TillCKNESS 20-30 mils TIIEORET! CAL COVERAGE AT 25.0 MILS 27 sq ft/ gal TlilNNER Water TEMPEllATURE RESISTANCE 120*F FIRE DATA (ASTM E-84-77) @ 20-25 MILS 20 (flame), 25 (fuel), 25 (smoke) l CllARCOAL FILTER 99.3% efficiency after 2 hrs. @ 40FPM face loading (methyl iodides DECONTAMINATION FACTOR 30100 (varies with substrate) 56
APPENDIX B. Continued g-
- 1146 - Page Two APPLICATION EQUIPMENT RECOMMENDATIONS:
- 1. Airless - Standard industrial spray equipment (stainless steel parts desired) such as Graco or Binks, using a 30:1 pump ratio with 50 60 psi inbound pressure and a.021" to.025" fluid tip with reversible cleaning i
head.
- 2. Holler %" lambs wool with phenolic core.
SAFETY EQUIPMENT HECOMMENDATIONS: #1146 is free of solvents and toxic materials. A par-ticulate (gauzel mask is recommended to prevent inhalation of overspray, Consult Material Safety Data Sheet.
APPLICATION PROCEDURE:
- 1. Flush all equipment with fresh we'er prior to i.se.
- 2. Stir material thoroughly before and throughou application.
- 3. Do not thin except for workability, and then with no more than 10% by volume with fresh water.
- 4. Flush all equipment immediately after use with fresh water. Use butyl cellosolve or similar solvent for final cleaning.
STORAGE CONDITIONS:
40* to 90*F. DO NOT FREEZE.
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NOTE: The technicaldata furnished is true and accurate to the best of our knou ledge; hou ever, no guaranter of accuracy is given or im-plied. We guarantee ourproducts to conform to imperial Qaality ControlStandards. We assume no responsibilities forits handling, use.
f storage. the results obtained or any snjury or damage resulting from its use.
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l APPENDIX C CUMULATIVE DOSE CALCULATIONS FOR THE ALARA 1146 STRIPPABLE COATING SAMPLES FROM THE TMI-2 REACTOR BUILDING GROSS DECONTAMINATION EXPERIMENT The procedures in this Appendix, which are based on the analysis given in Appendix A of Reference 42, outline the methods used to calculate cumulative absorbed dose as a function of time.
Table C.1 lists the radionuclides and pertinent decay parameters used for the calculatiors. A waste container is assumed to be a 55-gal drum the approxi-i mate dimensions of which are radius = 25 cm and height 100 cm.
The activity is assumed to be uniformly distributed throughout the volume of the drum. The ac-tivity densities used are from Table 4.6 and 4.7.
f Table C.1 Nuclides and Relevant Decay Data Used in Calculation of Dose to Strippable Coat!.ng Radionuclide A
t1/2 Ea pa i
(yr-1)
(yr)
MeV (rad em h-ImCi-1) 2 Sr-90 0.025 28 0.200 no y 5
(Y-90)b (95)
(.0073)
(0.931)
(no y)
Cs-134 0.33 2.1 0.152 8.7 Cs-137 0.023 30 0.195 3.3 Pu-239 24,100 5.19 no y aRef erence 1.
bDaughter of Sr-90; for the purpose of calculation, Y-90 was assumed to decay coincident with parent decay.
I b
The dose delivered by beta decay was calculated from the following equa-tions. The dose delivered by the alpha decay of the Pu-239 was calculated using the same formulation as for beta decay. The initial beta dose rate of the ith
- radionuclide, D, is:
i 6
Di= ACE.
ii 59
APPENDIX C, Continued CUMULATIVE DOSE CALCULATIONS FOR THE ALARA 1146 STRIPPABLE COATING SAMPLES FROM THE TMI-2 REACTOR BUILDING GROSS DECONTAMINATION EXPERIMEh7 Ci is the activity density.of the ith radionuclide, Ei is the average be_ta energy and A is a proportionality constant. When C is in mci /cm3 and Ei 3 rad cm3 gey-1 h-ImCi-and Di is ob-is in MeV, A equals 2.1 x 10 tained in rad p r h.
The total absorbed beta dose due to the decay of the ith radionuclide, D (=), is
'6
- B 3 h yr-1 D (=) = Di x 8.76 x 10 i
Ai where Ai is the decay constant of the ith radionuclide in years-1. The beta dose absorbed at any time may then be calculated by t
D (t) = D (=) (1-e i ),
i i
The dose delivered by gamma decay was estinated f rom the following equa-tio ns. The gamma dose rate is Y
Di = C rig-i where Ti is the gamma ray constant of the ith radionuclide and g is a geomet-ric factor, which assumes tissue equivalency.
Ti has the units rad cm mci-l -1, g has the unit cm-1 2
h The value of g = 136 was taken from the table of values of g given in Reference 2 for a cylinder of radius = 25 cm and height = 100 cm.
Y The total gamma absorbed dose, D (=), is i
D(=)=0{x8.'76x103 h
- y r-1 i
Ai i
(
and the cumulative
- gamma dose was obtained f rom l
Df(t)=Df(=)(1-e I)
E The total cumulative absorbed dose for all nuclei and decay types is shown in Table C.2.
60 l
APPENDIX C, Continued CUMULATIVE DOSE CALCULATIONS FOR THE ALARA 1146 STRIPPABLE COATING SAMPLES FROM THE TMI-2 REACIOR BUILDING GROSS DECONTAMINATION EXPERIMENT Table C.2 Total Absorbed Dose to the Average TMI Strippable Coating in 300 Years Assuming a 55-Gal Drum and the Activity Densities in Tables 4.6 and 4.7 Activity 3
y Radionuclide Denist D (300)
D (300) i i
(pCi/cm )
(rad)
(rad)
Sr-90 0.25 3.6 x 104 Y-90 0.25 1.7 x 105 Cs-134 0.47 3.9 x 103 1.4 x 104 i
Cs-137 56 8.7 x 105 9.5 x 105 Pu-239 2.6 x 10-6 75 Total = 1.08 x 106+ 9.6 x 105 Total = 2.04 x 106 rad References 1.
Bureau of Radiological Health and the Training Institute, Environmental Control Administration, Radiological Health Handbook, U.S. Government Printing Office, Washington, D.C. (1970).
- 2.
G. J. Hine and G. L. Brownell, Radiation Dosimetry, Academic Press, Inc.
New York (1956).
I 4.
r 61
'O U.S. NUCLEAR REGULATORY COMMsSSION NUREG/CR-3381 BIBLIOGRAPHIC DATA SHEET RNI-NifRFG 8i16R4
- 4. TlTLE AND SUBTITLE (Add Volurne Na, of appregnate)
- 2. (Leave blask)
Evaluation of Three Mile Island Unit 2 Reactor Building Decontamination Process 3 RECIPIENT'S ACCESSION NO.
l
- 7. Au fHOR(S)
- 5. DATE REPORT COMPLE TED M ON TH l YEAR
)
D. Doucherty. J.W. Adams May 1983
- 9. PERFORMING ORGANIZATION N AME AND MAILING ADDRESS (/nclude 2,p Code)
DATE REPORT ISSUED MONTH l YE AR Brookhaven National Laboratory Auaust 1983 Upton, NY 11973 e (teave bian4s 8 (Leave blanki
- 12. SPONSORING ORGANIZATION N AME AND MAILING ADDRESS (include Isa Codel Division of Waste Management Office of Nuclear Materials Safety and Safeguards
,,,,,y go.
U.S. Nuclear Regulatory Commission Washington, D. C. 20555 A 3162
- 13. TYPE OF REPORT PE RIOD COVE RE D I/nclusive dates)
Technical
- 15. SUPPLEMEN TARY NOTES 14 Ileave olana)
- 16. ABSTR ACT 000 words or less)
Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams.
Solid wastes being disposed of in commmercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings.
The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61.
It appears that much of the Epicor-II ion-exchange resin being disposed of in commercial land burial will be class B and require stabilization.
Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, sta-bility under immersion and biodegradability.
Results indicated that both radio-nuclide contamination and chelating agents leach from strippable coating waste.
- 17. KEY WOHDS AND DOCUMENT AN ALYSIS 17a DESCRIPTORS Decontamination, Radwaste, Ion-exchange resin, Strippable coatings, Waste classification 17tx IDENTIFIERS.OPEN ENDED TERMS
- 18. AVAILABILITY STATEMENT
- 19. SE CURITY CLASS (Tha report /
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