ML20108D060
| ML20108D060 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos Nuclear Center |
| Issue date: | 11/30/1984 |
| From: | Bowerman B, Dougherty D, Kempf C, Mackenzie D, Siskind B BROOKHAVEN NATIONAL LABORATORY |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| CON-FIN-A-3172 BNL-NUREG-51791, NUREG-CR-3864, NUDOCS 8412110033 | |
| Download: ML20108D060 (96) | |
Text
_. _ _ _ _..
NUREG/CR-386L l'
BNL-NUREG-51791 i
R Characterization of the 4
Low-Level Radioactive TNastes and i
Waste Packages of General Electric E
Vallecitos Nuclear Center F.inal Report yW 45 B3 Prcpared by C. R. Kompf, D. R. MacKenzie, B. S. Bowerman, 5
D. R. Dougherty, B. Siskind
' Brookhaven National Laboratory g
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U.S. Nuclear Regulatory g
Commission h'
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NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes s'ny warranty, expressed or implied, or assumes any legal liability of re.
sponsibility for any third party's use, or the re*ults of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owrmd rights.
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- 1. Ihe NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555 e
g
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
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purchased from the originating organisation or, if they are American National Standards, from the American National Standards Institute, t 430 Broadway, New York, NY 10018.
GPO Printed copv pree: MsE
NUREG/CR-3864 BNL-NUREG-51791 i
Characterization of the i
Low-Level Radioactive Wastes and Waste Packages of General Electric l
Vallecitos Nuclear Center Final Report l
M:nuscript Completed: June 1984 D:ta Published; November 1984 C. R. Kempf, D. R. MacKenzie, B. S. Bowerman,.
D. R. Dougherty, B. Siskind l
Department of Nuclear Energy Brookhaven National Laboratory Upton, NY 11973 Prrpared for Divisic t of Waste Management s
Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Wcshington, D.C. 20566 l
NRC FIN A3172 I
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ABSTRACT An evaluation of the low-level wastes and waste packages generated by
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Ceneral Electric Vallecitos Nuclear Center (CEVNC) was made on the basis of 10 CFR Part 61 criteria and on the Technical Position on Waste Form (TP).
In I
addition, a review has been performed of the handling and storage methods used by GEVNC for their transuranic wastes. Several options have been discussed 4
for management of these materials. This evaluation was the result of a study l
initiated by the U.S. Nuclear Regulatory Commission (NRC), in which GEVNC -
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participated.
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CEVNC generates radioactive vastes in hot cell processes which include examination of reactor fuel and components, and production of sources and radiopharmaceuticals. These wastes are usually Class B or greater. Class A wastes result from support activities which include maintenance of the hot cells. The dominant contaminating radioisotopes are Cs-137 and Co-60.
In addition, transuranic wastes result from examination and burnup analyses of fuel. The latter wastes are all currently stored on-site at GEVNC. The low activity Class A, Cs-137 and Co-60 dominated wastes are generally packaged in 55-gallon drums and wooden boxes, while those of higher activity (Class B and greater) are packaged in 84-gallon extended 17H drums that are grouted with cement. The Class A packages meet the requirements in 10 CFR Part 61. The Class B and greater grouted drum packages have been evaluated with respect to meeting the stability requirements in 10 CFR Part 61 and with respect to the guidance in the TP.
Based on the evaluation, overall, the waste forms of these packages are expected to maintain their stability, but a few concerns are identified and testing should be performed by CEVNC to demonstrate waste form stability.
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CONTENTS ABSTRACT...
iii CONTENTS...
v TABLES.....
vii ACKNOWLEDGMENTS....
ix 1.
INTRODUCTION....
1 2.
QUANTITIES OF LOW-LEVEL AND TRU WASTES GENERATED AT GEVNC......
3
?
3.
DESCRIPTION OF RADIOACTIVE WASTE STREAMS AT GEVNC.
7 3.1 E8t Cell and Support Activity Wastes.
7 i
3.1.1 General Description of Hot Cell Wastes and Waste 7
1 Pa cka g,e s...
3.1.2 Activity Determination for Radioactive Wastes at GEVNC..
10 3.1.3 Support Activity Wastes.
12 3.2 Summary of Was te Shipments From GZVNC.............. 13 3.3 Characterization of GEVNC TRU-Contaminated Waste.
16 3.3.1 Description of the Waste and Amounts.
16 3.3.2 TRU Concentra tions..
17 3;3.3 Classification of Waste.
18 3.4 Summary.............................
19 4.
EVALUATION OF GEVNC WASTE PACKAGES WITH RESPECT 77 NRC REQUIREMENTS. 21 4.1 Class A Wastes.
21
'4.1.1 10 CFR Part 61 Requirements for Class A Wastes.
21 4.1.2 Evalua tion of Class A Waste Shipments From CEVNC..... 23 4.2 Class B and Class C Wastes.
24 l
4.2.1 10 CFR Part 61 Requirements for Class B and Class C Wastes.....-.........
24 4.2.2 S truc tural S tabili ty................... 25 4.2.3 Evaluation of GEVNC Class B and C Wastes V' th Respect to the Guidance in the Technical Position. m Waste Fo rm.
32 5.
TRU WASTES..
51 5.1 Nea r-Surface Dis posal...................... 51 5.1.1 Waste Form Considerations.
52 5.1.2-Decontamination Processes.....
54 5.1.3 Cellulosics and Other Combustible Waste.
56 5.2 Grea ter Confinement Disposal (GCD)............... 57 5.3 Specific Applications to GEVNC TRU-Contaminated Waste.
59 Y
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CONTENTS, Continued 5.3.1. Waste Containing 80-100 nCi/g of TRU Isotopes.
.....- 59 5.3.2 Solidified Hot Liquid Waste Stream.
60 5.3.3 Solid Waste..
61 5.3.4 Summary of Options.
65 6.
CONCLUSIONS 67 6.1 Cla s s A Wa s te Pa ckage s.
67 6.2 Class B and Class C Waste Packages.
67 6.2.1 Minimum Requirements.
68 6.2.2 Stability Requirements.'.
68 6.2.3 Guidelines in' the Technical Position.
68 6.3 Evaluation of Additional Hazards in the GEVNC Wastes.
70 7.
REFERENCES.
71 APPENDIX A - 10 CFR PART 61 SECTIONS 55 and 56..
77 APPENDIX B - TECHNICAL POSITION ON WASTE FORM.
81 APPENDIX C - SAMPLE R ADIOACTIVE SHIPMENT RECORD FOR WASTES FROM GEVNC.. 85 APPENDIX D - CERTIFICATION OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES.
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TABLES 2.1 Summary of the CEVNC Low-Level Radioactive Waste Shipments Infor-mation From RSRs Supplied by CEYNC.
4 2.2 Radioactive Waste Shipment Summary for 1980-1983.
5 3.1. Summary of RSRs.
15 3.2 Breakdown of Class A Radioactivity Loadings and Waste Contents....
15 3.3 Summary of the Four Grouted Drums Shipped................16 3.4 GEVNC TRU Waste Inventory Information.
18 4.1 Concentration Limits on Long-Lived Radionuclides for Class A Wastes. 22 4.2 Concentration Limits of Short-Lived Radionuclides for Class A Wastes. 22 4.3 Concentration Limits of Short-Lived Radionuclides for Class B and Cla ss C Wastes.
25 4.4 GEVNC Tap Water Aaalysis.
34 4.5 Radionuclide Activities and Accumulated Doses for GEVNC Grouted Drum Waste Packages..
36 4.6 Dose Calculation Parameter Values for Principal GEVNC Vaste Radionuclides.
37 4.7 Activity Densities for the GEVNC Waste Radionuclides..
37
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i 4.8 GEVNC Grouted Drum Upper Class C Activity Dose Rate.
42 1
5.1 Generic Cos t Plus Risk Comparison Alternatives..
58 5.2 Possible Alternative for Treatment of Sorted GEVNC Solid Waste..
65 vii
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ACKNOWLEDCMENTS The authors would like to thank Drs. R. E. Davis and R. E. Barletta for review of the manuscript. In addition, the authors would like to extend sincere appreciation to Ms. Nancy Yerry and Ms. Mary McGrath for skillful l
preparation of the manuscript.
ix
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1.
l CHARACTERIZATION OF THE LOW-LEVEL RADIOACTIVE WASTES AND
- WASTE PACKAGES OF GENERAL ELECTRIC VALLECITOS NUCLEAR CENTER 1.
INTRODUCTION The low-level radioactive waste generated by many non-fuel cycle indus-tries and institutions is not as well characterized as that produced by nuc-lear power plants. As a part of a program to characterize non-fuel cycle waste shipped to commercial shallow land -burial, Brookhaven National Labora-tory (BNL) has identified, contacted and visited a number of non-fuel cycle wa ste generators. For selected generators, BNL has performed detailed evalu-ations of their low-level radioactive waste. These evaluations were performed with respect to 10 CFR Part 61, " Licensing Requirements for Land Disposal of Radioactive Waste" and included (1) an assessment of the chemical, physical, radiological and biological degradation mechanisms of the waste form and waste container which may affect the ability of the waste package to meet the accep-tance criteria for disposal, and (2) the identification of chemical hazards in the waste packages which by themselves or in conjunction with the radiological hazards may affect the behavior of the waste packages and the ability of the
' site to perform adequately. To date, three such evaluations have been per-formed. They are eya uations of the Class B waste packages of the New England Nuclear Corporation \\1, of the large quantity waste packages of the Union Carb dg Corporation (2), and of the wastes and containers of the 3M Corpora-tion 31 A fourth generator, the General Electric Vallecitos Nuclear Center (CEVNC), is the subject of this study. This study has been conducted in cooperation with GEVNC in order to provide the Nuclear Regula tory Commission with an iyaluation of GEVNC low-level wastes with respect to 10 CFR Part 61 criteria as well as the guidelines for Class B and C wastes specified in the Te,chnical Position on Waste Form [ Revision 0, May 1983] (TP). The relevant sections of 10 CFR Part 61 used in this study are those on waste classifica-tion and waste characteristics. These sections, 61.55 and 61.56, respec-tively, have been included in Appendix A.
The relevant sections of the TP are included as Appendix B.
The characterization of transuranic (TRU) wastes generated at GEVNC has been performed and the evaluation of potential waste management strategies for the TRU waste stored at GEVNC was also incorporated into this study.
Radioactive waste is generated at GEVNC during examination of reactor components and fuel and during the production of radiopharmaceuticals and ra-dioactive sources. In an effort to categorize the low-level wastes according to the weste classification scheme set forth in 10 CFR Part 61, a review of the wastes shipped from CEVNC has been performed by surveying selected radio-active shipment records (RSRs) from 1982 and 1983. The results of this survey are summarized n In addition, information has been included from a CEVNC report, 4)Section 2.
a letter from GEVNC including a tabular summary of the waste shipments,(5) and telephone conversations with GEVNC personnel.
Descriptions of the processes in which the low-level and TRU wastes are 1
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J generated and of the containers in which these wastes are packaged for either burial or storage have been written based on information provided by GEVNC.
These descriptions are given in Section 3.
There are three parts to the evaluations of the GEVNC low-level w'aste.
First, waste packages are evaluated to determine if they meet the minimum requirements for all wastes to be disposed of by commercial shallow land burial as well as the stability requirements for Class B and C wastes, as ap-propriate. This evaluation includes a consideration of pertinent degradation mechanisms, consideration of the requirements given in 10 CFR Part 61, as well as of the guidance given in the TP for the demonstration of waste form / package stability. The requirements for each class of waste and the evaluation of the GEVNC wastes in the context of these requirements are given in Section 4.
The second part of the low-level waste evaluation was an identification of those components of the GEVNC wastes which are either hazardous or which could af-fact the performance of the site in which these wastes are buried. As can be seen from the package descriptions and evaluations contained herein, it was felt that no such concerns existed with the GEVNC wastes themselves. The third part of this study involved the evaluation of potential strategies for managing the TRU wastes currently being stored by GEVNC. This assessment was performed to provide the NRC with background information on the types and forms of TRU wastes that are generated in processes at this type of facility.
This evalua tion is given in Section 5.
The evaluation of the low-level waste packages and the consideration of the TRU wastes have resulted in some concerns and recommendations. These are provided in Section 6.
l 2
2.
QUANTITIES OF LOW-LEVEL AND TRU WASTES GENERATED AT GEVNC In order to consider the impact of the then-proposed 10 CFR Part 61 on GEVNC's non-fuel-gycle waste, representatives from GEVNC, NRC, and BNL met in September 1982.(6) It was learned that GEVNC generates waste from three main sources:. hot cell activities, liquid waste evapora tor operations, and a research reactor (presently shut down).
The hot cell activities consisted mainly of examination work on reactor components and fuel, the fabrication of Co-60 and Cf-252 sources, and the manufacture of Xe-133, Cl-36, and C-14 radiopha rmaceuticals. The bulk of the waste was from these hot cell activi-ties and the primary waste radioisotopes were Co-60 and Cs-137.
In the pac t, GEVNC has shipped their waste to Beatty, Nevada, but they now ship to Richland, Washington.
In addition, GEVNC generates TRU-contaminated waste which is s tored on site.
As a result of more recent contacts,(7} GEVNC supplied BNL with the Radioactive Shipment Records (RSRs) for six waste shipments during the period from April through August of 1983 as well as an RSR for a shipment in March 1982. Some of the information from these RSRs is summarized in Table 2.1.
There were a total of 62 packages shipped, a total shipping volume of 3629 cubic feet, and a total activi ty of about 617 Ci.
One of the RSRs is repro-duced in Appendix C.
Most of GEVNC's annual waste activity was attributable to two isotopes, Cs-137 and Co-60, although C-14, Sb-124, and U-235 were also shipped. It was found in review of these RSRs that three packages contained waste for which the specific activity exceeded the Class B limit (according to the classification scheme of 10 CFR Part 61). These packages were all 84-gallon drums shipped in casks. All other packages listed in the RSRs are Class A under current regulations. It should be noted that the selected GEVNC RSRs supplied to BNL did not include all radioactive wastes shipped from GEVNC for the period covered. The RSR review indicates that GEVNC has shipped mainly Class A and Class C (and one greater than Class C) waste packages.
In telephone conversations with GEVNC staff, it has been stated that GEVNC does ship some Class B wastes.
In any event, all GEVNC Class B or greater wastes have been shipped in one type of container, the 84-gallon 17H drum grouted wi th cement. The waste streams at GEVNC and the detailed RSR information are discussed more fully in Section 3.
A summary of GEVNC radioactive waste shipments from 1980 through 1983 is given in Table 2.2.
GEVNC also supplied BNL with a tabular summary of waste shipment infor-mation and with a copy of a report (4) published by CEVNC which included the results of tests performed on higher activity GEVNC waste packages to demon-A strate the fulfillment of requirements necessary to apply for a certificate of compliance for these packages. The tabular summary indicated that GEVNC ships wastes in 55-gallon drums, in wooden boxes, and in cemen t-grouted 84-gallon
( 11.5-cu. f t. ) drums. The 55-gallon drums may be shipped in overpacks while the cement-grouted drums are shipped in the GE Model 1600 shielded shipping container. ( A copy of the'U.S. NRC Certificate of Compliance No. 9044, Revision 6 is Appendix D.)
In general, the tabular summary and the RSRs both indicate that mos t of the waste volume shipped from GEVNC is Class A (55-gallon drums, wooden boxes, and occasionally, a cement-grouted drum). As 3
l would be expected, the Class B or greater wastes comprise a much smaller frac-tion of the total volume shipped.
Several maximum activity values for specific grouted ' drum packages were found. The RSR survey yielded 103 Ci and 16.65 Ci for Co-60 and Cs-137, respectively. The tabular summary indicated values of 189 Ci and 81 Ci for these two isotopes, respectively, while the report published by CEVNC gave activity limit totals of 5000 Ci for aged mixed fission products (assumed to be ~50/50 Cs-137 and Sr-90), and 3000 Ci for Co-60.
For the purposes of the waste package evaluations, it is desirable that the activities used for cal-cula tions on radiolysis, dose, e tc., be conserva tively high. The totals from, the CEVNC report would be used were they not, for Cs-137 and Sr-90, in excess of the Class C limit for this size container. Hence, for the evaluation in this report, the activity values for the CEVNC grouted drum packages were a
taken as the upper Class C limit for both Cs-137 and Sr-90 or 1500 and 2300 C1, respectively, while 3000 Ci was retained as the activity limit for Co-60.
r Table 2.1 Summary of the CEVNC Low Level Radioactive Weste Shipments Information From RSRs Supplied by CEVNf RSR Date No.
Total Volume Total Activity Isotopes Package Packages (cu. f t.)
(C1)
Descriptions 03148 3/09/82 10 1144.5 216.05 Cs-137 1.S A wood boxes 7
Co-60 U-235 1097 4/18/83 18 135 2.025 Cs-137 55-gallon drums Co-60 21618 7/06/83 8
795.5 51.895 Cs-137 7 IJ A wood boxes Co-60 1 cook (Class C) 21615 7/25/83 11 17 5.5 1.7538 Cs-137 2 LS A wood boxes Co-60 9 55 gallon drums 21614 7/27/83 7
683.5 186.801 C 137 Cs-137 Sb-124 1 cask (> Class C)
C-14 21606 8/24/83 5
459.5 55.544 Cs-137 4 LSA wood boxes Co-60 1 cask (Class C) 21619 8/31/83 3
235.5 103.019 Cs-137 2 LSA wood boxes Co-60 1 cask 617.09 67 3629.0 aShipped to Beatty, Nevada, disposal site. All others shipped to Richland, Washington.
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Table 2.28 C
e CEVNC Radioactive Waste Shipment Summary for 1980-1983 4
1980 1981 1982 1983 Total Volume
,10,400 12,900 6,500 5,900 3
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of Shipments r
Activity Total (C1) 6651 1216.4 639.0 1643.2 Co-60 521 1146 515.8 1095.2 Cs-137 130.5 70.2 123.2
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"o the r 3.5 mci 100 C1 (Sb-124)
(U-235) 42.46 C1 (C-14) 127 mci (U-235) h r
Total Packages Drums (55-ga11on) 56 96 108 134 Boxes 90 106 54 44 3
Casks (11.5 f t )
11-11 4
9' l-aInformation provided by J. Tenorio of CEVNC in conference call with BNL staff on February 27, 1984.
bThe vast majority of the waste activity is concentrated in the ' cask packages. The 55-gallon drums and boxes contain essentia}1y trace amounts.
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The Class A waste packages are subject only to the minimum waste charac-teristic requirement.c of 10 CFR Part 61 and their evaluation is given in Section 4.1.
As stated earlier, those waste packages which are Class B or greater are, for GEVNC wastes, all 84-gallon cement-grouted drums. These are subject to requirements for structural stability in addition to the minimum waste characteristic requirements of 10 CFR Part 61 and they are evaluated in Section 4.2.
An overview of the TRU waste stored at GEVNC has also been provide 2.*
The TRU waste with activity much greater than 100 nCi/g consists of 9 one-gallon " paint cans" containing cement-solidified liquid from burnup analyses j
of nuclear fuel. The upper bound for the activity in these cans is 3 aci/g or about 30 Ci per can. The remaining TRU waste, consisting mostly of cellulosics and misceilaneous debris with activities in the range of 80 to 100 nC1/g, is stored in 91 liners with approximate volumes of 1.5 cubic feet (about 2/3 of the liners) or 5.5 cubic feet (about 1/3).
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- I fn ormation provided by J. Tenorio of GEVNC in conference call with BNL 1
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- Nost of the volume of wastes shipped by GEVNC is Class A and is packaged in 55-ga11on drums, wooden boxes, and occasionally, in a cement-grouted 84-gallon drum. Higher activity wastes, Class B or greater, are shipped only in the cement-grouted 84-gallon drums. Principal contaminating radioisotopes in Class B and greater wastes are Cs-137 and Co-60.
For the purposes of the higher activity grouted-drum waste package evaluation in this report, Cs-137 and Sr-90 have been _taken as waste contaminants at.the Class C activity limit
' (1500 and 2300 Ci, respectively, for this size container), and Co-60 has been included gt 3000 Ci per package. These values are all based on the GEVNC report,14/ which was the most conservative activity estimate available.
Prior to the establishment of the regulation waste classification scheme of 10 CFR Part 61, CEVNC shipped at least one package of activity greater than what is now the Class C limit. Such wastes would not generally be acceptable for disposal in a commercial shallow land burial site unless approved by -the NRC or by the appropriate licensing authority.
Other radioisotopes in GEVNC wastes, e.g., Xe-133, Cl-36, etc., have been stated to be at quite low activity levels compared to the Cs-137 and Co-60, i.e., these isotopes generally occur in GEVNC Class A waste packages.
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The C-14 and Sb-124 packages that were found in the RSR survey were one-time shipments and have therefore not been considered as a regular component in the GEVNC waste streams.
TRU wastes which GEVNC has divided into categories of 80-100 nC1/g and
>100 nCi/g are all presently stored on site.
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DESCRIPTION OF RADIOACTIVE WASTE STREAMS AT GEVNC GEVNC is involved in three main activities from which their radioactive j
. wastes are generated:
(1) hot cell activities which involve examination work on reactor fuel and components, fabrication of Co-60 and Cf-252 sources and manufac-ture of Xe-133, Cl-36, and C-14 radiopharmaceuticals. The primary waste radioisotopes from these hot cell activities are Co-60 and Cs-137.
(2) operation of a liquid waste evaporator. Liquid wastes from all non-TRU GEVNC radwaste generating operations and outside sources as well are evaporated.
(3) non-TRU operations generating solid wastes at other locations on the CEVNC site, and research reactor work, none of which is being done l
presently since the reactor is not in operation.
Radwastes from these activities are divided into three waste streams: (i) hot cell wastes, (ii) support activity wastes and (iii) TRU wastes. Hot cell and support activity wastes and their treatment prior to disposal are de-
[
scribed in Section 3.1.
A detailed discussion of information from GEVNC RSRs for shipments of hot cell and support activity wastes is given in Section 3.2.
j TRU wastes and treatment are presented in Section 3.3.
3.1 Hot Cell and Support Activity Wastes 3.1.1 General Description of Hot Cell Vastes and Vaste Packages Hot cs11 processes involve examination work on reactor fuel and compo-l-
nents, fabrication of Co-60 and Cf-252 sources and the production of radio-pharmaceu ticals. Wastes from these operations consist of irradiated metals, l
glass, and general cell trash. Radioisotopes identified in these wastes in-clude Co-60, Cs-134, and Cs-137.
Other radioisotopes are present in trace amounts that are not expected to exceed 1% of the values in Table 1, Column 1 of 10 CFR Part 61, Section 55.
These wastes'are mainly in solid form.
Com-pactible solid hot cell wastes are. placed in one gallon cans and compacted prior to disposal. The liquid waste volume from hot cell activities was esti-l mated to be approximately 30 gal per year. Liquid hot cell wastes are solidi-fled with cement in a 2:1 cement to liquid ratio by volume in one-gallon cans.
GEVNC has indicated that, according to the NRC waste classification scheme in 10 CFR Part 61, these hot cell wastes are expected to be Class B or Class C.
The container used for disposal of hot cell wastes consists of 11-1/2-cu. ft. 17H-drum with two inner perforated carbon steel baskets (84-gallon). These baskets have steel angle iron spacers attached to the outer sides and bottom which hold the basket approximately 3/4 inches away l,
from the drum itself. The waste is prevented, by this basket arrangement, from having any direct contact with the drum. Figure 3.1 shows two baskets i
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There is an open space 1-1/2-in wide at the top of each drum af ter the bas-ke ts have been inserted. A series of dose rate measurements is made on the waste-filled container., The highest in the series of readings is used in ap-plication of a conversion factor to obtain activities. GEVNC has stated this factor is generated by a computer code based on average dose rates and waste configurations (solidified, compacted, e tc.). This procedure is discussed in Section 3.1.2.
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Figure 3.1 Two perforated steel basket liners filled with simulated hot cell wastes.(4)
The hot cell wastes in the basket and drum assembly are then stabilized by a cement grouting trea tment. The grout consists of 600-lb Portland cement, 600-lb sand and 30-gal wa ter.
Cement is poured from the top into the annulus between the inner basket,and the inner drum side. The t'ow characteristics of the cement are such that it penetrates the openings of the baskets and fills nearly all the void space in and around the waste materials. The drum is vi-brated during tFis cement gouring process to ensure effective grouting.
Fig-ure 3.2 shows an 11-1/2 ft drum in a test stand ready for grouting. The vibra tor is also shown a ttached to the drum.
The cement is usually allowed to set up for two days. Drums are then checked for free liquid and concrete hardness and each drum is photographed prior to shipment) Figure 3.3 shows top of a grouted drum of simulated hot cell waste.(4 Following the the 8
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grouting process, the drums are closed with a ring-bolt-sealed lid.
Another series of dose rate measurements is made subsequent to the grouting for purposes of limiting the radiation exposure of transportation personnel.
Drums containing simulated waste have been given this cement-grouting treatment and then sectioned horizontally and vertically to allow inspection of the concrete-waste ma trix.(4) The concrete in these drums could be seen to have thoroughly filled the voids in the container, i
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4 Figure 3.2 Test s tand used for tes t drum grouting. A vibra tor, located near the middle of the drum, is used to promote grout pene tra tion.(4) i 9
l b
+
I l
l
{
t s
Figure 3.3 Top view of a grouted druu of simulated hot cell waste.
GEVNC provided an estimate of the average composition of one of their higher activity cement-grouted drums. Typical total weights range from 1600 to 2300 lbs of which ~300 lbs is waste. This waste is approximately 40% metal (ferrous, some aluminum, <10% lead), 10% glass and 50% paper and plastic. The plastic is mainly polye thylene, but the tubing is polyvinyl chloride.
3.1.2 Activity Determination for Radioactive Wastes at GEVNC The activity contained in GEVNC higher activity radioactive waste pack-ages is estimated using dose rape i
factors generated with the computer code, ISOSHLDt8)to activity convers on
, combined with actual dose rate measure-ments for waste packages, and smear measurements from the hot cells. The code was written to perform gamma ray shielding calculations. The input to the code consists of an activity loading and an assumed distribution of this ac-l tivity in a matrix (e.g., concrete) and in a specified geometric shape (e.g.,
l
' a cylinder in the case of a drum). The output of the code is a dose rate.
The procedures used by GEVNC to estimate the activity in a waste pack-age may be separated into four steps.
I Step 1: The dose rate is measured at various places around a waste package and the highes t dose rate is recorded.
10
Step 2: Smears are taken in the hot cells to determine the fractional contribution of Cs-137 and Co-60 in the activity.
Step 3: The ISOSHLD code was used to generate a dose rate per curie value for a unit Co-60 activity level for each of the waste containers used a t GEVNC. The activity content of the waste is estimated by multiplying the measured waste package dose rate by this conversion factor. For ungrouted drums, GEVNC assumes a homogeneous distribution of activity in a matrix with the density of 0.5 g/cm3 For the grouted drums, GEVNC assumes a homo-geneous distribution of activity in the drum in 3
a concrete matrix of density 2.35 g/cm.
Step 4: The smear data is used to partition the activity into Cs-137 and Co-60 fractions. No other isotopes are assumed to be present.
The accuracy of this procedure of determining the activity in a given waste package depends upon the representativeness of dose rate measurement, the accuracy of _ the ISOSHLD code for the conditions in which it is being used, and the correctness of the Cs-137/Co-60 fractional distribution from the smear measurement. It should be noted that if the Cs is a result of fuel examina-tion, other fission products, particularly Sr-90, may be present in the waste.
i Depending on the concentration of these other radionuclides, the waste classi-fication may change and some waste could potentially exceed the Class C limit.
It is, therefore, recommended that a more detailed characterization of the radionuclide inventory in the waste be performed by CEVNC.
I The technical correctness of the ISOSHLD code has been experimentally verified.(8) This was done by testing of the ISOSHLD code output agains t i
the measured dose rates for known radionuclide sources. The code yields con-sistentl'y correc t values. The mechanical correctness of the code can not be questioned from the point of view of distinguishing between heterogeneous and I
homogeneous activity distributions, since the code itself simply converts ac-tivities to dose rates for an assumed bomogeneous source.
In practice, CEVNC uses the highest dose rate measuremen't ' rom each waste package as the input to the code. This should result in a con.rvative overestimation of the activity j
contained in the package. The degree o f overestimation cannot be quantita-tively stated on the basis of the information available. However, it might be advantageous for CEVNC to determine this since they have shipped wastes of l
activities calculated to be in excess of the Class C limit. Depending on the magnitude of the overestimation, these wastes may not have been greater than l
Cla ss C.
l l
The correctness of the partitioning of activity into Cs-137 and Co-60 fractional contributions depends completely upon the correctness of the smear da ta. GEVNC conducts periodic smear tests of hot cell waste contamination and analyzes the data for the Cs-137/Co-60 ratio. The smear samples are taken from various (essentially random) locations in the non-compartmentalized hot cell area where fuel examination work is done. CEVNC has indicated that I
t 11 l
l
i
)
materials in.the hot cell make up a general mixture and that, since the work is not partitioned according to different steps in processing, there is no reason to believe that the contamination distributed throughout this hot cell work area is different from that which occurs in waste materials. The ISOSHLD ouput is partitioned according to this smear test ratio such that respective Cs-137 and Co-60 activities can be assigned to the waste drums. On occasion, GEVNC has some wastes that are only Co-60 contaminated, but.these batches are said to be well tracked. - In general, GEVNC feels that their smear. test anal-
'[
yses result in a Cs-137/Co-60 ratio that is indeed representative of that in the waste. GEVNC has not specifically performed an analysis to determine if the ratios obtain'ed from hot cell smears tests are representative of the radioisotope ratios in their waste packages. As mentioned earlier, it is recommended that a more detailed characterization of the radionuclide inven-1 tory in the waste be prepared by GEVNC. This could include tests to determine the correlation between the hot cell smears and radionuclide distribution in i
the waste.
3.1.3 Support Activity Wastes The second main waste stream at GEVNC includes wastes from what GEVNC d
terms " support" activities, i.e., waste from maintenance of all GEVNC facili-I ties in which radioactivity is handled. These wastes comprise the largest l
fraction by volume of GEVNC radioactive wastes. GEVNC estimated that 70% of the ~12,000 cu. ft. radioactive waste annually shipped for burial is low-level t
l waste resulting from these " support" efforts. GEVNC has indicated that all packages of these wastes are Class A.
Wastes from " support" activities are i
divided into two groups: solidified / treated and not solidified. These are discussed in the following sections.
3.1.3.1 Solidified / Treated Support Activity Wastes f
These wastes consist of three types: dewatered resins, solidified liquids, and non-compactible items. Resin dewatering is accomplished by pass-ing hot air through a mesh-like bucket containing the resins. This treatment requires about one day. The activity on the resins is occasionally counted, and a representative radioactivity loading is approximately 0.3 pCi of Co-60 per cubic centimeter. Only rarely are resins solidified with cement. Liquids originate from cleanup activities, fuel rod "de-crudding", hot water " leach-outs" from f uel elements, and evaporator bottoms. Generation of liquid wastes outside the hot cells is approximately 100,000 gals per year. Much of this is sent to the waste evaporator from which approximately 100-125, 55-gal drums of waste are generated in a year. All of these wastes are solidified with cement in a 2:1 (by volume) cement-to-liquid ratio. Many of the fuel rod "de-crudding" solutions are acidic and are neutralized prior to solidification.
Non-compactible items include equipment components, glass, etc. These items are manually placed in the waste container. All of this category of waste is packaged in 55-gal,17H drums. There are no mixes of materials in these pack-
- ages, i.e., one drum would not contain both dewatered resins and solidified liquids or non-compactible items. A typical package of this type weighs be-tween 600 and 700 pounds. All of these drums are lined with 50-mil plastic 12
' liners in which.the. wastes are placed, RTV* is applied and then the drums are ring-bolt sealed..
Non-aqueous chemicals (such as acetone) are present in GEVNC wastes only in trace amounts since the vast majority of solution processes are per-formed in aq'ueous systems. The comment was made that most of the work at GEVNC is non-destructive so there is actually very little " analysis-type" liquid waste.
Low-level liquid waste processing is offered as a service by CEVNC to contracted waste generators in the area. The example was given that liquid wastes have befn received from Rancho Seco Nuclear Generating Station. Spe-cific contaminating radionuclides and activities were not given, but it was stated that typical wastes registered a few thousand counts per minute. These wastes are sent to the GEVNC waste evaporator and the evaporator bottoms are then solidified with cement and packaged as are the solidified liquids dis-cussed earlier.
3.1.3.2 Non-Solidified Support Activity Waste These compactible general lab waste's are Class A, contaminated with co-60, Cs-134, and Cs-137, and are generally compacted into 112 cu. f t.
plywood boxes that meet low specific activity (LSA) packaging requirements.
These packages are loaded to minimize void spaces and, on the average, weigh 2000 pounds.
- Two other comparatively minor waste types from GEVNC contain Xe-133 and liquid scintillation vials (LSV). The Xe-133 waste arises from repack-aging bulk Xe-133 into glass ampules to be used for nuclear medicine.. The Xe-133 contaminates vacuum pump oil and this waste is generated at the rate of approximately 25 gal per year. The contaminated oil is solidified with
-"Enviros tone" (U.S. Gypsum Company). This waste is then considered solidified liquid and is packaged in a 50-mil plastic liner in a 55-gal,17H drum in a manner similar to the solidified liquida discussed earlier. Waste from liquid scintil.lation vials (LSV) is generated at a very low rate.
In two years, enough LSV waste will accumulate to fill a 55-gal drum in an arrangement in which the liquid scintillation vials are placed in layers alternating with dia tomaceous earth. The absorbent is, again, present in a volume twice that necessary to absorb the LSV volume.
3.2 Summary of Waste Shipments From GEVNC e
The material in this section is based on informa tion contained in a tabu-lar summary and a review of seven RSRs provided by GEVNC. One of the RSRs was dated 3/9/82 while the rest were selected from the period 4/18/83 to 8/31/83.
The first is discussed below, while the latter are discussed in the following sections.
- RTV is a tradename for a sealant that is marketed by the General Electric Com pany.
P 13
~
r L.,...
J i
e The RSR from 1982 listed waste which was shipped to the shallow land 1
burial site (SLB) a t Beatty, Nevada. It included LSA wastes consisting of3 j.
. compacted trash shipped,in ten boxes of various sizes ranging from 265 ft 3
to 6 f t. The two smallest boxes (6 f t3 and 9.5 f t ) contained special 3
3 minimum nuclear ma terial, i.e., U-235.
The remainder of the boxes (112 ft size) contained only Cs-137 and Co-60 and were all Class A.
The boxes con-T taining U-235 are also Class A.
The 9.5-ft3 box was listed as containing 10 g with a total activity of 3 x 10-4 C1. The 6-ft3 box contained 17 g of material; however, the activity level listed on the form was illegible.
Since no limit is specified in 10 CFR Section 61.55 for U-235 these two pack-l l
ages may also be considered Class A.
The following discussion summarizes' the information gleaned from RSRs selected from a five-month period in 1983 and dealing with shipments to the SLB at Richland, Washington, which is where CEVNC currently ships its wastes.
The majority of waste (by volume) is shipped in 112 f t3 wooden boxes 3 boxes and consists of compacted trash. There were two instances of 54 f t
{
being used.
l The remaining waste is shipped in drums. There are two sizes of drums
-rently in uses a standard S5-gallon drum, and an extended 17H (x17fi) e i
i i
t~84-ga11on) d rum.
In general, the S5-gallon drums are shipped without the addition of a solidification agent; however, in the RSRs examined, there were j
2 examples of 55-gallon drums in which cement' binder was used. One contained y
j
" crud in cement", and the other had liquid solidified in cement. Otherwise, 3 according to the RSRs) contained trash, I
the 55-gallon drums (each 7.5 f t dewatered resins, or liquids absorbed on diatomaceous earth.
Tables 3.1, 3.2, and 3.3 summarise the information contained in the RSRs.
From Table 3.1, it can be seen that 90% (by volume) of the waste is shipped in With one boxes, 8% in 55-gallon drums, and 2% in the extended 17H drums.
j exception, the only radionuclides listed in the RSRs were Co-60 and Cs-137.
(This appears to be a result of the manner in which GEVNC estimates the activ-ity distribution in their waste. See Section 3.1.2.)
Table 3.1 lists the y
j numbers, contents, and total activites of' boxes and 55-gallon drums, i
f Higher activity wastes are shipped in x17H drums which, following en-placement of the wastes, are filled with a cement grout. The grouted waste in i
l general consists oof metal oxides for the most part, but according to the RSRs,
)
miscellaneous trash, i.e., plastics and paper, as well as resins were also shipped in this manner. Four x17H drums were shipped during the period i,
surveyed. Table 3.3 details the waste characteristics and activity loadings of these higher activity packages. Included is the only package which was listed as containing radionuclides other than Co-60 and Cs-137. The other nuclides were Sb-124 (100 C1) and C-14 (42.46 C1) and, as mentioned in Section 2, this shipment was later clarified by CEVNC to be a one-time event.
i g-e d
h 4
e 14 h
--n--+->
~ ~ ~.,..,.
n-
t Table 3.1 Summary of RSRs e
Totals for 1983 ( April--August)
Boxes Drums SS-gallon 80-gallon Total Number Volume Number Volume Number Volume 3
3 3
Shipped (f t )
(f t )
(f t )
2484.5 ft3 21 2236.0 27 202.5
,4 46.0 Table 3.2 Breakdown of Class A Radioactivity Loadings and Waste Contents Package
-Waste Radioactivity (C1)
(No.)
Boxes (21) compac ted 1.029 0.446 1.475 trash Drums (15) compacted 1.183 0.426 1.609 (55-gallon) trash s
Drums (2) dewa tered 0.036 0.015 0.051 (55-gallon) resins i
Drums (8) liquid /
0.443 0.245 0.688 l
(55-gallon) diatomaceous earth l
Drums (2) crud / liquids 0.044 0.0 18 0.062 (in cement)
Totals 2.735 1.150 3.885 i
15 i
. _ _. _. _ _ ~. _ _. -. _
.n--
--- - - - -~
Table 3.3
~
summary of the Four Grouted Drums Shipped (From RSRs Covering 4/83 Throughout 8/83)
S hipment Contents
i-Datu (C1)
(C1)
(C1)
I 7/27/83-metal oxides /
31.0 13.3 142.46 186.76
>C irradiated hardware (Sb-124,C-14) 7/6/83 metal / plastic /
36.26 15.54 51.8 C
resin i
8/31/83 metal oxide 103.0 103.0 A
i 8/24/83 metal oxide 38.85 16.65 55.5 C
l 4
l aThe contents are as listed on the RSRs. There is not complete agreement l
between the description on the RSRs and that provided by CEVNC as summarised in Section 3.1.
bBased on comparison with 10 CFR Part 61 requirements.
j.
3.3 Characterisation of CEVNC TRU-Contaminated Waste l
3.3.1 Description of the Weste and Amounts Essentially all the TRU-contaminated waste generated by CEVNC arises from analysis or examination of fuel. Fuel elements are sectioned in a hot i
cell, using a diamond wheel cutter. Sections through a UO2 fuel pellet are i
taken for metallographic examination and a sample roughly 1/2 inch in length i
is taken for burnup analysis. The Zircaloy cladding is removed mechanically.
The sections for metallographic examination are transferred to a metallurgy l
cell for polishing. The sample for burnup analysis is dissolved in the hot cell to a final volume of 100 mL.
A small aliquot of this solution, such as i
0.1 mL, is diluted by a factor of 104 to make up the solution which is used l
for.the analytical work.
The " hot" liquid solutions produced by dissolving the samples of fuel are estimated to comprise roughly 2/3 of the TRU activity in the waste and this is contained in 5-6% of the volume. The dilute solutions on which the burnup analyses are performed contain only ~0.1% of the activity in the hot 1
liquid waste stream, but make up roughly 2/3 of the total volume. Both types of liquid waste solution (acidic when prepared) are neutralised and then so-i 11dified with cement in a volume ratio of 2:1 cement to liquid, in either pint or gallon cans.
i 16 i
I I
l i
The remainder of the waste, approximately one-third of the activity and 1/4 of the volume, consists of solid waste, which is placed in both pint and gallon cans, without encapsulation in cement. The waste contains the cuttings from the fuel sectioning (picked up on wipe papers), the cutting wheels, i
pieces of Zircaloy cladding, the metallographic polished sections in their j
plastic mounts, contaminated glassware and other contaminated equipment.
Another contribution to this waste comes from the polishing operations. It is l
made up of particles of fuel and polishing compound which collect in a sump.
The sump is cleaned out occasionally and the sludge is solidified in cement.
l The total volume of this waste is quite small, and the TRU level was estimated by GEVNC to, be considerably lower than that in the solidified hot liquid l
waste.
All TRU wastes are currently stored at GEVNC in their Hillside Storage Facili ty.
3.3.2 TRU Concentrations i
According to GEVNC personnel, the TRU concentration in the solidified waste from the dilute liquid waste stream is in the range of 80-100 nci/g.
l This estimate should have been made quite accurately, since TRU concentrations i
i in the solution would be determined during the burnup analysis, and known i
weights of cement and solution could be mixed to achieve the concentration i
range stated. If accurately known weights were not always used, of course, TRU concentrations in some cases may exceed 100 nC1/3 GEVNC personnel gave a figure of 3 aC1/g as an upper bound for the level of TRU activity in any of their waste. This value was for solidified waste from the hot liquid waste stream. The fuel from which the waste origi-nated was stated to be mostly ~3% enriched UO, BWR fuel in the burnup range 2
l of 5,000-50,000 megawatt days per ton (MWD / Ton), with most of it in the range of 15,000-30,000 MWD / Ton. Since GEVNC had a standard procedure for solidify-ing its hot solutions, it is assumed that the TRU isotope concentrations in l
the final cement product were proportional to their concentrations in the cor-I responding fuel sections, which in turn would be approximately proportional to the burnup. Thus, the TRU concentrations in this waste stream are estimated I
to be in the range 0.2-3 mci /g, with most in the range of 1-2 sci /s, and to average approximately 1.5 act/g. The exact values of the " average" or of the individual packages of this waste are not particularly important since they 4
are so much greater (a factor of some 10 ) than the Class C limit.
The solid waste would have considerable variation in its TRU content from container to container, depending on the particular components placed in each container. For example, waste from the metallurgy cell would be very hot if it contained any of the polished fuel pellet sections. Otherwise, it would be much " cooler". GEVNC did not provide information on the TRU content of this waste stream except to state that it was >100 nC1/3 Assuming three sec-tions were taken for metallographic examination when a burnup analysis sample was taken (CEVNC estimate), that the sections were 1/16-in. thick while the j
analytical sample was 1/2-in. long, and that all three sections went in a I
17 1
l l
single pint paint can, the TRU concentration of such a can would approach that j
j of the hot solidified burnup analysis solution from the same pellet.
It is probably not useful to try to estimate a range for the TRU con-tent of the solid waste since the variation would be significant and the j
amounts connected with the lower end of the range would contribute very little to the total TRU content of, the waste. However, an average TRU content can be i
i estimated on the basis of the hot liquid waste stream and the relative amounts
)
of fuel going into each stream. Assuming sample and section sises as given in the last paragraph (1/2 in. and 1/16 in., respectively), three sections for
- j each sample, and 1/64-in. saw cuts, the' solid waste would have approximately half the TRU content of the solidified hot liquid waste. This is contained in l
four. times the volume of the latter waste and is associated with a weight estimated by CEVNC (5-6 lb/ gal) as 1-1/4 times _ that of the hot liquid waste t
i stream. Thus, the average concentration of the solid waste stream is esti-mated to be a factoi of 2.5 less than that estimated for the hpt liquid waste.
These concentrations and other information about the waste streams are given 9
in Table 3.4.
l g
i Table 3.4 l
a csvue nu vaste inventory 1.formatt..
}
to t tas ted 1
Approstante Volume Line r Annual settested Averese -
+
i In Pa tnt Cane, Storage Liner Volume Producttee, 1R0 Aettvity, Activity 1RU Centent, v.et. st....
cati...
voi..., f L3 re.ett..
se ur, ci re.eit
.cus Salldified Hot Liquid 9
1.9 0.06 3
135 0.65 1.5 a 106 Free Dissolulut Fuel S..,ies Se!!dified bliote 90 22 0.6$
30 0.0 L 410*4 80 100 f
Selettone Frem l
Sgtney Analyste t
$stid veste -
38 10 0.29 13 72 0.35 6 a 105 f
l a
3.3.3 Classification of Waste l
Since CgVNC performs no chemical separations on either their hot or dilute solutions, the-isotopic ratios in the solidified liquid and in the I
solid waste are the same for any particular fuel. The enrichment and the usual burnup range quoted for their samples are very sinflar to those for commercial LWR fuel, so, to a first approximation, isotopic ratios in the CEVNC waste can be considered the same as those for LWR fuel on average. (N Thus, Sr-90 and Co-137 activities would be about 40 times TRU a-activity, f
and far lower than the concentration required to put waste which is Class C by l
virtue of its TRU content over the Class C ipits. However, Pu-241 activities would be about 30 times the TRU ct-activity, 91 and this would undoubtedly 4
5 c
l 18 i
c i
i put at least some of the GEVNC waste in the 80-100 act/s category (in terms of TRU activity) over the Class C limit on the basis of other radionuclides pres-ent in the waste given application of the sum of fractions rule in 10 CFR Sec-i tion 61.55.
If the Pu-241 TRU a-activty ratio were, in fact, 30, waste with a TRU content as low as 60 nC1/g would exceed the Class C limit. How much of the CEVNC solidified dilute liquid waste would exceed the limit would depend on specific values of the TRU concentration and the Pu-241:TRU ratio. (The latter apparently has not been determined.)
In any event, none of this dilute
{
solidified waste would greatly exceed the Class C limit.
l l
The average TRU concentrations of the solid waste and of the solidified I
hot liquid waste, as given in Table 3.4, are a factor of about 10 higher 4
than the Class C limit. All this waste is thus actual TRU waste, and not TRU-contaminated waste. It should be pointed out that the waste contains non-TRU j
isotopea in concentrations which are also orders of magnitude greater than the i
Class C limit. Plutonium-241 concentrations, about 30 times those of the TRU 4 over the Class C limit. Cesium-137 and i
isotopes, are a factor of about 10 Sr-90,(w ose concentrations are 40 times higher than the TRU concentra-
{
3 tions, 9 are present in amounts which exceed the Class C limits by 1 to 2 j
orders of magnitude.
l 3.4 Sussiary 1
]
GEVNC wastes consist of three main types:
(1) hot cell process wastes - generally Cs-137 and Co-60 at higher ac-l l
tivities and packaged in 84-ga11on cement-grouted drums, typically Class B or greater.
l (2) support activity wastes - relatively lower activity wastes con-taining a variety of isotopes, may.be packaged in wooden boxes,
[
SS-gallon drums or, on occasion, 84-ge11on cement-grouted drums, typically Class A.
(3) TRU-contaminated wastes - two main concentration ranges: 80-I 100 nC1/gm and >>100 nC1/ge, currently stored at CEVNC. Wastes in i
the first category are Class C on the basis of TRU content but may exceed the Class C limit on the basis of other isotopes present.
l I
Wastes in the second category exceed the Class C limit by orders of magnitude on the basis of TRU, Pu-241, Cs-137, and Sr-90
(
concentrations.
i.
(
j 1
?
I 19 i
l 1
.._,..-__,-.______.__-..,.c.
I_ :
~ _. - -
. _ - ~.
L i
i-1 4.
EVALUATION OF CEVNC WASTE PACKAGES WITH RESPECT.TO NRC REQUIRENENTS j
Low-level radioactive waste must meet the requirements specified in 10 CFR Part 61 if it is to be considered acceptable for shallow land burial.
The waste must be classified according to the scheme presented in Section 61.55.of the regulation, and also it must conform to the specifications re-j garding waste characteristics given in Section 61.56.
In this evaluation section a comparison is made between the requirements for low level waste l
and characteristics of the waste being shipped from CEVNC to shallow-land i
i burial sites. Information about the CEVNC wastes was obtained from CEVNC j
during the course of this work, and may not represent radioactive waste which will be shipped from CEVNC in the future.
i 4
' f 4.1 Class A Wastes
{
4.1.1 10 CFR Part 61 Requirements - for Class A Wastes' 4
}
Section 55 of 10 CFR Part 61 gives guidelines for classifying low-level l
radioactive wastes (LLW) according to the concentration and type of radioac-i tive species present in the LLV. There are three classes:
A, B, and C, and
[
I-these are determined for a particular waste package using the criteria listed in Section 55. The first consideration is whether the package contains any of j;
the long-lived radionuclides listed in Table 1 of Section 61.55. Table 1:also
}
. gives limiting concentrations for these radionuclides, and these are repro-l duced here in Table 4.1, in which the concentration limits for Class A wastes
[
j are explicitly presented in units of curies per cubic meter, as well as in units more amenable to comparison with valuesiteported by CEVNC in their radioactive shipment records (RSRs). If more than one of the radionuclides listed in Table 1 is present, then the sum of fractions rule is applied.
This rule can be represented as follows:
j SF =
_RNg
' I RNLi I
where RNg = radionuclide concentration in the waste package and, RNLg =
concentration limit for that particular radionuclide from Table 4.1.
As long as SF, the sum of the fractions calculated for the different radionuclides, is j
less than 1.0, the waste is Class A.
l
[
If the waste does not contain any of the long-lived radionuclides f
listed in Table 4.1, then the presence of short-lived radionuclides is consid-ered next. In 10 CFR Part 61, the concentration limits for Classes A, B and C l
t l
of several radionuclides are listed in Table 2 of Section 61.55.
The limits l
for Class A wastes are reproduced here in Table 4.2.
If none of the radionu-clides listed in Table 4.2 is present in the waste, it is Class A.
If a com-bination of the short-lived radionuclides is present, the sum of fractions rule must be applied, and the calculated value of SF must not exceed 1.0.
t I
L l
I t
I f
.,-,__.,___.m_._---__._._,-,
m.
~.
. ~ _
. -. ~
- _ - -. ~. ~. -
k 3
-Tabl'e 4.'1 a
Concentration Limits of Long-Lived Radionuclides for Class A Wastes L
Concentration Limit 4
Radionuclide C1/m3' C1/ft3
'C1/55 gal drumb C1/x17H drume C-14 0.8 0.023 0.17 0.26 i
- Ni-59 (IAM)d 22.0 0.62
'4.58 7.16 l
Nb-94 (IAM)d 0.02 0.00057 0.0042 0.006'5 Tc-99 0.3 0.0085
- 0.062 0.098 I-129 0.008 0.00023 0.0017 0.0026 TRU (tt/2>5 yr)*
10f Pu-241 350f Co-242 2000f 1
acalculated from values given in Table 1,10 CFR Part 61.
3 b55-ga11on = ~7.5 ft.
3 c 17H drum = 11.5 f t.
x d
IAM = in activated metal.
'TRU = a-emitting transuranic.nuclides-(half-1'ife greater than 5 years).
IUnits are nanocuries per gram.
~
~
r i
Table 4.2 i
a Concentration Limits of Short-Lived Radionuclides. for Class A Wastes i
Concentration Limit b
C i
Radionuclide Ci/m3 Ci/f't3',C1/55 gal drum C1/x17Hatrum i
All with t1/2 < Syd 700 19.8 145.6 227.9 H-3 40
-1 13 8.32
-13.0 l
Co-60 700 19.8 145.6 227.9 Ni-63 3.5 0.09 9 0.728 1.139 Ni-63 (IAM)*
35 0.99 7.28 11.39 j j 4
Sr-90 0.04 0.0011 0.008 0.013 Cs-137 1
0.028 0.208 0.32 aFrom Table 2 in 10 CFR.Section 61'.55.
b 3
55-ga11on = ~7.5 f t.
3 l
c 17H drum = 11.5 f t.
x d
.e., all radionuclides with half-life less than 5 years.
J i
~
i
- IAM = in ~ activa ted me tal.
i 22 i
-.i----
y--w-y-- - - - + - -
,--.-m
,-,c, wr
-c..-
-,,.,,,.-----,m._,
,.,,,,-,--,-,----y,,,r----
.-,,,---.--.,,_,,-,------,-w--,--.------.-
-,w.e,.
I If a combination of both long-lived and short-lived radionuclides is present in a waste package, the waste can be labeled C. ass A provided the limits listed in both Table i and Table 2 are not exceeded.
The waste characteristics requirements for Class A wastes are given in 10 CFR Section 61.56.
These requirements deal with the chemical and physical nature of the waste package. Section 61.56 specifies that cardboard and fiber-board boxes cannot be used for packaging wastes. Liquids are required to be solidified or packaged in an amount of absorbent sufficient to absorb twice 7
the volume of liquid. In solid wastes containing liquid, the liquid may not exceed one percent of the volume.
Chemical stability with, respect to detonation, explosive decomposition, and explosive reaction with water is also required of the wastes. Genera tion or containment of toxic gases, vapors or fumes which could be harmful to people is disallowed, as well as pyrophoric mate' rials. If pyrophoric mate-rials are present in the waste, these must be processed so as to be nonflam-mable. Hazardous, biological, pathogenic and infectious materials in wastes must be treated so that the potential hazards from these materials are reduced as much as possible.
Requirements for gaseous radioactive wastes are also prescribed. These must be packaged so that the internal pressure does not exceed 1.5 atmospheres
(~7.4 psig) at 20'C, and the total activity is limited to 100 Ci per container.
4.1.2 Evaluation of Class A Wast $ Shipments From CEVNC 4.1.2.1 Support Activity Waste As discussed in Section 3, waste from support activities is packaged in S5-gallon drums or wooden boxes while hot-cell waste is packaged in an 84-ga11on extended 17H drum which is grouted with cement. GEVNC further sub-divides support activity waste into two groups:
"s tabilized" and "not s tabil-ized." The " stabilized" wastes are solidified with cement and packaged in a 55-gal drum or x17H drum. These wastes can include dewatered resins, non-compactible items, and solidirled liquids.
"Not stabijized" waste is 3 plywood boxes.
compacted into 112 f t The major radioactive contaminants in support activity and hot-cell waste streams are Co-60 and Cs-137. As mentioned in Section 3, small amounts
~
of Xe-133 in pump oil are generated, and this is packaged in aecordance with 10 CFR Part 61 requirements. Other radionuclides may be present in low-level liquid waste processed as a contract service to waste generators in the area.
However, specific information is not available regarding which radienuclides and activity levels are present in this portion of the waste streams. Based on information supplied from CEVNC, these wastes are not greater than Class A and, giveh the absence of more specific information', they are,not considered in this evalua tion.
23 L.
f
-+
J The presence of hazardous chemicals, e.g. acetone, appears to Se a, minor concern, since these are only present in trace amounts.
From the. description of the waste streams given in Section -3 and a review of the RSRs provided by GEVNC, it appears that all wastes shipped in wooden boxes or 55-gallon drums meet the minimum Class A requirements with respect to radionuclide levels and waste characteristics. Liquids are either solidified in cement or absorbed on diatomaceous earth, -in both cases in a 2:1 ra tio, to the liquid volume.
4.1.2.2 Hot Cell Wastes Hot cel* wastes in general are packaged in 84-gallon extended 17H grouted drums and shipped in GE model 1600 ' shipping casks. These wastes con-tain higher activity levels than support activity wastes and are considered for the most 'part to be Class B or Class C wastes. However, as mentioned in Section~ 3, it appears that at least one of this type of package met the radio-nuclide -concentration limits for Class A waste (see Table 3.3). - In such a case, this package appears to meet 10 CFR Part 61 minimum waste requirements.
4 4.2 Class B atd Class C Wastes 4.2.1 10 CFR Part 61 Requirements for Class B and Class C Wastes The method for determining whether waste is Class B or Class C is de-tailed in Section 61.55 of 10 CFR Part 61.
As with Class A wastes, the pres-
- ence' of long-lived radionuclides is the first consideration. Class B wastes may not contain any long-lived radionuclides. If long-lived radionuclides are present in excess of the Class A concentration limits given in Table 4.1, the waste may be considered Class C provided the ~11mits given for this class are not-exceeded. These limits are ten times the values given in Table 4.1.
If short-lived radionuclides are present in a waste package, the guide-lines based on Table 2 of 10 CFR. Part 61 must be followed. The radionuclide limits for Class B and Class C wastes are reproduced here in Table 4.3.
If more than one radionuclide of either type, i.e., short. or long-lived,'is pres-ent, then the sum of fractions rule must be applied as with the Class A wastes.
s
?
m If the Class C limits of.either long-lived or short-lived radionuclides are exceeded, the waste is generally considered not. acceptable for shallow land burial.
The general characteristice; for Class A wastes discussed earlier also apply to Class B and Class C wastes, i.e.,
the waste may not contain free liquids (in excess of 1%), pyrophoric or explosive materials, or materials >
which will generate significant quantities o,f gas.
t i
a t.
i 14 4
= ~ -
,-<e5-y,
-r
'i 4
Table 4.3 Concentration Lir.its of short Lived Radionectides for Class B and Class C Westeaa Concentration t.imit Class B Class C Radionuclide C1/m3-C1/ft3 C1/55-galb C1/m17HC Cl/m3 C1/ft3 C1/55-galb C1/m17H drueC A11 with d
t1/2 < Sy e
e H-3 e
-=
e Co-60 e
e Nt-63 70 1.98 14.5 22.7 700 19.8 145.6 227.8 N1-63 (IAM)I 700 19.8 145.6 227.9 7000 198 1456,
2279 Sr-90 150 4.24 31.2 48.8 7000-198 1456 2279-r Cs-137 44 1.24 9.15 14.3 4600 130 957 1497 aCalculated from Table 2 in 10 CFR Section 61.55.
b 3
55-gal = ~7.5 ft.
Cx17H drue = 11.5 f t3 r
d.a..
all radionue11 des with half Itves less than 5 years.
t C No limits.
f!AM = in activa ted me tal.
4 In addition to the minimum requirements on waste characteristics given in.10 CFR Section 61.56(a), minimum stability requirements are specified in 10 CFR Section 61.56(b). These relate to structural stability, minimization of free liquid content and void spaces in the waste.
Structural stability means *. hat the waste will maintain its form and physical dimensions for a minimum of 300 years under expected disposal condi-tions,.which may include weight of overburden, moisture, microbial activity, radiation ef fects and chemical-changes. Stability can be provided by the waste form itself by processing to a stable form (e.g., by solidification-in a binder) or.by placing the waste in a container which can provide structural l'
stability. The limits on free liquid are 1% of the volume if a container is used, and 0.5% of the volume if the waste is processed to a stable form. Void l
. spaces in waste packages must be minimized to the greatest ' extent possible.
- 4.2.2 Evaluation of Class B and Class C Wastes from GEVNC With Respect to 10 CFR Part 61 Requirements 4.2.2.1 Minimum Requirements I
From the description of the support and hot cell activity waste
- streams and based on the review of the RSRs from GEVNC given in Section 3, it appears that only waste generated in hot cell activities contains sufficient levels of radionuclides to be considered Class B or Class C.
Some of these may'even meet Class A radionuclide concentration limits (see Section 4.1.2.2).
Table 3.3 lists the packages of this type shipped during the period 4/83 through 8/83. The right-hand column of Table 3.3 indicates the waste 25 l
4 m
-2.
.,,.--,,..~.,--.
m.-
-,m._..
J class for each of these according to the specifications given in 10 CFR Par t _61.. It'is apparent that the wastes from hot cell activities may go 'from t-
' Class A to unacceptable (greater than Class C).
It should be noted that the one. package which fails ' to meet the -
Class C limit does so because it exceeds the concentration limit for C-14.
a Further information regarding this particular package was obtained from-GEVNC,-
and it was. reported that the C-14' contaminated waste consisted of aluminum nitride pellets which had.been encapsulated. According to the RSR, the waste was grouted hot cell' waste. However, the aluminum nitride pellets were used for the production of C-14, which requires a high neutron flux, such as that in a reactor. Thus, the possibility exists that some waste generated at the GEVNC research reactor prior to its shutdown is still being shipped. At pres-ent, C-14 is not's major concern in wastes from hot-cell activities (see Section 3).
All hot-cell wastes are packaged according to the description given in Section 3, i.e., the wastes are placed in metal baskets,- the baskets are i;
placed in 84-gallon extended 17H drums, and the drums are filled with a cement grout. Simulated waste packages processed in this manner have been; prepared by CEVNC(4). Sectioning of these simulated packages has made it possible ta i
observe the absence of free liquids and void spaces in the cement matrix. As noted in Section 3, liquid wastes are solidified in paint cans before the cans are placed in the inner metal baskets. Assuming the radioactive. waste pack-i ages correspond to the simulated ones, it appears that the solidification of hot cell wastes in cement as practiced at GEVNC meets the free-liquid and void-space requirements given in 10 CFR Section 61.56(b) for Class B and Class C wastes. In addition, based on the information provided by GEVNC that any and all hazardous chemicals in these wastes are present in trace amounts, it is concluded that.the general requirements in 10 CFR Section 61.56(a). which I
L cover non-radiological hazards are fulfilled also.
The structural stability requirements for Class B and Class C wastes L
may not -be so readily -fulfilled, however. In 10 CFR Section 61.56(b) the
~
statement is givyn that "a structurally stable waste form will generally main-I tain its physical dimensions and form...." More specific guidelines are I
listed in the Branch Technical Position on Waste Form (TP), and ' these are dis-l.
cussed in more deta'il in Section 4.2.3.
Some general considerations of the l
waste package and its structural stability with respect to 10 CFR Part 61 will be given here. In particular, the issue of maintenance of monolithic form and physical dimensions will be discussed from the viewpoint of potential degrada-tive effects of spalling and cracking of the concrete as a consequence of cor-t rosion of the internal s teel components.
4.2.2.2 Structural Stability
(;
The grouted drum package can be treated as having three components.
The outermost is the carbon steel drum itself, the second is comprised of' the
[
, internal perforated mild s teel baske ts embedded in concrete, and the third is l
26 i
.m
~
I the waste itself solidified in cement.. A discussion of-the potential failure modes in which structural stability may be compromised is given in the follow-
- ing sections.
I 4.2.2.2.1 Carbon Steel Outer Container -
l The outer container ip an 84-gallon 17H carbon steel drum with a ring-bolt seal. Gause, et al.(l> have considered the stability of carbon steel drums in a trench enviroment. As they point out, it is not possible to accurately estimate the drum lifetime at a disposal site from existing ~ data on carbon steel corrosion in soil. They have, however, estimated a time to pit-ting of fran 2.5 to 9.6 years and a container lifetime of from 10 to 120 years.
- ~
depending on soil conditions. Thus, the carbon steel drum is expected to pro-l vide stability for only a relatively short time compared to the period over which structural stability is required.
4.2.2.2.2 Perforated Mild Steel Baskets Embedded in Concrete I
i The perforated mild steel baskets embedded in concrete may be con-sidered analogous to reinforced concrete with an overpack of 0.75 inches. A 4
possible degradative process that could take place in this section of the waste package involves corrosion of the metal followed by spalling of the con-t crete. The spalling is a result of the pressure generated by the greater i
volume occupied by metal oxide versus that occupied by the non-corroded metal.
It has been found that formation of rust on steel cembers embedded in concrete is accompan by a volume increase which can give rise to pressures up to 300 kg/cm.{e This corresponds to ~4300 psi which is in excess of the typical range of compressive strengths for concrete at 28 days curing time.(lll The issues of (1) whether or not such corrosion can be expected to occur and (2) the rate at which it occurs, are discussed in the following l
sections.
Factors Affecting Corrosion of Steel Embedded in Concrete Reinforcing s teel in concrete is covered by a passivating film which must be penetrated before corrosion of the steel can take place. Chemi-cal factors which strongly influence the depassivation of the film and, con-I sequently, the onset of corrosion include:
l (1) chloride ion i
r (a) the apparent threshold level of chloride at the j
steel surface in concrete that will cause breakdown I
of the passive film is be tween chloride by weight of concrete (0.025 and 0.035%12,13) for a con-crete with a cement factor of 700 lb/yd. GEVNC's cement is ~1500 lb/yd and it has been assumed that this level of chloride is necessary in this case as well.
27 l
- _ _. _ _ _ _, _, _,. ~,.
e
-(b) ' for corrosion o' reinforcing s teel to occur in.
a saturated, aera*.ed Ca(OH)2 solution, the threshold concentistion of chloride ranged from 0.02 to 0.03 L.(14,15),
(2) pH (a)' the high pH of concrete is geneEally the major factor in de termining the behavior of steel embedded therein; typical Portland cement con-crete pHs are 12 or above. This high pH 6ay be a corrosion-inhibiting factor.
(b) the possibility exists for pH cells to be set up be tween regions in the concrete, e.g., between the outer surface where contact with water may have lowered the pH and nearer the me tal surf ace where the pH may still be high. This may or may not be a mitiga ting f actor in any given
.~
reinforced concrete structure.
It has also been shown that the interaction of pH and chloride influences the. threshold chloride levels necessary for the initiation of cor-rosion, and once such initiation has occurred, the presence of oxygen is critical in supporting corrosion. Of course", the presence of chloride and oxygen at the metal surface depends in part on the perdeabiltiy of the con-crete. In other words, it is necessary that both initiators and supporters of corrosion diffuse through the concrete before reaching the reinforcing steel.
Conditions Expected in the GEVNC Grouted Drums and at the Hanford Burial Site The concrete overpack on the metal perforated baskets is 0.75-inches thick, which, were the chemical components necessary for initiation and propaga tion of corrosion (chloride and oxygen) present in sufficient quantities, would probably not present a significant barrier to these chemicals.
Diffusion of chloride ion to a reinforced concrete rebar has been documented to occur in less than one year for concrete 4-in, thick.
The question of whether chloride is present in sufficient quanti-ties to bring about initiation of the steel corrosion has been considered from two points of view:
l (1) the viewpoint of outside the package--once breach of the outer carbon steel drum has occurred, the j
grouted drum will be subjected to exposure to the l
trench soil environment which includes chloride.
l and l
28
[
(2) the viewpoint of inside the package--chlorides in the GEVNC wastes, or chloride present in the grout mixture itself may both be potential-sources of corrosion-initiating lon..
[
The chloride concentrations given earlier as the threshold for cor-rosion initiatica in a Ca(OH)2 solution (0.025.to 0.035 M) have been used
(
for comparison'here because it is believed these values are conservatively low.
It has been suggested that the amount of chloride needed to cause corro-sion in concrete is significantly in excess of that needed in Ca(OH)2 solu-i 7
.tions of similar pH due to the presence of a lime-rich layer on the surface of I
the steel.in concrete, which effectively acts as a source of " reserve alkalin-
.ity" and thereby increases)the chloride ion concentration necessary for pas-sive film breakdown.(16,17 1 - e Hanford soil (as mentioned in Section 2, GEVNC ships their wastes to the Hanford, Washington site) has begn (ound to have chloride present at 1.6 x 10-1 mg-eq per 100 grams of soil.(18/ An idea of a possible chlor-ide ion concentration that might contact these wastes at the burial site can i
be arrived at by assuming 100 mL of water were to contact 100 grams of soil, deple te - the soil completely of its chloride content, and then enter the
. grouted drum concrete monolith. The effective chloride concentration in this 100 mL would be ~0.002 M, or over than an order of magnitude lower than the threshold value of 0.025 M.
In addition, even were the chloride present in sufficient quantities in Hanford soils, there is eviden hat annual evapora-9 so tha t water tion potential at that site exceeds total precipitation transport of chloride to the waste form should be precluded.
7 GEVNC believes that chloride concentrations in their wastes are insignificant. Chloride concentrations in the grouting mixture would arise from the chloride present in tap water used to prepare the mixture. A table l.
of results from CEVNC tap water analyses is given (Section 4.2.2.1 and l
Table 4.4) and shows a maximum chloride value of ~60 mg/L (January 1983)..
l This~ converts to ~0.002 M chloride concentration in the tap water which, of course, is subsequently diluted further as the water is mixed into the l
. grouting material. It can be seen that this also is below the threshold
(
chloride concentration and, thus, concern about the tap water concentrations l
exceeding the chloride threshold for corrosion initiation can be eliminated.
m Additionally, constituents of concrete, specifically tricalcium aluminate (C A), can react with diffusing chloride, ther reducing " free 3
chloride" available to implement the depassivation step.
} The anount of C A in concrete is dependent on the type of cement and, for normal Portland 3
cement, C A constitutes 45% of the cement mix.(21) 3 l,
In summary, it appears that the initiation of corrosion of the per-l forated steel baske t in the grouted drum package may not occur due to insuffi-l cient chloride ion concentration. Were depassiviation of the embedded metal to occur, the possibility of continuation of the corrosion to the extent 29 i
d.
I necessary to bring about spalling is dependent on the presence of sufficient oxygen and water. Oxygen should be present 16 suf ficient amounts unless-conditions of the burial trench were to become anoxic (this is expected to be l
unlikely at Hanford), but water is expected to be extremely scarce at the Hanford site.
I Rates of Corrosion of Steel in Concrete General information available on the rates of corrosion of steel in Crowth of Fe3 4 on steel as a concrete is given here for completeness.
0 f unct1 n gf time and of the permeability of the concrete is given in work by' j
9 l
Tuuti.t22/ Corrosion of reinforced steel to the extent that failure of the l
con 9 rete cover occurred has been studied in palt water solutions for natu-ralt23) and impressed voltage situations.(24) The times to failure 'were i
l
~345 and 7-8 days, respectively. Steel, embedded in concrete and stored on the ocean floor for fif teen years as part of a low level radioactive waste
[_-
package, was found to have corroded, but not to a sufficient extent that spaft ling of the concrete occurred.(25) This may have been due to the fact that in seawater environments, a reaction can occur at outer surfaces of the con-crete, and in the' pores, whereby Mg(OH)2 is precipitated within the pores due to its decreased solubility product over Ca(OH)2 This evidently leads s
i to a decr ag d permeability, and thus diffusion rate, in concrete immersed in seawater.
Additionally, the level of oxygen on the ocean floor is sig-nificantly lower than at the surface and this lack of oxygen also may lower i
the steel corrosion rate.
i.
i Summary l
The structural stability of the grouted drum monolith is subject to compromise should spalling of the concrete occur as a consequence of corrosion I
of the embedded steel perforated baskets. This corrosion must be initiated j
and propagated. Conditions for this to occur depend on several chemical cons tituents; particularly chloride ion, oxygen, and water. It is anticipa ted l'
that insufficient chloride is present both at the waste package burial site (Hanford) and in the package itself for initiation of the corrosion of the steel to occur. Spalling of the concrete overpack due to corrosion of the perforated baske ts is thus not expected at the Hanford site.
4.2.2.2.3 Wastes in Concrete The third component of the waste package is comprised of the waste -
itself solidified in cement. The waste can consist of paint cans containing solidified liquids, compacted paint cans containing compactible solid items, and miscellaneous solid items such as plastic and rubber tubing, metal pieces,
' pape r, other pla s tics, 's te.
This inner level of the waste package may be con-sidered analogous to reinforced concrete due to the presence of metallic j
items, or it might also be treated as an analogue to a concrete mix containing a rather unique aggregate. This latter consideration may be the more signifi-cant with regard to the long-term structural stability of the inner component wa s te fo rm.
30
e In general,' the properties of ' concretes are dependent on a number 2
of. variables.. Among th'ese are relative proportions of cement,. aggregate, and wa ter, the type of cement, type of aggregate, gradation of aggregate paggi-cles, and distribution of aggregate particles within the cement matrix. tz7,28)
Aggregates used in concretes are generally of a mineral type, with a gradation of sizes from fine to coarse (usually no greater than 3 inches in the largest l
dimension). ( 28) Chemical characteristics of the aggregate, particularly at l
the surface, influence the strength and durability of the concrete mix as a l
result of cement-aggrega te surf ace interactions.
l The CEVNC wastes which are packaged in the extended 17H drum ob-viously constitute an inhomogeneous mixture of items. Extremely limited in-l formation is available on the properties, i.e., s trength and durabili ty, of cement forms containing cellulosic:, plastics, and metals other than steels t
i and aluminum. Some concerns that should be addressed are:
I[
(1) the influence of size, geome try, and distribution of the waste items on the waste form properties; (2) wh' ether, given the inhomogeneity of the waste stream, i
there is a limiting waste / cement binder ratio as has i
been found with other waste streams, e.g., lon-exchange resins;(29) t l
(3) whether chemical reactions between cement and the dif ferent waste materials can occur, and whether these 'rould compromise the overall integrity of the waste form.
These points have been raised to indicate that an evaluation of 'the s tability of the waste is not possible given the informa tion currently avail-t able. Stability and monolithic form are provided initially by the outer car-bon steel drum and perforated me tal baske t.
Af ter the ' cement grout has set, the outer drum need not be considered since. stability can be provided by the I
wa s te fo rm i tself. A ma tor question then appears to be whether the - baske t l
embedded in cement can.pt ovide the required stability as discussed earlier, l
since it is not clear thac the different types of waste encapsulated in cement
[
matrix provide stability in the absence of the metal basket.
l The requirement for structural stability has been addressed in L
terms of the different components of this package.
l
[
(1) The carbon steel outer drum is expected to corrode totally within 120 years af ter burial.
It cannot be relied upon to supply structural stability for the required period, f
(2) The perforated metal baskets embedded in concrete i
can be considered analogous to reinforced con-crete and, as such, are expected to be able to l
31 l.
i
~
6 i
supply structural s tability for a significant time. A principal failure mode, ~ spalling of the concrete due to pressures generated through the rusting of the embedded steel, hts been consid-ered and determined to be unlikely given the absence of sufficient chloride to initiate the corrosion process. - If some other chemical (s) capable of initiating the corrosion of the bas-ke ts were present, the probability _ of causing spalling is quite low due to the scarcity of
~
water, which is needed in the corrosion process.
(3) The component made up of the waste items themselves embedded in concrete is difficult to evaluate with respect to structural s tability. Since many of the waste items are steel, this component could also be considered similar to reinforced concrete and, hence, also subject to spalling but, it is believed that this can be ruled out on the same basis as in (2). However, the heterogeneity of the concrete /
waste item matrix may be the limiting factor in its ability to provide _ structural stability. Normal monolithic concrete forms with conventional size aggregate are expected to provide stability, but the behavior of concrete in which there are randomly-distributed, heterogeneous objects of sizes quite in excess of the concrete aggregate particle size cannot, at this time, he predicted.
4.2.3 Evaluation of GEVNC Class B and C Wastes With Respect to the Guidance in the Technical Position on Waste Form The evaluation of the CEVNC extended 17H cement-grouted drum waste packages with respect to the guidelines of the Technical Position on Waste Form follows. Each of the guidelines has been considered individually, and where it has occurred that there is overlap of factors being considered, reference to the pertinent section(s) has been made. It should be noted that, for the purposes of evaluation of the CEVNC waste packages themselves, two guidelines from the TP'do not directly apply (process control program and sam-pie testing size). These are, however, included for completeness.
Given the expectation that the carbon steel outer drum used for the GEVNC Class B and C grouted drum packages will not last beyond 120 years af ter burial, the waste form has been taken as the concrete-waste monolith that fills the carbon steel drum, including the perforated baskets and the wastes
-themselves. For the purposes of several calculations involved in the evalu-ation, the isotopes present in the package have been considered sequentially at conservatively high (Class C limit) values of Cs-137, Sr-90, and Co-60.
The values used have been based on statements made in a CEVNC report (4) to the effect that a grouted drum package may contain 5000 Ci of aged mixed 32
4 x
fission products or 3000 Ci of Co-60 (See Section 2).
Given that a reason"-
able activity distribution for aged mixed fission products is 50/50 Cs-137 and Sr-90, the 5000 Ci was halved to derive the individual isotope activities, but this proved to be in excess of the Class C limits. Therefore, Class C limits were taken for Cs-137 (1500 Ci) and Sr-90 (2300 Ci) for this size package, and for Co-60, GEVNC's estimate of 3000 Ci was used.
s 4.2.3.1 Process Control Program The TP states that radioactive waste generators should implement and maintain a process _ control program to demonstrate periodically that the solidification system is functioning properly and waste products continue to meet the 10 CFR Part 61 stability requirements.
Waste specimens should be prepared such that they are representative of the waste streams to be solidified.
The CEVNC waste streams that result in those wastes that are packaged in tne cement-grouted extended 17H drum are the hot cell process wastes and the " support" (principally, clean-up and maintenance) wastes.
In general, the hot cell process would be classified as Class B or greater (and hence, are required to have stability) while the " support" wastes usually are disposed of in packages with activities in the Class A range. As was discussed earlier, f_
no GEVNC wastes of activity high enough to exceed Class A limits are disposed of in regular 55-gallon drums or wooden crates.
The handling of radioactive wastes at CEVNC is subject to quite spe-cific guidelines in the form of written instructions. However, it is not clear that these instructions constitute a full process control program in-cluding explicitly stated directions and weight / volume ranges for the mate-rials used in the cement grout mixture.
It is known that CEVNC uses a set grout mixture which consists of 600 lbs of Portland cement, 600 lbs of sand l
and 30 gallons of water (tap water - sample analysis is given in Table 4.4) i per ba tch.(5) The amount of variation that can be tolerated in these values I
has been detailed in the GEVNC Radioactive Products and Services Operating Procedure, but it is not clear that the mix itself has been thoroughly characterized with regard to its use in a Class B or greater waste package.
i l
The process for packaging of GEVNC Class B and greater wastes in-
[
volves (1) placement of waste items in the two steel perforated baskets, (2) placement of the baskets in the extended 17H drum, (3) introduction of the
[
cement grout mixture, (4) vibration of the drum during (3) to ensure effective dispersal of the grouting material throughout the waste package, (5) setting and hardening of the grout, (6) inspection of the waste for free water, (7)
- I testing of the grout for hardness, (8) photographing of the waste package, and (9) placement and ring bolting (clamp ring) of the drum lid.
A simulated
[
grouted drum package has been tested for its ability to inhibit dispersal of the waste materials. During this testing, a sample package 'was subjected to a 30-f t drop which res gjed in crumbling of one corner and a " single fracture across the diameter."
(It is presumed that this crack represented the l
plane between the two perfora ted s teel baskets.) Additionally, the grouted i
i
(
(
33
--__-._._ -.~_- __-_-_ _,.__,_ _,,-
}
/
r cylinder containing simulated wastes was cut into four sections for examina-trated practically every space in the matrix. grout was found to have pene-tion of the degree of grout penetration. The
)
Table 4.4 m.
GEVNC Tap Water Analysis (1983)
Influent Nonradioactive Constituents (mg/L)
Chlorides Chromium Copper Lead Mercury Zine pH January 59.4 0.002
'<0.0001a 0.005 0.0001 0.005 7.2 February
<0.5 0.0.03 0.0029 0.01 0.0006 0.008 8.25 March b
0.008 0.0035 0.0006 0.0002 0.038 b
April 2.0 0.005 0.0031 0.003 0.0002 0.010 7.9 May 3.0 0.001 0.0026 0.013
<0.0001 0.035 8.15 June 3.,2 0.0038 0.0046 0.0004 0.0004 0.0117 7.85 July 4.0 0.001 0.0029 0.002 0.0002 0.006 8.75 5.4 0.008
<0.0001 0.01 0.0004 0.046 7.4 August Sep tembe r 2.35 0.,01 0.0012 0.012 0.0003 0.011 7.3 October 2.0 0.013
<0.0001 0.034 0.00015 0.016 9.6 November 2.0 0.001 0.0001 0.01 0.0007 0.05 9.3 Dec embe r 3.5
<0.005
<0.0017
<0.001
<0.0001
<0.01 8.9
~
a< indicates less than the detection limit for the measurement method.
2 bLost sample.
9 Summary l
At GEVNC there are explicitly written handling instructions, but it is not clear that there is a full process control program. GEVNC has estab-lished procedures to ensure a uniform cement grout is obtained. It has not been demonstrated that these procedures will ensure compliance with the recom-mendations given in the TP.
Ideally, the proceus control program implemented by the generator would be one such that the end-process waste packages were consistently structured according to a design that produced a package / waste form with the required s tability.
4.2.3.2 Co'apressive Strength The TP states that solidified waste specimens should have compressive Con-strengths of at least 50 psi wpen tested in accordance with ASTM C39.
l crates of varyi"ug compositions 11) are expected to have compressive strengths of 2650 to 4000 psi at 28 days., Typically, the GEVNC grouted drum i
concre,te is allowed to set for two days prior to hardness testing and it in 1
34
~
i
not known whether 28 days elapse prior to shipping of these packages to the burial site.. It is expected, huwever, that concrete alone would have a com-pressive strength of at least 50 psi. The GEVNC grouted drum package has one potentially significant influencing factor, namely, the presence of "non-concrete" waste materials in the concrete monolith. The waste materials them-selves (as outlined in Section 3) may be metal, paper, glass, or plastics which have been placed in me tal paint cans and compacted prior to grouting.
These items may or may not represent locations of potential cracking or non-adhesion of the grout such that there may be failure upon compression. Such questions can only be answered by actual testing of representative waste forms and, as mentioned earlier, performance of such testing is planned by GEVNC.
4.2.3.3 Radiation Stability The TP states that waste form specimens should remain stable af ter being exposed in a radiation field equivalent to the maximum level of exposure expected from the proposed wastes to be solidified. Specimens should be ex-8 posed to a minimum of 10 rad and specimens should have a minimum compres-sive strength of 50 psi following irradiation.
GEVNC has indicated that " aging" tests have been performed on cement
, cylinders in which the grou t mixture has been duplicated. Apparently, no me tal, glass or plastic waste items were present in the test samples. The 8 rad and no breakdown was observed.
cylinders were exposed to a dose of 10 Thus, it* may be expected that the grouting mixture is stable to this radiation dose. However, two important factors must be considered before it can be con-cluded that the GEVNC grouted drum waste packages are expected to have radia-tion stability (1) The maximum accumulated doses which GEVNC wastes may experience are on the order of 109 rad and thus, the ef fects of this dose on a representative waste form must be tested, and l
(2) the presence of waste items, particularly of paper and plastics, f
may influence the behavior of the grouted waste cylinder in a radiation field because the radiolysis products of these mate-rials have the potential to be degradative.
I Activity loadings and expected accumulated doses at 300. years based on these loadings which may occur in the GEVNC grouted drum packages are given in Table 4.5.
The procedure and assumptions made for the dose calculations are discussed below.
35 s
c-.
4 h
Table 4.5 kadlonuc ttde Activt ties anJ Accumulated Doses fo r CEVNC Grouted Drum Weste Packegee Weeteb Upper
- Up pe r Sh t paen t Clase C a
d d
Ac tivi ty Summa ry Activi ty g
Y to tal 8'
Y*
- to t al*
a.
R adionuclide t
Le t-
. D( 0 r) D(
r) D(
r) D(
r) D(
r)
D(
r)
Aged MFP S000 none given taken as $0/50 Co-137 2500 81 1500 1.2x109 1.54x109 2.7x109 7.2x108 9.2m108 1.6a109 aqee 1.1x109 1.0x10 none 1.0x109 9
Sr-90 2500 none given 2300 1.1x109 Co-60 3000 159 not 1.2x108 1.25a109 1.4x109 1.2mt08 1.25a109 1.4x109 spectitcally set
- Values from CEVNC Report Ref erence 4.
b alues f rom Weste Shipment Summary, Attachment to Reference S.
V
- 10 CPR Part 61 values extended to 84-ge11on container, d alculated for activity values from (a).
C
- Calcula ted for activity values from (c).
Radiation Dose Calcula.tions(30)
The equation used for the beta dose calculation is 3
AC4 E, x 8.76 x 103 h yr-1 t (300 yr)==
D Ag (1 -e-A t) l where
'A is a proportionality constant equal to 2.1 x 103 rad cm3 MeV-mci +cm-3 Ci is the activity density of the ith radionuclide in the waste form in mCL cm-3 Et is average beta energy of the ith radionuclide in MeV.
At is the decay constant for the ith radionuclide in yr-1 t is the time period of interest (for these calculations, 300 yrs).
Pertinent values for all of these parameters for the radionuclides concerned are given in Table 4.6.
Substitution of the parameters given in Table 4.6 and of appropriate values of'Ci (given in Table 4.7) for the corresponding activity limits, yields the total accumulated beta dose for 300 years. The waste activity has been assumed to be homogeneously distributed throughout the container.
36
. f' Table 4.6 Dose Calculation Parameter Values for Principal GEVNC Waste Radionuclides A
t1 Et rt Radionuclide (yr-1)
(y/2 (rad
- cm h-ImCi-1) 2 r)
(MeV)
Co-60 0.132 5.25 0.094 12.8
.Sr-90 0.025 28 0.200 no y Cs-137 0.023 30 0.19 5 3.3 Table 4.7 Activity Densities for the GEVNC Waste Radionuclides b
Ca Ci Radionuclide i
(mci /cm3)
(mci /cm3)
Cs-137 7.7 4.6 Sr-90 7.7 7.1 Co-60 9.2 9.2 abased on values in Reference 1.
bBased on values in 10 CFR Part 61.
I The equation used for the gamme dose calculation is CgP g x 8.76 x 103 h
- yr-1 D T(300 yr) =
i g
-A t) t (1 - e A1 i
l where Ci has the same meaning as before (see Table 4.7)
Pt is the gamma dose constant for the ith radionuclide (see Table 4.6) g is the geometry factor for the particular three-dimensional structure under consideration, conserva tively taken as 160 cm(31) for the 84-gallon extended 17H drum used at GEVNC.
At is the decay constant for the ith radionuclide.
37
Tissue equivalency has been ' assumed. Subs titu tion of the appropriate values from Tables 4.6 and 4.7 yields the 300-year accumulated gamma dose for the particular radicnuclides considered (given in Table 4.5).
Radiation Exposure Tests As can be seen in Table 4.5, the total accumulated dose for each of 9 rad.
In general, if the Class C the radionuclides individually exceeds 10 limit activities of these radionuclides are expected to be pre,sent, the mini-mum radiation testing accumulated dose should be at least 109 rad.
Radiolysis Effects The radiation exposure experienced by the waste materials in the GEVNC grouted drum package may lead to several gaseous and liquid radiolysis products. The ef fect of the radiation dose on the metal and glass waste com-ponents is expected to be minimal, but the ef fects on paper (cellulosic) and plastic materials may be significant and is discussed in the following sections.
Radiolysis of Cellulosic Components of GEVNC Grouted Drum Packages As mentioned in Section 3, the average composition of a CEVNC grouted drum package includer ~50% by weight paper and plastic. Given ~300 lbs of waste per grouted drum, this represents ~150 lbs or 68 kg of paper and plas -
tic.
For the purposes of the radiolysis calculations to be performed here, it is assumed that the paper and plastic component of the waste is divided 50/50 into paper (cellulosics) and plastics.
Additionally, this material has been taken to have a density of 1 g/cm3 and the accumulated doses corresponding to radionuclide loadings at the Class C limit have been used.
Gaseous Radiolysis Products Radiolysis of the cellulosic component of these wastes is expected to result in hydrogen, carbon dioxide and carbon monoxide gas production. The amounts expected have been calculated based on the G value (total gas) of 0.63 molecule /100 eV,(32) and on the total accumulated radiation dose (y and
- 8) for a package containing only Cs-137 at the Class C limit (~2 x 109 rad).
The total doses expected for packages with Sr-90 or Co-60 alone or, in combi-nation with Cs-137 such that unity is not exceeded on application of the sum 9 rad.
of fractions rule for Class C wastes, are all on the order of 1-3 x 10 Given the total accumulated y and 8 dose of 2 x 109 rad, i t is ex-pected that ~9 moles or ~200 L(STP) of gas would be generated by radiolysis over the 300-year period. Given the poFosity of concrete, and the lack of a gas-tight seal on this package, this gas production is expected to be of little consequence. There is, however, one conern as discussed below.
38 T
E
(
~.
Should any water be present in the vicinity of CO2 Production by radiolysis, a solution of carbonic acid in a localized region could result.
Carbonic acid has been shown to enhance the corrosion of rebars in reinforced concrete.(33) Thus, it may be expected that'the corrosion of ferrous metal waste items in the vicinity of cellulosic waste items all placed in the grouted drum package could be aggravated as a consequence of pene tration of the concrete by wa ter.
Howeve.r, the extremely high pH of concrete makes this unlikely.
Liquid Radiolysis Products It has been found(34) that radiolysis of cellulosics may also lead to the production of carboxylic acid group-containing molecules.
The G-value for this is 3.6 molecules /100 eV and the major acids produced are:
fo rmic (G =.2.3, 64%), glucuronic, 2-ketohexanoic, and 3 unspecified "5-ketohexanoic or uronic acids."(34) There is a potential for acceleration of netal waste component corrosion by these components.
It has been assumed that the G-value of 3.6 applies to both y and 8 radiation exposure. The accumulated dose of 2 x 109 rad should yield a total of ~50 moles of organic acids, of which ~30 moles would be formic acid.
This organic acid production may be of significance in regard to acceleration of the corrosion of me tal waste components and, consequently, potential accel-eration of cracking /spalling of the concrete waste form. Formic acid (anhy-drous or 10-85% solution) (in contact with carbon steal leads to corrosion
>50 mils pene tration/ year. 35) Since it is possible that cellulosic wastes may be in direct contact with metal wastes (i.e., not physically in contact with concrete and, hence not necessarily neutralized by the high pH-producing hydroxide in the concrete) in the grouted drum package, this corrosion ef fect of the organic acida produced by radiolysis, and particularly of the formic acid, has the potential to be degradative. The quantitative ef fect of low-molecular-weight organic acids on concrete does not seen to have been documented. Acids such as acetic, citric, malic, and lactic, but not oxalic, have been found to at gg concrete, of ten having "a marked action" within a few months to a year.
When compared to acetic acid in its effect on concretg36) formic acid has been described as corroding concrete more slowly.t On the other hand, it has also been described as being more des truc tive. ( 27 ) There is thus no consensus on the possibility of signifi-cant damage to the concrete grout as a result of attack by organic acids pro-duced in the radiolysis of the cellulosics in these' wastes. The presence of wa ter is not clearly indicated as a necessity for these destructive interac-tions to occur.
Radiolysis of Plastics Component of CEVNC Crouted Drum Packages The plastics fraction of these wastes ( taken as 50% of the paper and plastic component, ~68 kg average per drum) w9uld be expected to generate gas, predominantly hydrogen, in a radia tion field.t 2) The gaseous products ex-9 rad) pected to be produced at the total y and 8 accumulated dose (2 x 10 based on the G-value for polyethylene [3.7 molecules per 100 eV absorbed \\37)} should amount to approxima tely 103 L of gas (STP).
39
r As mentioned earlier, the porosity of the concrete in this package and the lack of a gas-tight seal on the outer drum should allow escape of this radiolytically-produced gas.
Summary The radiation stability of the GEVNC grouted waste packages is predi-cated on the stability in a radiation field of each of the waste package com-ponents: concrete, metal and glass, and paper and plastics. The grouting mix concrete has been tested by CEVNC in a radiation field to 108 rad with no signs of degrada tion evident. The higher activity loadings that these pack-ages may contain, however make testing at higher (~109 rad) doses necessary.
In addition, the ef fects of the radiation field on the waste components them-selves must be considered. The cellulosics component of these wastes may re-sult in production of up to 200 L (STP) of gas (H, CO, and CO) and 2
2
~50 moles of organic acids. The plastics component may result in production 3
of up to 10 L(STP) of gas (predominantly hydrogen). The gases are expected to escape the waste package. The presence of CO2 gas and water (should it be present) may lead to carbonation and, if neutralization by the hydroxide in the concrete is incomplete, consequently, to accelerated corrosion of the ferrous metal components of the package (and, subsequent to this corrosion to cracking /spalling of the concrete). The organic acids produced through radiolysis of the cellulosics may be destructive in contact with either the concrete or carbon steel waste components or both.
4.2.3.4 Biodegrada tion Ef fects The TP states that specimens for each proposed waste stream formula-tion should be tested for resistance to blodegradation. GEVNC has indica ted that blodegradation testing is planned for the grouted drum waste form, but at this time, no informa tion on this testing is available.
Blodegrada tion of the CEVNC concrete-grouted package materials has been considered from two points of view (1) outside the container, i.e.,
in the trench soil environment, and (2) inside the container, i.e.,
in the wastes themselves. In neither case is blodegradation expected to be a primary prob-
- lem, i.e., direct biodegradation of the concrete and carbon steel is not ex-pected since neither material supplies a carbon source.
In addition, the high pil of concrete precludes most microbe growth. From both viewpoints, however, blodegradation by-products may be of concern with respect to corrosion of the container. A discussion of the complexity of the composition and behavior of a system of microorganisms which may exist either in the soil or in the wastes and also of the different chemicals they may consune or produce has been given in cause et al.(2) For the case at hand, it should be noted thatt (1) from the point of view of the soll environment, the considera-tion of corrosion of the container from outside (see Section 4.1) has been based on published measured corrosion rates for i
40
me tals in soils which contained microorganisms (i.e., not sterilized) and thus the effect of microbial activity on corro-sion is reflected in this soil corrosion data.
.and (2) from the point of view of. biodegradation of :the wastes, there is
- the potential for self-sterilisation within the first year for wastes at contamination levels at the upper Class C limit.
The consideration of sterilisa tion of the wastes by radiation from waste radioisotopes must include several factors as summarized below.
(1) Sterilisation has been shown to occur at accumulated doses up to 5 x 106 bacteria); 3{) the. dose rate effect has not yet been totally that.necessary for sterilisation of sporulating r
es tab lished.
(2) At dose rates greater than 104 rad /h, it appears that the sterilisation effect is independent of the dose rate,(38) and the exact lower bound on dose rate has not been de termined.
(3) For the Class C limit Cs-137 activity considered for the CEVNC grouted drum packages the initial dose rates (8 and y) are in excess of 103 rad /h. These dose rates will decrease exponen-tially at a rate dependent on the radionuclide decay constant.
Since the lower limit for sterilisation to have dose rate independ-ence is not known, it is possible that those packages which produce a dose rate within an order of magnitude of the known 104 rad /h upper threshold may also ef fectively startlise the wastes once the necessary accumulated dose has -
been reached.
For the upper Class C activity limit CEVNC grouted drum waste pack-ages, biodegrada tion may not occur until (1) The radiation dose rate has fallen below the " threshold" of 103 rad /h and l
{
(2) the waste package has been re-inoculated with microbes from outside the package.
The intervals over which the waste package may experience the l
103 rad /h or greater dose rates are given for the particular waste isotopes l
in Table 4.8.
Both the y and 8-dose contributions have been included in these calcula tions. Inclusion of the 8-dose implies the assumption that the 8" emitting activity and the microbes are homogeneously distributed through-out the wastes. For each of these' radionuclides (Co-60, Cs-137, and Sr-90) at
[.
the activities given in the table, the necessary accumulated doses for self-j startlisation are reached within the first year. To be specific, at a dose
?
}
41 a
4
~
rate of 103 rad /h 000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or 210 days are needed to reach the necessary accumulated dose (58 of 5 x 106 rad (self-sterilisation). Thus, as long as the dose rate remains at 103 rad /h or greater, a self-s terilisa tion may require 210 days. This is important in reference to the pote'ntial inocu-lation of the waste form from outside following breach of the outer carbon steel drum used in the,CEVNC grouted drum package.
As 64s mentioned earlier, the estimated period to firs t pitting of carbon steel drums in a trench envi-ronment has been given as 2.5 to 9.6 years and drum lifetime estimates range from 10 to 120 years depending on the soil conditions.(1) Once the package is buried, there is expected pitting and eventual total corrosion of the outer,
drum within the time period f rom' 2.5 to 120 years.
Essentially, the drums with activities given earlier will initially be self-sterilized until first pitting and then, even given the possibility of influx of water and the intro-duction of any microbes that may be present therein, the packages should re-main sterilizing (this is assuming, of course, virtually total containment of the activity in the waste form, which for Cs-137 contaminated packages, may be questionabic given the high diffusivity of this isotope in many concretes) until 23, 44, and 64 years have elapsed for Co-60, Sr-90, and Cs-137 packages, res pe ctively.
Table 4.8 CEVNC Grouted Drum Upper Class C Activity Dose Rate Radionuclide/ Activity Loading Interval With Dose Rate >103 rad /h (C1)
(yr)
Co-60/3000a 23 Sr-90/2300b 44 Cs-137/1500b 64 aActivity given in Reference 3.
bUpper Class C limit values.
In general, self-sterilized packages should experience minimal ef-f acts from blodegrada tion. Following the interval during which the waste radionuclide activities are high enough to render the packages sterile, how-ever, the wastes may be subject to inoculation from outside. There could thus be a replenishment of the microogranism population responsible for blodegrada.
i tion. For the GEVNC grouted drum packages, the microorganisms may me tabolize the cellulosics and, possibly, the plastic portions of the wastes. The by-products of this metabolism could potentially degrade the concrete and/or the metal components of the waste form and thereby lead to failure.
42 j
f Several uncertainties exist which preclude the assignment of quanti-tative degradative effects of biodegrada tion. These includes s
(1) when inocu'lation from outside the package will occur, (2) likelihood tha t microorganisms', even should they enter the pack-age once it is no longer capable of self-sterilization, will find, in addition to a location immune from the extremely alkaline pH of the concrete grout, a sufficient carbon source and other necessary growth-sustaining conditions.
(3) the degree to which the ra tes of blodegradation are dependent on the particular types of microorganisms, the ambient gas condi-tions (anaerobic va aerobic), the amount of water present, the type and amounts of nutrients present, etc.
S ummary CEVNC grouted packages may be capable of self-sterilization at least initially, depending on the radionuclides and activities.
It is expected that the high pH of the concrete will inhibit microbial growth, but a description of the ef fects of blodegrada tion cannot be quantitatively given.
4.2.2.5 Leach Testing The TP states that leach testing should be pe f rmed for a mininun of 90 days in accordance with the procedure in ANS 16.1.
Specimen sizes should be consistent with the samples prepared for the compressive tests (Section 4.2.2.2).
Leaching tests should be carried out in a variety of leachants besides demineralized water synthesized seawater teachant be used. specifically, it is pre {gg ed that The leachability index should be grea ter than 6.
CEVNC has indicated that leach testing on the grouted drum waste forms is planned but no da ta is yet available.
The establishment of the teach testing procedure to be used on the CEVNC grouted drum waste form should include consideration of three factors (1) The heterogeneity of both the waste compositions and sizes in the CEVNC wastes will probably make it difficult to ensure that the teach test sanples (and all test samples, for tha t matter) are indeed representative of typical hot cell piecess waste fo rms.
(2) Each particular grout mixture has characteristic radionuclide retention characteristics which makes testing necessary unless the grout mixture exactly duplicates that of a previously tested solidification medium.
Example published Cs-137 leach rates for various cement mixtures are 43
[
r i
2 x 10 12 x 10-2 gje 2. day for Co-137 contami-(a)_ - nated liquid wastes in a cement waste form with a 3:4 solu-tion: coment ratio and 7.5% bentonite clay added.(40)
(b) 10-2 10-3 gje 2 day for Cs-17 in wastes cemented with ordinary Portland cement. 41)
I (c) 10-7 g/cm2. day for Cs-137 in hydrofracture grout
- made of cements fly ash:
0.5 ratio.(42)pulgite-150:
Atta Crundite in a 2.5: 2.5: 1.0:
A process control program could address this issue. Length of curing time has also been found to af fect the teach rates, par-ticularly at initial stages.
l (3) The presence of waste items that are relatively large with re-spect to the components of the solidification medium may change the aggregate character of the concrete, for example, the ad-hesion interf ace be hveen waste item and concrete may be af-facted such that crevices or channels develop between the two phases. Such inhomogenetties may signficantly influence teach I
test results.
E Aside from these issues, the physical distribution of the waste radionuclide activity in the grouted form may be expected to influence the amounts of material released in the teaching tests. For the OgVNC hot cell process' wastes, the radionuclide contamination is in and on the solid waste components themselves and not homogeneously distributed throughout the cement grout mixture during its preparation. The guidance in the Technical Position" that waste forms should have a teachability index greater than 6 is presumably directed toward monoliths containing uniformly distributed activity. The ap-plication of this guidance to a non-homogeneous waste form may be based on the assumption that the regulatory purpose is limitation of radionuclide releases from wastes.
It should be determinable which waste form has better radions-clide retention characteristics through comparison of the teach. test results for two different waste forms, one with a homogeneous activity distribution l
and a teachability index greater than 6, and the other heterogenous (or, con-i taining concentrated point sources) and not having a known or even specifically-defined leachability index.
I
- Portland cement Type 1.
Fly ash (a possolanic material) obtained from the coal-fired Kings ton Steam Plant, Kingston, TN.
Attapulgite-150 is the trade name of a clay product from the polygorskite group of cla/ minerals (with the general formula 5 Mg0*8SiO *9H O).
Grundite is the trade name of a i
2 2
clay product from the 1111te group of clay minerals [with the general formula (OH)4 Kn(A1 *Fe4*Mg4*Mg6) (81 -x*AIx) 020j from Grundy County, Illinois.
4 8
P 44 i
An idea of the expected form of the leach test results for the two types of waste forms jus t mentioned may be obtained through consideration of the following discussion.
The assumptions pertaining to the leaching conditions ares (1) the teaching is dif fusion-limited and hence, the transport of material is driven by the concentration gradient, (2) there is essentially infinite teaching solution available and/or the concentration of the diffusing substance in the leaching solution is zero, and (3) the radionuclide activity initially exists as point sources in the heterogeneous package.
The heterogeneous waste form, with its point sources of activity, will tend toward homogeneity.
Each of the sources of activity will be initially surrounded by a region in which the radionuclide concentration is much less. Thus, due to this concentration gradient, activity will diffuse radially away from the source. Given suf ficient time the activity will dis-tribute itself throughout the form.
In time, release can occur from portions of the forms that did not initially contain the radioisotopes of interest.
Modeling the release from such a form is dependent upon the particular activ-ity distribution in a given package and may not be possible in a generic sense. llenc e, leach testing of this form is required.
It should be noted that there is a potential for the release observed f rom such a teach test to be initially low due to the initial depletion of activity in the surface of the form.
It is important that teach test results which may, a t firs t glance, appear more favorable for the heterogeneous waste form not be misinterpreted such that the waste stabilization medium is credited with the better radionuclide retention abilities. The leach tes t results may be simply reflecting the " induction" period for this type of activity distribu tion.
A more pragma tic approach to the question of the leaching behavior of the CEVNC grouted drum package leads to the two statements:
(1) The fluidity of the CEVNC groacing mixture and the existence of smearable contamination on the waste items will lead to spread (homogenization) of the radionuclide activity throughout the package during the grouting procedure, and (2) that activity which is not dispersed by the grout (i.e., not smearable) will likely remain as a fairly intact point source (e.g., activa ted me tal) and, in any case, probably be rather slow to teach.
45
t Summary Leach testing should be performed on representative samples of the GEVNC grooted drum packages. The tests must be thoroughly and rigorously designed and the results of such tests must be carefully interpreted to ensure
.r that the true behavior of the wastes and grouting material is understood. It is not clear that the heterogeneity of the GEVNC grouted drum monolith will lead to increased cumulative radionuclide release through teaching as compared with a homos noously distributed waste form.
s 4.2.3.6 Immersion Testina The TF recommends that waste specimens should maintain a minimum com-pressive strength of 50 poi as tested using ASTM C39 or ASTM 1074, following
[
immersion for a minimum period of 90 days.
The immersion testing of the GEVNC grouted waste forms is planned but has not yet been completed. The actual behavior of these forms under such conditions cannot be predicted but the observations may be made thats (1) the presence of water may influence the corrosion of the metallic waste items (either directly or in conjunction with corrosive products of radiolysis discussed in Section 4.2.2.3) and, consequently enhance the spalling/ cracking of the concrete (reinforced concrete samples immersed in aqueous chloride solu-tions have been found, depending on conditions, to spell or crack af ter periods of from 7-8 days (24) to 345 dayst23))
and (2) water may cause swelling of the waste celtulosics and some of the waste plastics as well. This may lead to cracking'and crumbling of the waste form.
Incorporation of waste items of sises much different from the cement components may affect the strength / integrity of the concrete.
4.2.3.7 Thermal Dearadation Degradation of concrete with changes in temperature is. caused by sev-eral f actors.(27,43) These include the breakdown of the c'encrete structure due to the dif ferent thermal expansion coefficients of various constituents embedded in the concrete, fr'on s tresses caused by temperature dif ferences between the surface and interior parts of concrete structures during tempera-ture changes an4 f rom freese-thew cycling when water is present in concrete.
4' By far,. the most damaging conditions occur when concrete naturatsd with water undergoes repeated freesing and thauing. Dry concrete per se is not af fected to ensure that no free water remains in the grouted drums.(4) rums are designed by frost. procedures used.by GEVNC to grout their radwestg d 46
In general, temperature changes cause stresses to develop in concrete due to differences in thermal expansion coefficients between the binder and bound solids. The binder in the CEVNC grouted waste is cement mortar composed of a til mix (by weight) of Portland cement and sand. The bound solids in-clude metal articles, compacted 1-gallon cans, glass and other hot cell trash (cellulosics, etc.) placed in the perforated steel baskets. The thermal ex-F(27,43) pansion coef ficient of mortar is typically 8-10 x 10-6 per 0 (i.e., 14-18 x 10-6 per CC).
While for carbon steel it is approximately 11 x 10-6 per CC, for glass about 8-10 x 10-6 per CC, for rubber about 80-110 x 10-6 pe r OC, for lead about 27 x 10-6 O
less steel it is approximately 17-20 x 10-6 per OC.(4I')and for stain-per C Steel rein-forcenent of concrete is a common practice and materials with thermal expan-sion coefficients similar to that of steel would presumably cause no problems due to thermal cycling. Thus, glass and stainless steel should not be af-fected by thermal cycling, but lead might cause some problems for "large" tempera ture expansions.
(It is not known how large a temperature change would be required before expansion of a piece of lead would significantly damage the surrounding mortar *.)
No quantita tive es tima tes of tolerable tempera ture varia-tions as thermal expansion coef ficients for concretes have been found. Cen-erally, dif ferential thermal exgansion of aggregate in concrete is not con-sidered to be a major concern.( 3) Solid rubber objects might cause some degradation of the surrounding mortar upon thermal cycling due to rubber's rela tively large thernal expansion coef ficient. Ilowever, it seems doubtful that tubing or hollow rubber objects would cause a problem.
Tharmal stresses may also arise from temperature dif ferences between the interior and the surface of concrete structures. The largest stresses occur for largo, rapid temperature changes. Stresses are intially incorpo-rated into concrete from the heat generated by the setting of the wet cement.
For normal Portland cement structures not over a few feet in thickness, the heat generated by setting is dissipated rapidly enough that excessive tempera-ture dif ferences between the inner and outer portions of the structure do not occur. Therefore, it is not expected that large stresses will be incorporated into the CEVNC grouted drum, which is a relatively small structure and whose internal volume is largely occupied by waste, f rom the heat generated by the setting mortar. External temperature changes, which cause dif ferential tem-perature ef fsets through a concrete structure have been said to be important in some climates.t28) Dif ferential temperature ef fects were considered to be of concern in concrete paving staha.g28,45) Curbing stresses caused by tempera ture differences be tween the upper and lower surfaces of a concrete slab result in cracking when slabs are made too long between expansion joints.
The separation between expansion joints tolerable for reinforced coneg slabs was found to be 35-80 f t compared to 15-20 for plain concrete.
It may follow that the reinforced structure of the CEVNC grouted waste, which is primarily attributable to the concrete / steel banket layer, should help pro-teet the waste form from cracking and degradation due to temperature gradients f rom environmental temperatures changes.
47 a
Summary It is not expected that thermal effects would compromise the stabil-ity of the grouted drum waste.
In particular, the reinforcement of the grouted drum provided by the steel baskets, which contain the waste, should help stabilize the waste form against degradation due to thermal cycling.
However, thermal cycling testing of representative waste forms should be pe rfo rmed.
4.2.3.8 Free Liquid The TP states that waste specimens should have less than 0.5 percent by volume of the waste specimen as free liquid as measured using the method described in ANS 55.1.
Free liquid should have a pH between 4 and 11.
CEVNC has indicated that grouted drum packages are checked for free liquid prior to closure of the ring-bolt lid. The method used by CEVNC is believed to be a visual inspection such as that in Section 4.2 of ANS 55.1.
The check described in Section 4.3 of ANS $5.1 that involves breach of the container by drilling and then observation of the opening for flowing or dripping of free liquid from the breach has not been carried out. The test described in Section 4.4 of ANS $5.1 involving sectioning of the waste con-tainer contents has been completed by CEVNC for a simulated waste package and has indicated total penetration of the grout and no free liquid.
In general, CEVNC solidifies waste solutions prior to emplacement in the grouted drum pac ka ge. Thus, free 11gulds, should they exist in this package, would be ex-parted to result from excess water used in the grout mixture.
i Summary The CEVNC simulated grouted drum waste form has been found to contain no free liquide by visual inspection and observation following sectioning of the waste form (Sections 4.2 and 4.4 of ANS 55.1).
The test involving breach of the container by drilling and subsequent checking for draining of fluids (Section 4.3 of ANS SS.1) has not yet been performed.
4.2.3.9 Testing Sampin Size The TP indicates that if small, simulated laboratory size specimens
, are used for the testing recommended in the Technical Position, test data from sections or cores of the anticipated full-scale products should be obtained to correlate the characteristics of full-sise products with those of simulated laboratory size specimens.
4 As was discussed in the section on teach testing (Section 4.2.2.5),
the production of representative samples other than full-scale may be dif fi-cult for the case of CEVNC grouted drum packages. The presence of the various waste items in their dif ferent sizes, degrees, locations of cortamination (smearable, integral to the material, e.g., activated metals, etc.), and com-positions (glass, metal, paper) leads one to believe that the local behavior 48
r-4 of different areas in the form may strongly depend on the type of waste items present. A consequence of this observation is that it may become difficult to justify the representativeness of small-scale sample size waste form specimens.
Summary Testing of the CEVNC grouted drum vaste form must be carefully planned to include assurance that samples are indeed repterentative.
It is likely that full-scale sample testing may be required since sampling of a heterogeneous waste form is naturally subject to question with respect to representation of " typical" waste forms.
4.2.3.10 Pomogeneity of Compressive Strength The TP states that waste samples from full-scale specimens should be destructively analyzed to ensure. that the product produced is homogeneous to the extent that all regions in the product can be expected to have compressive s treng ths of at least 50 pai.
Testing of this type has not yet been performed by CEVNC. As was men-tioned in earlier sections (4.2.2.2, -5 and -6) the cement grout itself would be expected to have a uniform compressive strength of at least 50 psi but the grout / waste item combination may have properties that dif fer significantly f rom those of the grout alone. Class and metal waste components would prob-ably have sufficient compressive strengths, but the compressive strengths of plastic and cellulosic wastes in the concrete matrix cannot easily be predicted.
Summary The testing to determine whether the grouted drum waste form has suf-ficient homogeneous compressive strength should be performed by CEVNC. Aside from the difficulties of setting a representative test specimen size, there may be problems in determining the necessary scale on which the destructive testing recommended here should be performed (i.e., should waste items themselves be destructively analyzed?) Testing to show that sufficient aggregation and adhesion of the concrete grout occurs in the vicinity of the different types of waste items may make possible asserance that the waste form has the necessary uniform compressive strength (given that individual waste items such as paper and plastic do not compromise this property).
Section Summary The guidance given in the TP has been considered as it pertains to the CEVNC grouted drum waste package, and concerns and information gaps have been identified.
In most instances, concerns can be addressed through testing and this is recommended in several cases. CEVNC has very detailed waste handling procedures, but it is not clear that they have a process control pro-gram in which specific instructions, complete with limits of error and value 49
t l
s l
l ranges, e tc. are given.. Compressive I
T is radi a
a lit of th e we e fo e representative waste forje,
}
rad for w stes with the highest Class C activities needs to be tested at 10
{
of contaminating radioisotopes. gadiolytic production of gases is not ex-escape t e package. There is a con-ce ab t th ffe t of rbon e and organic acids which may result from radiolysis of the wastes. Biodegradation by-products may be able to affect the weste form stability but quantitative discussion of effects is not possi-ble since several information gaps esist. Self-sterilisation may occur in some of the higher activity packagest this could extend the time during which the package is not subject to biodegradation. giodegradation testing should be performed. Leach. testing should be performed on represonatative samples of these waste forms.
It is not clear that the heterogeneity of the GgVNC waste form activity will lead to increased cumulative radionuclide releases compared to those from a homogeneously distributed waste form. Thermal degradation of the vaste form is not anticipated but thermal cycling testing should be per-formed. Free liquid testing should be completed. The establishment of repre-sentative testing samples may require significant justification but this needs to be done before recommended testing can be performed. Homogeneity of com-pressive strength throughout the waste form should be tested as well.
i 1
l 50 '
o u
5.
TRU WASTE At present, CEVNC stores all TRU-contaminated waste with a level of TRU isotopes >10 nCi/g. According to 10 CFR Part 61, wastes with a level between 10 and 100 nCi/g could be disposed of as Class C low-level waste, provided that any other nuclides listed in Tables 1 and 2 of 10 CFR 61 which migh't be present were not in concentrations such that Class C limits were exceeded.
Those wastes containing such other radionuclides in excess of the Class C limits or having TRU isotopes present at levels >100 nC1/g, would be con-sidered "not generally acceptable" for shallow land burial, under Section 61.55 of 10 CFR Part 61.
Section 61.58 allows for authorization by NRC of other provisions for waste exceeding the Class C limits, providing the Commission finds reasonable assurance of compliance with the performance objectives given in Subpart C of 10 CFR Part 61 (Sections 61.40 through 61.44). Section 61.7 (b) (5), recog-nizing that there may be instances where waste with concentrations greater than permittted for Class C would be acceptable for near-surface disposal "with special processing or design," provides for evaluation of such waste on a case-by-case basis. It is under these sections, then, that the GEVNC waste containing >100 nCi/g of TRU isotopes would have to be considered by URC for near-surf ace disposal.
DOE has recently begun programs (45-49) to develop a concept for treat-ing wastes with radioactivity levels greater than the Class C limit. The con-cept is known as greater confinement disposal (CCD), and includes such alter-natives as improved waste form, deeper burial, and underground engineered ba rrie rs. These alternatives address the need expressed in Section 61.7 for special processing or design to enable waste to be considered by NRC on a case-by-case basis. So far as is known, however, there is as yet no move to provide CCD facilities at any of the commercial LLW burial sites, and no proposed rules regarding CCD have been issued by NRC.
This section of the report, then, reviews methods proposed in the li te ra-ture for treating TRU wastes, and discusses their applicability to GEVNC's stored waste.
In Section 5.1, alternatives relevant to near-surface disposal are considered.
Specifically, waste forms potentially suitable for TRU waste are discussed in Section 5.1.1, and decontamination methods useful for treat-ing TRU waste are described in Section 5.1.2.
Section 5.1.3 discusses the options for dealing with combustible organic waste. GCD options are con-sidered in Section 5.2.
In Section 5.3, possible alternatives for handling CEVNC's specific waste s treams are discussed.
5.1 Near-Surface Disposal For standard shallow land burial of wasta exceeding Class C limits, NRC may allow other provisions for the waste classification, but it is clear from Section 61.7 (b) (5) that the waste should be accepted for near-surface disposal only af ter special treatment - "with special processing or design."
Alternatively, wastes could be treated to reduce the TRU levels to either Class C or Class A, and be disposed of as standard LLV.
However in the 51
- ~_
-. = -
)
,a i
process, a secondary TRU waste. stream of potentially higher concentrations
- would be-produced and would have to be disposed of.
Possible alternative waste forms and special treatmants are discussed below in Section 5.1.1 and 2
5.1.2.
j -
It should be pointed out tha[ f.a container is relatively unimportant for TRU waste, since no container cat se expected to Last more than a small frac-tion of the hazardous lifetime of the long-lived TRU activity. This applies not only to regular containers, but also to HICs, both those presently li-censed and any that are likely to be.
Thus, HICs probably should not be con-sidered as an op tion for near-surf ace disposal of TRU waste in forms which are j
thought to be unsuitable for use.in regular containers, since they will prob-ably not provide adequate long-term containment of TRU activity.
5.1.1 Wjaste Form Considerations DOE work on waste form development for immobilisation of TRU waste was discontinued in 1981. Waste forms developed up to that time were evaluated for NRC in a BUL report.on alternative technologies for geologic disposal of TRU wastes.(461 The waste form evaluation included comparisons of ease of preparation, ability to incorporate reasonable loadings of a wide variety of-TRU wastes,. physical and chemical durability, radiation st-sbility, and leach-abili ty.
This last was treated as the single most important property, and the report assessed the different waste forms on the basis of their ability to pass. the 10-5/yr release rate criterion required of the waste in a high-i
~
{
1evel waste (HLW) geologic repository. This was a conservative approach, since it was the repository as a whole which had to meet that criterion. Thus if a waste form could meet it, whatever the other engineered barriers and the L
geology could add would be a bonus. The emphasis on waste form rather.than-engineered barriers and geology is particularly apt for near-surface disposal j.
since the container is not a factor in the case of TRU waste, there is very -
f little in the way of engineered barriers, and the geological pathway to the-l surface is short.
5.1.1.1 Forms Involving Conventional and Relatively Simple Processing Much of the earlier work on waste forms specifically applied to TRU waste was based on rather simple concepts. These generally involved mixing solid TRU waste, such as incinerator ash, dried sludge, scrap metal or used filters, in a steel drum with a liquid or slurry and letting the mixture set.
The principal binder or encapsulants were bitumen, urea formaldehyde resin,
~
and ordinary hydraulic cement.
I In the NUREG/CR-2333 evaluation,(50) waste forms prepared with o
l bitumen and urea formaldehyde were considered unacceptable for use in a geo-logic repository because of gas production from radiolysis and blodegradation.
The same would apply to vinyl ester-s tyrene polymer, which was not considered in NUREG/CR-2333.because it had not been used as a TRU waste form. Cas gen-eration may not be a problem for shallow land burial, and these organic forms may be considered capable of providing sufficient stability for Class C 52
,,,=---,w.
,v--e w.
[
T wa s tes. However, in view of the adverse ef fects of radiolysis and blodegrada-
- tion over the long term (i.e., periods >> 300 years), it is doubtful if any organic waste form could be relied on to provide the very long-term performance required for waste with concentrations of the long-lived TRU iso-topes exceeding Class C limits. 'Use of urea formaldehyde, in any case, has been discontinued as a waste form because of production of excessive amout ts i
of free standing liquid, expressly forbidden by Section 61.56 of 10 CFR i
Part 61.
In NUREG/CR-2333, ordinary cast concretes or hydraulic cements were considered to have excellent leachability with respect to Pu loss, but genera-tion of gas (particularly hydrogen) due to radiolysis of pore water was con-sidered a weakness. For this reason, it was recommended that further work on cement as a TRU waste form for repository emplacement be restricted to 'spe-cially prepared concretes with essentially no unbound water. Experimental work with some of these special concretes had shown acceptably low levels of gas generation in a number of tepts before the programs were discontinued.
In fact, gas generation from concrete appears not to be a serious problem for shallow land burial, at least under circumstances which can be readily envisaged.
In that case ordinary cas t concrete might be a suitable TRU waste form if TRU leach rates were deemed low enough to meet the perform-
&nce objectives of Section 61.41.
This would, of course, have to be demon-strated with the particular formulation of concrete to be used, but is is worth noting that one formulation containing actual TRU waste has demonstrated a release rate of <10-5 yr.(51) Aside from SYNROC, which was primarily a HLW form,. it was the only waste form reviewed in NWLEG/CR-2333 which was judged to have met that release rate criterion.
l 5.1.1.2 More Advanced Ceramic and Mineral Phase Waste Forms i
Several waste forms were reviewed in NUREG/CR-2333, which required i
(
considerably more advanced technology than simply mixing waste and binder at i
. or near ambient temperature. These included iron-enriched synthetic basalt l
(at one time the reference form for immobilization of DOE's stored TRU waste at INEL), borosilicate glasses, specially prepared concretes and cementitious forms, synthetic monazite, and SYNROC. This last form had been developed for l
use with HLW, but testing (including leach testing) had been done with samples containing TRU isotopes in concentrations comparable to those found in actual l
TRU waste.
Borosilicate glass was, of course, being considered for the refer-I ence HLW form, however certain formula tions were developed specifically for application to TRU waste streams.
It was concluded in NUREG/CR-2333 that the more advanced forms gen-(.
erally had not unde rgone sufficient testing, particularly leach tes ting, to demonstrate suitability as forms for immobilizing TRU waste in geologic i
r eposi tories. However, certain SYNROC formulations apparently me t the release rate and other criteria applied in NUREG/CR-2333, even though it had been -
developed as a high level waste form.
f 53 l-f
~.-
l 4 -
While the more advanced forms developed specifically for TRU waste were judged as not completely ~ demonstrated for geologic disposal. it is -possi-ble they could be. considered suitable for shallow land burial of reasonable l
amounts of TRU waste. Certainly SYNROC is in that category since it appears to _ be accep table for geclogic disposal. They are all generally quite stable with respect to thermal degradation, biodegradation and to expected amounts of radiation, and can assuredly be described as having received "special process-ing or design," as required by Section 61.7 (b) (5) for. case-by-case f
evaluation by NRC.
In fact, the fault of mos t of them umy be that they re-quire too much "special processing" in their preparation, and might, there-fore, be considered too expensive.
5.1.2 Decontamination Processes One possible alternative has been mentioned (Section 5.1) of treating a TRU waste stream exceeding Class C limits by reducing or removing surface con-tamination so that the waste can be treated as Class A, or at least Class C.
At the same time, the removed TRU contaminant gives rise to a secondary waste s tream. Although this requires disposal, it will be in a much more compact form than the original and one which should be easier to treat.
Decontamination processes are generally designed for use with metals.
Some can be applied to ceramics, and even to rubber and plastics, but such applications are limited. Methods developed up to 1981 were reviewed in NUREG/CR-2333. Since that time, the only DOE-sponsored work connected with TRU isotope contaminants has been carried out under the Civilian Nuclear Waste Treatment Program. Relevant processes from these sources are described in the following sections.
4 i
5.1.2.1 Electrolytic Methods These are applicable only to contaminated metals. The general method of operation in electrolytic decontamination is to remove the contamination along with a surface layer of metal. This layer is removed by passing an electric current through a suitable electrolyte in a cell in which the piece to be decontaminated is the anode. Depending on the choice of electrolyte and the current-voltage conditions, the me tal can be rela tively uniformly dis-solved and the surface lef t in either a highly polished state or in a sonewha't l
usually H PO, alone or in combina-roughened condition. S trong acids, 3 4 tion, are used to achieve a highly polished surface, and the process is called electropolishing. In terms of contamination removal, it can be considered as one type of electrodecontamination.- Another type, producing a roughened sur-f ace, uses mildly basic strong salt solution as the electrolyte.
A great deal of research and development work has been done on these electrolytic methods, both in the U. S. and in other countries, and the tech-3 nology is in a well-developed s tate. Decontamination to well below the 10 nCi/g level is routinely achievable, so that the decontaminated metals is Class A low-level waste. However, a t the time of the NUREG/CR-2333 review, satisfactory treatment of the secondary TRU waste stream from electropolishing 54 I
-.-m-
had not beenidemonstrated. f TRU isotopes in basic electrolyte precipitate as hydroxides along with the dissolved meta?. being decontaminated, and can be easily filtered to provide a small volume of secondary TRU. waste. Since there has-been no DOE ef fort in electrolytic methods since that time, the basic electrolyte must be considered the more attractive from the strictly waste management point of view.
An electrolytic method developed in the U.K.,
which does not involve l
removal of metal from the contaminated surf ace was described in NUREG/CR-2333.
This was called electro-cleaning, and involved removal of surface contamina-tion by microbubbles of electrolytically-generated gas, with no metal dissolu-tion. To the extent this method was applicable to a given waste stream, it would be preferred over use of either of the dissolution processes, since it 3
l would give an even smaller secondary TRU waste stream. The current status of development of the electro-cleaning me thod is not known.
5.1.2.2 Vibratory Finishing Vibratory finishing is an industrial process used for surf ace finish-4 ing of both small and large metal pieces. Its firs t waste management applica-tion was to' pretreat metal surfaces for subsequent electropolishing.
It was shown to have general applicability, removing scale, rus t, grease, paint, and organic films of' all kinds. In the process it was found to remove most of the l
TRU contamination, and was therefore developed as a decontamination method -in i
its own right. Early work used ceramic cutting media, but this contributed I
relatively large proportions of both ceramic and metal particles to the secon-dary ~was te s tream. By the time of the review carried out in NUREG/CR-2333, metal media were used almost exclusively, particularly case-hardened carbon steel and s tainless steel ball cones, which have the subs tantial advantage of producing no secondary waste due to media wear.
i The process is carried out in a liquid, commonly 10% NaOH solution, l.
from which the secondary waste can be easily filtered. The method is appli-cable 'to rubber and plastic as well as metal (unlike electrolytic methods),
and experience,has been that the presence of rubber and plastic pieces facili-tates ' processing of metal pieces. Decontamination to <10 nCi/g has usually been obtained. Recently in the Civilian Nuclear Waste Treatment Program, the vibratory finishing process has been applied to decontamination of Zircaloy cladding hulls with similar good results.(52) l 5.1.2.3 Chemical Decontamination Removal of radioactive contamination from solid surf aces by treating
~ ith various chemical solutions is standard practice at U.S. and foreign nu-w z
clear ins tallations, particularly as applied to decommissioning operations and cleaning contaminated equipment. Some procedures have been developed for use with TRU-contaminated equipment, but the principles are the same regardless of i
the nature of the radioactive contaminant.
In general, the aim is to remove only the oxide layer from a contaminated me tal surface, with the expectation that any contamination would have been held in the oxide layer and would be
- removed along with it.
55 i
1 In some ins tances, contamination might penetrate deeper than the oxide layer and a layer of metal might have to be removed in order to remove the contamination. Electrolytic methods could remove a surface layer of 2
metal, but removal is not uniform with irregular shaped pieces.' A method developed by HEDL(53) for treating Pu-contaminated stainless steel used strong nitric acid solutions containing 0.1 M Ce(IV) at somewhat elevated temperatures (approxima tely 900C).
While the process was very effective in removing Pu contamination, it employed rather aggressive conditions and pro-duced a secondary waste stream requiring considerable treatment prior to disposal.
- i. '
The milder treatments which do not remove metal have the advantage that ~ they are simple to use, since application of the cleaning solutions and l
collection of the contaminants do not require highly specialized equipment.
i The method can be applied to " washing" of large structures and pieces of equip-ment, or to immersion of smaller pieces. Good results with cleaning Pu' glove boxes and cell liners af ter 12 years of service were demons trated at SRL(54), and their process could presumably be adapted to small metal pieces. Chemical treatment consisted of washing with alkaline permanganate and oxalic acid solutions, with intervening water flushes, all at ambient tem-p era ture. Two cycles reduced the contamination level to <10 nCi/g.
i It can be concluded that relatively simple treatment may be capable i
of reducing TRU contamination levels on metal surfaces to the point where the me tal is Class A waste. The SRL procedures, for example, achieved this with' the use of relatively small amounts of innocuous chemicals. Use of various complexing and chelating agents which are included in standard procedures for certain other types of contamination and contaminated equipment is not recom-mended; in fact, their use could presumably only be permitted if the secondary l
waste were treated to ensure their complete destruction before disposal.
l l:
5.1.3 Cellulosics and Other Combustible Waste L
Properties of TRU waste in temporary storage at INEL and still being l
generated by DOE were reviewed in NUREG/CR-2333 with particular attention to leachability and gas generation. Leach data for this existing stored waste l
was minimal and covered a range of many orders of magnitude. More work had l
been planned by DOE, but was not done. In any case, it would be difficult, if L
not impossible, to obtain da ta quantitatively rela ting leachability to some standard waste due to the very wide range of materials, especially organics, making up the waste. On the basis of the available info rma tion, it was con-cluded that "no credit can be given to as-generated TRU waste as a major bar-rier in the controlled release of actinides.a(50) Thus, while containnent i
of activity from this type of waste in a geologic repository might be possible I
because of the major barrier imposed by the geology, it could not be expected l
in a shallow land burial situation, where the pathway to the surface is short
[
and the possibility exists for relatively rapid migration due to chela tes.
l Production of gas from radiolysis of pore water in concrete has been mentioned (Section 5.1.1.1).
In general, gas generation from contaminated
)
l 56 y
---n---
,n.---
-.~.-n
-. - -.. - - - ~ - - - - - + - - - - - - - - - -
4 l
cellulosics and other organics cons titutes more of a problem, for a given TRU 4
activity level, since they are subjected to production of gas from biodegrada-tion as well as from radiolysis. Gas production from radiolysis is in direct proportion to the amount of radioactivity present, whereas that from biodegra-dation 'is not connected with the level of contamination. Thus biodegradation can be, and has indeed of ten been found to be, the cause of the greater pro-duction of gas. This will, of course, depend on the exact nature of the or-ganic waste, and the conditions (temperature, moisture, etc.) to which the was te. is subjected. The'particular mixture of gases, as well as the amounts l.
formed, will also vary considerably in biodegradation.
t Because ;of the problems associated with contaminated organic TRU-contaminated waste as a waste form, particlularly those connected with leach-ability and gas generation, it was recommended in NUREG/CR-2333 that.combus-tible TRU waste not be accepted in geologic repositories. Such waste is not excluded from shallow land burial as Class A Waste, provided it is not capable
.of generating quantities of toxic gases harmful to persons trans po rting,
handling, or disposing of it.
F6r any combustible TRU-contaminated waste I
(other than Class A), it would undoubtedly be advisable to process the waste (by incineration or acid digestion) to an inorganic form, as recommended in
. NGLEG/CR-2333 for geologic disposal. For actual TRU waste (TRU content l
.>100 nCi/g) avoidance of combustible organics would be even more important due j
to the.more s tringent long-term s tability requirements for waste exceeding i
Class C limits. As discussed in Section 5.1, use of HICs for disposal of any i
TRU waste, including combustibles, is considered unacceptable.
5.2 Greater Confinement Disposal (GCD)
L
[
This ' term refers to disposal in such a manner that confinement of the disposed radionuclides will be greater than that provided by standard shallow l
Land burial. It is meant to be applied to non-high-level wastes considered l
unaccep table for shallow land burial, i.e., those that contain such high con-1.
centrations of radionuclides and/or quantities of long-lived radionuclides
~ th*L standard shallow land burial would result in a dose to a member of the general public exceeding the performance objectives given in 10 CFR Section 61.41.(47} Emplacement.in a deep' geologic repository is ruled out by its high cos t.
Thus GCD has been proposed as a safe alternative to shallow land burial, and an economic alternative to deep geologic disposal.(46) l DOE for several years has been involved in work with GCD concepts, and L
criteria were published in 1981.(45) More recently, a generic study has been completed (47) which analyzes the costs and risks of a number of GCD alternatives, comparing them with thosu for standard shallow land burial and.
HLW burial in a geologic repository. The results are summarized in Table 5.1.
j
.They indicate that risks for all the GCD options analyzed are likely to be several orders of magnitude less than those for regular shallow land burial, while cos ts should be within a factor of 2 or 3 greater.
i 1
57 i
7-
Table 5.1 L
- Generic Cos t Plus Risk Comparison Alternatives" (Taken From Table,7-of Reference 47) 6 6
6 Facility Type Cost ($10 )
Health Risk ($10 )b Total ($10 )c SLB ' reference facility -
8 5000 5000 Deep - trench 14 1
1.5 l
' Improved waste form 27 0.06 27 Engineered a trueture:
concrete-walled trench 22 0.009 22 4
intruder barrier 15 1
16.
- Augered shaft:
. southeast region 19 0.4
-19 southwest region 19 0.004 19
- Hydrofractured 24
<0.01 24 HLW repository
->>50
<0.01
>>50 abased on disposal of 10,000 m3 of warm GCD waste and 62.5 m3 of hot CCD waste at a facility co-located with a SLB f acility for LLW that meets 10 CFR Part 61 or equivalent criteria for near-surface disposal.
bAll health risk estimates have been rounded to one significant digit.:
cThe ' total is the ranking parameter which is the sum ' of cost plus health risk.
Cos ts and ranking parameter values have been rounded to two significant digits.
j.
dFor liquid waste.
It is emphasized that this study was generic, and' that the particular al-te rna tive to be. chosen for a given site would have ~ to be based on a site -
specific analysis. One factor which is of great importance is the depth of -
the water. table,(47) and alternatives requiring greater depths will not.
(
necessarily be better at any given site.
A point about the ' study which should
(
also be kept in mind is that-the tuo reference waste streams whose compasi-tions were used for analysis of nost of the options (all except hydrofracture).
L contained insignificant amounts of a-activity (<1 nCi/g). There is~thus some doubt as to the relevance of the analyses of these options for TRU wastes con--
taining rela titely high TRU concentrations. The third reference waste stream had a TRU concentration well above the Class C limit. Its composition was used for analysis of the hydrofract' re op tion, so that anlaysis, a t least, u
should be relevant for TRU waste.
I'
. A study has been reported by Pacific Northwest Laboratories (PNL)(55)
L on possible."TRU advanced disposal systems" for burial at Hanford of some of -
the TRU waste for which there will not be room in the Waste Isolation Pilot
(~
u 58 H
l-h
't s
w my,
f y
s.
Plant (WIPP).
It identifies several techniques, including. grouting and in situ. vitrification, which were censidered to provide " greater confinement" against intrusion than that provided by shallow land burial. An example systems analysis was performed with assumed performance. objectives and Hanford-specific disposal sys tems,. waste forms, site characteristics, and engineered barriers. Preliminary waste disposal criteria for Pu-239 were l
determined by applying the Allowable Residual Contamination Level (ARCL) method. The dependence of. exposure on depth for Pu-239-contaminated soil
- derived from this' analysis was such that allowable soil concentrations of
~
Pu-239 were 0.5 nci/g be tween the surface and a depth' of 1 m, 2200 nCi/g at a depth of 5 m, and.10,000 nCi at a depth of 10 m.
The two general options for GCD are exemplified by the 2nd and 3rd alter-natives :in Table 5.1, namely deeper burial and im;;oved waste form. The other alternatives in that table involve.special combinations of deeper burial with
. engineered barriers. Field work has been commenced by DOE on several GCD Laboratory (SRL) (humid).(49)concep ts at-two sites, one in Nevada (arid),(48) an Savannah River Work at the Nevada Test Site
. signed to demons trate the use of large diame ter bore holes (3 m diame ter, 37 m depth), and is in the nature of a relatively long-term tes t, with instruments in.the shaf t and in monitoring holes to monitor migration of tracers and i
radionuclides. At SRL, the emphasis is more on actual disposal of waste being i
genera ted. (49) It has been found that 95% of the activity of this waste is contained in 5% of the volume, and this high-activity fraction will be dis-posed of by GCD in boreholes and concrete-lined trenches.
l 5.3 Specific Applications to GEVNC TRU-Contaminated Waste f
.Three different waste streams containing TRU isotopes were identified and l
. described in Section 3.3.
Each is characterized by a particular form and a l
range of TRU isotope concentrations. Possible methods of dealing with each type of waste to allow for its disposal are discussed'below. This discussion does not constitute a set of recommendations to GEVNC for disposal of their waste, but describes some possible options.
5.3.1 Weste Containing 80-100 nci/g of TRU Isotopes -
t This waste stream accounts for about 2/3 of the total. waste volume and t
(
- an insignificant fraction of the total activi ty in the current inventory. (see L
Table 3.4).-
All the cement-solidified dilute burnup analysis. solutions be-l~
long in this. category, and essentially no other waste. As discussed in Sec-i tion 3.3.3, it is likely that a good deal of this waste exceeds the. Class C limit significantly, bu t not to a great extent, because of its Pu-241 content.
Specifically, applying the sum of fractions rule (10 CFR Section 61.55) would probably yield a value significantly >1, but <l.8, for a good deal of the l-waste. That part of it would thus be officially TRU waste, rather than simply
.TRU-contaminated LLW, but it is close enough to Class C LLW that it could be expected to 'be considered by NRC as a special case' for near-surface disposal l
(see Section 5).
l r-l 59 l
6 e
At least some formula tions of ordinary hydraulic cement (51) have been shown to have excellent Pu leach resistance, as discussed in Section 5.1.1.1.
If the GEVNC formulation can be shown to behave similarly, there is reason to expect that NRC would consider the form acceptable for near-surface disposal of this low concentration TRU waste. An even stronger case could.be made if
. the small waste form, preferably without paint cans, were placed in a larger container, such as a 30-gal or 55-gal drum, and made into a solid monolithic form by encapsulation in concrete. Documentatation of the waste form's good leach. performance would presumably be required.
In addition, in order that this waste be accepted as Class C stabilized waste, data would have to be pre-seated by GE to show that the properties of the waste form were consistent
.with -10 CFR Part 61 and the TP for Class C s tabilized waste.
d As far is known, no work is being done at commerical LLW burial sites on development of GCD methods and no facility for GCD is in the active planning stage at these sites (or elsewhere). This situation will presumably continue until NRC indicates a need for such a facility, and provides guide-
. lines and criteria for its construction and operation. Some non-DOE waste i
(e.g., the GEVNC waste under discussion) exceeds Class C limits, and more will undoubtedly be produced. It is probable that some fraction of it will be considered unsuitable for regular shallow land burial but acceptable for GCD.
The volume of this fraction may not be large enough to warrant building a 1
special GCD site, but one or more facilities at existing LLW burial sites may be required. This need, in term, implies a need for criteria for GCD disposal.
i j
5.3.2 Solidified Hot Liquid Waste Stream l
Inventory informa tion for this waste stream is given in Table 3.4.
The waste represents 67. of the total volume and about 2/3 of the TRU activity in the current inventory. The only waste in this waste stream is that obtained by cement solidification of the hot liquid waste arising from dissolution of l
samples of irradiated fuel used for burnup analysis. As discussed in Section 3.3.2, the TRU content is estimated to be in the range of 0.2 to 3 mci /g, with mos t of it in the range 1 to 2 mci /g (1 to 2 x 106 nCi/g). Plu tonium-241 concentrations are probably some 30 times these levels, so both TRU and Pu-241 activities are of the order of 104 times the respective Class C limits.
Also Cs-137 and Sr-90 concentrations are 1 to 2 orders of magnitude higher than the Class C limits.
It seems clear that this waste, while not equivalent to HLW, is so much more radioactive than the hottest low-level waste generally acceptable by NRC for land disposal that it is not expected to be considered -
i acceptable for disposal at a commercial LLW site, even as a special case.
I j
One possible option for dealing with the waste would be to arrange for
(
its transfer to DOE.
DOE has several possible. alternatives for handling it, l
which are not available to non-DOE generators. These include emplacement in the Waste Isolation Pilot Plant (WIPP) and s torage at a facility such as that at INEL until a suitable CCD facility operated by DOE became available.
60 i
I
..-,.,,,..---__._x
_.. - -. -,, - - -,. - - ~ ~,.., - - -.,
(
The present inventory of 9 gallons could presumably be repackaged, with-out paint cans, in one or two 55-gallon drums lined with 6 in or more of ce-ment, which could then be filled with cement. Dilution of the activi ty to this extent would probably be acceptable, but would still leave it about 103 times the Class C limit. However, the procedure would provide a significant barrier of nonradioactive cement for the waste, and might be acceptable for one of the GCD op tions.
It will, of course, be some years before any suitable GCD option is likely to be available at a commercial burial site and criteria are not yet available to judge the adequacy of this was te form.
The current inventory of this waste stream is already solidified, but for hot liquid waste produced in the future the option exists of treatment to prepare one of the special waste forms discussed in Section 5.1.1.2.
A s ui t-able SYNROC formulation or iron-enriched synthetic basalt should provide good long-term stability, and SYNROC in particular has been shown to have very low leach rates for Pu.
These qualities address the concern for "special process-ing or design" expressed in 10 CFR Part 61.
Thus such waste forms might qual-ify for regular land burial at somewhat greater depth than normally used at commercial LLW burial sites. At least a good case could be made for their disposal with some form of GCD.
However, the applicable criteria do t.ot exist at present to de termine the accep tability of processed wastes' for GCD.
From the preceding discussion, it can be concluded that no acceptable method of disposing of waste containing very high levels of TRU activity appears to exist in the priva te sector. When the concepts for a high level waste repository were described in the original version of 10 CFR Part 60(55), it was considered that TRU waste would be emplaced in a licensed HLW 60(gository, along with the HLW.2) deals only with HLW, it does not explicitly rule out emplaceme res Although the present version of 10 CFR Part TRU waste in an ALW repository.
In fact, waste with levels of TRU activity such as those in the GEVNC waste may still be required to be emplaced in a geologic repository. However, if not, it seems clear that if such waste is to be disposed of at commercial sites, development of suitable GCD alternatives at one or more of these sites will be necessary.
5.3.3 Solid Waste The principal components of this waste stream are described in Section 3.3.
In terms of activity it represents about 1/3 of the to tal in-ventory while contributing about 1/4 of the volume. The bulk of the activity is concentrated in the sections of fuel pellets used for me tallographic exami-nation, and in the cuttings formed when they are removed from the fuel ele-ments by a diamond cutting wheel. The cuttings are cleaned up with paper wipes. The remainder of the activity is associated with pieces of Zircaloy c lad ding, pieces of discarded equipment, including the diamond-encrusted brass cutting aheels, used glassware, and the small amount of solidified sludge resulting f rom polishing operations.
61
m e
s The solid waste varies widely in.its TRU content, but on average is a factor of only 2.5 lower than that of the solidified hot liquid waste (see Table 3.4).
.Thus much of it will be at a high enough level that the same
~
situation exists for it as for the solidified hot. liquid waste discussed in Section 5.3.2, i.e., there appears to be -no way of disposing of it at a com-
'mercial site.
In its present form (loose scrap in paint cans), it would be even less acceptable than the. solidified hot wastes. However, since it ijt loose, it should be possible to sort it into its comp # ents and to devise methods of treating it to obtain some acceptable waste forms for shallow land burial. Some of these possible methods as they apply to the mixed waste and to different components of the solid waste stream are discussed below. - On the basis of the information on this waste stream at our disposal, there :seems to be no reason why the methods could not be applied to the present inventory.
They could obviously be applied to future waste generated, and it might be advantageous to do so.
At least the waste could be segregated into its sepa-rate components as it was generated, in' case the option to treat components separately was chosen at a la ter time.
5.3.3.1 Encapsulated Mixed Solid Waste The solid waste could be encapsulated in cement as is (without the paint cans, or at least without the lids)'in cement-lined 55-gallon drums as described in Section 5.3.2 for blocks of solidified hot liquid waste.
Al-though there is a larger volume of solid waste, the dilution factor upon en-cagsulation would be similar, so the TRU concentration would s till be some 10 to 103 times higher than the Class C limit. As is the case with the solidified hot liquid waste encapusluated in this way, it might be accep table for one of the GCD options.
5.3.3.2 Fuel Pellet Sections This component of the solid waste stream' contains the largest frac-tion of the activity and would be easy to separate from the rest of the waste.
An apparently simple method of treating the polished sections would be to re-move them from their plastic (presumably Lucite) mounts and dissolve them as the-larger samples for burnup analysis are dissolved. The resulting acidic solution could be neutralized and cement solidified in the same way as the burnup analysis solutions have been handled. Alternatively, this solution could be treated to convert the radioactive isotopes into a chemical form from which one of the special waste forms discussed in Section 5.1.1.2 could be prepared. The discussion of special waste forms included in Section 5.3.2 applies here also.
5.3.3.3 Paper Wipes with Fuel Cuttings The second largest fraction of activity in the solid waste is that arising from cutting the fuel pellets. The small particles of fuel end up mostly on paper wipes, which could easily be separated or kept separate, from the rest of the waste. As discussed in Section 5.1.3, it is advisable to ex-clude organic combustible waste from Class C LLW.
When encapsulated in a 62
solid waste form, it would almost certainly detract from the form's long-term stability due to its decomposition by radiolysis and biodegradation. There is even more reason to exclude it from waste forms containing highly active TRU waste, where long-term stability is crucial.
In any case, organics can easily be destroyed by such. methods as incineration and acid digestion, so it,seems reasonable to require this treat-ment - for the paper wipee. Considering their small volume, incineration might be -impractical, but ashing in a crucible could presumably be subs tituted,' and the ash could be incorporated in cement or in one of the special waste forms discussed in Section 5.1.1.2.
Acid digestion would also be easy to apply on a small scale, particularly to cellelosics, but would leave the TRU isotopes dissolved. in sulfuric acid solution. Treatment to prepare special waste forms would still be feasible, and preparation of ordinary hydraulic cement would be s traigh tfo rward. In fact, for the la tter form, the acid solutions from this stream, from the dissolving of the fuel pellet sections and from the hot liquid waste stream could all be combined if desired.
5.3.3.4 Miecellaneous Solids Surface-contaminated solids (other than wipe papers) account for mos t of the volume in the solid waste stream but only a small fraction of the 13tU activi ty, pe rhap s 10%.
It is possible that a great deal of this waste could be decontaminated to levels where it would be Class A or Class C LLW, and thus could be sent to shallow land burial. In the process, a secondary TRU waste stream would be produced, but with TRU concentrations much less than those in.
the solutions resulting from processing burnup analysis samples, metallo-graphic examination sections, or wipe papers. This secondary waste stream, when converted to a solid waste form, would s till have TRU levels much lower than waste forms prepared from the hot liquid wastes, quite possibly in a range where the waste could be considered on a case-by-case basis by MRC for near-surface disposal.
Specific components of the GEVNC solid waste and possible treatments are listed below and presented in tabular form in Table 5.2.
Although results obtained with decontamination procedures will vary with the exact nature of the waste, the methods considered have all been shown to be capable of re-ducing TRU levels to <10 nCi/g on actual (rather than simulated) waste. To be applicable to the GEVNC waste, they would have to be capable of operating on a small scale, since the amounts of waste involved are relatively quite small.
1.
Contaminated Equipment Made of S teel a.
This could be decontaminated electrolytically by one of the methods described in Section 5.1.2.1.
Electrolytic me thods can readily be applied on a small scale. Use of basic elec-trolyte would be preferable from the point of view of dis-posing of the secondary waste.
b.
Chemical decontamination methods, such as those discussed in Section 5.1.2.3, can also be applied on a small scale.
63 l
l
Vibratory finishing (Section 5.1.2.2) works well with metals c.
and also plastics, but might not be feasible for the CEVNC waste, depending on the shapes and sizes of the waste to be decontaminated.and its amount. Adapting to the small scale required might be impractical.
2.
Zirealoy Cladding a.
Chemical decontamination is applicable on the small scale called for here. A simple treatment with HNO3 in a beaker might be suitable to dissolve any fuel s till adhering to the cladding.
b.
The volume of waste is so small that vibratory finishing would almost certainly be impractical for this waste by itself.
3.
Diamond Cutting Wheels a.
These are made of brass and could probably be decontaminated readily by simple treatment with HNO.
3 b.
Since the number of wheels entering the waste stream is ex-pected to be small, the remarks under 2b apply here also.
4.
Plastic Meta 11ographic Section Holders It is quite likely that chemical decontamination would work a.
for this waste. If it were not satisfactory, the plastic could be ashed, or decomposed by acid digestion. All of these treatments are suitable for small-scale work.
b.
The plastic pieces could also be treated by vibratory finish-ing, but again the small volume of waste might make this method unsuitable.
5.
Contaminated classware a.
The only obvious and simple way of decontaminating glass waste is probably by chemical trea tment, which can readily be applied to whatever volumes are generated.
b.
An alternative me thod of handling gla ss was te might be to pulverize it and incorporate the powder in cement or another suitable waste fo rm.
6.
Overall Waste S tream a.
Although chemical decontamination is applicable to all known components of the solid waste, several different agents would 64
almos t certainly be required for ef ficient treatment of the dif ferent materials. This ef fectively necessitates that each component be handled separately. As pointed cut in Section 5.1.2.3, chelating agents should not be used unless they are completely renaved from the secondary waste stream.
b.
All the components except glass can be decontaminated by vibratory finishing. The method would probably be imprac-tical for any one component, but if they were combined, it is possible the waste volume might be large enough to make the me thod f easible.
Table 5.2 Possible Alternative for Treatment of Sorted GEVNC Solid Waste (Applicable technology indicated by x)
Trea tment Method Decontamination Method S tanda rd Acid Vib ra to ry Component Dis so lu tion Digestion Ashing Electrolytic Chemical Finishing
- Metallographic x
Sectione Pape r Wipes x
x S teel a
x x
Equipment Zircatoy x
x C la d di r.g Cu t ting x
a Wheels Plastic Ha1Jera x
x s
a Classware x
- 1sthod works well for types of vaste indicated. Might be impractical for small amounte of any one type produced, but might be feasible if all combined.
5.3.4 Summary of Options The GEVNC solidified dilute burnup analysis solutions can prob-bly be handled at a commercial LLU burial site. The remainder of the waste, both current inventory and that produced in the f 2ture will require some other trea tmen t.
Most of its radicactivity will probably not be able to be handled by shallow land burial, even on a special case-by-case basis. On the other hand, most of its volume could be decontaminat.ed and disposed of as LLU at a commerical disposal site.
65
L t.
o Thei solid waste in both the current inventory and that resulting from future operation could be. sorted into components. Sections of fuel and-wipes of fuel _ cuttings contain the bulk of the activity in a small. volume, and could be' treated as discussed in Sections
- 5.3.3.2 and 5.3.3.3.
Possible methods for treating the contaminated
. solids, which account for. only about 5% or less of GEVNC's, total TRU activity, have been described in Section 5.3.3.'4.
The object of all these methods is to convert the solid waste volume to Class A waste,
~
with_ production of a secondary waste stream which _can readily be in-corporated into a solid waste form exhibiting long-term stability and ilow TRU leachabili ty.
The levels of TRU activity in such secon-dary waste might be low enough that it could also' te considered for.
near-surf ace disposal on a special case basis.
It was pointed out in Section 5.3.3.3 that all the hot waste solu-e tions resulting from dissolving fuel sections (for burnop analysis
- and metallograhic examination), and from acid digestion of paper wipes,_could be combined..This would allow treatment of 95% or more of GEVNC's total TRU activity in one waste stream. The TRU concen-trations of a waste form prepared from such solutions would be sev-eral orders of magnitude higher than the Class C limit, but no greater than those in the current solidified hot if quid waste.
As discussed in Section 5.3.2, there appears presently to be no ac-e ceptable method in the priva te sector for disposing of wastes con-taining such high levels of TRU activity. One possible option would be to arrange to transfer them to DOE.
If this op tion is not avail-
'able, they would either have to be emplaced in a future licensed geologic repository, or in a suitable GCD facility at a LLW burial site once such a facility has been developed. Both of these alter-natives appear to be some distance in the future.
The potential need for GCD facilities for non-DOE wate was pointed e
out in Section 5.2.
Guidelines and criteria would have to be deve-loped by MRC for construction and operation of such facilities.
Criteria would also have to be developed for the types of waste and waste forms to be handled by GCD.
66
6.
CONCLUSIONS The waste generated by GEVNC during the examination of reactor components and fuel and in the production of sources and radiopharmaceuticals has been characterized.
It was found that GEVNC produces wastes which can be classi-fled as Class A, Class B, and Class C according to the waste classification sys tem of 10 CFR Part 61.
GEVNC also has shipped wastes with radionuclide concentration in excess of the Class C limit.
A review of RSRs for waste shipped by GEVNC to Hanford indicates that GEVNC shipped its wastes in three main types of packages. These are wooden crates, carbon steel drums (55 gallon) and 84-gallon extended 17H drums grou ted with cement. Class A wastes have been packaged in each of the con-tainer types.
Based on the review of the RSRs, Class B and greater wastes are packaged in concrete-grouted drums. A total of five isotopes were spe-cifically identified in the RSRs reviewed although it is known that much of the was te is actually composed of mixed fission products. GEVNC has indicated that its wastes are dominated by Cs-137 and Co-60.
Class A, B, or C wastes may contain both of these isotopes, or occasionally only Co-60.
GEVNC de termines the concentrations of radionuclides in its waste from dose rate measurements, the distribution of isotopes in smears and a computer code which converts the dose rate measurements to activi ties. This procedure was reviewed.
It was recommended that GEVNC perform a more detailed charac-teriza tion of the radionuclide inventory in its grouted drum waste as the waste classification could be af fected by this inventory.
Each of the waste packages used by GEVNC was evaluated with respect to the appropriate sections of 10 CFR Part 61 on waste characteristics, as well as the Technical Position on Waste Form.
In addition, the components of the GEVNC was te packages were reviewed to de termine if ma terials were present which were hazardous or which could compromise burial site performance. The results of these evalua tions, as well as concerns and needs for additional information, are summarized below.
6.1 Class A Was te Packages The largest volume (~707.) of waste shipped by GEVNC is Class A waste.
These wastes appear to meet the minimum requirements for Class A was tes.
6.2 Class B and Class C Waste Packages CEVNC currently ships its Class B and C wastes in cement-grouted 84-gallon extended 17H drums. The waste itself consists of miscellaneous components from hot cell operations. These include me tal, glass, paper and plastics. These wastes are placed in perforated carbon steel baskets which are then put in the outer drum and the entire baske t/ waste array is grouted with cement. The outer carbon steel drum is not expected to provide stability and be recognizible for 300 years in the trench environment. The inner concrete / perforated basket / waste monolith, however, may have these character-istics and it is this monolithic waste form that has been evaluated with 67
i I
res pect to the guidelines given in the TP for waste forms.
In addition, this waste package was evaluated to see if it met the minimum requirements for was tes given in 10 CFR Part 61.
6.2.1 Minimum Requirements The Class B and C waste packages evaluated appear, in general, to meet the minimum requirements for waste given in 10 CFR Part 61 Section 56.
6.2.2 Stability Requirements Spalling and cracking of the concrete / basket / waste monolith in the GEVNC high activity waste package as a consequence of the pressures generated in corrosion of embedded steel was considered for the GEVNC package as gen-ersted and in its trench environment. This spa 11fng/ cracking process is a potential failure mode that is not directly addressed in the Technical Posi-tion. Such spalling and cracking could lead to loss of monolithic form of the waste. It was determined that insufficient chloride was present in the wastes, grout, or trench environment to allow initiation of the metal corrosion proc-ess.
If other agents capable of initiating the corrosion of the baskets were present, the probability of these causing spalling is quite low due to the scarcity of water at the Hanford site. If these packages were disposed of in a humid site, an assessment of the expected corrosion rates in the presence of these agents would be necessary before conclusions about lifetime stability could be drawn.
6.2.3 Guidelines in the Technical Position The GEVNC grouted drum concrete / basket / waste monolith has been evalu-ated with respect to the guidance on processed waste provided in the TP.
A major factor that enters into this evaluation is that very little actual test-ing in accordance with the TP has yet been performed by GEVNC and, the deter-mination of proper representative test sample sizes is not straightforward be-cause CEVNC has heterogeneous waste.
The TP guidelines have been considered individually as they apply to the GEVNC high activity Class B and C waste form and the evaluation summaries are given as follows:
GEVNC high activity wastes are packaged according to a well-defined s
procedure which cannot strictly be called a process control program since it lacks specific information on weights / volumes. It does, however, provide tolerances for the grout mixtures. It has not been demonstrated that this procedure will ensure compliance with the recommendations given in the TP.
The variability in the waste s tream is considered a limiting factor in production cf consistently uniform processed wa rte packages.
The compressive strengths of the concrete grout and of the waste e
metal and glass are all expected, individually, to exceed 50 psi.
68
The compressive strength of compacted paper and plastics present in the wastes is not known. However, the compressive s trength of a monolith produced by combination of these materials cannot be simply determined by means other than actual testing.
e The radiation stability of the CEVNC grouted waste monolith is de-pendent on the radiation stability of the individual components.
The concrete, metal and glass waste items are expected to be stable in a radiation field. GEVNC has run tests on the grout mixture con-crete to 108 rad exposure in which no signs of degradation were evident, but testing of repre6entative waste samples should be performed a t 109 rad since this is the dose that occurs with the Class C waste activity limits.
The waste plastics and cellulosics are expected to undergo radioly-sis. The radiolysis products are expected to include both gases (H, CO, and CO ) and liquids (organic acids). Gaseous radioly-2 2
sia products should, under normal conditions, escape the package.
The liquid organic acids may attack the concrete and/or metal waste items depending on the distance between these materials. The large amount of acid-neutralizing hydroxide present in concrete is ex-pected to reduce the capacity of radiolysis-produced acids to compromise the waste form s tability.
e The high pH of the concrete will tend to inhibit microbial growth.
The potential exists for radiation self-sterilization of the high activity grouted wa ste forms. Acids are the principal biodegra-dation by-products expected to be waste form-degradative but acid neutralization will occur in the concrete (until its capacity is exceeded) and, thereby the potential ef fects of these acids may be lessened.
e Leach testing should be performed on representative GEVNC waste samples. Leach tests of these forms must be carefully designed and their results carefully interpreted to ensure appropriate assignment of leaching properties. The leachability index of the grouting material and/or the form itself should be determined by CEVNC through tes ting.
e Immersion testing should be performed on these waste forms.
The presence of free liquid in the waste package has been ruled out e
on the basis of observations during tests performed by GEVNC. How-ever, the check by breach-drilling of the containst for drainable free liquid has not yet been done and it should be performed, The establishment of test sample size for the GEVNC heterogeneous o
high activity grouted waste monolith must ensure that samples are indeed representa tive.
It may be found that only full-size testing is appropriate and justifiable.
69
The homogeneity of compressive strength throughout the waste form e
monolith is questioned on the basis of the heterogeneity of waste i tems. Testing must be done to show that this monolith has the necessary strength throughout.
Conclusions and Recommendations Regarding GEVNC's TRU-contaminated waste, the solidified dilute e
burnup analysis solutions can probably be handled at a caemercial LLW burial site. The remainder of the waste, both current inventory and that produced in the future, will require some other treatment.
Most of its radioactivity will probably not be able to be handled by shallow land burial, even on a special case-by-case basis. On the other hand, most of its volume could be decontaminated and disposed of as LLW at a cannercial disposal site. In order to accomplish this, sorting of the solid waste into its components would be neces sa ry.
Combining all the hot waste solutions resulting from dissolving fuel o
sections and from acid digestion of paper wipes would allow treat-ment of 95% or.more of GEVNC's total TRU activi ty in one waste stream. The TRU concentrations of a waste form prepared from such solutions would be several orders of magnitude higher than the Class C limit, but no greater than those in current solidified hot liquid waste.
There appears presently to Se no accep table me thod in the priva te e
sector for disposing of waste containing such high levels of TRU ac tivi ty. One possible op tion would be to arrange to transfer them to DOE.
If this option is not available, they would either have to -
be emplaced in a future licensed geologic repository, or in a suit-able GCD facility at a LLW buiral site once such a facility has been developed.
It is recommended that guidelines and criteria be developed by MRC e
for construction and operation of GCD facilities. Criteria should also be developed for the types of waste and waste forms to be handled by GCD.
6.3 Evaluation of Additional Hazards in the CEVNC Wastes Based upon the information obtained from CEVNC regarding the contents of their wastes, as well as the evaluations performed by BNL, there do not appear to be any materials present in the Class B and C wastes which pose a signifi-cant non-radiological hazard. Further, there appear to be no materials pres-ent in these wastes in sufficient quantity to compromise long-term performance of the burial site.
70 1
f 7.
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E. P. Gause et al., Brookhaven National Laboratory, "Characteriza tion of the Class B Stable Radioactive Waste Packages of the New England Nuclear Corporation," NUREG/CR-3018, BNL-NUR EG-51607, 19 83.
2.
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3.
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6.
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7.
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9.
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Corrosion Bulletin, Volume 2, Number 3, May 1982, K. S. Rajagopalan, Ed.
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R. F. Stratfull, "The Corrosion of Steel in a Reinforced Concrete Bridge," Corrosion 13,173t (1957).
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K. C. Clear, " Time-to-Corros ion of Reinf o rcing S teel in Concre te S labs,"
FHWA-RD-76-70, Federal Highway Administration, Washingtor., DC,1974.
71
f 14.
H. A. Berman, "Ihe Ef fect of Sodium Chloride on the Corrosion of Concrete Reinforcing Steel and on the pH of Calcium Hydroxide Solution,"
FHWA-RD-74-1, Federal Highway Administration, Washington, D.C.,1974.
15.
D. A. Hausmann, "S teel Corrosion in Concrete: How Does It Occur,"
Materials Protection, p.19, November 1976.
16.
C. L. Page, Nature 258, 514 (1975).
17.
C. L. Page, "The Corrosion of Reinforcing Steel in Concrete:
Its Causes and Control," Bulle tin 77, Ins titution of Corrosion Science and Tech-nology (U.K. ), p. 2., November 1979.
18.
1:. L. Piciulo and others, Brookhaven National Laboratory, " Analyses of Soils at Low-Level Radioactive Waste Disposal Sites," BNL-NUREG-31388, June 1982.
19.
U.S. Energy Research and Development Administration," Final Environmental Statement," Waste Management Operations, Hanford Reserva tion, Richland, Washington, December 1975.
20.
J. E. Slater, Corrosion of Metals in Association With Concrete, ASTM Special Technical Publication 818, ASRM (PCN) 04-818000-27, Philadelphia, 1983.
21.
E. Veakis and E. P. Cause, Brookhaven National Laboratory, " Effects on Burial Site Performance and Burial Site Monitoring of New England Nuclear Corporation Radioactive Waste, BNL-NUREG-30765R, October 1982.
22.
K. Tuutti, " Corrosion of Steel in Concrete," presented at 6th European Congress on Metallic Corrosion, London, 1977.
23.
R. F. S tratfull, " Half Cell Potentials and the Corrosion of S teel in Concrete," California Division of Highways, CA-HY-MR-5116-7-72-42, prepared for Federal Highway Administration, November 1972.
24.
R. P. Br own and R. J. Ke s sle r, "An Accele ra ted Labora to ry Me thod f o r Corrosion Testing of Reinforced Concrete Using Impressed Current,"
Florida Department of Transportation, FHWA-FL-78-206, October 1978.
25.
P. Colombo and others, Broo'khaven National Laboratory, " Analysis and Evaluation of a Radioactive Waste Package Retrieved from the Atlantic 2800 Meter Disposal Site," EPA 510/1-82-009, BNL-51102, May 1982.
26.
H. H. Haynes, "Pe rmeability of Concrete in Seawa ter," in Performance of Concrete in Marine Environment, ACI SP-65, American Concrete Institute, De troi t, MI, p. 21, 19 80.
27.
F. M. Lea, The Chemistry of Cement and Concrete, Chemical Publishing Co.,
Inc., New Yo rk, 19 71.
72
28.
G. E. Troxell, H. E. Davis, J. W. Kelly, Composition and Properties of Concrete, 2nd Edition, McGraw Hill Book Co., New York, 1968.
29.
H. K. Manaktala and A. J. Weiss, Brookhaven National Laboratory, " Prop-erties of Radioactive Wastes and Waste Containers, Quarterly Progress Report, January-March 1980," NUREG/CR-1514, BNL-NUREG-51220. Hay 1980.
30.
D. R. Dougher ty and J. W. Adams, Brookhaven Na tional Laboratory, "Evalu-ation of the Three Miles Island Unit 2 Reactor Building Decontamination Process," Nu1EG/CR-3381, BNL-NUREG-51689, Augus t 1983.
31.
G. J. Hine and G. L. Brownell, Radiation Dosimetry, p. 858, Academic Pres s, Inc., Pew Yo rk, 19 56.
32.
N. E. Bibler, Savannah River Laboratory, '1Radiolytic Cas Production During Long-Term S torage of Nuclear Wastes," USDOE Report DP-MS-76-51 (Conf. 761002-3), October 1976.
33.
J. L. Clarke, The Behavior of Cattle Sla ts Under Biaxial Loading, A Comparison Be tween Used and Unused Units," CCA-TR-541, December 1980.
34.
B. Bludovsky and V. Duckacek, "Some Aspects of the Mechanisms of Cellulose Radiolysis," Radiochem. Radioanal. Letters 38, 21-30 (1979).
35.
P. A. Schweitzer, Corrosion Resistance Tables, pp. 473-475, Marcel Dekker, Inc., New York,1976.
36.
I. Biczok, Concrete Corrosion-Concrete Protection, Akademai Klado, Budapest, 19 72.
37.
A. Charlesby, " Breakdown of Organic Ma terials Under Irradiation," IEEE, Trans. Nucl. Soc. 16, 153-159 (1969).
38.
International Atomic Energy Agency, " Training Manual on Food Irradiation Technology and Techniques," Technical Report Series No. 114, p. 69, Vienna, Austria (July 1970).
39.
" Measurement of Leachability of Solidified Low-Level Radioactive Wastes,"
Third Draf t of a S tandard, American Nuclear Society S tandards Committee Working Group ANS-16.1, November 16, 1983.
40.
A. I. Nazarova and others, "The Fixation of Radioactive Wastes in Cement," ORNL-TR-4 418, 19 65.
41.
D. J. Le e and D. J. B r own, " Factors Af fecting the Leachability of Cesium and S trontium From Cemented Simulant Evaporator Was tes," United Kingdom Atomic Energy Au thority, AEEW-R-1461, Augus t 1981.
42.
J. G. Moore and others, " Leach Behavior of Hydrof racture Grout Incorpo-rating Radioactive Vastes," Nuclear Technology 32 (1977).
73
d a
f 43.
Handbook of Chemistry and Physics, The Chemical Rubber Publishing Ca...
Ohio, 42nd Edition, 2241-2245, 1968.
44.
J. Eisenmann, " Analysis of Restrained Curling Stresses and Temperature Measurements in Concrete Pavements," in Temperature and Concrete, American Concrete Institute, Detroit, Michigan, 235-250 (1971).
45.
D. H. Card, P. ~ H. Hunter, J. A. Adam, and R. B. White, "Cr'iteria for Greater Confinement of Radioactive Wastes at Arid Western Sites,"
NV0-234, May 1981.
46.
P. H. Hunter, " Technical Concept for a Greater-Confinement-Disposal Testing Facility," DOE /NV/10253-2, March 1982.
47.
T. L. Gilbert and C. Luner, " Analysis of Alternatives for Creater Con-finement Disposal," p. 291-307, Proceedings of the Fifth Annual Partici-pants' Information Meeting, DOE Low-Level Waste Management Program, Augus t 30-September 1,1983, Denver, CO, 00NF-8308106, December 1983.
48.
P.
T.~ Dickman and J. R. Boland, " Greater Confinement Disposal Test, Fiscal Year 1983 Progress," p. 316-323, Proceedings of the Fifth Annual Participants' Information Meeting, DOE Low-Level Waste Management Program, August 30-September 1983, Denver, CO, CONF-8308106, December 1983.
49.
O. A. Towler, J. R. Cook, and D. L. Peterson, " Greater Confinement Dis-posal Program a t the Savannah River Plant," p. 308-315, Proceedings of the Fif th Annual Participants' Information Meeting, DOE Low-Level Waste Management Program, August 30-September 1983, Denver, CO, CONF-8308106, December 1983.
50.
C. Bida and D. R. MacKenzie, Brookhaven National Laboratory, " Nuclear Waste Management Technical Support in the Development of Nuclear Waste Form Criteria for the NRC Task 2, Alternative TRU Technologies,"
NUREG/CR-2333, BNL-NUREG-51458, Vol. 2, Feburary 1982.
51.
R. M. Neilson, Jr. and P. Colombo, Brookhaven National Laboratory,
" Waste / Rock Interaction Technology Program, Status Report on Solidified TRU Waste From Plutonium Leach Tests," BNL-28856, August 1980.
52.
J. L. McElroy and H. C. Burkholder, Compilers, "Commerical Waste Trea tment Program, Annual Report for FY 1983," PNL-4963, February 1984.
53.
P.. E. Lerch, C. R. Allen, and M. D. Crippen, " Division of Waste Manage-ment Programs Progress Report, January-June 1979," HEDL-TME-79-60, October 1980.
54.
J. H. Crawford, "Eecontamination of TRU Glove Boxes," DP-1473, March 1978.
74
i 55.
W. E. Kennedy, Jr. and R. L. Aaberg, Iransuranic Advanced Disposal Systems: Preliminary Pu-239 Waste-Disposal Criteria for Hanford,"
PNL-4254, Sep tember 1982.
56.
U.S. Nuclear Regula tory Commission, "10 CFR Part 60,: Technical Criteria for Regulating Geologic Disposal of High-level Radioactive Waste, Pro-posed Rule," Federal Register, Vol. 45, No. 94, May 13,1980, pp. 31393-31408.
57.
U.S. Nuclear Regulatory Commission, "10 CFR Part 60,: Disposal of High-Level Radioactive Wastes in Geologic Repositories Technical Criteria, Final Rule," Federal Register, Vol. 48, No.120, June 21,1983, pp.
28194-28220.
b 75
I APPENDIX A 10 Cm PART 61 SECTIONS 55 and 56 77
PART 61 o LICENSING REQUIREMENTS FOR LAND DISPOSAL OF RADIOACTIVE WASTE disposal site before 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 e1.84 Alternative requiremente fo' disposal methods must be different, and disposal.
Wal n aM OPereuone.
in general more stringent, than those (v) For wastes containing mixtures of 9
The Commisslort may, upon request or specified for Class C waste. In the the nuclides listed in Table 2. the total on its own initiative. authorize absence of speci!!c requirements in this concentration shall be determined by provisions 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 section.
segregation and disposal of waste and approval, pursuant to I 61.58 of this for the design and operation of a land part.
TAats 2 disposal facility on a specific basis. Ifit (3) Cla ssification determined by long.
finds reasonable assurance of
!!ved radionuclides. lf radioactive weste C',".'*'*f'$'"
compliance with the performance contains only radionuclides listed in ce U objectives of Subpart C of this part.
Table 1. classification shall be C8 '
s a
determined as follows:
~
ls1.ss waste steulficsuon.
(1)If the concentration doee not f**' 8 8 "*** ** *" '"a 5 (a) Classification of waste for near exceed 0.1 times the value in Table 1
- -3"'*"*
'E N
N surface disposal, the waste is Class A.
c>eo rm it is (1) Considerations. Determination of (ii)If the concentration exceeda 0.1 U.,. e,,
d'
,5 /"
the classification of radioactive waste tJmes the value in Table 1 but does not s~a e o*
iso rom c*isi i
.4
.e=
involves two considerations. First, exceed the value in Table 1. the waste is considerstion must be given to the Clase C.
.n...,
m.a.s e., e.
%,8,,",c,j,,,m,,,==,n.ig,,77g['=ao,wa concentration of long. lived (iii)If the concentration exceeda the wa,n,,m, m
radionuclides (and their shorter. lived value in Table 1. the waste is not ga,a,.,a-*s,,-g,*y=lg w sg precursors) whose potential hazard will generally acceptable for near surface w
=
n,.non..
noon t.o e
persist long after such precautions as moposal.
@ ** "** * ** C'*" C **P*aaa' * ""
institutional controls, improved waste (iv) For wastes containing mixtures of
- form, and deeper disposal have ceased
. radionuc!! des listed in Table 1. the total (5) Classification determined by both
- to be effective.These precautions delay
- concentretion shall be determined by long-and short lived radionuclides. If the time when long. lived radionuclides
- the sum of fractions rule described in
- radioactive waste contains a mixture of a
f could cause exposures,in addition, the g paragraph (a)(7)of this section.
- radionuclides. some of which are listad magnitude of the potential dose is s in Table 1. and some of which are listed yAets,
- limited by the concentration and
- in Table 2.claulfication shall be availability of the radionuc!!de at the c
determined as follows:
time of exposure. Second. consideration naman E"'".
(i)If the concentration of a nuclide must be given to the concentration of listed in Table 1 does not exceed 0.1 g
=
shorter lived radionuclides for which times the value listed in Table 1. the e
requirements on institutional controls.
c.u o
class shall be that determined by the 3
s form, and disposal methods are
- egifm, g
concentration of nuclides listed in Tuble
=
pe se a um,eise mass es (2) Classes of waste. (1) Class A waste tem (ii)If the concentration of a nuclide "E
Z.m,. e.a s,.,.,p,,.,,mu,,,,,,,,,,,
listed in Table t exceeds o.1 times the
.E is waste thal is usually segregated from other waste clasus at the disposal site.
p
. iw value hated in Table 1 but does not M
The physical form arid characteristice of Neu e s'"
exceed the value in TableJ. the wasto Class A waste must meet the minimum shali be Class C. provided the j
requirements set forth in 161.56(a). If
'was se aans==== >= y==
concentration of nuclides listed in Tablo
.m Class A weste also meets the stability (4) Classification determined by short.
2 does not exceed the value shown in
-J requirements set forth in i 61.56(b). It is lived radionuclides. If radioactive waste Column 3 of Table 2.
=
not necessary to segregate the waste for does not contain any of the (6) Classification of wastes with n
disposel.
radionuclides listed in Table 1.
radionuclides other than those listed in M
(ii) Class B waste is weste that must classification shall be determined band Tables 1 and 2. lf radioactive waste meet more rigo'ous requirements on on the concentrations shown in Table 2.
docs not contuin any nuclides listed in g
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 (e)(6)of this section.lf radioscuve (7)The sum of the fractions rule for
'Eid characteristics of Class B waste must waste does not contain any nuclides mixtures of radionuclides. For 2
meet both the minimu.a and stability listed la either Table 1 or 2. lt is Class A.
determining classification for waste that g
requirements set forth in 161.46.
(1)If the concentration does not contairts a mixture of radionuclidss. It is (iii) Class C weste is waste that not exceed the value in Column 1. the waste necessary to determine the sum of only must meet more rigorous is Class A.
fractions by dividing each nuclide's requirements on weste form to ensure (ti)li the concentration exceeds the concentration by the appropriate limit stability but also requires additional value in Column 1.but does not exceed and adding the resulting values. "Ito j
measures at the disposal facility to the value in Column 2. the waste is appropriate limits must all be taken protect against inadvertent intrusion.
ClassB.
from the same column of the same table.
The physical form and characteristics of (111)If the concentrati9n exceeds the The sum of the fractions for the column Class C waste must meet both the va19e in Column 2. but does not exceed must be less than 1.oif the waste class minimum and stability requirements set the value la Column 3. the weste is is to be determined by that column.
i forth in i 61.56.
Class C.
Exumple: A waste contains Sr.00 in a 1
l 4I i
December 30,1982 3
b h
?)
I I
d' 4 E
PART 61 e LICENSING FEQUIREMENTS FOR LAND DISPOSAL OF RADIOACTIVE WASTE concentration of 50 Ci/m* and Co-137 in reanimum extent practicable the owned in fee by the Federal or a State a concentration of 22 Cl/mL Since the potential hazard from the non-government.
concentrations both e=ceed the valua radiological materials.
(b) Institutionalcontml. ne land in Column 1. Table 2 they must be (b) The requirements in this section owner or custodial agency shall carry compared to Column 2 values. For W are intended to provide stability of the out an Institutional control program to fraction 50/150-0.33, for Co-1M waste. Stability is intended to ensure physically control access to the dispoeal fraction. 22/44-0.5, the sum of the that the weste does not structurally site following transfer of control of the fractions = 0.83. Since the sun. Is less degrade and affect overall stability of disposal site from the disposal site than 1.0, the waste is Class A the site through slumping. collapse or operator. The institutional control (8) Determinatipn of concentmtions In other failure of the dimosal unh and program must also include, but not be wastes. The concentration of a I thereby lead to w.%r irMtration.
limited to, carrying out an r dionuchde may be determincJ,o/
bbtlity is n'..o a ; actor Y lo iting environmental moaltoring program at n
indirect methods such as use d acabi espoone t > an inadn rtent inta. der, the disposal site, periodic surveillance, factors which relate ire inferred
,m it povides a recognir.able and minor custodial care, and other concentration of one radionucliue to nond5rsible waste, requirements as determined by the 7ttother that is measured. 4 (1) Weste must have anctural Commission; and administration of radionuchde material wcountability, if stability. A structurally stable wsste funds to cover the costs for these there is reasonable assurance that I io form will generally maintain ha phye. cal avivities. De period of institutional indirect methods can be correlated vitth dimensions end its form, vrder the controls will be determined by the sctual measurements. %e concentratioc expected disposal conditivns such n Commlulon, but institutional controla
.1 of a radionuchde may be average ' over weight of overburden er.d compact' 2n may not be telled upon for more than the volume of the waste or weight of the equipment. the presence of motature, 100 years following transfer of control of waste if the units sie expressed as and microbtai activity, and 1Alernal
'.he disposal site to the owner, nanomries per gram.
factore Juch u radiation effects and chemical cher ee. Structural stabdity i Subpart E-F1nancial Assurances s
i 41.5e wasta charactertatks.
can be prov'.ded b the waste form 5 41.61 Appricant qua:tftcatione and (a) The follow'ng recements are minimum requirements fm all classes of itself proessMg t e waste to a stable assurances.
f rm. or placing ths. waste in a disposal Each applicant shall show that it waste and are intended.o facilitate contamr or suucture that provides either possesses the necessaiy funds or
. handling at the dispm ! site and provide
- protection of health ned safety of
- stabi.*y sher d weal
- has reasonable assurance of obtaining (2) Notwithstanding the provisions in the necessary funds, or by a
- personnel at the dispcsal site.
(1) Waste must not be packaged for
' il 61.%(a) (2) and (3). liquid wastes, or
' combination of the two, to cover the s
wastes ;ontaining ligeld, must be
$ estimated costs of conducting all disposa!in cardbod or fiberboard a
converted into a form nat contains a'
~ licensed ae'ivities over the planned
- b-a xes (2) Liquid waste must be solidified or little free standing erd ns zrrosive operating ife of the project. including packaged in sufficient abscrbent liquid as is reasonably achmable, but coste of constructiori and dispnaal.
material to absorb twice the volume of in no case shall the 'lquH r<ced 1't. of the liquid.
the volume of the wa.'e whm the wme j 41.62 Funding for dhposal alto closure and stamen, (3) Solid waste containing liquid shall is in a disposal container d esgned to contain as httle free standing and ensure stability, or 0.5% of the volume of (a)The applicant shall provide noncorrosive hquid a i is reasonably the waste for e arocessed to e assurance that sufficient funds will be achievable, but in no ca se shall the stable form.
available to carry out disposal site liquid exceed 1% e'.Le volume.
(3) Vol
- spaces within the waste and ekeure an I stablitration, including: (1)
(4) Waste must nM be readily capable between the wwte nnd ce package m at DecontaminaHon or di.manllement of of detorstion or of explosive be reduced to the extrat practicable land # iposal hellity structures, and (2) closure and stabilization of the disposal decomposition or reaction at normal I'
""9-site so that followins transfer of the pressures ard tenweratures, or of Eachgacka e of waste enust be disposal site to the site ownc, the need explosive reactior, with water.
to identify whether it is for ongoing active maintennnce Is (5) Waste mus< not contain, or be clearly abele Cines t waste, Class D waste, or rh es C eliminated to the extent practicable and capable of generating, quantities of toxic we sts in Jecordance with j 61.55-only minor custodial cere, surveillance.
gases, vapore, or fumes harmful to persons transporting, handung, or U i.58 Mterneuve requirements for waste and monitoring are required. These assurances shall be based on disposing of the waste Tiis does not dassmcauon and characteri toes.
Commission. approved cost estimates spply to radioact.ve tair Jus waste 4e CommissP 1 may, urn t.
iest or reflecting the Commission. approved packaged in accordanc, with paragraph on its own initiative, authorlo er plan for disposal site closure and (a)(7) of this section.
p:ovisions for the vansif caNi and stabillwation. The applicant's coat (8) Weste must not be pyrophoric, characteristics of we on a specific esumatu must take into account total Pyrophork rasterials contained in waste basis, if, after evalunts a of the specific enNat conts that wmuld be incurred if shall be treated prepared, and packared chaia.cleristice c' ur wwste, disposal ari llependent contractor were hired to b
to be nonnammable.
=lte. end mew of disposal, it finds perforrr the closure and stabihrauon (7) Wasa in a gaseous form must be ru s' 4.;c asurance of complier.ce packard et a pressure that don not e 'w p#ritence objectives in "O
(b) In rder to avoid unnecenary exceed 1.5 nmospheres s' 20*w. Total Subpart C of this part.
duphcatinn ard en pensa. th-activity must not exceed 100 curies per container.
l 61.59 instt%ttonal requirementa-Commission will accept Nnci11 (6) Weste containing basardous.
(s)I.ond ownership Disposal of sureties that have been o, sohdated biolog' cal. pa'hogertic, or infectio1s radioactive waste recelud from other with earmarked financle' or surety
=i material must be treated to reduce to the persons may be permitted only on land arrargaments estabhshed 'o meet l
J 74 Decembm 30.1%7 n
My
APPENDIX B TECHNICAL POSITION oN WASTE FORM 2.
Stability Guidance for Processed (i.e., Solidified) Class B and C Wastes a.
The stability guidance in this technical position for processed wastes should be implemented through the qualification af the individual licensee's process control program.
Generic test data may be used for qualifying process control programs.
Through the use of a well designed and implemented process control program, frequent requalification to demonstrate stability is, expected to be unnecessary. However, process control programs should include provisions to periodically demonstrate that the solidification system is functioning properly and waste products continue to meet the 10 CFR Part 61 stability r'equirements. Waste specimens should be prepared based on the proposed waste streams to be solidified and based on the range of waste stream chemiitries expected. The tests identified may be performed on radioactive or non-radioactive samples.
b.
Solidifiedwastespecimensshouldhavecompressivestreggthsof at least 50 psi when tested in accordance with ASTM C39.
Compressive strength tests for bitumino products should be performed in accordance with ASTM 01074 Many solidification agents will b' easily capable of meeting e
the 50 psi limit fop properly solidified wastes.
For these cases, process control parameters should be developed to achieve the maximum practical compressive strengths, not simply to achieve the minimum acceptable compressive strength.
The specimens for each proposed waste stream formulation should c.
remain stable after being exposed in a radiation field equivalent to the maximum level of exposure expected from the proposed wastes to be solidified.
Specimens for each proposeg waste stream formulation should be exposed to a minimum of 10 Rads in a gamma irradiator or equivafent.
If the maximum level of exposure is expected to exceed 10 Rads, testing should be perfonned at the expected maximum accumulated dose.
The irradiated specimens should have a minimum compressive strength of 50 psi folloying irradiation as tested in accordance with ASTM C39 or ASTM D1074.
81
3 d
../
' i APPENDIX B, Continued
. TECHNICAL POSITION oN WASTE FORM d.
' Specimens for each proposed waste stream formulation should be tested fgr resistance go biodegradatioh in accordance with both ASTM G21 and ASTM G22. No indication of culture growth should be visible.
Specimens should be suitable for compression testing in accordance with ASTM C39 or ASTM D1074.
Following the biodegradation testing, specimens should have compressive strengths greater than 50 psi as tested using ASTM C39 or ASTM D1074.
For polymeric or bitumen products, some visible culture growth from. contamination, additives.or biodegradable components on the specimen surface which do not relate to ove'rall substrate integrity may be present.
For these cases, additicnal testing should be performed. If culture growth is observed upon completion of the biodegradation test for polymeric or bitumen products, remove the test specimens from the culture, wash them free of all culture and growth with water and only light scrubbing. An organic solvent compatible with the substrate may be used to extract surface contaminants. Air dry the -
specimen.at room temperature and repeat the test. Specimens should have observed culture growths rated no greater than 1 in the repeated ASTM G21 test, and compressive strengths greater than 50 psi. The specimens should have no observed growth in the repeated ASTM G22 test,'a.nd.a compres'sive strength greater than 50 psi. Compression testing' sho'ild'be performed in accordance with ASTM'C39 or ASTM D1074.'
If growth is observed following the extraction procedure, longer term' testing of at least six months should be perform 9d to determine biodegradation rates.
The Bartha-pramer. Method is acceptable for this testing. Soils used should be representative of those at burial grounds. Biodegradation extrapolated for full-size waste forms to 300 years should produce less than a 10 percent loss of the total ' carbon in the waste form.
. Leach testing should be performed for a mfnimum of 90 days in e.
accordance with the procedure in ANS-16.1.
Specimen sizes-should be consistent with the samples prepared for the ASTM C39 or ASTM 01074 compressive strength tests.
In addition to the demineralized water test specified in ANS 16.1, additional testing using'other leachants specified in ANS 16.1.should also be performed to confirm the solidification agents leach resistance in other leachant media'.
It is preferred that the 82
m v
i APPENDIX ~ B, Continued TECHNICAL POSITION ON WASTE FORM synthesized sea water leachant also be tested. In addition, it is preferable;that radioactive'. tracers be utilized in performing the leach tests. The leachability index, as
~
calculated in ai:cordance with ANS 16.1, should be greater-than r
6.
f.
Wast'e specimens should maintain a minimum compressive strength of 50 psi as tested using ASTM C39 or ASTM D1074, following inanersion for_a minimum period of 90 days.
Immersion testing may be performed in conjunction with the leach testing.
g.
Waste specimens should be resistant to thermal degradation.
The heating and cooling chambers used for the thermal degradation testing should_ conform to the description given in ASTP B553, Section 3.
Samples suitable for performing conipressive strength tests in accordance with~ ASTM C39 or ASTM 01074 should be used._ Samples should be placed in the test 4
chamber and a series of 30' thermal cycles carried out in accordance with Section 5.4.1 through 5.4.4 of ASTM B553. The high temperature limit should be 60C and the low temperature limit -40C.
Following testing the waste specimens should have.
compressive strengths greater than 50 psi as tested using ASTM C39 or' ASTM D1074.
h.
Waste specimens should have less than 0.5 percent by volume of L
the waste specimen as free liquids as measured using the method -
described in ANS 55.1.
Free liquids should have a pH between 4 i
and 11.
I 1.
If small, simulated laboratory size specimens are used for the above testing, test data from sections'or cores of the anticipated full-scale products should be obtained to correlate the characteristics of actual size products with those of-r
. simulated laboratory size specimens. This testing may be performed on non-radioactive specimens. The full-scale f
specimens should be fabricated using actual or comparable solidification equipment.
t j.
Waste samples from full-scale specimens should be destructively analyzed to ensure that the product produced is homogeneous to L
the extent that all regions in the product can expect to have L
compressive strengths of at least 50 psi.
Full-scale specimens raay be fabricated using simulated non-radioactive products, but should be fabricated using actual solidification equipment.
83
=
APPENDIX C-SAMPLE RADIOACTIVE SHIPMENT RECORD FOR WASTES PROM GEVNC 9
e -e - Luj lo.s I 13. 11 1 A.
.n. Al
.o m 21614 Aon 1
.o n_r_.mA33.eg..a u _gester US ECOLOGY, INC.
.. _no. - enintra es saave==a==
8 a_ a aan EXECUTIVE OFFICE:(602) 426-7100 gja P.O. BOX 7246 e LOUISVILLE. KENTUCKY 40207 v.14'- '
m E!A 8'a- E/A
- ~--
Illinois Office:(415) 454 2370 0"
PS --- ^-
svare talif!gynta RAEAA 1
c,n couract N/A j
cow,ac, J. f. Tggeria e Consigned To:
suo e m1A fA11) mas.eest IXP.O. Box 638 O P.O. Box 578 unre peninv uo.
k1A j
Richland, WA 99352 beatty. NV 80003
' sis.esi E
-ears J/27/83 g
mo - - --
(509) 377 2411 (702) 553-2203 a
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Total Actevery Total Volume TOTAL QUAMYI tPER de CFR 172.1011 uneen We Pouts 05 The shomene The shgm.nt The Shement l Radioactive Device, N.O.S. - Red.oective Meterol UN29t t 7
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APPENDIX D CERTIFICATE OF COMPLIANCE FOR RADI0 ACTIVE MATERIALS PACKAGES Perm NRC418 U.S. NUCLEAR REGULATORY COMMISSION CERTIFICATE OF COMPT. LANCE cRn For Radioactive Materials Packages 1.tbl Revis No.
3.lcl ag i i
io No.
1.(d) P s No. l.(el Toa No. Pages l.tel Certi uneer
- 2. PREAMELE 2.lal This certificate is issued to satisfy Sections 173.393a.173.394.173.39s. anus 173.396 of the Departrnent of Transportatant Harardous Meierials Regulations 149'CFR 170-189 and 14 CFR 1031 and Sect' ions 146-19-10s and 146-19-100 of the oeoartment of
.. Transportation Dangerous Cargoes Regulations (46 CFR 146-149), as amended.
The packaging 'and contents described in item s below. meets the safety standards set forth in Suboast c of Tiefe 10. Code of 2.tni Federal Regulations. Part 71. " Packaging of Radioactive Meteriefs for Transport and Transportation of Radioactive Material Under Certain Conditione.-
This certificate does not re;leve the consignor from comotience weith any requirement of the regulations of the U.S. Deoer*, ment of 2Jcl Transoortation or other acolicable regulatory. agencies, including the government of any country through or into which the package unit 1pe transoorted.
- 3. This certificate is issued on the basis of a safety analysis rooort of the package design or application-Etel Prepared by (Name and address):
3.fb)
Tide and identification of report or application:
GIneral Electric Company General Electric Company appitcation dated P. O. Box 460 January 8, 1969, as supplemented.
Pieasanton, CA 94566 71-9044 3,,,
CONotTioNS This certeficate is conditional upon the fulfitring of the reaviremones of Suboort O of 10 CFR 71, as applicable and the conditions specified isiitem 5 below.
- s. Description of Packaging and Authorized Contents. Modet NumDer. Fissale Class!Otner Conditions and
References:
(a) Packaging l
(1) Model No.: GE-1600 1
(2) Description Steel encased lead shielded shipping cask. A double-walled steel cylinder protective jacket encloses the cask during transport.
It is bolted to a steel pallet. The cask is closed by a lead-filled flanged plug fitted with a silicone rubber gasket and bolted closure. The l
cavity is equipped with a drain line and the physical description is as follows:
Cask height, in 67 $
Cask diameter, in 38.5 Cavity height, in 54.0 i
Cavity diameter, in 26.5 Lead shielding, in 5.0 Protective jacket height, in 81.4 Protective jacket width, in 68.0 Packaging weight, Ibs 23:050 87 i
e
. ~..
. - = _ -
i APPENDIX D,-Continued CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MAT 1! RIALS PACKAGES PIge 3 - Certificate No. 9044 - Revision No. 6 - Docket No. 71-9044 5.
(b)' (1) Coatents (continued)
(vi) Solid nonfissile irradiated metal hardware, reactor control rods (blades), and reactor start-up sources.
(c) Fissile Class III Maximum. number of packages (i) Contents 5.(b)(1)(1), 5.(b)(1)(fi),
per shipment or 5.(b)(1)(iii):
Two (2): or (ii) Contents 5.(b)(1)(iv):
One (1) 4 6.
The U-235 equivalent mass is determined by U-235 mass plus 1.66 times U-233 mass plus 1.66 times Pu mass.
7.
For packaging of neutron sources, the cavity drain Ifne shall be closed with a plug with'a melting temperature of 200*F and the cask cavity shall be filled with water with a 5-inch air space within the cask cavity. When needed, sufficient antifreeze in the cask shall be used to prevent damage to any component of the package due to
[
freezing.
8.
For packaging of other than neutron sources, the. cask shall be delivered to a carrier dry and the cavity drain line shall be closed with a plug which will main-l tain its seal at temperatures up to at least 620*F.
l
- 91. Shoring shall be provided to minimize movement of contents during accident condi-l tions of transport.
10.
Prior to each shipment the silicone rubber lid gasket (s) shall be inspected. This gasket (s) 'shall be replaced if inspection shows any defects or every twelve (12) months, whichever occurs firts. Cavity drain line shall be sealed with appropriate sealant applied to threads of pipe plug.
- 11. For packaging of neutron sources, measurements shall be made to determine that the dose rate does not exceed 1,000 arem/hr at 3 feet from the surface of a dry cask with no additional shielding within the cask.
- 12. The contents described in 5.(b)(1)(v) shall be transported on a motor vehicle, railroad car, aircraft, inland water crafts, or hold or deck of a seagoing vessel assigned for sole use of the ifcensee.
13.
The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CFR $71.12(b).
- 14. Expiration date: December 31, 1980.
I 88 l
l
1 APPENDIX D, Continued CERTIFICATE OF COMPLIANCE FOR RAD 40 ACTIVE MATERIALS PACKAGES Page.2 - Certificate No.-9044 - Revision No. 6 - Docket No. 71-9044 5.'
(a) Packaging (continued)
E
'(3) Drawings The packaging is. constructed in accordance with the following General-Electric Company Drawing Nos.:
212E255, Rev. 3 135C5598, Rev. I 10603986, Rev. I 10603973,-Rev. 1 174F237 Rev. 1 (b) Contents (1) Type, form and maximum quantity of material per package Plutonium in excess of twenty (20) curies per package must be in the form of metal, metal alloy or reactor fuel elements; and i
(1) Byproduct material and special nuclear material as solid metal or oxides. Decay heat not to exceed 600 watts. The radioactive material shall be in the form of fuel rods, or plates, fuel assemblies, or meeting special form requirements of 10 CFR 571.4(o).
500 gm U-235 equivalent mass; or (ii) Neutron sources in special form.
L 500 gm U-235 equivalent mass. Decay heat not to exceed 50 watts; or 1
I-(iii) Irradiated Pu0 and U0 fuel rods clad in zircaloy or stainless 2
steel. Decay heat not to exceed 600 watts. All fuel rods shall be j
contained within a closed 5-inch Schedule 40 pipe with a maximum useable length of 39-5/8 inches.
I 1,200 gm fissile material with no more than 300 gm fissile material j
per 5-inch Schedule 40 pipe.
(iv) Irradiated UC and ThC fuel particles clad in graphite and contained within a standard HTGR hexagonal cross-section graphite block.
Decay heat not to exceed 600 watts.
Each graphite block shall be l
contained within a sealed cylindrical inner container constructed in l
accordance with General Atomic Company Drawing No. 021583, Issue A, with three, 1/2-inch by 4-1/2-inch radial fins to provide centering r
l within the cavity.
i 1,400 grams U-235 equivalent mass in each inner container with no more than one inner container per package.
(v) Process solids, either dewatered, solid, or solidified in a secondary sealed contaii.er meeting the requirements for low specific activity radioactive..aterial.
89
APPENDIX D, Continued CERTIFICATE OF COMPLIANCE FOR RADIOACf1VE MATERIALS PACKAGES Page 4 - Certificate No. 9044 - Revision No. 6 - Docket No. 7.1-9044 REFERENCES General Electric application dated January 8, 1969.
Supplements dated:
February 12, 20, and 27, and March 10 and 24,1969; November 20, 1970; January 29 and March 12c 1971; July 3 and November 15, 1973; and August 26, 1980.
Nuclear Plant Services supplement dated: July 7, 1975.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION Charles E. MacDonald, C ief (
Transportation Certification Branch Division of Fuel Cycle and Materials Safety OCT 0 31363 Date:
90
'NiAC rgMM 335
- 1. HLPOHT NUMl'EH '%ss+eny UUCI U.S. NUCLE A 3 EE!ULATORY COMMISSION BIBLIOGRAPHIC DATA SHEET hk hJRhG-9I 4 TITLE AND SuuTITLE LAdd Volume Nn.s!noproprosse)
- 2. ILeave OIMk)
)
Charreterization of the Low-Level Radioactive Wastes and
/
Wuta Packages of General Electric Vallecitos Nuclear Center
- 3. HECIPIENT'S AQCESSION NO.
\\
/
- t. AulilO HISI te. DATE HEPMtT COMPLE TED C. R. Kempf, D. R.
Kenzie, 3. S. Bowerman, you y, j, l nan D. R. Dougherty, and Siskind
.Tiahe 19R4 U. PEHFOllMING OHGAN1/ATION NA AND MAILING ADDRESS / Include les Codel DATfREPOHT ISSUED l"^"
Brookhaven National Labo tory
- November 1984
.D:partment of Nuclear Ener, Upton, New York-11973 f*"""*#
b 4. (Leave WMhl
- 12. SPONSOHING OHGANilATION NAME AND M LING ADOHESS (include I,a Codel Division of Waste Management Office of Nuclear Material Safety d Safeguards f
II. FIN NO.
U.S. Nuclear Regulatory Commission A3172 Wuhington, D.C.
20555 g/
51 I YPL UI 144 Pulli Pk pitou covt HL u //nctusere detest l
J lla. SUPPLEMENTAHY NOTES
- 14. (Lasve o/m&J V
- 10. AUSTH ACT (200 wo<ers or less)
An evalua tion of the low-level wastep nd waste packages generated by General Electric Vallecitos Nuclear Center
,EVNC) was made on the basis of 10 CFR Part 61 criteria and on the Techdical osition on Waste Form (TP).
In addition, a review has been performed 5f the ndling and s torage methods used by GEVNC for their transuranic wastes [ Severa options have been discussed for management of these materials. This evalua on was the result of a stuoy initiated by the U.S. Nuclear Regula1ory Commiss n (NRC), in which GEVNC participated. CEVNC genera tes ClasTi B or grea ter adioactive wastes in hot cell processes which include exami$ation of reacto fuel and components, and productionofsourcesandradiop[h/rmaceuticals.
Th dominant contaminating radioisotopes are Cs-137 and Co- 0.
In addition, tr suranic wastes which result from examination and but p analyses of fuel a. all currently s tored on-site a t CEVNC. Class B and reater wastes are pack ed in 84-gallon extended 17H drums that are gr uted with cement.
Base on the evalua tion, overall, the waste forms of t ese packages are expected naintain their stability, but a few concern are identified and testing. ould be performed by CEVNC to demons tra te wa s form s tability.
4 / KE Y WOHOS AND DOCUMENT AN ALYSIS I /4 DE SCHiP TOHS evaluation with respect to low-level, Cs-137, Co-60, routed concrete waste forms, packa Jechnical Position on Was > Form, low-level TRU waste managemen options I /It IDENTIFIE HSiOPEN kNDE D TEHMS 11 AV AIL ABILITY ST ATEMENT
- 19. SE CURITY CLASS I/h.s report /
yi Nu of PA(,I5 Unclassified Unlimited zu sEcum rY CL Abb (/Arapet
- PHIO Unclassified 5
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