ML19305E828
| ML19305E828 | |
| Person / Time | |
|---|---|
| Issue date: | 05/14/1980 |
| From: | Bell M, Rohrer D, White L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| References | |
| RULE-PR-60 NUDOCS 8005200489 | |
| Download: ML19305E828 (18) | |
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80 052 0 OVF9 l
I REGULATION OF GE0 LOGIC REPOSITORIES FOR THE DISPOSAL OF HIGH-LEVEL RADI0 ACTIVE WASTES l
Lawrence A. White i
Michael J. Bell David M. Rohrer High-Level Waste Technical Development Branch Division of Waste Management 2
U.S. Nuclear Regulatory Commission j
Washington, D.C.
20555 i
INTRODUCTION 4
The U.S. Nuclear Regulatory Commission (NRC) is vested with licensing and regulatory authority over certain U.S. Department of Energy (00E) facilities by Sections 202(3) and (4) of the Energy Reorganization Act of 1974.
These sections refer to:
- 1) facili-ties used primarily for the receipt: and storage (including-dis-posal) of high-level radioactive wastes (HLW) and 2) retrievable surface storage and other facilities authorized for the express purpose of long-term storage of HLW.
Geologic repositories would not be licensed as " production" or " utilization" facilities.
Rather, they would be licensed under those provisions of the Atomic Energy Act dealing with receipt and possession of " byproduct" and "special nuclear" materials.
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The NRC has the, responsibility to evaluate for all aspects of repository performance which could impact -the public health and safety. As a first step in carring cut its responsibility, the NRC _ staff is developing the. regulations under which a geologic repository will be-licensed. This regulation will be known-as 10 CFR Part'60 - Disposal of High-Level Radioactive Wastes in Geologic Repositories.
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2 PURPOSES ~CF THIS PAPER The primary purposes of this paper are to:
(1) discuss the NRC staff's proposed strategy for regu-lating the disposal of radioactive wastes in geologic repositories; (2) discuss the basic technical requirements being con-
.sidered in developing the regulation; (3) discuss the rationale for the major technical require-ments with regard to engineered systems; and (4) solicit comments and recommendations from interested parties.
PROPOSED NRC STAFF STRATEGY FOR REGULATING GEOLOGIC DISPOSAL OF RADI0 ACTIVE WASTES The overall performance objective for the repository (i.e.,
the maximum allowable release of rMionuclides to the biosphere) will be established by the U.S. Environnental' Protection Agency (EPA) in their Environmental Radiation Protection Standards. ' The -
fRC staff will implenent these requirements in developing its regul ations.
.The NRC staff has considered a-number of alternative approaches by which geologic disposal could be licensed.
For example, the staff could merely state that the standards estab-lished by the EPA must be satisfied.
In that case the NRC would regul ate. on an ad-hoc basis. Another alterntive is for the staff to develop a general regulation, without specific numerical performance objectives or criteria..Both of these alternatives offer substantial flexibility; however, they provide neither the DOE nor the public any substantive guidance as to how the NRC (and its licensing boards) will make their findings. A third-alternative, which the NRC staff has chosen,-is to develop a more detailed regulation including specific numerical performance requirements.
This approach may constrain D0E's options; how-ever, it does provide guidance as to what the NRC will find acceptable.
Before'I discuss our proposed regulatory approach, I would like :to _eaphasize that our regulation at this point is tentative.
..The technical requirements I am going to discuss are still evolving and are subject to change.
Our technical support and research programs will be aimed at refining these technical requirements.
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Multi-Barrier Approach Until very recently the designs of deep geologic reposi-tories have placed almost total reliance for containment of the radionuclides on the site characteristics and surrounding geology.
As a result, little attention has been given to waste forms, containers and other engineered barriers to significantly contri-bute to the containment and isolation of the radionuclides for extensive periods of time.
For example, credit was not normally given for the waste form and its associated packaging being able to contain the radionuclides for more than a few tens of years.
One does not have to search very far in the scientific literature, regarding the geologic aspects of disposing of HLW, before discovering that there is still a great deal of scientific uncertainty surrounding many of the assumptions that need to be made to predict long-term geologic performance. First, we cannot define geologic conditions exactly and therefore must make certain assumptions which cannot be completely verified.Second, the earth scientist is called upon to describe how the geologic environment will change in the future. Such predictions intro-duce another source of uncertainty. The Interagency Review Group confirmed this assessment of the uncertainties in making geologic assessments in their report to the President in March 1979.
In order to compensate for the uncertainty in predicting the behavior of geologic systems over icng periods of time, the NRC staff has adopted a conservative multi-barrier approach.
In this. approach the staff views the repository to be composed of three major barriers: (1) the waste package, (2) the engineered repository structure, and (3) the site and its environs. The NPC will establish minimum perfcrmance objective: for each of these j
major barriers.
While we may foreclose some options, we do not intend to abandon the " systems approach" by setting performance objectives for these major parts of the system. 00E would have flexibility in the details of design of the waste package and repository structure so-long as the performance objectives are met and the waste package, repository design and site are compatible with one another.
The " Multiple Barrier Approach" was adopted for a number of reasons. -It provides " targets" for scientists and engineers
- working on specific parts of the overall problem of engineered and geologic isolation. The engineers and material scientists know what is required of the engineered system and the earth scientists know what,is required of the site. Also, from the NRC's point of view, the approach permits inclusion of safety margins at convenient points in the system to compensate for uncertainty in assessing performance of the overall system.
4 The three major barriers are defined as follows:
The " Waste Package" is the first major barrier and consists of the physical waste form, and its container, as well as any ancillary enclosures, shielding or overpacks such as an outside container.
The " Engineered Repository Structure" is the sum total of the many design features and engineered barriers of the reposi-tory which act to protect the waste package from the geologic environment and to retard the novement of radionuclides into the surrounding geologic environment. Engineered features could include such items as the design and layout of undergrou.nd openings, treatment of the rock, structural support systems, and materials emplaced to retard groundwater movement and to prevent radionuclide migration. Borehole and shaft seals are also part of the engineered repository; however, a separate objective will be established for their performance since they are primary potential pathways for water or nuclide movement into or away from the repository structure.
The " Site and Its Environs" is the sum total of the many natural geological, hydrological, and geochemical barriers which provide long pathways for groundwater movement and act to retard the movement of radionuclides along those pathwcys from the engineered repository to the accessible environment.
These barriers could include such items as low hydraulic gradients, low permeability rocks, high geochemical retardation capability, the. ability to dissipate heat, and a geologic setting which is simple to evaluate, highly stable, and has the necessary charac-teristici needed to permit sealing of boreholes, shafts and underg, ound openings.
PERFORMANCE OBJECTIVES In general, our planned approach is to set performance objectives, first, for the period when the hazard is dominated by the fission products and, second, for the long-term performance when the hazard is relatively constant and dominated by the acti-nides.
The staff is putting emphasis on: 1) engineered contain-ment of radionuclides during the fission. product pulse when the hazard is the greatest; 2) assurance of a controlled release thereafter. This simplifies analysis and reduces uncertainties introduced into the analysis of the total system. During the period of engineered containment of the waste, the site geology
'should provide sufficient backup to account for those scenarios which may result in loss of engineered containment.- Thereafter, the site geology should also have the capacity to reta-d the-movement of the long-lived radionuclides to the accessible environment so that the EPA standard is not exceeded.
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5 The staff views the waste form and container as a major component of the engineered containment system. A major advantage of putting emphasis on tha waste package is that it can be manu-factured under closely controlled conditions and can be physi-cally tested to verify its performance. Other engineered barriers can be designed to have uniform properties and can be placed under controlled conditions; however, complex systems models may be needed to project their performance. The site geology is the most difficult component to deal with in that we cannot accurately define the geologic environment nor test the gross characteristics, given a limited number of exploratory boreholes over the several thousands of square miles that will be involved in the transport analyses. In summary, the waste package and repository design are inherently knowable and under man's control whereas the site may never be thoroughly understood and certainly is not under man's control. We are therefore encouraging DOE to (1) develop and experiment with alternative waste forms and containers to see exactly how much reliance can be put on this component, and (2) put more emphasis on other engineered barriers as part of the repository design to protect the waste package over the short-term and help control releases over the long-term. However, we are not neglecting the site.
We anticipate that DOE will choose very stable sites which will inherently have multiple natural barriers to radionuclide migration.
With regards to the long-term performance objectives for the engineered systems, the staff is considering several alterna-tives generally along the approach of specifying a maximum near-field release rate, applied either to the whole engineered system or to the waste package. Placing a release rate restraint on the waste package has the advantage that the release rate can be physically measured under aggravated or accelerated conditions.
On the other hand, applying a release restraint on the total engineered system permits flexibility in giving credit for other engineered barriers in retarding radionuclide migration.
With regard to the site, the performance objective will be largely controlled by the EPA standard. The site should provide sufficient protection to back up the engineered system considering all credible scenarios and the conditions of the engineered system for those scenarios.
In addition to specifying performance objectives for the site, the NRC staff is considering specifying preferred site characteristics and exclusion requirements _ to help guide DOE t'o selection of sites that meet the performance objectives.'
The performance objective for the period during which the fission products dominate the hazard is one of cnntainment within the engineered barriers. The combination of the waste
6 package and repository structure is intended to provide assurance that radionuclides will not be released into the near-field geologic environment for at least the first 1000 years. For credible scenarios which breach the engineered system, the site should provide sufficient backup and rentention of radionuclides so that releases to the accessible environment are within the EPA standard.
In order to ensure engineered containment of the radionu-clides we plan to specify a minimum performance objective for the waste package.
Each waste package should be designed to provide reasonable assurance of containment of radionuclides for at least 1000 years and for as long as reasonably achievable beyond that, assuming that the repository becom'es saturated shortly after closure. Other assumptions concerning groundwater chemistry, flow rates and in situ stresses, which are necessary to define the envircnment in which the waste package must meet its performance objective, should be based on site conditions and the state of the repository structure under appropriately chosen scenarios.
The intent here is to envelop potential perturbation to the site and design so as to establish design basis conditions.
The repository structure should be designed to protect the waste package and to provide reasonable assurance that radio-nuclides will nct enter the geosphere for at least 1000 years in the event any of the waste packages should fail to meet their performance objective. The idea is to design the repository so as to compensate for any credible scenario which results in degradation of the waste package by itself.
Under conservative assumptions on the conditions of the waste package for such scenarios, there should be suff.icient barriers in the repository structure to mitigate the consequences so that there is high assurance that radionuclides will not be released into the geologic environment for the first 1000 years.
This can be accomplished by:
providing barriers to retard the movement of groundwater to and from the waste; zoning backfill materials so as to altogether prohibit water from' coming in contact with the waste; or by providing zones of high ion ex.hange. materials to capture radionuclides. Also, designing a waste package that would be extremely leach resistent, even though it may struc-turally degrade or become chemically altered, could mitigate the consequence of package failure.
Borehole and shaft seals are also a part of the repository design which will affect the rate at which groundwater from overlying or underlying aquifers may reach the repository structure or the waste package, as well as affect the rate of movement of radionuclides to the biosphere once they are released to the geologic environment.
The performance objective
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for borehole and shaft seals, therefore, is to provide a barrier to fluid migration equivalent to that of the undisturbed section of rock through which the borehole or shaft passes.
The perfomance objective for the engineered system after the first 1000 years is to limit the release of radionuclides to the geologic environment. After the first 1000 years, the waste package and other engineered barriers together, should retard tha movement.of the radioactivity and limit its release from the confines of the engineered system to as icw as reasonably acgievably but to at most a yearly release rate of one part in 10 to 10 of the package inventory.
It is recognized that it may ultimately be difficul~t to demonstrate that this perfomance objective is met because of difficulty in analyzing the performance of engineered barriers over the long-term. One way of reducing the uncer-tainty is to limit the release rate of the waste fom or package since this can be directly measured by accelerated testing.
Therefore, we are encouraging DOE to examine the release resis-tance of alternative waste foms and to see how much emphasis may be placed on the waste fom to limit the release of radionu-clides over the long-tem.
Appendix A evaluates the influance qf limiting the release of radioactivity from the engineered system. This derivation is based on a simple model in which a constant number of grams of waste is released per year per gram of waste initially emplaced in a. repository. While a more realistic mcdel would need to include such factors as temperature and the release rate of the waste fom, this simple model seems adequate as a rough estimate of the effectiveness of a release rate restriction.
Lowering the release limit allows more of the initial radio-activity to decay and thereby reduces the amount of radioactivity available for transport to the environment. As Appendix A indicates, release limits greater than 10- per year result in little reduction in the fraction of long-lived nuclides which are ultimately released.
In order to achieve virtually total containment of the plutonium nuclides, it would be necessary to limit the release to a value of about 10~6 per year. Similarly, substaptial containment of Tc-99 would require a release limit of 10- per year or lower. Because the plutonium nuclides and Tc-99 could be the major long-lived contributors to the popula-tion-dose from waste dispgsal, thy.NRC staff considers a release limit on the order of 10 to 10 per year to be appropriate in order to minimize the impact of these nuclides.
The selection of good sites is a prerequisite to meeting the design perfomance objective. Repository sites should be
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1 chosen which (1) provide a high level of stability so as to enhance the engineered containment of radionuclides, (2) permit i
design of a stable repository,.and (3) provide capability to contain the waste in the event of breach of the engineered system for the first 1000 years, and to isolate the waste so as to meet the EPA standard thereafter.
In addition to the above performance objectives, there is another performance requirement for the design of the reposi-to ry. The repository should be designed to preserve the option to retrieve the waste throughout operation of the repository and for 50 years thereafter. The length of time, 50 years, is tentative.
The purpose of this requirement is to ensure an extremely stable repository design that will preserve the option to monitor the performance of the repository after it is filled with waste, will allow corrective measures, if necessary, and, as a last resort, will allow all or part of the waste to be removed.
RATIONALE FOR MAJOR TECHNICAL REQUIREMENTS The 1000 year performance requirement on the waste package was established because the waste form and packaging are engineered items which can be physically tested, their perfor-mance and reliabili'ty evaluated, and the system optimized for the specific emplacement environment. Also, the manufacture of these ccaponents can be strictly controlled. Such performance of the waste form and packaging during the period of time when the ~ wastes would be subjected to the greatest potential for detrimental alternation (i.e.,-the period of high thermal flux
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and of potential chemical interaction) would ccmpensate for the-uncertainties in predicting the complex waste / rock interactions 4
and their consequences during this time period.
I The staff also feels that requiring the waste package to 1
i have a reasonably extended expected life is one of the easiest
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ways to compensate for the inevitable uncertainties in predicting l
the behavior of the geologic systems over long periods of time.
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More specifically, containment of the radionuclides by the pack-age for the first 1000 years is desirable from the standpoint that:
(1) The thermal output of the. waste packages will have dropped by 'over two orders of magnitude.
(The leach rates of waste forms are highly temperature dependent.) ~
(2) The temperature of the emplacement environment near.
the waste package will have reached a maximum and have begun to d rop. Mcwever, under. most conditions temperatures will not have n
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For this reason, we are asking DOE to examine the possibility of developing a waste package which would last much longer than 1000 years. For spent fuel, it will take on the order of 10,000 years for the rock to return to near ambient temperatures.
(3) The total radioactivity (curies) in the packages will have decreased by two orders of magnitude and the radiotoxicity of the wastes will have decreased by over three orders of magnitude.
The NRC staff does not believe that the changes in the waste package needed to meet the thousand year objective will be unduly expensive.
In any case, DOE's present estimate of the costs associated with a repository shows the waste encapsulation (including the waste container) accounts for only about nine percent of the total cost of disposal.
Thus, the financial impact of the 1000 years expected life for the package on total waste isolation costs is expected to be a small fraction of the cost and therefore a small price to pay for the protection.
Recently the NRC staff has acted, through a contract at the Brookhaven National Laboratory, to see whether the expected life of the package might reasonably be increased to 10,000 years or more.
In this effort, the NRC staff will continue to ask what is necessary to achieve waste isolation without limiting its requirements to those which can be met using only the presently demonstrated technology.
"It should be noted that on September 26, 1979 at the OECD Nuclear Energy Agency (NEA) meeting in Paris, the British announced that they are also placing major emphasis on the waste package.
Their intent, which was developed quite independently of the NRC, is that the waste package provide redundant protec-tion for at least 1000 years. Further, they are also actively working to see if that number can be pushed out to 10,000 years or more.
The rationale for emphasizing the need for a well designed engineered repository is rather simple. We are applying practi-cal engineering principles in a conservative manner to gain the confidence needed that the repository will work the way it is intended.
We. view a repository as civil construction as opposed to a convential mine. As a civil structure, the repository performs the functions of (1) protecting the waste package during the 1000 year containment, (2) mitigating the effects of any geologic ~ events or processes which may affect the waste package during this period, and (3) over the long-term acts with the waste-package to limit the release of radionuclides to the geologic environment.
It also provides for the retrievability
1 10 of waste until such time as a decision can be made that we have 4
dcne everything right and the repository sdll work in containing the waste as planned.
It is our intent to encourage the construction of a struc-ture which (1) is stable, (2) provides a hydraulic barrier to groundwater movement, and (3) captures radionuclides or provides a long tortuous path for nuclides to reach the geologic environ-ment.
In view of the uncertainty involved in systems modeling, a well designed engineered structure will provide the needed confidence that releases are in fact within the EPA standard.
The NRC staff believes that current geotechnial engineering principles can be used to design and construct a repository to meet its performance objectives. For example, rock fracturing can be reduced to an acceptable level by control of the construc-tion process and heat load. We believe there is much room for technology transfer from experiences gained in underground civil const ruction. Research and development should be aimed at testing these design principles as applied to a repository.
The purpose of the long-term performance objective of limiting the release of radionuclides to the geologic environment is similar to the containement objective for the first 1000 years.
That is, to gain confidence that the EPA standard will be met.
The toxicity of the waste decreases with time, which is fortunate because so does our ability to predict repository perfonnance. Our goal is to provide a level of safety com-mensurate with the hazard over the long-term. We also want to ccmpensate for uncertainty in systems analyses to be consistent with the degree of protection that is needed. The toxicity of the waste after 1000 years has been reduced to a level near to that.cf a natural are body. However, the waste is much more concentrated and the inventory of nuclides is different.
In fact, some nuclides. are known to be more mobile than those of a natural are body.
The intent of limiting the release at this point is to (1) to build in engineered barriers to reduce the mcbility of the radionuclides as much as practical and (2) ccmpensate for the uncertainty in predicting how well the geology will contain the radionuclides. The characteristics of the geologic and geochemical environment around an ore body is the reason the ore body was formed in the first place. There
- fore, there are natural site characteristics which contain the waste. Repository sites, however, will be selected'by man, and we may never know for sure how well the site may perform.
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11 In addition to the perfonmance cbjectives describcd in the paper, the NRC staff is developing guidance for the technical community regarding how to meet these objectives. Appendix B contains preliminary supporting guidance on waste forms and packaging for your purusal and comment.
CONCLUSION It must be enphasized that the performance objectives pre-sented in this paper are preliminary. They are subject to change in light of information we receive frca our technical contractors and from comments and se"L;*tions we receive frca both the technical community at large
- a. well as the int'erested public. We urge your review and critique of these positions and look forward to receiving your comments.
ACKNOWLEDGMENT The authors wish to express their appreciation to Dr. Daniel J. Fehringer of the NRC staff for his input on Appendix A of this paper and to Mr. John B. Martin, Director, Division of Waste Management for his encouragement.
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APPENDIX A The influence of the release rate on containment of wastes can b'e evaluated as follows. Denote the Release Rate by RR, and assume for the sake of simplicity a constant release rate over time, i.e., for each gram of waste initially present, RR grams of waste will be released each year. The total amount of waste will then be released in a time period called the. Release Time (RT) where RT = 1/RR. Combining this relationship with the law of radioactive decay, one can calculate the fraction of a specific radionuclide inventory which will be released before the radionuclide decays. This relationship is given by the equation:
1.443 FR =
RT/t1/2 5,here FR is the iraction of the initial nuclide inventory-rel eased before decay, t is the half-life-of the nuclide and RT is the release tidd2 The relationship between FP, and RT/t indicates that, in order to provide any significant cont 1hment of a nuclide, the release time must substantially exceed the half-life of the nuclide.
If, for example, the release time of the waste is 15 times the half-life of a nuclide, approximately 90% of that nuclide will decay before being released to the geologic environment.
Table 1 evaluates'the influence of the release time on the releases of several long-lived nuclides present. in high-level 6
wastes.
The table shows that a release time of 10 years would provide virtually total containment of the plutonium nuclides 7
and about 70 percent containment of Tc-99. A release time of 10 years would be necessary to achieve nearig canplete containment of Tc-99, and a release time exceeding 10 years would be necessary to provide substantial containment of I-129.
This information can be use'd to evaluate the effectiveness of proposed release rate requirements.
If, for exampl e, it -is considered desirable to achieve virtually total contai; mat of the Tc-99 and I-129'nuclides within the waste fom, it woulg be necessary to limit the release rate to a value of.about.10- and 10- per year, respectively. Alter that release rates greater than 10 gatively, Table 1 indicates per year would provide little. containment of these. long-lived nuclides, and.that nearly all of the initial radioactive-inventory of these nuclides would ultimately be released. from the waste form.
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13 Ta' l e 1.
Relationship Between Release Time and Total Releases a
From tha Engineered Repository 5
Release Time = 10 years fluclide Hal f-l ife Fraction Released 3
C-14 5.7x10 y 0.08 3
Pu-240 6.6x10 y 0.10 4
Pu-239 2.4x10 y 0.27 5
Tc-99 2.1x10 y 0.38 7
I-129 1.7x10 y 1.0 6
Release Time = 10 years Nuclide Hal f-life Fraction Released 3
C-14 5.7x10 y 0.008 3
Pu-240 6.6x10 y 0.010 4
Pu-239 2.4x10 y 0.035 5
Tc-99 2.1x10 y 0.29 I-l ?.9 1.7x10 y 0.98 7
Release Time = 10 years flucl ide Half-life Fraction Released 3
C-14 5.7x10 y 0.0008 3
Pu-240 6.6x10 y 0.0010 4
Pu-239 2.4x10 y 0.0035 b
Tc-99 2.1x10 y 0.030 7
1-129 1.7x10 y 0.82 p
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14 Appendix B Waste Fona Acceptance Criteria Solidification. Pursuant to 10 CFR $0, Appendix F, and 10 CFR 71.42(a), all high-level liquid radioactive wastes shall be converted to a dry solid and placed in a sealed container prior to transfer to a Federal Repository. This criterion shall extend to all radioactive vastes (including TRU wastes and other non-high-level wastes) emplaced in a repository.
Stabilization.
Finely divided waste forms shall be stabi-lized by incorporation into a containing matrix so as to minimize the production and availability of respirable fines during any accident condition. Particles of sizes smaller than about 10 microns present a potential for serious health hazards should the container be breached during a transportation, handling or emplacement accident. Such fines can become airborne, enter the lungs through inhalation, and reside in the pulmonary region of the lung for very long periods of time resulting in doses to the lung in excess of those allowed under 10 CFR 20.
Further, the aggregate surface area of the finely divided particles is large relative to the mass of the particles, and if'a process which is surface area dependent (such as dissolution or leaching processes might be), the rate of reaction can be accelerated by the wastes being in the form of finely divided particles.
Free Liouid. Radioactive wastes containing free liquid shall not be accepted at the repository. The removal of these free, unbound liquids is required to (1) reduce the potential of radio-nuclide release during any accident which may breach the waste c'ontainer, (2) lessen the potential for or extent of pressuriza-tion of the waste container due to hydrolysis and radiolysis, (3) reduce the potential of any criticality event taking place in the waste, (4) decrease the potential for internal corrosion of the waste container, and (5) reduce the potential for dissolu-tion or leaching of the wastes.
Combustibles. All solid or solidified radioactive waste classified as combustible shall be incinerated or otherwise reduced to a non-combustible " ash" which shall be fixed in a stabilizing matrix, or the original combustible wastes and their associated packaging shall be such that:
(1) the incidence of a fire and/or explosion involving the wastes shall not unduly affect the health and safety of the reposi-tory operating personnel, and
-(2)_ a fire and/or explosion involving a single container cannot i
migrate to involve'other containers, and k
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15 (3) the organic content of the material shall be shown to be l
incapable of interacting with the available radionuclides in such a manner as to fom organic complexes which may make those radionuclides chemically mobile.
l If combustible wastes are to be allcwed in thi apository at all, the consequences of any potential fire and/or explosion event upon the health and safety of the operating presonnel is of para-mount importance and must be within acceptable limits (i.e.,10 CFR 20, etc.).
Combustible wastes also present the potential for two other i
distinctly different detrimental effects which must be avoided.
i First, problems should be avoided which could arise from a fire or explcsion taking place within the waste which could cause an j
uncontrolled release of radionuclides, volatilize parts of the l
waste, or spread the fire to involve other nearby wastes.
The second major effect is the potential for organic compounds within the vaste to be altered by the chemical, thermal and i
radiation environments and subsequently fom compounds which i
could react with the wastes or the radionuclides and make them l
more mobile.
j Exolosive, Pyrophoric, and Toxic Materials.
There shall be 1
no known explosive or pyrophoric materials or conditions existing in the radioactive waste, nor shall there be any gross quantities of highly toxic chemical wastes. The exclusion of explosive and phyrophoric materials is a straight forward requirement for i
safety reasons.
The exclusion of toxic materials is a more difficult problem.
i The Staff's intention is that the wastes not contain gross quantities of highly toxic chemicals such as cyanides and other l
similar material which could become an occupational safety problem if the container were to be breached by an accident 2
during the operational phase of the repository.
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Chemical, Thermal and Radiological Stability. The waste fom and its associated packaging shall be chemically non-reactive i
to the maximum extent practicable (including leachability and solubility) when exposed to the emplacement environment. The stability of the waste package shall be analyzed under expected in situ conditions considering the characteristics and properties j
of the waste fom itself, the primary containment and subsequent i
packaging, the emplacement medium, the surrounding groundwater and the radiation and themal fields.
. Phase changes, loss of mechanical integrity, enhanced chemi-cal activity, and offgassing are only a few examples of condi-tions which could adversely' affect the ability of the waste form i
16 and package to isolate the waste and must be avoided.
The waste fem and its associated packaging must be mutually compatible and, to the extent possible, compatible with both the emplacement environment and the repository design in order to provide substantial radionuclide isolation.
Container Design Recuirement Physical Dimensions and Weight.
The physical dimensions and weight of the container and its contents shall be such that hand-ling of the material at the repository can be performed safely with a minimum potential for handling accidents. Addi tional consideration should be given to the techniques and equipment required to retrieve the waste container should that become necessary for safety reasons. The maximum allcwable weight is controlled by the hoisting capacity of the lifts used to transport the waste container from the repository surface facilities to the emplacement horizon and by the carrying capacity of the empl acement equipment itsel f.
The container size and shape are limited by the physical layout and dimensions of the emplacement rocms and corridors.
Mechanical Strength, Heat Resistance and Fabrication.
The con-tainer and packaging shall be designed and fabricated to the specification of acceptable codes and standards ( ASME Boiler and Pressure Vessel Code, ANSI, ASTM, D0T, etc.) where they are applicable to existing containers of a similar type and function.
The mechanical ability of the waste container to survive an accidental drop as well as routine emplacement operations and the stresses of retrieval operations without loss of its design per-fomance capability is important to the protection of the public and operational personnel alike.
Further, since corrosion rates are strongly temperature dependent, the temperature at the ccntainer surface is an important design parameter.
The heat fluxes and temperatures can be controlled by limiting the amount, geometry, and age of nuclear waste in the container.
The Staff feels that existing design and fabrication ccdes and standards developed by the ASTM, the ASME and other similar standards organizations for other purposes, may well contain areas which could be appropriate for these waste containers.
Materials of Construction and Corrosion Control.
The materials used to fabricate the container and packaging shall meet the specifications of acceptable codes (10 CFR 71.31, ANSI, ASTM, ASME, etc.) where they are applicable to existing con-tainers of a similar type and function.
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3 17 The primary failure mode of the enplaced container is expected to be corrosion. The requirement for retrievability i
severely constrains the selection of container materials.
Appropriate packaging materials shall be selected to assure compatibility with the container material and the waste fom, thereby reducing the potential for corrosion.
Some design may rely on corrosion to fom in::aloble, impermeable barriers.
j However, care must be taken *) avoid situations where the corrosion products could adversely affect the geochemical retardation of the radionuclide transport in the medium, increase the corrosion rate of the packaging material, or increase the dissolution rate of the waste form in the groundwater. Any of these situations could significantly degrade the waste isola, tion perfomance of the repository.
Mechanical Handling. The waste container shall include features and devices which enhance the capabilities for safely j
lifting and moving the container and its contents. These j
features and devices should be in compliance with 10 CFR 71 r
Subpart C and 10 CFR 71.31 such that they do not provide a means 4
for easily damaging the container should a handling accident occur.
i The case of handling of the waste container during transport and emplacement is important to the safety of the operations per-sonnel should an accident take place resulting in the breach of the container and release of the radionuclides.
The use of appropriate lifting and handling devices and features could reduce the potential for handling accidents. However, these devices and features must be such that they themselves do not increase the potential for damage to the container during such a handling accident.
Criticality Control. The maximum allowable quantity of fissile material in individual waste container shall meet the requirements of criticality safety established in 10 CFR 71, 49 CFR 173 and other applicable standards. Calculations and measurements made by the shipper to assure criticality safety shall be available to the repository operator upon request. The possibility of a criticality accident in the repository opera-tions area must be avoided. This may be done by limiting the total inventory of fissile materials in each container, the room itself, or by using neutron poisons, criticality safe package i
geometries and enplacement arrangements.
Surface Contamination. There shall be no significant i
removable radioactive surface contamination on the exterior of the container.. Removable (non-fixed) radioactive contamination is considered significant if the level of contamination exceeds the requirements of 49 CFR 173.397.
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18 Penetrating Radiation. The penetrating radiation level from an unshielded container shall be controlled to provide for the health and safety of the general public and the operational personnel at the repository. The provisions and requirements of 10 CFR 20,10 CFR 71, applicable D0T standards, and other applicable codes and standards shall be followed.
Maximum Ccntainer Wall Temocratures.
The maximum allowable centainer wall temperature shall not exceed a level at which either the chemical activity of the waste or interactions between the container and the enplacement enviroment would impair the gbility of the waste package to perfom ijs task (i.e.,
100 C). At above about a temperature of 100 C the physical and chemical mechanisms involved in leaching or mass dissolutioning of the components of the waste package are not well understood. Unless a better understanding of these phencmenona and their controlling mechanisms is gained in the near fut fre the Staff feels that the maximum wall temperatures must be limited to a conservative value.
Unicue Identification. A label or other means of identi-fication snall be installed on each container.
The label shall not impair the integrity of the container or make the surface irregular and shall be attached in such a manner that labels and descriptions thereon will be legible at least to the end of the retrievable storage period.
Each label shall contain specific infomation considered necessary for identification and tracea-bility of the packages in addition to a unique serial number for reference to a set of permanent records which contain detailed infomation regarding the specific packages.
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