ML19347F368
| ML19347F368 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1981 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML19347F361 | List: |
| References | |
| FRN-46FR13971, REF-10CFR9.7, RULE-PR-60 SECY-81-267, NUDOCS 8105190082 | |
| Download: ML19347F368 (81) | |
Text
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0 RATIONALE FOR PERFORMANCE OBJECTIVES AND REQUIRED CHARACTERISTICS OF THE GEOLCGIC SETTING:
TECHNICAL CRITERIA FOR SEGULATING GEOLOGIC DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTE Division of Waste Management April 1981 U.S. NRC-Enclosure J 8105190 0N
4/30/81 Alexander 5/e Table of Contents Page Introd Jcti o n...........,..............
1 Sa'. action of the Regulatory Approe:h..............
8 Selection of the Major Barriers 16 Major Barrier Performance 26 Retrievability.........................
46 References...........................
54 Tables...................
53 Figures 58 Glossary............................
73 f
l l
1 Enclosure J
5 s.
4/30/81 Alexander Ele Li_st of Figures Ficure g
1.
Vertical Temperature Distribution in Typical Bedded Salt Formation (ref.
2)..........................................
2.
Vertical Temperature Distribution in Typical Granite Formation (ref.
2)..........................................
3.
Vertical Temperature Distribution in Typical Salt Dome (ref.
2)....................................................
4.
Maximum Repository Temperature Versus Time in Three Geologic Formations or Reprocessec waste (ref.
2)...........
5.
Maximum Repository Temperature versus Time for 10, 20, and 50-Year-01d Waste in Bedded Sal t (ref. 2)..................
6.
Maximum Repository Temperature Versus Time for 10- and 20-Year Old Waste in Granite (ref.2)..._......................
7.
Vertical Temperature Distribution in Typical Granite Formation (ref.
2)..........................................
8.
Effect of Fuel Cycle on Maximum Recository Temperature Versus Time for 10 year-old Waste,in Gedded Salt Formation (ref.
2)....................................................
9.
Temperature Rise Versus Radius at Canister Mid-Pl ee for a Repository in Domed Salt (ref.
2).........................
10.
Radioactivity as a Function of Decay Time for High-Level Waste from PWR Throwaway Cycle (ref.
3).....................
11.
Radioactivity as a Funct b of Decay Time for Reprocessed High-Level Waste from PWR Uranium Recycle (ref.1)...........
12.
Radioactivity as a Function of Decay Time for Reprocessed High-level Waste from PWR Mixed 0xide Recycle (ref.1).......
13.
Decay Heat as a function of Time for High-Level Waste from PWR Throwaway Fuel Cycle (ref.
1)...........................
14.
Decay Heat as a Functior of Time for Reprocessed High-Level Waste from PWR Uranium Recycle (ref.
1).....................
15.
Decay Heat as a Function of Time for Reprocessed High-level Waste from PWR Mixed Oxide Recycle (ref.
1).............
ii Enclosure J
e 4/30/81 1
Alexander 5/A RATIONALE FOR PERFORMANCE OBJECTIVES AND REQUIRED CHARACTERISTICS OF THE GEOLOGIC SETTIFG I.
Inti action.
High-level radioactive waste (HLW) is a byproduct of the irradiation of nuclear fuel in nuclear reactors.
In the United States, commercial nuclear reactors are principally light water reactors, whcse fuel consists of pellets of UO.
The 2
uranium is a mixture of isotopes that is mostly U-238, but includes about 2 to 3 percent U-235 ud trace amounts of other isotopes of uranittm.
During irradiation, the U-235 vissions and releases energy.
During irradiation, some of the U-238 is converted to Pu-239, which, like U-235, can fission and release energy.
After the fuel has been removed from the reactor, the Pu-239 and remaining uranium can be removed (fuel reprocessing) and recycled.
This has led to the conceptual development of several fuel cycles.
Three fuel cycles have been considered by the Environmental Protection Agency (EPA) (1979a) in developing its high level waste standard and the Department of Energy (00E) (1979) in its draft Generic Environmental Impact Statement (GEIS) on Commercial Waste i
Management.
1.
The " throwaway" cycle in which low-enrichtd uranium as U0 is irradiated 2
in a light water reactor (LWR), with direct disposal of the spent fuel as waste; 4
Enclosure J
4/30/81 2
Alexander 5/A 2.
Uranium-only recycle in which low-enriched UO is irradiated in an LWR, 2
toe spent fuel is reprocessed, the recovered uranium is recycled, and the plutonium is stored; i
3.
Mixed-oxide recycle in which mixed Pu0 and low-enriched 00 is irradi-2 2
ated in an LWR, the spent fuel is reprocessed and the recovered uranium and plutonium are recycled.
In its final GEIS, DOE (1980) deleted the uranium only recycle case because of the low likelihood it would ever be implemented.
Storch and Prince (1979) discuss several other potential fuel cycles that are found in the literature, but are not being commercially developed at present.
The different fuel cycles are significant because they result in different waste products.
The EPA, to develop its standard, has been considering the threwaway, uranium-only recycle, and mixed-oxide fuel cycles for characterizing radioc:tive waste.
The NRC staff has also used these same cycles, since we consider their wastes have characteristics that will bound these of wastes from any fuel cycle likely to be commercially developed.
The more important nuclides in radioactive waste result either from fission or from neutron capture in actinide isotopes.
Both processes occur during irradiation of the fuel in the reactor. When the spent fuel is removed from the reactor, it consists principally of fission products and actinides in addition to some activation prelucte in the fuel assembly 5tructures and fuel Enc'esure J
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Alexander 5/A
. cladding (00E,,1979; AOL, 1979a).
Each has certain general characteristics
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that are described in ADL (1979a, 1979c) and 00E (1979, 1980).
For the most part, the fission products are relatively mobila in the geologic environment, have short half-lives, and high specific activities.
In consa" quence of their high specific activities, the fission products generate heat at a significant rate.
In contrast, the actinides tend to be relatively immo-bile, have long half-lives, and lower specific activities.
In addition, thay generate heat at a lower rate than the fission products and, like uranium ore,
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they emit primarily alpha radiation.
In the throwaway cycle, the spent fuel assemblies are the waste.
The uranium-only recycle'and the mixed-oxide fuel cycle both involve reprocessing of the spent fuel.
That operation is generally carried out by the PUREX process, which produces a nitric acid solution containing the fission product:: and various amounts of actinides as a wasta stream.
This waste can be solidified before final disposal and is called reprocessed high-level waste.
High level radioactive wastes from these fuel cycle will need to be disposed of in a manner that does not represent a hazard to public hetith and safety.
Three Federal agencies have major roles in the national program for disposal of high-level radioactive wastes.
The EPA is responsible for establishing generally applicable environmental standards for protection of the general environment from radioactive matenal.
The standards apply to all uses of radioactive materials, including disposal of high-level wastes.
The ~)epartment Enclosure J
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4/30/81-4 Alexander 5/A
-ofEnergyhastheresponsibilitytode.elopthetechnolohandtoselectthe i
sites for safe disposal of high level wastes.
The Nuclear Regulatory Commission (NRC) is responsible for developing technical criteria to be used in implementing
^
EPA's standard.
The technical criteria that the NRC proposes includes the following performance objectives and required characteristics of the geologic
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setting:
660.111 Performance cbjectives.
(a) Performance of the geoloafe repository operations area through l
permanent closure.
(1) Protection against radiation exposures and releases of radio-logical material.
The geologic repository operations area shall be l
designed so that until permanent closure has been completed, radiation
, exposures and radiation levels, and releasss of radioactive materials I
to unrestricted areas, will at all times be maintained within the limits specified in Part 20 of this Chapter and any generally applicable envi-ronmental standards established by the Environmental Protection Agency.
l (2) Retrievability of waste.
The geologic repository operations area shall be designed so that the entire inventory of waste could be retrieved on a reasonable schedule, starting at any time up to 50 years after waste emplacement operations are complete.
A reasonable schedule for retrieval is one that requires no longer than about the same overall period of time than was devoted to the construction of the geologic repository operations area and tha emplacement of wastes.
(b) Performance of the geologic repository after permanent closure.
(1) Overall system performance.
The geologic setting shall be selected and the subsurface facility designed so as to assure that releases of radioactive materials from the geologic repository following l
permanent closure conform to such gene ~ ally applicable environmental standards as may have been established by the Environmental Protection Agency.
(2) Performance of the engineered system.
(i) Containment of HLW.
The engineered system shall be designed so that even if full or partial saturation of the underground facility were to occur, and assuming anticipated processes and events, the waste packages will contain all radionuclides for the first 1,000 years after permanent closure and for as long thereafter as is rearanably achievable.
1 i
Enclosure J
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4/30/81 5
Alexander 5/A This requirement does not apply to TRU waste unless TR'U waste is emplaced close enough to HLW that the TRU release rate can be significantly affected by the heat generated by the HLV.
(ii) Control of releases.
_ u.
(A) For HLW, the engineered system shall be designed so that, after the first 1,000 years following permanent closure, the rat'e of release of radionuclides from the underground facility is as low as is reasonably achievable. As a minimum, the design shall provide that the annual
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release of any radionuclide does not exceed one part in 100,000 of the maximum amount of that radionuclide calculated to be present in the underground facility (assuming no release from the underground facility) at any r.ime after 1,000 years following permanent closure.
(B) For TRU waste, the engineered system shall be designed so that following permanent closure the rate of release of radionuclides from the underground facility is as low as is reasonably achievable.
As a minimum, the design shall provide that the annual release of any radionuclide does not exceed one part in 100,000 of the maximum amount criculated to be present in the underground facility (assuming no telease from the under-ground facility) at the time of permanent closure.
(3) Performance of the geologic setting.
(1) Containment period.
During the containment period, the. geologic setting shall mitigate the impacts of premature failure of the engineered system.
The ability of the geologic setting to isolate wastes during the isolation period, in accordance with paragraph (b)(3)(ii) of this section, shall be deemed to satisfy this requirement.
(ii)
Isolation ceriod.
Following the containment period, the geologic setting, in conjunction E th the engineered system as long as that system is expected to function, and alone thereafter, shall be capable of isolat-ing radioactive waste so that transport of radionuclides to the accessible environment shall be in amounts and concentrations that conform to such generally applicable environmental standards as may have been established by the Environmental Protection Agency and thereby will not result in cignificant doses to any individual.
For the purposes of this paragraph, the evo'iution of the site shall be based upon the assumption that those processes operating on the site are those which have been operating on it during the Quaternary Period, with perturbations caused by the presence of emplaced radioactive wastes superimposed thereon.
660.112 Required characteristics of the geologic setting.
(a) The geologic setting shall have exhibited structural and tectonic stability since' the start of the Quaternary Period.
Enclosure J
4/30/81 6
Alexander 5/A (b) The geologic setting shall have exhibited hydrogeologic, geo-chemical, and geomorphic stability since the start of the Quaternary Period.
(c) The geologic repository shall be located so that pre-waste-- --
emplacement groundvater travel times through the far field to the accessible environetnt are at least 1,000 years.
INTENT The design of a geologic repository for disposal of high level wastes must pro-vide for protection of public health and safety during two periods:
(1) during repository operations, when the prirmipal concern involves exposure of operators or releases of radioactive materials to unrestricted areas in liquid or gaseous effluents, and (2) after repository sealing, when the principal concern 17volves long-term migration through groundwaters to the accessible environment.
The performance objectives of 10 CFR 60.111 apply to both of these periods.
(1) Releases during operation Several operations will be needed prior to disposal of waste in the geologic repositorj.
These operations could include storage of materials, chemical ceparatic.9, solidification of waste, packaging and emplacement.
Each of these operations could result in releases to the operations area or to the general environment.
The EPA is preparing a standard that will limit ambient levels of radioactive materials resulting fret caeration of a high level waste disposal facility.
Part 20 of the Commission's regulations wi;l implement this standard.
Part 20 alreacy contains limits for doses to the operating personnel.
The performance objective s960.111(a)(1) requires that the geologic repository operations area Enclosure J
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Alexander 5/A bo designed to operate within the Ifmits of Part 2G. -Since the problems likely to be encountered during the operational phase will be similar to those dealt
_ with in similar radioactive materials handling facilities, no sp'ecial performance objectives in addition to those contained in Part 20 are considered to be necessary.
(2) Releases after decommissioninq The objective of waste disposal is to isolate the waste from the envircnment for a long period of time.
The EPA is preparing a standard that w1il set limits on ambient levels of radioactivity in the general envi onment from any disposal system. While the EPA standard has not yet been published in croposed form, we expect that it will require quantities of radioactive materials released to the environment over a long period of time to be limited to very small amounts.
The performance objective gg60.111(a)(2) requires that the geologic repository be designed in a way that provides reasonabic assurance that ambient levels of radioactive materials will be within limits that EPA may establish.
Disposal of radioactive waste in a manner that will assure safety for many thousands of years represents a unique problem not previously dealt with in other NRC or EPA standards.
The NRC staff has considered several performance objectives to address this unusual regulatory problem.
The remainder of this thapter provides the technical bases for the performance objactives selseted as well as evaluation of alternatives considered for siting and design of the repository to assure effective long-t 'mn isolation of the wastes.
Section II discusses the alternatives considered in selecting a regulatory approach and Enclosure J
4/30/81 8
Alexander 5/A the rationale for the approach selected.
Section III describes the alternetives considered for the major barriers of the waste isolation system and the rationale-for the required barriers.
Section IV describes alternatives considered for the specific performaace objectives for the major barriers:and required character-istics of the geologic setting and provides the basis for the numerical valves that were selected.
Section V describes the rationale for requiring the repository to be designed so that the option to retrieve the wastas is p eseryed_and.gives the basis for the numerical value selected for the period this option shall be presirved.
II.
Selection of the Regulatory Approach A.
Need for Numarical Models While the EPA stan'dard has not beon cast in its final form, its icplementation will require quantitative predictions of radionuclide releases to the general environment.
NAS (1979) notes that this can be done only through the use of numerical modeling of the repository system because of the very long tima frame involved.
The IRG (1979) also concluded that the degree of long term isolation provided by a repository can only be assessed through analytical modeling.
Predictive modeling of the repository will require that postulated releases be traced from the deeply buried waste through the geologic and hydrologic environ-ment to those parts of the general environment thJ are accessible by people.
Thus, the procedura for repository evaluation will involve the determination of release scenarios, characterization of the geologic and hydrologic environment, Enclosure J
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Alexander 5/A and nue rical modeling of the many physical and chemical processes involved in
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the release and transport of radionuclides.
B.
Sources of Uncartainty in Numerical Models
~ _ _ _ _ _ _... _
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- Bredehoeft and others (1978), IRG (1979), NAS (1979), Craig (1979), Da'vis (1980),
and many others have each noted uncertainties associated with at least one of the steps of this procedare for repository evaluation.
The use of numerical modeling methods introduces errors and uncertainties through the use of approxi-mative techniques, undiscovered logic errors in complex comouter coces, and undiscovered errors in algorithms.
Bredehoeft and others, LBL (1979), and Davis discuss a second contribution to overall uncertainty:
uncertainties that are attendant to site charactertiation.
Davis points out that uncertainties in the methods used to determine data and uncertainties associated with undetected features will contribute to the overall uncertainty in repository performance.
A third contribution to the overall uncertainty arises from the uncertainties that exist in our understanding of the basic physical processes from which the release scenarios that form the basis for the evaluation of performance are to be constructed (Davis, 1980).
In that connection, Bredehoeft and others discuss the complex perturbations on the geologic and hydrologic environment caused by the presence of the waste and the repository.
DOE (1979) also discusses the uncertainties associated with waste-rock interactions and notes that it is a major area of concern.
Enclosure J
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Bredehoeft and others point out an additional area of concern:
the potential for unanticipated interactions in complex systems.
They observe that unantici-pated interactions have occurred in many engineering systems whose components were tnought to be well characterized.
They also observe that-other-investigators-------
have argued that because of the complexity of identiff ed and unidentified poss-ible interactions between processes in the earth sciences, long-term prediction
^
is unreliable and impossible to perform with high colifidence.
Taken together, the uncertainties in site characterization, in basic physical processes, and in the possible interactions in complex systems, suggest that the evaluation of repository performance will be subject to considerable uncertainty.
In view of the above, IRG (1979) recommended that the EPA recognize the large range of inherent uncertainty involved in determining the performance of waste management systems and permit the NRC to account for it in its implementation and licensing process.
In principle, uncertainties in the numerical models, uncertainties in characterization of the site and engineered elements, and uncer-tainties in basic physical processes can be estimated and bounded.
Therefore, it might be possible to account for them directly in determining whether the EPA standard will be met.
However, a direct accounting of uncertainties has not been done in any modeling to date.
The potential for unanticipated inter-actions and occurrences cannot be bounded even in principle.
P Enclosure J wi
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4/10/81 11 Alexander 5/A Compinsation for uncertainty that weuld ctherwise confou~nd adequate' demonstration of compliance with the EPA standard is an essential part of the NRC staff's regu-latory approach.
Since any licensing proceeding will in'volve~the question of adequately demonstrating compliance with an EPA standard, the NRC staff has placed primary emphasis on selecting approaches to facilitate resolution of this issue.
Other approaches which might prove useful in the long run, but which are hard to demonstrate compliance has been achieved and could involve years of delay in a licensing proceeding, have been de-esphasized.
C.
Alternatives Three alternative approaches to regulating geologic disposal of HLW were considered in the development of the technical criteria of 10 CFR Part 60.
Each was examined in light of its ability to compensate for the major undertainties in the quantitative prediction of the performance of geologic disposal.
The alternatives considered were:
1.
Regulation of repository systems by setting a single overall performance standard that must be met by the system.
The performance standard in this case would be the EPA standard; 2.
Regulation of repository systems by setting minimum performance standards for each of the major system elements as well as requiring the overall system to conform to the EPA standard ; and Enclosure J
4/30/81-12 Alexander 5/A 3.
Regulation of repository systems by setting numerl criteria ca critical engineering attributes of the system.
The NRC staff has examined each of these alternatives from the standpoint of its ability to compensate for uncertainty ir evaluating compliance with the EPA standard in a licensing proceeding.
The NRC staff further examined each alternative with two objectives in mind:
(1) providing as much guidance and detail as may be warranted by generic considerations; and (2) avoiding undue constraints upon system design.
The alternative of setting a single system parformance standard f s often referred to as the " systems appreach."
It has as its principal advantage the fact that regulation would be through a single figure of merit, overall system performance.
This leaves maximum flexibility for the designer to make trade-offs among compo-nents of the system.
The systess approach can include the concept of multiple barrier design to compen: ate for uncertainty in overall system performance (see for example 00E, 1979).
Unfortunately, the systems approach as interpreted above is not practical from a regulatory point of view. As noted earlier, a quantitative assessment of the expected performance of a geologic repository is a complex and difficult task.
The results of such an assessment contain the uncertainties described above.
Compensatinn for uncertainties can be achieved, however, without imposing ancillary requirements on the systems approach by introducing conservatism.
Either the measure of performance can be made more stringent than is truly needed, Enclosure J
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4/.30/81.
13 Alexander 5/A or the metnod of evaluation, or both.
Unfortunately, estimates containing worst case and bounding assumptions necessarily depart from reality and therefore become less certain than the already uncertain realistic estimate.
Further~,
conservative standards need a reasonably precise realistic measure against which to be conservative.
Hence, neither method of introducing conservatism affords a very clear picture of just how much conservatism has been introduced.
Hence neither method gives a very clear picture of how much gain in confidence should be realized from that conservatism.
The second alternative establishes major subelements of the repository system, called barriers, and assigns minimum performance objectives to cach while maintaining the EPA standard as the measure of overall system performance.
This alternative has two advantages over the systems approach.
First, if the barriers are ' chosen judicicusly, the uncertainty in the evaluation of repository performance can be reduced by requiring the barriers to perform in ways which redcce their relative contribution to the uncertainty.
Second, by judicious choice again, multiple barriers can be prescribed which act independently and thereby enhance confidence that the wastes will be isolated.
As is discussed in subsequent sections, the NRC staff has secured these advantages through performance objectives which 1) serve to reduce the effective source term for the repository evaluation, using reasonably verifiable engineering methods thereby r~ lucing calculational uncertainty, and 2) independently provide confidence that the wastes will not reach the environment during the period when they present the greatest hazard.
Enclosure J
4/30/81 14 Alexander 5/A An additional benefit follows t,cm the establishment of major barriers and their associated performance cbjectives.
Once the barriers and objective are known, that knowledge can be used by the DOE as input to the design of the repository.
Design issues related to repository performance can be' addressed early on,' reducing the potential for major design changes resulting from the licensing process.
Yet since only the major subsystems and their performance are specified, design flexibility is retained.
The third alternative, use of numerical criteria for certain engineering attributes of the system (a peak canister wall temperature, for example) has two major advantages.
It would provide clear guidance to designers as to exactly what is required for licensing.
Secondly, the criteria can be selected to compensate directly for uncertainty by introducing conservation into the acceptable levels for each significant attribute of the system.
The approach also has several disadvantages.
Of the three alternatives, it is most restrictive of design flexibility.
In fact, it begins to force the regu-lator into a designer role.
In addition, criteria must be set on the basis of existing knowledge to be effective.
Therefore, the approach cannot fully accom-modate the benefits of future research and development work.
During the development of its regulatory approach, the NRC staff received peer comments from two workshops in addition to the public comments received on the Advance Notice of Proposed Rulemaking published in the Federal Register on May 13, 1
1980 (46 FR 13971).
Tha workshops were sponscred by the Keystone Center for Enclosure J
b 4/30/81 15 Alexander 5/A Continuing Education (see Craig,1979) and the University of Arizona (see Davis, 1980).
The participant at both workshops supported the use of minimum performance standards fo barriers.
The Keystone group emphasized the importance of multiple barriers-to compensate for gaps in understanding of the response of deep geologic forma-tions to the disposal of high-level radioactive waste.
The Keystone group further noted that minimum performance objectives can reasonably be placed separately on the major parts of the system.
The University of Arizona workshop also supported an approach based on minimum performance objectives.
In its report, the workshop stated that the multibarrier concept and common sense approach to the establishment of performance objectives is a practical way to achieve a viable regulation.
t Several commenters on the Advance Notice of Proposed Rulemakinc pointed to,the strength of the systems approach as interpreted in alternative one as providing the designer with flexibility to make trade-offs between system elements, provided that the overall performance standard is ret.
These commenters argued that a regulatory approach based on minimum performance standards for individual barriers is unnecessarily restrictive.
To the contrary, however, the United States Geological Survey (USGS, 1930)
I commented as follows.:
"In particular we believe that section 60.111(c), Performance of Required Barriers and Engineered Systems, represents a sound approach to licensing.
It is sometimes stated that only the performance of the total waste isolation system is relevant to licensing and per-formance requirements.
But assessing the total system, whether by models or some other approach, is an extremely complex undertaking Enclosure J
4/30/81 16 Alexander 5/A 4
eabject to considerable uncertainty as the supplementary information points out.
By requiring each major element in the waste isolation system to independently meet certain performance objectives, the proposed rule breaks the problem down into more Lanegeable parts and allows for uncertainties in the performance of some components."
No commenters supported the third alternative, the use of numerical criteria for engineering attributes of the system.
The NRC staff considers that alternative two, based on minimum per.ormance standards, achieves the test balance between the need to compensate for uncertainties in demonstrating Coppliance with the EPA standard in the licensing process and the need to preserve flexibility for the designer.
III.
Selection of the Major Barriers The staff considered three alternatives for regulation of tne system of barriers:
1.
Rely entirely on the natural barriers of the site to meet the system perform-ance standard; 2.
Rely entirely on engineered barriers to meet the system performance standard; and 3.
Rely on a combination of engineered and natural barriers to meet the system performance standard.
In considering the alternatives the staff gave particular emphasis to reducing or limiting uncertainty in assessment of system performance over the long term.
Enclosure J
4/3 0/81 17 Alexander 5/A The first alternative, that of total reliance on the site for isolation, has a straightforward rationale:
if a geologic formation has been undf sturbed for
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many millions of years, there is reason to believe that it will remain undisturbed into the future even if it is mildly disturbed by placing waste in it.
In principle, isolation by the geologic medium can be accomplished by placing the waste at depth in a tectonically, hydrologically and mechanically stable medium that is essentially free of and isolated from mebile groundwater.
In addition, the medium would be capaole of absorbing radiation and diffusing heat without impairing the integrity of the formation.
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Large areas of the North American continent have been tectonically stable for millions of years.
Moreover, some of these stable areas contain geologically old rock salt formations that are deep enough and thick enough to be able to hust a gaologic repository.
Because salt is highly soluble in water, the very existence of an old salt formation suggests that it has been isolated from mobile gfoundwater for a long time. Other rock formations such as shale or granite can have low enough intrinsic permeabilities to be able to host a repository and can also be found in these stable areas.
Certain natural analogues also seem to suggest that geologic format'sas can be found cnat can effectively isolate the waste.
Most uranium ora deposits in the United States were formed many millions of years ago at sites having peculiar geologic and geochemical conditions.
During the time since ' formation, radio-nuclides from these deposits have dispersed only very slowly.
Although high-level radioactive waste contains many radionuclides that are not found in uranium Enclosure J 1
M30/81 18 Alexander 5/A ore, another natural analogue, at Oki', in Gabon, West' Africa, suggests that o
such additional radionuclides might be isolated by the geologic medium (Cowan,1976).
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About 1.8 billion years ago, criticality occurred in a uranium ore deposit at Oklo.
Fission of U-235 in this natural reactor produced the full spectrum of radionuclides found in high-level radioactive waste. With the exception of some of the fission products, little migration of radionuclides appears to have occurred even though the reaction stopped more than a billion years ago.
These are the types of considerations that originally led to consideration of geologic dispcssi for permanent isolation of high-level wastes.
There are two major uncertainties involved 'with adopting alternative one for regulation of the system of barriers, (1) construction of the repository and emplacement of the wastes disturos the natural systems in a number of ways that are difficult to evaluate and that have the potential to compromise the ability of the site to isolate wastes; and (2) our ability to characterize and rigorously predict the performance of the large regional hydrologic a:id geologic systems depended on for isolation is relatively limited.
Bredehoeft and others (1978) observe that perturbations resulting from the emplacement of waste will affect the host rock anu included water for a long time.
They identify three distinct types of perturbation:
(1) stress and mechanical effects from excavation, (2) chemical effects frem changes to the chemical equilibrium by adding the waste, and (3) thermal effects from the decay heat generated by the emplaced waste.
Enclosure J
4/30/81 19 Alavander 5/A The hydrologic flow at a site can be affected by the following:
(1) creation
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and circulation of convection cells; (2) induced pressure gradients; (3) decreased viscosity; (4) changes in the bulk permeability of the rocks; (5) chan~ges in the rock stress field; and (6) changes to solubility characteristics.
Each of these will be affected by the thermal load that will result from waste emplacement-(Bredehoeft and others, 1978; DOE, 1979, 1980).
A number of chemical reactions can occur that are significant with respect to repository performance:
hydration, dehydration, formation of strong acids, sorption, solution, buffering reactions, reactions that produce gasses, reactions that prt, duce volume changes, and the formation of concentrated brines in salt.
Each of these reactions will be aff,ected by the temperature at which it occurs.
Heat breaks down hydrated minerals to release water and hot moving fluids may alter existing minerals, causing changes in permeability.
In additie,n to the uncertainties that arise from perturbations of the site by esp'.acement of waste, there will also be uncertainties in the characteristics of the site.
The characterization of geolog1c parameters can be a difficult task.
Questions arise regarding the transferability of data from one site to another.
Considerable difficul.y often arises in characterizing and quantifying important geologic conditions or features.
t With respect to geomechanical characterization, there are limitations in testing i
and exploration technology.
In characterizing the thermal and mechanical response l
of the rock mass, the rock's ther.nomechanical properties, time-dependent properties, Enclosure J
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the distribution and the influence of fractures, potential for movement of gaseous l
-or liquid inclusions, determination of in situ stress, the validation of laboratory -
and in situ experiments, and the development of instrumentation for monitoring I
present particular difficulty (Wawersik,1978).
In the groundwater system, field techniques for measuring and characterizing important parameters such as hydraulic conductivities in tight rocks, dispersion, i
l and fracture flow are neither well developed nor well understood, IEC (1979),
Golder (1977).
(
Sensitivity analyses done by Heckman and others (1979) show that the geochemical system is a critical site component in isolating waste.
However, as discussed by Isherwood (1978), it is the least understood.
A wide disparity in our knowledge of the geochemistry of radionuclides exists, EPA ad hoc (1978).
There are some 30 to 45 sign.ficant radionuclide isotopes in spent fuel or high-level waste, Cloninger (1979), Heckman (1979), ADL (1979c). Of these radionuclides, the majcr potential contributors to radiological dose (under more favorable conditions) appear to be Tc-99, I-129, C-14, Np-237, and Ra-226, (Hill, 1979; ADL, 1979c).
Under less favorable conditions where path length is short, groundwater velocity high, or sorption low, other nuclides such as Sr-90, Sn-126, U-234, Pu-239, Pu-240, Am-243, and Cm-245 have been identified as being potentially significant contributors to dose.
4 l
The applicability of data obtained in laboratory experiments over short times and using small sample sizes to geologic situations over long time periods and l
Enclosure J
4/30/81, 21 Alexander 5/A
- path lengths of kilometers has not been demonstrated, Serne (1977).
Addition-ally, little work has been done regarding retardation at elevated' temperature.
Thus, little work is applicable to the disturbed zone of a repository.
In the field, the problem is compounded by having to define the behavior ~of a number ~
of nuclides, each with respect to a number of rocks, each of which must, in f
turn, be taken in context of different groundwater chemistries.
Significant variations in measured sorption in the same rock, but taken from different depths in boreholes, have been reported (Erdal ard others, 1978).
Members of the NRC staff visited five national laboratories during June,1980 (see Robbins and others, 1980).
During these visits, the staff investigated experimental work on geochemical retardation and found little progress in reducing uncertainties in this area. While it is likely that geochemical retardation will contribute to waste isolation, the magnitude of its contribution will be dificult to quantify now or in the foreseeable future, based on current DOE programs.
In summary, emplacement of the waste will modify the mechanical, hydrological, and chemical properties of the host rock through a variety of phenomena, some which are as yet not entirely understood, and some which the data do not now exist to adequately describe their effects under expected repository conditions.
Also, which methods sheuld be used to obtain the site specific data needed to assess the ability of a site to isolate wastes is still a point of discussion within the scientific community.
Thus, the NRC staff has concluded that the undertainties associated with the prediction of site performance are likely to Enclosure J
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4/30181 22 Alexander 5/A
~~ be so great that it would be difficult to conclude a licensing prcceeding, and that independent criteria are needed to allow the Commission to find that ther is reasonable assurance that the health and safety of this public are protected.
The second alternative, that of ralying on engineered barriers alone to meet the overall systets performance objectives, has the advantage of being less dependent on site related characteristics, with their associated uncertainties.
Under this alternative, the waste disposal system would be designed to incorporate very leach resistant waste forms, high integrity packages capable of containing the wastes for long periods of time, sorptive backfills capable of retarding nuclide migration, and low permeability plugs and seals that prevent intrusion cf groundwater and release of radionuclides.
Engineered barriers have the advantage that materials can be selected and barriers can be designed to perform specific functions.
Once designed, prototypes can be set up and tested under conditions that provide increased assurance that the design objective will be met.
Finally, engineered barriers can be fabricated and emplaced under rigorous standards of quality assurance to increase confidence they will perform as designed.
On the other hand, with geologic systems, it l
is hard to know for sure what the system even is, since much of the information must be obtained indirectly.
In addition, it is difficult to characterize the properties of the geologic materials we are dealing with, even when they are i
accessible for testing.
To this is added the difficulty of predicting how natural systems will perform to isolate the waste.
Therefore, it is possible in theory Enclosure J
4/30/81 23 Alexander 5/A
-to obtain a relatively higher degree of confidence by rep 1~ acing the isolation capability of the site with reliance on engincored barriers.
It is impossible, however, to nake the engineered system enti' rely independent-of the si.te, since the site provides the environment in which the engineered system is constructed.
Further, there are also uncertainties involved with assessing the perrormance of the engineered barriers under conditions affected by the emplacement of the waste and over the ic9g periods required for isolation of high level wastes.
The third alternative, that of supplementing the isolation capability of the geology with engineered barriers that are designed te contain the waste for a period and then control the rate at which radionuclides are released, has the advantages of both the preceding alternatives.
In addition, the use of engineered barriers can provide the means of compensating for the uncertainty in our ability to assess isolation of the wastes by the site.
i Others who have considered the problem of geologic disposal have reached similar conclusions.
Bredehoeft and others (1978) point out that the waste l
form, the host rock, and the groundwater flow path provide potenti carriers.
AOL (1979b) suggests four principal barriers:
the waste foria, the container in which the waste form is packaged, the geologic envi- : ment, and adsorptive phenomena in the geologic environment.
Ringwood (1978) describes essentially the same barriers as AOL (1979b), but considers adsorptive phenomena to be a part of what he calls the geologic barrier.
i Enclosure J
. ~ -,
,yy
_,,,,,v,.-,.r-,-
,w,,,....-,,,._,.,_,_,_..#,,..., - _.. _ _. _,,..,,,, - -
s 4/30/81 24 Alexander 5/A The NRC staff finds that the physical nature of the prpblem lends itself to classificction of the major barriers as the waste package, the underground facility, and the geologic setting.
In this scheme the geologic setting is equivalent to Ringwood's geologic barrier and the waste package is equivalent to ADL's waste form and cortainer considered as a system.
In identifying. the underground facility as a potential major barrier, the NRC staff has recognized tne need for careful excavation of the repository to avoid creation of pathways to the biosphere and the potential for placement of additional engineered barriers in the underground excavation before sealing of the repository.
The uncertainties in evaluating the performance of the system caused by emplacemer1t of the waste are to a large degree time dependent.
Many of the perturbations that are expected to occur are the result of the increased temperature in the host rock due to radioactive decay heat.
Temperatures peak and begin to fall within the first few hundred years after the waste has been emplaced (ADL, 1979b).
During the same period total radioactivity of the waste decays by several orders of magnitude (AOL, 1979a).
As the temperature decreases, many of the uncertainties in near-field behavior decrease as well.
The decrease in total radioactivity represents a decrease in the source term available to be released as well.
Our approach, for this initial period of high temperatures and radionuclide inventcry is to contain the wastes within a corrosion resistant package that confines the radionuclides within a physkal boundary.
Such " waste packages" can be designed to provide assurance of their ability to perform to specifica-tions under anticipated near field conditiors.
Thus, this alternative provides Enclosura J
4/30/81 25 Alexander 5/A
. a reasonably verifiable barrier to compensate for geologic = uncertainty during Ithe period when the specific activity of the waste is hight and the perturbations of the natural systems are large.
~
Engineered barriers can also be designed to limit the rate at which radioactive materials are released from the engineered system after the containment' period and thereby supplement the geologic system in limiting the rate of release to the environment.
The rate at which radionuclides are released to the site can be limited by using waste forms and overpacks that limit releases from the package to some maximum rate; by emplacing materials (e.g., backfill) around the waste that have chemical properties that retard or inhibit radionuclide transport; or by some combination of the above.
Either way, in principle, the source term to the geologic system can be maintained at a low level and can be tested to verify release rates under anticipated conditions.
The NRC staff has considered these three alternative approaches to selecting the major barrier:: in light of their ability to compensate for uncertair!ty in assessing system performance without unduly constraining system design.
Alternative three, that of supplementing the isolation capability of the site with engineered barriers, is considered by the,NRC staff to be superior in that w1 ~
it allows the flexibility of a combination of erigineered and natural barriers g
that compensates for the major sources of uncertainty in the natural system.
Enclosure J
4/30/81 26 Alexander 5/A i
1 This approach has also been adopted by DOE in its program for geologic disposal of high level radioactive wastes (DOE 1980).
In arriving at its conclusion the NRC staff received important guidance from peer reviews by the Keystone Radioactive Waste Discussion Group and a panel of earth scientists convened by the University of Arizona: The sense of both groups was in agreement with a subdivision of the repository system into tScee major barriers: waste packages, the engineered underground facility, and the geologic environment.
The Keystone group suggested that the entire engineered system should be considered the barrier which limits the rate of release of waste.
Following their suggestion, the NRC staff decided to set a long-term release rate for the underground facility and waste packages working together as opposed to a release rate for the waste form alone.
Both workshops emphasized the importance of a prcgram of testing and verification to evaluate barrier performance.
IV.
Major Barrier Performance The next issue considered by the NRC staff was "What minimum performance criteria should be set for the major natural and engineered barriers in light of the uncertainties in predicting system performance over long periods of time under repository conditions?"
For the purpose of assessing repository performance, the multiple barriers are treated as a series of elements that form the repository system.
For a given initial inventory, the overall performance of a geologic repository with respect W a MM >
J to releases to the biosphere;;; : :r + :q by three characteristics:
(1) the 1
Enclosure J
4/30/81 27 Alexander 5/A length of time after closure during which radionuclider are contained,'(2) the time over which radionuclides are released from the engineered system after containment fatis, and (3) the travel time through the-geologic ' setting for rsdionuclides to reach the biosphere (see Burkholder, 1976; DOE, 1979; Cloninger, 1979). The effects of disruptive events on system performance can be considered in terms of reducing these times.
The performance of the individual barriers can also be specified in terms of these three characteristics.
The performance of waste packages or the under-ground facility is determined by specification of a containment time and a-fractional release rate that is aqual to the reciprocal of the release time.
Site performance is determined by the travel time.
Thus, all of the major barrier performance standards can be specified in terms of an appropriately chosen containment time, release rate, and travel time.
In order to evaluate reasonable minimum performance criteria for the individual barriers, we next considered in more detail the properties of the wastes as a function of time and the uncertainties associated with containment and isolation of high level wastes.
As noted earlier, thermal effects of decay heat generated by the waste are one of the principal causes of uncertainty in predicting the performance of the repository system.
These have also been assessed in our consideration of performance objectives for the major barriers.
Enclosure J
o.
4/30/81 28 Alexander 5/A Several investigators have calculated the temperature distribution in space' and time that would result from waste emplacement (see'for example 00E, 1979; Wang and others, 1979; AOL, 1979b).
Figures 1 through 9 have been taken from AOL (1979b) to illustrate the following qualitative cEaracteristics of temperature
--~
distributions of a geologic repository for HLW:
1.
The magnitude of the maximum temperature at the canister mid plane and the time at which it is reached depend on the age'of the waste before burial (Fig. 5), the planar heat density (Fig. 6), and the fuel cycle (Fig. 8).
On the other hand, the magnitude of the maximum temperature and the time at which it occurs are relatively independent of the host rock type (See Fig.4).
, 2.
The maximum temperature of the repository as a whole is reached during the period 100 to 500 years after emplacement and near maximum temperatures persist for a few hundred more years (See Fig. 4).
Aged wastes and wastes with a higher concentration of long-lived materials (Mixed oxide and throwaway fuel cycles), reach maximum temperatures at the later timer (See Figs. 5 and 8).
3.
Maximum temperature gradients in the host rock occur within 100 years for all fuel cycles and host rocks.
Enclosure J
.4/30/81 29 Alexander 5/A 1
4.
After 1,000 years, both temperatures and thermal gradie'ats in the reposit.ory
~'
hr.ye peaked and are decreasing for all fuel cycles, and by~10,000 years, temperatures and thermal gradients are near pre-emplac~ement conditions.
~
Table 1 is from AOL (1979a) and shows the differences between the general characteristic of HLW generated by the each of three fuel cycles.
Because of these differences, the time behavior of the waste characteristics also differ from one fuel cycle to another.
ADL (1979a) has characterized the source term as a function of decay time for HLW from each of the three fuel cycles.
The results of their calculations are displayed graphically in Figures 10-15.
For each fuel cycle, the following data are plotted:
1.
Radioactivity versus decay time (Figures 10-12); and 2.
Decay heat generation versus decay time (Figures 13-15).
In all cases the fuel was assumed to have been irradiated in a pressurized water reactor (PWR) rather than a boiling water reactor (BWR) because PWR fuel is irraciiated to a higher burnup before refueling.
Higher burnups yield i
higher fission product inventories per unit of fuel, and therefore provida an l
upper bound on the rauioactivity and decay heat rates from a light water reactor.
1 Enclosure J l
4/,30/.81 30 Alexander 5/A Examination of Figures 10-15 leads to the following observations:
1.
In all cycles for the first few hundred years, the fission product activity-is tha principal contributor to the total radioactivity.
Sr-90 and Cs-137 are the principal contributors to the fission product activity.
2.
The radioactivity levels, and the decay heat generation are similar for all three fuel cycles during the first few hundred years.
This is because fission product activity per unit of energy produced is largely independent of fuel cycle.
3.
The actinides become the dominant isotopes after the first few hundred years.
Differences in actinide content in the waste from the three fuel cycles then cause significant differences in the properties of the wastes.
4.
During the first 1000 years, the radioactivity, and decay heat generation rate of the fission products in the wastes decrease by five to six orders of magnitude and then level off.
5.
During the first 1000 years, the radioactivity, and decay heat generation rate of the wastes from the three fual cycles decrease by three orders of i
magnitude.
I l
Enclosure J
4/30/81 31 Alexander 5/A (1) Waste Package Performance u _.. _
In light of the above information on repository ther.ral conditions and waste characteristics as a function of time, the staff examined a range of containment times as performance objectives for the design of the waste package.
Our objective is to require the waste package-to be designed to contain the wastes during the period when the perturbatio'ns in the near field due to emplacement of the waste are large and would cause unacceptably large uncertainty in our ability to predict waste isolation performance.
Our intent is that during this period the waste would be contained within the waste packages. We namined the following alternatives ~for the waste package containment time:
(i) 300 years; (ii) 1,000 years; and (iii) 10,000 years.
(i) Containment of the wastes for 300 years, as suggested by 00E in its comments on our Advance Notice of Proposed Rulemaking, would prevent releases from occurring until the bulk of the fission products would have disappeared by decay and the heat generation rates will have decreased by about 2 orders of magnitude for wastes from all fuel cycles.
Containment for 300 years is within the range that DOE is Enclosure J
, ~.
~
4/30/81 32 Alexander 5/A considering for repositories in bedded salt and aNeamto 'be~achievabte' at reasonable cost (Magnani and Braithwaite, 1980).
~
~
A minimum containment time.f 300 years has the disadvantage, however, that packages fail and releases begin to occur when temperatures in the repository are near their peak and when the thermal gradients that provide the driving force for convective transport are still relatively high.
Under these conditions of high temperature and high thermal gradients, hydrothermal reactions of the waste form and mineral phase changes of the backfill materials and near-ff' eld host rock will be most severe, and the leaching and transport of radionuclides through the underground facility will be most difficult to evalute.
A contain-ment time of'300 years presents considerable uncertainty in the predic-tion of the releases from the underground facility which constitute the source term for the far field transport models due to the effects of temperature on leach rate, hydrologic flowpaths viscosity, rock l
permeability and geochemistry.
l (ii) Containment for 1,000 years would prevent releases from occurring until the fission products will have essentially disappeared and decay heat generation rates will have decreased by three orders of magnitude.
More importantly, containment for 1,000 years has the effect of delaying l
releases until temperatures in the underground facility are past their peak and are decreasing and until thermal gradients in the underground facility and surrounding rock have decreased substantially from the i
Enclosure J l
j 4/30/81 33 Alexander 5/A first few hundred years.
Lower temperature and temperature gradients allow release rates and radionuclide migration rates to be predicted with greater confidence under these conditions.
Containment fo'r 1,000 years also requires only extrapolation by a small factor beyond what 00E has already been considering for repositories ~in bedded salt (Magnani and Braithwaita, 1980).
(iii) Containment for 10,000 years would prevent releases frcm occuring until the bulk of the fission products and some intermediate-lived transuranics (e.g., Am-241, half life 450 yr) would have decayed to negligible levels.
Heat generation rates would have decreased by over four orders of magnitude and temperatures and thermal gradients
- in the repository and host rock would have nearly returned to pre-waste emplacement conditions.
Under these conditions, we consider that many of the transport processes can be modeled with some confidence and analoges between the transport of actinides and their daughters and migration from are bodies are more reasonable.
However, design of a package to contain wastes for 10,000 years requires a considerable extrapolation beyond those concepts DOE has considered in the past i
and for which any test information exists.
Costs for such a package are uncertain and may not be justified by the reduction in uncertainty that might be achieved.
i Enclosure J
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s 4/30/81 34 Alexander 5/A
~-
The staff considers that a containment requirement for the waste package of 300' years is insufficient to increase confidence in long-term performance credic'-
tions.
If packages fail and migration bagins after 300 years, in order' to evaluate overall performance, it will be necessary to consider transpor't from
~
~
the waste packages through the disturbed zone undhr environmental conditions that will make calculation of the source term for the transport through the geologic setting highly uncertain.
On the other hand, containment for 10,000 years would delay the onset of radionuclide migration until temperatures and temperature gradients in the disturbed zone had returned to near pre-emplacement conditions, and the source term for migration could be predicted with much less uncertainty.
The staff considers that if containment for 10,000 years could be~ achieved, it would reduce uncertainty in prediction of long-term performance by reducing the source term available for migration; by better control of the chemical form of the waste when migration begins; and by delaying the start of migration until the perturbations in the geologic environment due to temperature have suostantially decreased.
At present the amount of the reduction in uncertainty cannot be quantified and the costs to achive containment for 10,000 years are very tenuous.
However, the staff considers that 00E should be encouraged to investigate the practicality of a package with a 10,000 year life.
Therefore, we have framed our performance objective for the waste package such that DOE is required to design the package to provide reasonable assurance of containment for at least 1,000 years and as long as is reasonably achievable thereafter.
We consider that containment for 1,000 years will substantially reduce the hazard associated with a release from the package and will increase our confidence in our ability to evaluate the effectiveness of the disposal Enclosure J
a.
4/.30/81 35 Alexander 5/A system to maintain releases to the environment to within:'the EPA standard.
We further consider that such a requirement is achievable'at1~reas'nable cost by a o
reasonably straightforward extrapolation of current 00Ei programs. ~ Howev' er, we consider containment for periods as long as 10,000 yea ~~s'to be a desirable goal-r H
znd consider that DOE should continue to develop information~ on the performance and costs of packages for long-term containment and to include them in the
~
repository system if found to be reasonably achievable.
Since specific designs that would result in a favorable licensiti2 decision are'not available now ce likely to be available in the near future, we do not consider that a detailed balancing of costs and benefits of longer lived packages can reasonably be performed now, but should be considered by DGE in its application.
(2)
Long-Term Performance Objective for the Engineered System- _ _ _
In order to evaluate reasonable minimum performance objtctives for the engineered system after the initial period of containment, the NRC staff evaluated the following information.
In the Draft Generic Environmental Impact Statement on Management of Commercially Generated Radioactive Waste (GEIS) (DOE,1979), DOE evaluates the lifetime (50 yr) accumulated total body doses to maximum individuals as a function of time of release and release rate for spent fuel and reprocessed UO2 wastes.
The calculations are performed for a repository that has a 100 year water transit time to the environment and employs sorption equilibrium constants (Kds) typical of subsoils at the Hanford site.
The calculations show that for approximately the first 1000 years after breach of containment, lifetime doses to the maximum Enclosure J
4/30/81 36 Alexander 5/A
.i.ndividual are approximately proportional to the release rate (leach rate)...
For individuals exposed one million years after breach of containment, the release rate showed little effect on lifetime dose because doses were due to daughters.of U-238 which had been entirely released during the previous one million year period.
Cloninger (1979) calculated potential dose to individuals who may be exposed to radioactivity released from a repository in salt via a groundwater leach /
transport pathway.
A sensitivity analysis was performed of the waste form ieach rate, the delay prior to groundwater contact with the waste, aquifer flow velocity and flow pathlength.
He concluded that even for a site with fairly good hydrologic characteristics, there is benefit in providing a leach resistant waste form or some equivalent engineered system that will limit the_, _
,l rate of release of the nuclides into the flowing groundwater. The results also show that for a well intrusion event, reduced leach rate causes a significant reduction in the lifetime dose commitment to the maximum individual.
The NRC staff has calculated the effect of the annual release rate on the fraction of long-lived nuclides released from a repository system (White, et al., 1979).
Limiting the release rate from the engineered system compensates for uncertainty in the prediction of long term performance by reducing the source term that is i
l available for transport through the hydrologic systems.
The calculations show that annual release rates in the range of 10 s to 10'7 per year result in a significant reduction in the fraction of several environmentally significant Enclosure J
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4/30/81 37 Alexander 5/A long-lived isotopes that could potentially be released from the repository,
~~
which could result in corresponding reductions in population doses.
Based on the above considerations, the NRC staff considered the following alter-natives for the criterion for the release rate from the engineered system after the containment period:
(i) a range of 10 3 to 10 4/yr, which is typical of leach rates of many borosilicate glasses at low temperature; (ii) a release rate of 10'5/yr; and 1
~
(iii) a release rate of 10 7/yr.
(i) Typical leach rates of borosilicate glasses being tested by DOE are in the range of 10 s to 10 8 2
g/cm / day (Weed and others, 1980).
It is expected that the glasses will crack due to thermal and mechanical stresses during heating and cooling in the repository to fragments on the order of ten centimeters on a side.
These parameters result in a range of annual release rates of 10's to 10 4 of the waste inventory.
Dissolution rate of UO 2
fuel pellets in simulated repository groundwaters are also in this range.
~
Thus, annual release ?ates after package failure of 10 3 to 10 4 of the waste inventory appear achievable based on current DOE programs, considering the leach rate of the waste form as the only e.gineered barrier controlling
~
the release rate.
However.
annual release rate of 10 3 to 10 4 of the Enclosure J
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4/30/81 38 Alexander 5/A waste inventory is insufficient to achieve much reduction'in the quantities of long-lived material that would be released, and we would still be in' the position of relying almost entirely on the geology and the far field geochemistry to provide isolation for the long-lived radionuclides in the waste.
(ii) Leach rates of high-temperature nepheline syenite glasses are 2 to 3 orders of magnitude lower than borosilicate glass (Walton and Merritt,1980), as are leach rates of a number of ceramic and composite materials being considered by DOE for high-level waste forms.
Some newly developed boro-silicate glasses may fall into this range also.
In addition, Nowak (1980) has described commonly available clay backfill materials that have the potential to delay breakthrough of Pu and other transuranics for 10,000 to 100,000 years. We consider that, based on technology currently being
~
developed by DOE, annual release rates of 10 5 of the waste inventory are achievable at reasonable cost using combinations of waste forms and engineered barriers.
In addition, a release rate after containment failure of 10 s of the waste inventory per year, while not adequate to isolate waste on its own merit, is long enough that significant decay of long lived species takes place before release.
This limit will contribute to reducing doses to both populations and the maximum individual, and will :ubstantially reduce our reliance on less certain geochemical retardation to limit releases to the accessible environment.
Enclosure J
4/30/81 39 Alexander 5/A (iii)An annual release rate of 10 7 of the waste inventory after containtant
~
failure will reduce doses to individuals and releases to very low levels-with little or no reliance on geochemical retardation.
An engineered system that could meet this criterion would best satisfy our_o'bjective of reducing reliance on being able to characterize and model the behavior of the far-field ~
geochemical system and placing reliance on known materials whose properties can be controlled and tested.
However, DOE has not yet' demonstrated whether such a release rate is achievable and the costs are very uncertain.
~
The staff considers that an annual release rate after package failure in the
~
~
range 10 3 to 10 4 of the package inventory is insufficient to achieve our objectives, since little reduction is achieved in the quantity of long lived radioact.ive material that vould be released, and the repository system would rely almost entirely on the site to provide long term isolation.
The staff considers that if an annual release cate from the engineered system as low as 10 7 of the package inventory at 1000 years could be achieved, it would compensate for uncertainty in the calculation of the transport of radionuclides through the groundwater pathway by limiting the source term to a relatively low value.
Maintaining the release rate at a value this low would result in decay of most radionuclides within the engineered system.
At present the amount of the reduction in uncertainty cannot be quantified, and the costs to achieve a ralease rate this low are very uncertain.
However, the :;taff considers that 00E should be encouraged to investigate the practicality of maintaining release rates at very low levels.
Therefore, the staff developed a minimum
~
performance objective of an annual release rate no larger than 10 5 of the l
Enclosure J t
.4/30/81 40 Alexander 5/A package inventory and as long as is reasonably achievable thereafter. We consider that a release rate of 10 s per year is low enough that appraciable benefit will be gained by radioactive decay before release, and is achievable' at reasonable cost by methods currently being developed by 00E.
However we consider a release rate of as low as 10 7 per year to be a desirable gest and consider that DOE should continue to develop information on materials and costs to achieve such low release rates and should include them in the repository system if found to be reasonably achievable.
(3) Minimum Performance Objectives and Required Characteristics for the Geologic Setting Engineered barriers designed to minimum performance standards can provide reason-able assurance that the everall performance objective of the HLW disposal system will he met for an initial period of time.
After containment failure, engineered barriers can be designed to limit the rate of release of radioactive materials from the repository.
However, once materials are released from the engineered system, the site must provide whatever additional isolation is needed in order to meet environmental standards.
Reliance on the geology to provide ene of the major barriers to releases also introduces diversity into the waste disposal system that can compensate, in part, for any unanticipated failures of the engineered system, as well as acting ao one of the system barriers.
The geologic setting is characterized by a variety of parameters that could themselves be regulated.
Enclosure J
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4/30/81 41 Alexander 5/A i
i Examples of such parameters that could be considered are'~ permeability, inter-stitial groundwater velocity, and equilibrium sorption coefficients, to name a few.
However, all of these parameters combine to determine two' characteri'stii:s:
of the geologic setting, assuming radionuclides have escaped the engineered system:
(1) the transport time of groundwater from the underground facility to the accessible environment and (2) the transport time of individual radio-nuclides from the underground facility to the accessible environment.
The second characteristic dif,fers from the first in that it-takes into' account the geochemical characteristic of the medium and acounts for retardation of the nuclides by precipitation and ion exchange.
Based on the above, we considered three alternatives for setting performance objective for the geologic setting:
(i) require the nuclide travel times from the underground facility to the accessible environment under repository conditions to exceed some minimum value; (ii) require the groundwater travel time for the undisturbed geologic setting to exceed some ininimum value; and 1
(iii) not specify a minimum value but simply require the geologic setting to provide whatever margin is needed to complement the engineered barriers to ensure that the overall performance criterion for the disposal system is met.
Enclosure J
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4/30/81 42
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Alexander 5/A (1) In order to implement a requirement of the type that the geologic setting provides a minimum nuclide travel time, it would be required ~
that a model for the hydrogeologic system be developed that could
=
predict the behavior of 'the flow system and the geochemical system under the thermal field of the repository, and of the far field geochemical system.
Such a model would be subject to many of the use types of uncertainties that modeling of the entire disposal system involves.
A performance objective of this type would not achieve our regulatory objective of bounding or eliminating uncertainty in the analysis and increasing confidence in the performance of the system.
(ii) A requirement that the undisturbed geologic setting provide a minimum travel time to the' accessible environment avoids the need to model the thermal effect or the hydrologic system and the geochemical impacts of nuclide transport.
It requires only the seasurement of parameters and modeling of aquifer flow that is commonly done in water resource analyses.
Computer codes for these types of analyses are commonly used by the USGS and in the oil industry.
Some uncertainty will result because the number of boreholes for measuring permeabilities and hydrau-lic heads will be limited because of the desire to preserve the integrity of the site, but the uncertainty will be less when compared to measuring geochemical parameters and modeling nuclide transport.
The objective of this requirement is that for the long term when the site plays a major role in isolation,. the perturbations due to emplacement of the Enclosure J
4/30/81 43 Alexander 5/A waste will have died down, and the site can be relied on with greater confidence to provide isolation.
In order for this to-be a useful approach for regulating repository perfomance, the geologic setting-must be stable to provide confidence that waste will continue to be isolated.
Also a complementary requirement is needed on the engineered system.
This requirement is that the underground facility not provide a preferential pathway that bypasses or short circuits the hydrologic flow system, providing a direct pathway to the accessible environment.
(iii)
A requirement that the geologic setting provide whatever margin is needed to ensure that the overall system performance criterion is met is an implicit performance requirement, since this would always be required.
It is subject to the same uncertainties as alternative (i), since it would requjre an assessment of overall system performance.
Also, this alternative does not in any way bound or reduce uncertainty in predicting the performance of the systes and does not increase confidence that the overall performance objectives will be met.
Based on the above reasoning, we have selected alterrative (ii) as *.he framework for establishing performance objective for the geologic setting. We next considered what the minimum travel time should be.
Travei times of a hundred years or less would require considerable reliance on the geochemical system to ensure that the overall performance objective for the system is met. While geochemical retardation is expected to be a strong i
factor in providing waste isolation, there will be considerable uncertainty in f
Enclosure J
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the magnitude of its contribution.
This uncertainty results from the fact that it is very difficult to know how much geochemical retardat.sn will occur.
There is currently no agreement among the scientific community on how 'such an evaluation ~
can be made.
A rigorous, agreed on correlation between laboratory data and real site performance doesn't yet exist.
This would likely be a major source i
of contention in a licensing proceeding.
A travel time of only one hundred years does not provide any margin to compensate for uncertainties.
Also, from gectndwater dating studies, travel times well in excess of 100 years are known to be achievable in a variety of hydrogeologic environments and we would not consider a travel time for an unperturaed site as low as 100 years to be suitable for a repository. We, therefore, considered longer times, viz 1,000 and 10,000 years.
A travel time for groundwater from tHe repository to the accessible environment of 10,000 years would be sufficient for many shorter-lived nuclides to meet i
the system's overall performance objectives with no reliance on site geochemistry.
For several long-lived nuclides, e.g., Pu-239, Tc-99, some reliance on geochemical retardation would be required, but considerable margin would exist between equilibrium distribution coefficients (Kds) measured in the laboratory and actual site geochemistry performance required to meet the release limits of the EPA standard.
We are uncertain, however, to what extent such a groundwater travel l
time is achievable. We do not want to rule out otherwise good repository sites by unnecessarily restrictive requirements.
However, this could be used as a goal.
i l
Enclosure J
4/30/81 45 Alexander 5/A Groundwater travel times from repository depths to the accessible environment of 1,000 years are achievable in many hydrologic systems.
For a groundwater travel time of 1,000 years, sorption equilibrium coefficients of 100 ml/g or less are sufficient to prevent most of the principal contributors to dose fron, reaching the accessible environment.
Sorption equilibrium coefficients measured in the laboratory for the actinides and other nuclides that-are principal contri-butors to dose are in the range of 102 104 ml/g, so that'some margin is provided to compensate for the uncertainty in actual values of Kd under repository conditions.
Because of the greater confidence in our ability to measure hydraulic rather than geochemical parameters, snd the conservatism that is introduced, it seems prudent to sciact the water travel time rather than Kd to meet the overall performance standard.
Therefore, we have framed our site performance objective so that the travel time from the repository to the accessible environment be at least 1,000 years and we intend that DOE consider during site screening that sites with longer water travel time are preferred.
1 If sites with long enough water transport times are selected as potential repository sites, some of the major uncertainty in site evaluation can be resolved.
Licensing issues will then mainly be restricted to ensuring that the proposed repository does not disrupt the hydrologic flow pathways such that shorter travel times to the environment are created, and the adequacy of engineered barriers dealing with disruptive events and natural processes that could result in shorter flow pathways.
Enclosure J
i i
t.
4/30/81 46 Alexander 5/A l
V.
Retrievability In its licensing procedures for disposal of high-level radioactive waste in geologic repositories, the NRC has adopted a step-by-step approach that consists of four stages:
(1) Site characterization, during which detailed studies of alternativa candidate sites are conducted prior to selection of one of the sites for develop-ment as a repository.
l (2) Construction cuthorization, during whicn NRC reviews a license application prior to construction that contains a detailed design and analysis of the performance of the repository based on the site specific information obtained during site characterization.
(3) License application, when an application for a license to receive waste at the facility f 4 reviewed again prior to operation.
At this time, the repository design and performance assessment are updated in light of new information obtained i
about the site during construction of the repository, i
(4) Decommissioning, or permanent closure, at which time an application i
to terminate eparations and seal the repository is submitted.
The application will again contain updated analyses of the performance of the repository in light of:
(1) information obtained about the site during the operation of the repository; and (2) data collected about the performance of the engineered system to verify that performance is within design limits.
1 Enclosure J
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4/30/81 47 Alexander 5/A
.This step-by-step approach and continuing re-evaluation is consistent with earlier recommendations of NAS (1979) and IRG (1979).
~ "
NAS (1979) has recommended that repository development "... be a continuing process that includes evaluations of site suitability and satisfactory repository performance before construction, reevaluations during construction and prior to emplacement of wastes, and a findi assessment before emplaced wastes are committed to disposal.
Corrective actions, including removal of emplaced wastes and site abandonrent, should be available options until final qualification and closure of the repository."
At the decommissioning stage, the Commission will determine whether the DOE's comprehensive program of testing, monitoring, and verification indicate that the repository will work as planned.
Unless the repository is designed to preserve the option to retrieve the waste starting at any time prior to permanent closura, an action reserved to the Commission could be foreclosed, and an unsafe condition could be transmitted to future generations.
A number of the public comments on the draft criteria publishcd with the NRC's May 13, 1980 Advance Notice of Proposed Rulemaking addressed the issue of retrievability.
Several commenters suggested that retrievability be maintained for a period of time Efter waste emplacement sufficient to conduct a monitoring program of repository behavior.
Most of those commenters sucgested a period of 10 to 15 years to be satisfactory for this purpose.
One commenter (AIF) i suggested that retrievability be required only during the emplacement period Enclosure J
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., - - _ _.. -..,,, - - - + -.,. -.
c,...
,.__rr.-.---.
., _,----..,yem
, =,,
.mr,---.c e,
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l 4/30/81 48 Alexander 5/A and until all or a part of the waste disposal. facility is defined'as a permanent repository.
Several ccmmenters interpreted the draft critoria to preclude backfilling of the mined areas until decommissioning.
Along with the commenters, the NRC staff considers that he option to retrieve i
the wastes must be preserved long enough to complete a' program of monitoring and verification of repository performance.
The design. mutt also ensure that the option is preserved long enough to permit a decisibn to decommission the repository or take corrective actions based on the evaluation of the results of the verification program, including the timo required to retrieve all or part of the wastes, if shown to be necessary by the results of the monitoring program.
Since some of the assumptions and issues,that will need to be verified and resolved by the monitoring program may not be identified until the underground facility is excavated, it is not possible to specify, prior to construction the content of the verification program or how long it will take. We expect the verification program to evolve throughout the operating lifetime of the repository.
On the other hand, important design decisions will need to be made prior to submitting an application.
Some of these design decisions will affect the length of time available to take corrective action or conduct retrieval, if found to be necessary.
For example, the thermal loading of the waste in the emplacement areas will affect the temperature of the host rock and the stability of the underground structure.
The items will have a large effect on the ability to retrieve the wastes, since the structure could become too unstable or the rocks Enclosure J
~
4/30/81 49 Alexander 5/A too hot to safely recover the wastes.
Therefore we co~ncluded th.at a retrievability period must be chosen early in the design process to permit the design to go-forward.
The staff considered how long might be required to carry out a monitoring and verification program that would provide the information to support a decision to decommission the repository or to decide that some correcti've action need be taken.
One of the key parameters that needs to be monitored is temperature.
Temperature is an important variable affecting package co~rrosion rates, fl~uid flow rates, geochemical reaction rates, stress in the rock mass and brine migration rates in salt.
For conceptual repository designs being considered by 00E in slat, granite, shale and basalt, maximum rock temperatures in the underground facility occur at approximately 35 years after emplacement for reprocessed wastes and at 75 years after emplacement for disposal of spent fuel.
By 100 years after emplacement, near-field rock temperatures have started to slowly decrease for both waste types in all four media.
Also, estimates of repository resaturation times for granite, basalt and shale range from a few years to the order of 100 years (EPA 19809).
- Finally, experimentally determined (Roedder and Belkin, 1980) and calculated (Cheung 1980) brine migration rates indicate that measurable quantities of brine would accumulate in emplacement holes in a salt repository in a few decades.
- Thus, within a period of about 50 years after termination of waste empiacement, it is possible to obtain field measurements of the geochemical, hydrologic and Enclosure J
d 2
4/30/81 50 Alexander 5/A geochemical environment in the underground facility under what will likely be the most severe repository conditions that will affect'the waste packages and engineered barriers.
A monitoring period of only 10 to 15 years after emplacement, as suggested by some of the commenters, may not be sufficient to provide the information needed to make a decision to decommission.
The design must also allow for the time required to thoroughly investigate problems that may be identified during the monitoring program, to evaluate the results of the program, and to take corrective actions, including retrieval of part or all of the waste, if fount; necessary.
The design of the facility must provide access for the time necessary to carry out these operations or else the ability to conduct these activities'may be precluded.
Therefore, we have required that the repository be designed so that the waste could be retrieved on a reasonable schedule starting at any tina up to 50 years after waste emplacement is complete.
We consider a resonable schedule is one where the waste could be retrieved in the same overall time that the repository was constructed and wastes were emplaced.
We do not intend to preclude a decisica to decommission the repository before 50 years has elapsed, if sufficient data are available to support an earlier decision, and if tne people charged with the decision to seal the repository are satisfied.
- However, we do not want the underground facility design to be such that retrieval would be so expensive or difficult or entail such high occupational exposures that the option is for6 closed and needed corrective actions cannot be taken.
Enclosure J l
=..
4/30/81 51 Alexander S/A Two commenters (AIChE, DOE) incorrectly inferred that the requirement to design the repository to preserve the option to retrieve the wastes would pass an expense and a responsibility on to future generations that should be borne by the present generation.
These commenters have misinterpreted our requirement.
We only~
require tha*. the design of the repository preserve the oction to retrieve the wastes for reture decision-makers.
The persons in charge at the time emplacement is complete will have the opportunity to decide whether to decommission and seal up the repository or to continue to monitor its performance.
We only require that the design be such that they have this option.
We consider that if NRC's regulations do not require that the option be preserved, there is a potential to pass on to future generations an unsafe repository for which corrective actions could be taken only at enormous costs both in dollars and in occupational radiation exposures that far outweigh the costs to design the repository to preserve the option to retrieve the wastes. Maintaining th2 cption to retrieve the wastes does not entail keeping the mined areas open, although DOE may choose to do so in some geologic media.
A design in which the emplacement rooms were backfilled and sealed, but corridors and shafts were kept open and surface handling facilities were maintained could be acceptable, provided that the rooms could be remined and the wastes removed, if necessary.
Remining of the backfill should not be precluded because of high temperatures or because it was needed for structural stability.
Trade-offs between keeping rooms open and ventilated, backfilling, and areal heat densities are design options that DOE must consider in meeting his requirement.
The proposed rule does not require that retrieval be the reverse of emplacement. We can foresee no situation where protection of the public health and safety would require the waste to be removed very rapidly.
Enclosure J
l 4/30/81 52 Alexander 5/A
.Rather, we envision that as the results of years of data collection and analysis, a decision is made that the site or design is not adequate to isolate the wastes for the long term, and corrective actions would be required.
These operations could be performed over a period of years or decades without an iminent health and safety hazard.
Therefore, the proposed rule requires that if a decision to retrieve is made, the design shocid be such that the inventory of wastes could be removed in about the same number of years in which it was emplaced.
We intend for DOE to have considerable flexibility in the design of the repository in meeting these requirements.
A repository designed to permit retrieval of the waste has advantages in addition to the limiting case of preserving a Comission option to order abai,Jonment of the site at as late a stage as decomissioning.
From the time waste emplacement starts until decomissioning any of a variety of eventualities may require corrective action.
Examples might include repair or replacement of cannisters that prove to have manufacturing defects, changes to more effective backfill, or penaps installation of additional barriers in the tunnels.
Design of the repository for retrievability of the waste assures that it will remain practical to take corrective actions should they become necessary.
Enclosure J
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4/30/81.
53 Alexander 5/A Table 1 (From AOL, 1979)
HIGH-LEVEL WASTE CHARACTERIZATION Fission Product Actinide
^
Case Characterization Characterization Comments __-__a_.
(1) Throwaway All fission products All actinides and 1.
Potentially Cycle and daughters daughters most radiotoxic high-level waste per unit fuel weight of may LWR D02 or mixed oxide cases.
(1-3) 2.
Oecay heat rate per unit fuel weight highest of any of the LWR UO2 or mixed oxide cases (1-3)
(2) Uranium Only All fission products All actinides Least raiotoxic and Recycle and daughters and daughters least heat producing Less L.ess bulk of U waste of cases (1-3).
Some percentage of recycled.
Pu 1.
Gaseous Elements separated and (Xe, Kr) stored for future 2.
Volatile Elements use (may be stored (I, Br) contaminated with 3.
Tritium fission products),
or may be made part high-level wastes.
(3) Mixed-Oxide Same as (2)
Same as (2),
1.
Waste produced Recycle except bulk from reprocessed of Pu as well U0 assemblies 2
as U is recycled.
different (and less radiotoxic at longer cooling times) than that produced from reprocessed mixed-oxide assemblies.
2.
Potential radio-toxicity at longer 1
cooling times from ;
equilibrium mixed-oxide cycle waste per unit fuel weigh approaches that of case 1.
4/30/81 54 Alexander 5/f REFERENCES 1.
AOL (1979a), Technical Support for Standards for High-Level Radioactive Waste Management:
Source Term Characterization.
EPA 520/4-79-007A, prep. by A. D. Little, Inc. for USEPA.
- - - - -AOL (1979b), Technical Support for Standards for High-Level Radioactive
~"
Waste Management:
Engineering Controls.
EPA 520/4-79-007B, prep. by A. D. Little, Inc. for USEPA.
3.
AOL (1979c), Technical Support for Standards for High-Level Radioactive Waste Manageraent: Migration Pathways.
EPA 520/4-75 007c prep by A. D. Little, Inc. for USEPA.
4.
Bredehoeft, J. D., A. W. England, D. B. Stewart, N. J. Trask, and I. J. Winograd (1978), Geologic Disposal of High-Level Radioactive Wastes--
Earth-Science Perspectives.
Geological Survey Circular 779, U.S. Geological Survey, Arlington, VA.
5.
Burkholder, H. C. (1976), Management Perspectives for Nuclear Fuel Cycle Wastes.
Nuclear Waste Management and Transportation Quarterly Report, January through March, 1976, Battelle Pacific Northwest Laboratories, Richland, WA.
6.
Burkholder, H. C. (1981), The Technical Approach to Uncertainty Analysis in the National Waste Terminal Storage Program, Gatlinburg, TN (to be published in proceedings).
7.
Cheung, H. and H. H. Otsuki (1980), Post Closure and Retrieval Considerations of Spent Fuel and Transuranic Waste Disposal in Salt, NUREG/CR-1365.
8.
Clark, L.
L., and A. D. Chockie (1979), Fuel Cycle Cost Projections, NUREG/CR-1041, prepared by Battelle Pacific Northwest Laboratory for USNRC.
9.
Cloninger, M. O., (1979), A Perspectives Analysis on the Use of Engineered Barriers for Geologic Isolation of Spent Fuel.
Paper presented at the National Waste Terminal Storage Program Information Meeting October 30 -
November 1, 1979, Columbus, OH.
10.
Cowan, G. A. (1976), A Natural Fission Reactor.
Scientific American, V. 235, N.1.
11.
Craig, R. W., (1979), letter report to John B. Martin, Director, Division of Waste Management, USNRC on results of peer review of draft 10 CFR Part 60.
A copy has been placed in the NRC Public Document Room.
Enclosure J
p a r 4/30/81 55 Alexander 5/f 12.
Davis, S., (1980), letter report to Kellogg Morton Chief, Research-Contracts Branch, USNRC, on results of peer review of draft 10 CFR Part 60.
A copy has been placed in the NRC Public Document Room.
13.
DOE (1979), Draft Environmental Impact Statement on the Management of Commerically Generated Radioactive Waste, 00E/EIS-0046-0~,'U.S. Department of Energy, Washiogton, DC.
14.
00E (1980), Final Environmental Impact Statement on the Management cf I
Commercially Generated Radioactive Waste, 00E/EIS-0046FF U.S. Department of Energy, Washington, D.C.
15.
EPA ad hoc panel of Earth Scientists, (1978), State of Geologic Knowledge Regarding Potential Transport of High-Level Radioactive Waste from' Deep Continental Repositories, EPA 520/4-78-004.
. t 16.
Erdal, R. B., and others (1978), " Sorption - Desorption Studies on Argillite,"
'i PNL-SA-7352, WISAP Second Annual Information Meeting.
4 1
l 17.
Golder Associates, Inc., (1977), Development of Site Suitability Criteria for the High-Level Waste Repository, UCRL-13793.
4 18.
Heckman, R.
A., and others (1979), High Level Waste Repository Site Suitability Study - Status Report, NUREG/CR-0578.
i 19.
Hill, M. D., (1913), Analysis of the Effects of Variations in Parameter Values on the Predicted Radiological Consequences of Geologic Disposal of High Level Waste, NRPS-R-86.
I 20.
IEC, (1979), " Review of Geotechnical Measurement Techniques for a Nuclear Waste Repository in Bedded Salt, UCRL-15141 (International Engineers Co.,
Inc.).
i 21.
IRG (1979), Report to the President by the Interagency Review Group on Nuclear Waste Management, National Technical Information Service, Spring-i field, VA.
22.
Isherwood, D., (1978), Geochemistry and Radionuclide Migration, UCRL-80841.
23.
Isherwood, D., (1981), Geoscience Data Base Handbook for Modeling a Nuclear Waste Repository, NUREG/CR-0912, 2 vol.
I 24.
LBL (19O, Geotechnical Assessment and Instrumentation Needs for Nuclear Waste Isolation in Crystalline and Argillaceous Rocks; Symposium Proceedings, July 16-20, 1978, Lawrence Berkeley Laboratory, Berkeley, CA.
1 25.
Magnani, N.
J., and J. W. Braithwaite (1980), " Corrosion-Resistant Metallic Canisters for Nuclear Waste Isolation," in Scientific Basis for Nuclear Waste Management, Vol. 2, C. J. Northrup, Jr., Ed., Plenum Press, New York.
Enclosure J
.,o=
1 4/30/81 56 Alexander 5/f 26.
NAS (1979), Implementation of Long-Term Environmental Radiation Standards:
-the Issue of Verification, National Academy of Sciences, Washington, DC.
27.
NAS (1980), A Review of the Swedi:h KBS-II Plan for Disposal of Spent Nuclear Fuel, National Academy of Sciences, Washington, OC.-
28.
Nowak, E.' J., (1980), "The Backfill as an Engineered Barrier for Nuclear Waste Management," Scientific Basis for Nuclear Waste Management, Vol. 2, C. J. Northrup, Jr. ed., Plenum Press, New York.
29.
OWI (1978), Technical Support for GEIS, Radioactive Waste Isolation in Geologic Formations, Groundwater Movement and Nuclide Transport, Y/0WI/TM-36/21, Prepared for US 00E.
30.
Ringwood, A. E., (1978), 5afe Disposal of High-Level Nuclear Wastes:
A new strategy, Australian National University Press, Canberra, Australia, and Norwalk, CT.
31.
Robbins, G. and others (1980), " Review of DOE / National Laboratory Geochemical Retardation Programs, USNRC, A copy has been placed in the NRC Public Occument Room.
32.
Roedder E. and H. E. Belkin (1980), Thermal Gradient Migration of Fluid Inclusions in Single Crystals of Salt from the Waste Isolation Pilot Plant Site (WIPP)," Scientific Basis for Nuclear Waste Managemant, 2, C. J. Northrup, Jr. e_d, Plenum Press, New York.
33.
Serne, R. S., and others (1979), Preliminary Results on Comparison of Adsorption - Desorption Methods and Statistical Techniquas to Generate K Prediction Equations, PNL-SA-7245.
O l
34.
Storch, S. N., and B. E. Prince (1979), Assumptions and Ground Rules in Nuclear Waste Projections and Source Term Data, ONWI-24 Office of Nuclear Waste Isolation, Columbus, OH.
35.
USGS (1980), U.S. Department of Interior, Geological Survey, letter to Secretary of the Nuclear Regulatory Commission, dated July 10, 1980, Comment #12, PR-60.
- 36. Walton, F. B., and W. F. Merritt (1980), "Long Term Extrapolation of Laboratory Glass Leaching Data for the Prediction of Fission Product Release Under Actual Groundwater Conditions," in Scientific Basis for Nuclear Waste Management, Vol. 2, C. J. Northrup, Jr. ed., Plenum, Press, New York.
- 37. Wang, J. S. Y., C. F. Tsang, N. G. W. Cook, and P. A, Witherspoon (1979),
A Study of Regional Temperature and Thermohydrological Effects of an Under-ground Repository for Nuclear Wastes in Hard Rock, LBL-8271, prepared by Lawrence Berkeley Laboratory for U.S. 00E.
Enclosure J
'. i <.
4/30/81 57 Alexander 5/f 38.
~ '-
Wawersik, W., (1978), " Nuclear Waste Disposal" in liaitations of Rock ~
Mechanics in Energy-Resource Recovery and Developa.entJa11onaLResources.
Council / National Academy of Sciences.
- 39. Weed, H. C., and others (1980), " Leaching Characteristics of Actinides from Simulated Reactor Waste, Part 2," in Scientific Basi ~s for Nuclear Wa'ste
--~ -
~- ~ Management, Vol. 2, C. J. Northrup, Jr., Ed., Plenum P_ttis,_New Ycrk.
- 40. White, L. A., M. J. Bell, and D. M. Rohrer (1979), " Regulation of Geologic Repositories for the Disposal of High-Level Radioactive Wastes," in Scientific Basis for Nuclept Vaste Management, Vol. 2,_C.. J. _Northrup, Jr...__ _ __
Ed., Planum Press, New York.
Enclosure J
.o a.
_-_ _. _ __ _ _ Surfaes 0
~ " ~ -
D.
150 410 10y 31t4hr
=
1,000y I
5
'5 L5 j
10,000y E
S
{
H 3
A L
E Note: Based on 10 year.oid reprocessed wests with planer heat density of 150 kW/ acre at burial time.
Waste at 480 m (1500 ft).
' Sedimentary deposits.
440 0
100 200 300 Temperature (*Cl Figure 1 VERTICAL TEMPERATURE DISTRIBUTION IN TYPICAL BEDOED SALT.eCRMATION (ref. 2).
=
4 4
Surfese O
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O.
150 1,000y 410 k-1%
~
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Weste at 460 rn (1500 fth
~
- Sedimentary deposits.
2.430 0
jd0 2$0 300 Temperature (*C)
Figure 2 VERTICAL TEMPERATURE OfSTRIBUTION IN TYPtCAL GRANITE FORMATION (150 kW/ ACRE) (ref. 2).
}
O..u.
4
~
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O.
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IU
~
~
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of 150 kW/ acre at buried time.
N Waste at 460 m (1500 ft),
I T
' Sedimentary deposits.
E i
2,430 0
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Figure 3 VERTICAL TEMPERATURE DISTRIBUTION IN TYP! CAL SALT DOME (ref. 2).
i
- ., ' 'a I
1
)
g a
/s-~~N
/
/
\\
/
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0
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~l l
1 I
Figure 8 MAXIMUM REPOSJTORY TEMPERATURE VS. TIME FOR 10-YEAR-OLD t
WASTE IN BEDOED SALT FORMATION (ref. 2).
i
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~
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S.*
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Waste at 440 m (1500 ft).
' Sedimentary deposats.
2.430 '
O jd0 250 300 Temperature (*C) %
Figure 7 VERTICAL TEMPERATURE CISTRIBUTION IN TYPICAL GRANITE FORMATION (60 kW/ ACRE) (ref. 2).
1
I s*,
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92 200--
Horizontal Specing Keyed to
/
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l s
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6 i
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- i ii,i 10 100 1,000 10.0UO 17me (years)
Figure 6 MAXIMUM REPOSITORY TEMPERATURE Vt. TIME FOR
- 10. AND 20 YEAR.Cl.D WASTE IN GRANirs (ref. 2).
't g e e
e e
e b
o e
~ ~.. _..
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~
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Time (years)
Figure 5 MAXIMUM REPOSITORY TEMPERATURE VS. TIME FO 10, 20, AND 50 YEAR OLD WASTE IN BEDDED sal.T (ref,2).
PQOR ORIBlHS
... ~ _..,
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pioner heet density of g*
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/
-Qw 1,
time. Westes at Salt Dome N2 460 m (1500 ft).
I O
,,, i Time (years)
Figure 4 MAXIMUM REPOSITORY TEMPERATURE VS. TIME IN THREE GEOLOGIC FORMATIONS (REPROCESSED WASTE) (r$f. 2).
~..
- w. m,.
ai m. m.no;ill P00R ORIGINAL
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"D 1
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waste twilli planar liest density of g*g y,,,
Ilie kW/ aces at lunial tiene.
g
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N 10A00 Yems f
g'y
~
m 5
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s*
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TEMPERATUHE RISE VS. RAD 20S AT CANISTER MID PLANE (Z-0)
IDH A 1100-ACHE S10HAGE FACILITY IN A SALT ME l
(itESULT OF 2 D ANALYSISl (ref. 2).
l
~
t
, =,... -
i I
10
~
Note:
The strucuits espresenu the decay of activation producs of all non fuel components of the fu-8 assemWy.
8
\\
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N jf.
s N
N Actinides k
~
and
\\
\\-
Doughters s
4 N
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s is
.,i N
'd-
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h Structure [
(Nots)
(
l f~ Fission Producs
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j
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3
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.I
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\\
s %~
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10 -*
N 1
s
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x O
~. N to -
N \\
N N
10*' d
\\
10'2 0
1 2
2 4
5 10 10 10 10 10 10 3;5 Decay Time From Disc::arge (Yrs)
FIGURE 10 - Radioactivity as a Function of Cecay Time for H4gn-Levei Waste frcm DWR Threwaway Cycle (ref. 1).
inu;;:.i0 !!i P00R.0RIGINAL
7
~
_ _ 10 _.
Note:
The structure represena the demy of activetion producu of all ncn fuel components of the fuel assembly.
8 l
10 -
=
l Total 105_
10 -kN[
4 Structure N
(Note)
I s
N l
h N\\
Fission Products
.s N
u N
3 3
Actinides 1g -
O and
\\
2
\\
Daughters N
%O s
i N
\\
N s
5 102-N 3
s N
i
\\\\N s s~x N' %,
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l 1
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g
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l
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~ -. _
s 100-N
% h N'
l
\\
10 \\
l 8
i 1
0 1
3 6
l 10 10 10' 1
1 10 10 Oecay Time From Cischarge (Yrs)
FIG'JRE l.1 - Radioactivity as a Function of Decay Time for Reprocessed l
'. m. h b,,.,,l.;
s..><
f Note: Tha structure represents the deny of activetion n
prooden of all non fuel components of the fuel
- assembly, s\\.
8 10 -
s Total
\\
- 105, g
N
\\
N s s
N 104 -N s
\\
}
\\
\\
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's i
l N
't 5
103-I E
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l
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t I
\\
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g
)
102 4-N \\
Daughters g
s 5
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k, Fission,Predues, g
101-
\\
\\
N N
\\
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s.
Structure
\\% getel 10
% N N
N N
10*I ~
\\
5 10 0
I 2
3 5000 10' 10 10 5
3 10 10 50 10 500 10 Oecay Time From Cisenarge (Yrs)
FIGURE 12 - Radioactivity as a Function of Cecay Time for Reprocessed High-Level Waste from PWR Mixed Oxide Recycle (ref.1).
.. ; k '. '.
<l -
.,e.
~.__
Nom: The structure represenu the decay of activation products of all non fuel components of *:he fuel assembly.
5_
~
10 4-10 s
sN
-Total 103-N g
's,_,
\\%s
~
~
h Actinides
=
102_
~
.nd
\\ c.ve an n
3 E
i
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_3 101-k
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2
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.i 100~
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10*l Fission Producs
\\
\\
g;mg x
10'2 ~
~
\\
\\
\\
\\
II2 e
i i
0 I
2 3
5 5
10 10 10 10 10 10 10 Decay Time From Oisenarge (Yrs)
FIGURE 13 - Decay Heat as a Function of Time for High-Level Waste fecm PWR Throwaway Fuel Cycle (ref. 1).
Ar;ygn anne P00R ORIGINAL
- 2. m
i r
9 Nom: The structure represents the decay of activation
~2 y..
producs of all non. fuel components of the fuel assembly.
104d 103,,,
Actinides and
\\ s
. Daughters 1g2 NN NN s
E
\\
's
\\
j
\\
N 5
Total s
_2 101-N is E.
k
's R
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l 3
g
=
to -
\\
\\
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\\
\\
\\
,0,
x's__,_ "
g
~
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and Daughtert 10-2,,,
\\
s
\\
Structure
\\
(Note)
/
N,
\\
10 3 i sg
\\
\\.
\\
a
.G i
i 0
I 2
3 4
5 6
1c 10 10
$0
$g
$c gc Decay Time.crom Oise::arge (Yrs)
FIGURE 14 - Cecay Heat is a Function of Time for Recrecessed High-Level Waste frem PWR Uranium Racycle (ref. 1).
sa hg n g r,,um 10.s 01 P00R ORIGINAL
3
~ 10 Note: The stnicture rfpresenu the deosy of activation produce of all non. fuel ::omponenu of the fuel assembly.
\\
~
3 10
\\
- Total s
3
's s
10 s,
Actinides
's \\
and N \\
Daughters
- N'
~ '~~'
\\
s 10 ' -' N
~
\\
.D s s N
1, N
1 3-
\\
10 O
\\
\\
\\
l
\\
n 2'
\\
\\
I 10
\\
i j
E
\\
\\
Fission Produca 3
{:
l I~
10
\\
~
Structure t
(Notel
\\
N 0-
\\
N 10
\\
s N
\\
10*l 9
\\-
10~2 3h 0
1 2
3 4
5 10
- 0 10 10 10 10 Decay Time From Discharge (Yrs)
FIGURE 15 - Decay Heat as a Function of Time for. Reprocessed Higt1-Level Waste from PWR Mixed Oxide Recycle (ref. 1).
P00R DillGINAL
4/30/81 73 Alexander 5/h GLOSSARY OF TERMS Accessible Environment - means those portions of the ertyironment direct.ly_in contact with or readily available for use by human beings.
It includts the earth's atmosphere, the land surface, surface waters, and the oceans.
It also includes presently used potable aquifers and those which t tve been designated as underground sources of drinking water by the E.vironmental Protection Agency.
Anticioated Processes and Events - means those natural._ processes _and events _that _ __
are reasonably likely to occur during the period the intended performance objective must be achieved and from which the design bases for the engineered system are derived.
Barrier - means any material or structure that prevents or.substantially delays movement of water or radionuclides.
Containment - means the act of keeping radioactive waste within a designated boundary.
Decommissioning or oermanent closure - meens final backfilling of subsurface facilities, sealing of shafts, and decontamination and dismantlement of surfac:.
facilities.
Enclosure J
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f( (pl l Ar. t..;
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- s *3
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-.,,,n.
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- ' * - ~
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^
4/30/81 74 Alexander 5/h Disposal - means the isolation of radioactive wastes from__the biosphere.
Disturbed zone - means that portion of the geologic setting that_is significantly affected by construction of the subsurface facility or by the heat generated by the emplacement of radioactive waste.
Engineered system - means the waste packages and the underground. facility.
Far field - means the portion of the geologic setting _that_ lies beyond the disturbed zone.
Geologic repository - means a system for the disposal.of.. radioactive wastes in excavated geologic media.
A geologic repository includes (1) the geologic repository operations area, and (2) the geologic setting.
Geologic repository operations area - means an HLW facility that is part of a geologic repository, including both surface and subsurface areas, where waste handling activities are conducted.
Geologic setting or site - is the spatially distributed geologic, hydrologic, ano geochemical systems that provide isolation of the radioactive waste.
High-level radioactive waste or HLW - means (1) irradiated reactor fuel, (2) liquid wastes resulting from the operation of the first cycle solvent extraction Enclosure J
~. -
i
.,;. a.-
4/30/81 75 Alexander 5/h system, or equivalent, and the concentrated wastes from subsequent extraction cycles, or equivalent, in a facility for reprocessing irradiated reactor fuel, and (3) solids into which such liquid wastes have been converted.
o HLW facility - means a facility subject ta the licensing _and_related regulatory authority of the Commission pursuant to Sections 202(3) and 202(4) of the Energy Reorganization Act of 1974 (88 Stat 1244).
Host rock - means the geologic medium in which the waiteils. emplaced.
l Hydrogeologic unit - means any soil or rock unit or subsurface zone that has a distinct influence on the storage or movement of ground water by virtue of its porosity or permeab'ility.
Isolation - means inhibiting the transport of radioactive material so that amounts and concentrations of such material entering the accessible environment will be kept within prescribed limits.
Medium or geologic medium - is a body of rock characterized by lithologic homogeneity.
Overpack - means any buffer material, receptacle, wrapper, box or other structure, that is both within and an integral part of a waste package.
It encloses and protects the waste form so as to meet the performance objectives.
Enclosure J
, -p..p.
4 4/30/81 76 Alexander 5/h Radioactive waste or waste - means HLW and any other r.adioactive. mater.ials.other-than HLW that are received for emplacement in a geologic repository.
Site - means the geologic setting.
Site Characterization - means the program of exploration _and_research, both in the laboratory and in the field, undertaken to establish the geologic conditions and the ranges of those parameters of a paraceters of a particular site relevant to the procedures under this part.
Site characterization includes a program of borings, surface excavations and borings, and in situ testing at depth needed to determine t,he suitability of the site for a geologic repository, but does not include preliminary borings and geophysical testing needed to decide whether $tte characteriza' ion should be undertaken.
t Stability - means that the nature and rates of natural processes such as erosion and faulting have been and are projected to be such that their effects will not jeopardize isolation of the radioactive waste.
Subsurface facility - means the underground portions of the geologic repository operations area including openings, backfill materials, shafts and boreholes as well as shaft and borehole seals.
Transuranic wastes or TRU wastes - means radioactive waste containing alpha l
emitting transuranic elements, with radioactive half-lives greater than one year, in excess of 10 nanocuries per gram.
Enclosure J
^
,c,,..,-
4/30/81 77 Alexander 5/h Tribal organization - means a Tribal organzation as defined _in the_. Indian Self-Determination and Education Assistance Act (Public Law 93-638).
Underground facility - means the underground structure,_ including. openings and backfill materials, but excluding shafts, boreholes, and their seals.
Unrestricted area - means any area access to which is not_contralled_by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, and any area used for residential quarters.
Waste form - means the radioactive waste materials and_any encapsulating or stabilizing materials, exclusive of containers.
Waste package - means the airtight, watertight, sealed container which includes the waste form and any ancillary enclosures, including shielding, discrete backfill and overpacks.
l Enclosure J
_ _. _ _ _ _ _