ML20204G820

Draft Generic Technical Position on Items & Activities in High Level Waste Geologic Repository Program Subj to 10CFR60 QA Requirements

ML20204G820
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Issue date:	07/31/1986
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
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ML20204G785	List: ML20204G779 ML20204G798 ML20204G820
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t DRAFT GENERIC TECHNICAL POSITION

, ON i ITEMS AND ACTIVITIES IN THE HIGH-LEVEL WASTE GE0 LOGIC REPOSITORY PROGRAM L

SUBJECT TO 10 CFR PART 60 QUALITY ASSURANCE REQUIREMENTS 4

Division of Waste Management Office of Nuclear Materials Safety and Safeguards

! U.S. Nuclear Regulatory Commission i

July 1986 i

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TABLE OF CONTENTS Page

1.0 INTRODUCTION

......................... 1 i

2.0 BACKGROUND

.......................... 2 B 3.0 REGULATORY FRAMEWORK ..................... 2 T

l 4.0 DEFINITIONS ......................... 4 5.0 STAFF POSITIONS 5.1 Quality Assurance Criteria for Licensing . . . . . . . . . 5 5.2 Identification of Items Important to Safety ....... 6 5.3 Identification of Barriers Important to Waste Isolation . 8 5.4 Staff Information Needs ................. 8 5.5 Graded Application of QA Measures ............ 9 6.0 DISCUSSION 6.1 Quality Assurance Criteria for Licensing . . . . . . . . . 10 6.2 Identification of Items Important to Safety ....... 13 6.3 Identification of Barriers Important to Waste Isolation . 17 6.4 Staff Information Needs ................. 20 6.5 Graded Application of QA Measures ............ 22 7.0

SUMMARY

........................... 26

8.0 REFERENCES

.......................... 27 9.0 BIBLIOGRAPHY ......................... 28

APPENDIX A - Glossary ....................... 29

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1.0 INTRODUCTION

The requirements which apply to radiological protection of public health and safety and the environment from disposal of high-level radioactive waste (HLW) -

in a geologic repository are defined in the performance objectives and other criteria in 10 CFR Part 60. These requirements address a pre-closure phase, which includes design, construction, waste emplacement, and possible retrieval of waste, and a post-closure phase, which includes containment and long-term isolation of waste. In the pre-closure phase, structures, systems, and components essential to the prevention or mitigation of an accident that could result in an off-site radiation dose of 0.5 rem or greater are termed "importanttosafety"(10CFR60.2). In the post-closure phase, the barriers which contribute to meeting the containment and isolation requirements of 10 CFR Part 60 are defined as "important to waste isolation." These structures, systems, components, and barriers (items), and the activities related to their characterization, design, construction, and operation require quality assurance (QA) measures to provide confidence in the performance of the geologic repository. The list of the items and activities important to safety and waste isolation is referred to as a "Q-list" and comprises the scope of the QA program specified in 10 CFR 60 Subpart G.

In order to obtain a license for a geologic repository the U.S. Department of Energy (D0E) must demonstrate that all requirements in 10 CFR Part 60 are or will be met. This demonstration includes the identification of items and activities important to safety and waste isolation, and the implementation of a QA program for verifying and documenting the quality of work performed to support licensing findings. The purpose of this draf t Generic Technical Position (GTP) is to provide guidance to DOE on approaches the NRC staff considers acceptable for identifying items and activities important to safety and waste isolation. It also addresses measures to assure the quality of all items and activities that will be used to demonstrate compliance with the licensing requirements in 10 CFR Part 60. NRC staff positions on the QA criteria for licensing, the types of analyses appropriate to determine which items and activities are important to safety and waste isolation, staff information needs to assure adequate and timely staff involvement, and graded application of quality assurance measures to items and activities important to safety and waste isolation are provided. The emphasis of this GTP is on pre-closure structures, systems, and con.ponents since the number to be considered is large. In contrast, the number of natural and engineered barriers for post-closure is few and all or most are expected to be important to waste isolation. Though the barriers important to waste isolation may te relatively few, the related activities are many and comprise a large number of the activities to be conducted during site characterization.

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2.0 BACKGROUND

The identification of items having safety significance has been an important issue in the nuclear power reactor program. It has been litigated in various reactor licensing hearings and has been responsible in part for extensive delays in schedules and large increases in cost for reactor plant construction.

In the reactor program safety-related items are subject to the QA program requirements in 10 CFR 50 Appendix B and comprise the reactor "Q-list."

Through reactor licensing experience the NRC has developed a body of practice for defining which items are on this Q-list. For example, a series of design basis accidents, a source term for release of radionuclides to the atmosphere, and meteorological conditions to be assumed during an accident have been defined. Listings of safety-related items have also been developed at the system level as guides based on years of staff and industry experience with nuclear power reactors.

In contrast to the reactor program where fairly prescriptive criteria have been developed, the principal criteria for identifying the Q-list for a geologic repository are broad performance objectives. In the pre-closure phase of the repository, structures, systems, and components important to safety are those whose failure could lead to an off-site dose of 0.5 rem or greater. There are no explicit design basis accidents, source terms for releases, meteorologic conditions, or generic lists of items identified in the requirements or guidance, thus allowing 00E some flexibility in developing approaches for establishing this infonnation. For postclosure,10 CFR Part 60 also provides DOE flexibility in determining what specific barriers will be relied upon to meet the performance objectives and technical criteria described in 10 CFR Part 60.

3.0 REGULATORY FRAMEWORK This section contains a summary of the applicable regulations and formal staff guidance which provide the regulatory basis for the staff positions which follow.

Structures, systems, and ccn.ponents important to safety are those items essential to the prevention or mitigation of an accident that could result in a radiation dose to the whole body, or any organ, of 0.5 rem or greater at or beyond the nearest bcundary of the unrestricted area at any time until the completion of permanent closure (10 CFR 60.2).

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The overall system performance objective for a geologic repository follcwing permanent closure specifies that the geologic setting be selected and the engineered barrier system, shafts, boreholes, and their seals be designed to assure that releases of radioactive materials to the accessible environment

following permanent closure conform to such generally applicable environmental standards for radioactivity as have been established by the Environmental Protection Agency (EPA) in 40 CFR Part 191 (10 CFR 60.112). The performance objectives of particular barriers after permanent closure are specified in 10 CFR 50.113. The barriers that contribute to meeting the specific and overall '

containment and isolation requirements of 10 CFR 60.112 and 60.113 are considered "important to waste isolation."

1 To help ensure protection of radiological health and safety of the public and l

the environment, 10 CFR 60 Subpart G requires that DOE apply a QA program based on the criteria of 10 CFR 50 Appendix B "as applicable and appropriately supplemented" to "all systems, structures and components important to safety,

to design and characterization of barriers important to waste isolation and to activities related thereto" (10 CFR 60.151). The QA program to be implemented by DOE is to include "all those planned and systematic actions [ including quality control] necessary to provide adequate confidence that the geologic repository and its subsystems or components will perform satisfactorily in

! service" (10 CFR 60.150). DOE must provide a description of the 10 CFR 60 l Subpart G QA program in the Safety Analysis Report included with the license application (10CFR60.21(c)(4)). DOE should also include a list of those

' items and activities to which the QA program applies (a Q-list) in this i description. The 10 CFR 60 Subpart G OA program contains provisions for verifying that work has been performed correctly, for correcting any deficiencies identified, and for documenting the results of completed work, as required in 10 CFR 50 Appendix B.

i Additional staff guidance on the 10 CFR 60 Subpart G QA program is contained in j the "NRC Review Plan: Quality Assurance Programs for Site Characterization of i

High Level Waste Repositories" (USHRC, 1984a). The scope of the QA program is discussed in Appendix A, Section 2 of that document. Additional guidance on

use of performance assessment techniques for identifying barriers important to i waste isolation is provided in the " Draft Generic Technical Position on

! Licensing Assessment Methodology for High-Level Waste Geologic Repositories"

(USNRC,1984b).

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' In addition to items and activities important to safety or waste isolation.

other items and activities will be associated with demonstrating that DOE meets j all of the 10 CFR Part 60 licensing requirements. For example, 10 CFR Part 20 i requirements, which are referenced in 10 CFR Part 60, will need to be addressed

! in the license application. Altheyh these additional items and activities are '

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not covered by the 10 CFR 60 Subpart G QA requirements (which applies only to items and activities important to safety and important to waste isolation),

assurance measures are needed to provide confidence that the requirements have been met. Certain assurance measures, such as use of written procedures, documentation of c'ampleted work, and monitoring of radiation levels, are currently prescribed in the regulations and, although not explicitly stated as quality assurance requirements, provide a basis for demcnstrating compliance with the licensing requirements.

4.0 DEFINITIONS This section provides a definition of significant terms used in the staff positions presented in this GTP. Definitions for other commonly used terms are provided in Appendix A, the Glossary.

" Activities" related to items important to safety and barriers important to waste isolation include: site characterization, facility and equipment construction, facility operation, performance confirmation, permanent closure, and decontamination and dismantling of surface facilities (10 CFR 60.151).

Activities to be identified en the 0-list that will be conducted during site characterization include, for example, waste package and exploratory shaft testing, and performance assessments. Specific activities within the 10 CFR Part 60 Subpart G QA program, such as control of design, purchasing, fabrication, inspection, and maintenance (10 CFR 60 Appendix B, Introduction) need not be identified on the Q-list.

" Barrier" means any material or structure (including bulkheads and seals) that prevents or substantially delays movement of water or radionuclides (10 CFR 60.2).

" Barriers important to waste isolation" include the site, engineered barrier system, shafts, boreholes, seals, and any other structures, systems, or components which contribute to meeting the related performance objectives in 10 CFR 60 Subpart E.

" Containment" means the confinement of radioactive waste within a designated boundary (10 CFR 60.2).

" Credible event" or " credible accident" refers to an event or accident scenario which is sufficiently likely to warrant consideration in design of the ceologic repository in order to prevent or mitigate the consequences of its occurrence.

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"Important to safety" includes those engineered structures, systems, and components essential to the prevention or mitigation of an accident that could result in a radiation dose to the whole body, or any organ, of 0.5 rem or greater at or beyond the nearest boundary of the unrestricted area at any time until the completion of permanent closure (10 CFR 60.2).

"Q-list," as used in the geologic repository program, is a list of structures, systems, and components important to safety, barriers important to waste isolation and related activities that must be covered under the QA requirements of 10 CFR 60 Subpart G.

5.0 STAFF POSITIONS 5.1 Ouality Assurance Criteria for Licensing (a) Criteria for Q-list Items and Activities DOE shall apply a QA program which meets the 10 CFR 60 Subpart G requirements to all systems, structures, and ccmponents important to safety, barriers important to waste isolation, and related activities (10 CFR 60.151).

(b) Criteria for Non-Q-List Items and Activities In addition to items and activities important to safety and waste isolation,10 CFR Part 60 contains requirements for other items and activities, such as those associated with meeting the design criteria contained in 10 CFR 60,131(a) for radiological protection for workers. While these items and activities are not subject to the QA requirements in 10 CFR 60 Subpart G, DOE should implement a program to assure and demonstrate that these requirements are met. Additional guidance on QA for some non-Q-list items and activities is provided in Regulatory Guide &

4.15 " Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment" (USNRC,1979).

It is anticipated that DOE will apply QA programs to items and activities which are not needed to support licensing, but are important to repository operation based on occupational health 5

and safety, reliability, cost, and other programmatic considerations. The staff will not evaluate the QA measures applied to non-licensing related items and activities, but will review such items and activities to assure that they have been correctly classified.

(c) Information Not Collected unoer 10 CFR 60 Subpart G QA Data collection, interpretation, analysis, and other work not conducted under the DOE 10 CFR 60 Subpart G QA progrom but needed for licensing should be evaluated to determine its suitability for use in licensing. This information includes data collected by sources external to 00E, such as oil companies and universities. Staff guidance on this subject is provided in the " Draft Generic Technical Position on Qualification of Existing Data for High-Level Nuclear Waste Repositories,"

(USNRC,1986).

5.2 Identification of Items Important to Safety The threshold for determining which structures, systems, and components are important to safety is a radiation dose to the whole body, or any organ, of 0.5 rem or greater at or beyond the nearest boundary of the unrestricted area before permanent closure. The design basis should also .

consider 0.5 rem as the off-site dose limit for a credible accident.

(a) Analysis Pr'.ababilistic risk assessment (PRA) techniques should be used to the extent practicable to support the identification of structures, systems, and components important to safety in the license application. Engineering judgment and conservative assumptions will be required for identification of items and activities on the Q-list presented in the Site Characterization Plan (SCP) due to the limited data base available at that time.

The PRA techniques used at licensing should include:

(1) System modeling to depict the combinations of safety function and system successes or failurcs which constitute accident sequences. Fault trees and event trees should be utilized in system modeling.

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(2) Consequence analysis to evaluate the movement and deposition of radioactive materials released from the HLW facility. Source terms for individual events or accidents should be identified and justified.

All structures, systems, and components of the geologic repository that could, irrespective of the probability of failure, initiate an accident which if unmitigated could cause an off-site radiation dose of 0.5 rem should be on the Q-list.

(b) Redundancy DOE should, as a minimum, em)loy redundancy in those areas specified in 10 CFR Part 60 [e.g., 10 CFR 60,131(b)(5)(ii) and 60,131(b)(10)(iv)]. The need for additional redundancy, if any, should be determined based on the analyses performed to identify items and activities on the Q-list.

(c) Non-Mechanistic Failures In the development of the Q-list, DOE may use conservative bounding assumptions such as nonmechanistic failures where data are insufficient to support probabilistic risk assessments.

(d) Use of Previously Established Criteria 00E may utilize existing nuclear power reactor criteria for initiating events (e.g., regulatory guides covering the design basis earthquakes, tornado wind velocities, and floods) in the identification of items and activities important to safety where these criteria can be shown to be applicable to the geologic repository.

(c) Retrieval Process 00E should analyze the retrieval process to identify items and activities that may be on the Q-list and subject to the 10 CFR 60 Subpart G QA program in the event that retrieval is necessary.

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5.3 Identification of Barriers Important to Waste Isolation Items and activities important to waste isolation should include the barriers relied on to meet the performance objectives related to post-r,lnsure performance of the repnsitory system (e.g., the site, the waste package, engineered barrier system, shafts, boreholes, and their seals). DOE should allocate performance among the various components of the natural and engineered barrier systems to provide the basis for determining which barriers may be important to waste isolation and how much each contributes to isolation. The amount of credit assumed for individual barriers in initial allocations of performance will provide a basis for the amount and type of testing needed during site characterization.

5.4 Staff Information Needs (a) License Application DOE should submit with the license application a description of the QA program to be applied to items and activities following construction authorization [10 CFR 60.21(c)(4)]. DOE shall identify the structures, systems, and components important to safety [10 CFR 60.21(c)(1)(ii)(E) and 60.21(c)(3)], the barriers important to waste isolation, and related activities falling under the 10 CFR 60 Subpart G QA progran to be described in the license application.

(b) Site Characterization Plans DOE should submit with the site characterization plans a description of the QA program to be applied to items and activities during the site characterization phase. A preliminary 0-list should be provided in this description identifying major structures, systems and components irportant to safety, barriers important to waste isolation, and all major related activities to be conducted curing the site

' characterization phase, such as waste package and exploratory shaft testing. Individual activities related to the Q-list should also be described or referenced in the SCP. The SCP should contain a schedule for refinement of the preliminary 8

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Q-list based on design advancements and collection of new information.

All site characterization activities should be considered to be within the scope of the 10 CFR 60 Subpart G QA program, unless DOE can derronstrate they are not related to items important to

safety or waste isolation. DOE should have in place a 10 CFR 60

! Subpart G QA program prior to start of site characterization activities.

I In addition to the Q-list information, plans for development and implementation of the program to assure and document that non-Q-list requirements are met should be described in the site characterization plans.

i 5.5 Graded Application of 0A Measures i Graded QA measures may be applied to items and activities important

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to safety or waste isolation based on the following considerations:

} o The impact of malfunction or failure of the item, or the impact 4

of erroneous data associated with data collection activities, to '

! safety or waste isolation.

4 o The ccirplexity of design or fabrication of an item or design and

! implementation of a test, or the uniqueness of an item or test.

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! o The special controls and surveillance needed over processes, tests, and equipment.

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! o The degree to which functional compliance can be demonstrated by. .

inspection or test, j

! o The quality history and degree of standardization of the item or ,

j test.

Additional guidance on the grading of quality assurance program elements

! is provided in Appendix 4A-1 of NQA-1 (ANSI /ASME, 1983).

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6.0 DISCUSSION This discussion section provides the rationale for and anplification of the positions in Section 5.0 and is organized to follow the same headings.

6.1 Quality Assurance Criteria for Licensing The purpose of the geologic repository program is to pennanently dispose of high-level nuclear waste in a manner which will protect radiological health and safety of the public and the environment.

Requirements for licensing a repository to meet this goal are specified in 10 CFR Part 60. These requirements describe the performance objectives and other technical criteria to assure safe operation during waste emplacement and retrieval (if necessary), as well as effective containment and long-term isolation of waste following permanent closure of the geologic repository. In addition, 10 CFR Part 60 incorporates other standards such as 10 CFR Part 20 and the generally applicable environmental standards established by the EPA for the release of radioactive materials.

In order to obtain a license for receipt and possession of radioactive material at the geologic repository, the DOE must demonstrate that the repository system will function as required to protect radiological health and safety of the public and the environment. The 10 CFR Part 60 specifies requirements for the performance of structures, systems, and ccmponents important to safety (during pre-closure) and the performance of barriers inportant to waste isolation (following waste crplacement). 10 CFR 60 Subpart G specifies the QA program for these items to assure that their characterization, design, construction, and operation comply with the requirements of 10 CFR Part 60.

(a) Criteria for Q-List Items and Activities The 10 CFR 60 Subpart G QA requirements apply to items and activities on the Q-list. These 0A requirements include, as applicable and appropriately supplemented, the 18 criteria of 10 CFR Part 60 Appendix B for siting, design, construction, operation, and decommissioning of nuclear power plants. These criteria address, in general terms, the basic elements of a QA program, such as organization, design control, test control, 10

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inspection, and records management. NRC staff guidance on the application of the Appendix B QA criteria to the site characterization phase of the repository program is provided in the "NRC Review Plan: Quality Assurance Programs for Site Characterization of High Level Nuclear Waste Facilities,"

i (USNRC,1984a). It is important to note that while DOE must comply with the QA requirements in 10 CFR Part 60 in order to obtain a license, the basic objective of the QA requirements is to assure that work associated with protection of public health

- and safety is of adequate quality.

In addition to the QA requirements in 10 CFR 60 Subpart G, items t important to safety are subject to the design criteria of 10 CFR

60,131(b). These added criteria help to provide assurance that the margins of safety during normal and accident conditions are adequate throughout the life of the facility. They include protection of items important to safety against natural phenomena and environmental conditions, dynamic effects of equipment failure, and fires and explosions, as well as special emergency capabilities, criticality control, and shaf t conveyance features.

2 (b) Criteria for Non-Q-List Items and Activities Items and activities that are not important to safety or waste isolation must also be addressed in the license application to demonstrate compliance with 10 CFR Part 60 requiremerts. These i include, for example, items and activities associated with meeting the design criteria contained in 10 CFR 60.131(a) for radiological protection of workers. Although these items and activities may not be included in the scope of the QA program described in 10 CFR 60 Subpart G, they will be subject to NRC review in the process of licensing a 9e ologic repository.

To demonstrate in the licensing process that the non-Q-list 3

requirements are met, assurance measures (such as procedures for documentation, testing, and inspections) should be implemented.

Scme of these assurance measures are to be described in the Safety Analysis Report [10 CFR 60.21(c)] and include, for example, qualifications and training requirements for personnel t

at the operations area; and maintenance, surveillance and periodic testing of structures, systems, and components of the geologic repository operations area [10 CFR 60.21(c)(15)].

Other assurance measures are specified in 10 CFR Part 20 and

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include, for example, precautionary procedures, instruction of personnel, and records of surveys and radiation monitoring. The staff considers the guidance on QA for radiological monitoring programs during normal operation provided in Regulatory Guide 4.15, " Quality Assurance for Radiological Monitoring Programs-(Normal Operations) - Effluent Streams and the Environment" (USNRC, 1979) to be applicable to such programs for a HLW facility.

DOE should consider a graded 10 CFR 60, Subpart G QA program (s) ordevelopanalternateprogram(s)fornon-Q-listitemsand activities needed to support licensing findings. The staff recognizes that DOE intends to apply QA in a graded fashion to all program activities through its classification of items and activities into D0E quality levels I through III. DOE's level II applies certain NQA-1 requirements and may be an acceptable approach for non-Q-list items and activities used to support licensing findings. DOE should provide to the staff additional information on the development and implementation of the quality levels program for review prior to submittal of the SCPs.

NRC review of the DOE repository program is limited to 10 CFR Part 60 requirements and criteria. However, the staff will review all items and activities in the program to assure that those identified by DOE as related to 10 CFR Part 60 are complete.

(c) Information Not Collected Under 10 CFR 60 Subpart G QA All data collection, interpretations, analyses, and other work to be used to support findings in the licensing process trust be technically and procedurally defensible by the D0E. The staff expects that some information collected outside the 10 CFR 60 Subpart G QA program may be used or referenced by DOE in the licensing process to support findings against items'and activities important to safety and waste isolation. Information from sources such as oil companies, university research programs, and DOE or DOE contractors which was generated prior to implementation of a Subpart G QA program may be necessary for licensing. DOE should review the adequacy and traceability of all information used to support a license application to determine whether it is defensible. The staff has developed a staff position entitled " Draft Generic Technical Position on Qualification of Existing Data for High-Level Nuclear Waste 12

Repositories" (USNRC, 1986), which addresses specific methods for reviewing and qualifying these data. These include the use of peer reviews, consideration of corroborating data, and use of Subpart G confirmatory testing. DOE will need to develop its own programs for reviewing these data and determinn.g which are suitable for use.

In addition to existing information, some materials that may be important to safety or waste isolation may already have been purchased without a 10 CFR 60 Subpart G QA program (e.g.,

materials for exploratory shaft construction previously purchased for the Basalt Waste Isolation Project). DOE needs to evaluate these materials to determine whether 10 CFR 60 Subpart G applies and, if necessary, what actions are necessary to qualify them for use.

6.2 Identification of Items Important To Safety items important to safety are those items essential to the prevention or mitigation of an accident that could result in a radiation dose of 0.5 rem or greater to an individual in unrestricted areas (10 CFR 60.2). The 0.5 rem value is, therefore, the threshold for determining what structures, systems, and components shall be on the Q-list as items important to safety. The rationale behind placing a system, structure, or component on the Q-list is to assure, via application of additional QA and design requirements, that accidents which may cause an off-site dose of 0.5 rem or greater are prevented (made not credible) or mitigated to less than 0.5 rem. Therefore the 0.5 rem value should also be considered the dose limit for design basis accidents.

(a) Analysis 00E should use PRAs to the extent practicable in the determination of items important to safety for the HLW facility.

PRAs have been shown to be a useful licensing tool for assessing the safety of nuclear facilities. In addition, use of this approach for the operations phase of the program is consistent with the approach prescribed by the EPA standard (40 CFR Part 191) for the period following emplacement of waste in a geologic repository. PRAs also provide a framework for grading of QA measures based on the risk associated with the failure of 13

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1 individual components. Since the geologic repository will be a first-of-a-kind facility, the staff recognizes that complete data will not be available for initiating events and failures.

In cases where data are limited, conservative bounding assumptions should be used. Even less data will be available at the time of the SCP, and greater use of engineering judgment and conservative assumptions will be required to identify the preliminary Q-list presented in the SCP.

The PRA should utilize the following techniques:

(1) System modeling to depict the combination of safety function and system successes or failures which constitute accident scenarios. Two modeling techniques often used are event tree analysis, which identifies the sequence of events that may result in an accident, and fault tree analysis, which determines how failures in safety systems may occur. Both techniques are analytical tools which organize and characterize potential accidents in a -

methodical manner.

Event tree analysis first identifies all conceivable events that could lead to an accident. These are referred to as

" initiating events." Following this, all significant sequences of events that could follow the initiating event are examined.

A fault tree examines the various ways in which a system designed to perform a safety function can fail. Each safety system identified in the event tree as involved in an accident is examined to determine how failures of components within that system could cause the failure of the entire system.

To provide an illustrative example of an accident analysis, one potential initiating event is the failure of the hoist used to lower canisters to the emplacement areas of the repository. If a hoist failure is postulated, subsequent events that need to be examined in an event tree will be the response of a canister to the fall (i.e., whether it breaches on impact), the type and amount of radionuclides J released in the vicinity of a canister if it does breach, l

the ability of monitoring systems to detect a release and activate mitigating systems, and the response of the l

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i mitigating systems to a release. -If failure of a mitigating system could contribute to an off-site dose,

!~ individual componer,ts within-the mitigating system need to be reviewed, using fault tree analysis, to determine the effect of their failure on performance of the overall i system. For example, individual components in the i ventilation system which might be analyzed include dampers, motors, and filters.

1 To the extent practicable, probabilities of scenarios and

-releases will need to be developed. The combination of a hoist failure, canister breach, detection system failure, and ventilation system failure may be so unlikely that this scenario does not need to be considered. One of the i

difficulties in this phase of the program is that there 4- will be little reliability data for structures, systems.

- and components unique to the repository program for use in

• risk assessments. In such cases, the staff expects that conservative bounding assumptions will be made and j justified to demonstrate that the subject equipment will be reliable.

! (2) Consequence analysis to evaluate the consequences of i

accident scenarios identified in event / fault tree analyses i

to determine the amount and kind of radionuclides which may

reach the unrestricted area and contribute to an off-site

> dose. Consequence analysis includes identification of a i source term for radioactive releases and evaluation of 1 mechanisms for movement and deposition of radioactive 1

materials released from the HLW facility. The energy, magnitude, and timing of radiological releases resulting from various accidents need to be considered in this l analysis.

All structures, systems, and components of the geologic repository whose failure could initiate an accident which, if unmitigated, could cause an off-site radiation dose of 0.5 rem

should be on the Q-list. The staff also considers that ,

structures, systems, and components whose' failure may initiate

such an accident cannot be removed from the Q-list due to the i addition of mitigating features. It is possible, however, that systems, structures, and components capable of performing accident mitigating functions can be eliminated from the repository design or removed from the Q-list if DOE can show j

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that the need for such mitigative functions is so unlikely as to be incredible.

< For the purpose of design basis accident calculations, justification such as reliability data, large design margins, and in-service and qualification testing will need to be

• provided by DOE in order to demonstrate that failure of an item '

is not credible. The validity of su:h justification will need to be assured by application of a 10 CFR 60 Subpart G QA

program.

The NRC Office of Nuclear Regulatory Research has an ongoing l study of pre-closure accidents. The latest report published under this study (Harris, et. al.,1985) contains exan.ples of 1 initiating events, accident scenarios, and discussions of possible consequences using a specific design for a HLW facility.

(b) Redundancy The use of redundant structures, systems, and components is one 3- method for providing additional assurance that necessary safety functions will be performed if an accident occurs. In a redundant system, the failure of one train of the system will not compromise or prevent the safety functions from being performed. For the high-level waste repository, several sections of the regulations in 10 CFR 60 [60.131(b)(5)(ii) ano

60,131(b)(10)(iv)] refer to the need for redundancy, but for the most part, the decision is left to the designer. The use of PRA should indicate those areas where structures, systems. and components are not sufficiently reliable to perfom their safety

functions without redundancy.

(c) Non-Mechanistic Failures

' 'Non-mechanistic failures, (i.e., postulated failures which are i not based on previously observed failure modes but which are

conservatively assumed to maximize consequences) need not be

used in safety assessments. In cases where data.are insufficient to support PRAs, however, non-rechanistic failures

, may need to be assumed, i

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(d) Use of Previously Established Criteria Many criteria and standards have been developed in the reactor program and other nuclear programs which may be applicable for the geologic repository program. For example, there are regulatory guides covering design basis earthquakes, floods, and tornado wind velocities which may be used in designing the HLW facility and developing the associated Q-list. The staff recognizes that certain of the standards for nuclear power reactors may not be directly applicable to the preclosure phase of a geologic repository. However their use, to the extent practicable, would eliminate the need to develop acceptable new approaches.

A recent DOE-sponsored publication, Evaluation of Regulatory Guides Potentially Useful to Geologic Repository Development" (Chang,1986), although not yet endorsed by DOE or NRC, is a useful source of information and provides a basis for future discussions between DOE and NRC staffs.

(e) Retrieval The option for retrieval of waste is addressed as a performance objective in 10 CFR 60.111(b). Analyses of retrieval operations need to be conducted by DOE to identify Q-list items and activities and to assure that, in the event retrieval is 4 necessary, the retrieval process will not expose the public to radiation levels above 0.5 rem. These analyses need to include, for example, re-mining operations, conveyance shafts, and

, equipment which Iray be used in the retrieval' process. These

! analyses should be conducted in a timely manner so that the items and activities related to planning, characterization, design, and construction of these items can be evaluated and covered by the applicable QA requirements.

6.3 Identification of Barriers Important to Waste Isolation i

The term " barriers important to waste isolation" (10 CFR 60.151) refers to those natural or engineered barriers that' contribute to meeting the containment and isolation performance objectives of 10 CFR 60 Subpart E. The four primary performance objectives for waste 17

isolation atter permanent closure are stated in 10 CFR 60.112 and 60.113 and include:

o ground water travel time, o waste package containment period, o maximum yearly release rate from the engineered barrier system, o the overall system performance objective in 10 CFR 60,112 for release of radioactive materials to the accessible environment (the EPA standard in 40 CFR Part 191).

The items and activities important to waste isolation consist of the barriers relied on to meet these performance objectives and should include:

o components of the engineered barrier system (waste package and underground facility), shafts, boreholes and their seals, o components of the natural barrier system (e.g., host rock, and geochemical retardation characteristics),

o items and activities necessary to support the determination of whether the performance objectives will be met, including collection of data to characterize the site or performance of engineered barriers, and o items and activities in the preclosure phase that could affect post-closure performance.

The identification of the barriers on the Q-list is relatively simple for the post-closure phase since there are so few. Identification of activities related to these barriers is more difficult as the activities to be considered include, for example, characterizatian of the site, design and testing of engineered barriers, and specific work activities such as drilling, inspecting, and data reduction.

The broad performance objectives for waste isolation provide DOE with some flexibility in allocating credit among the various components of the natural and engineered barrier systems to meet each objective.

For example a 300 to 1000 year lifetime for the waste package might be achieved by a combination of performance from each of the components in the waste package or by a single component, such as the 18

canister. The allocation of performance among the various components of the natural and engineered barrier system will provide the basis for determining which barriers may be important to waste isolation.

Performance assessments shall be conducted on these barriers to ascertain that those relied on will meet the waste isolation and containment performance objectives of 10 CFR Part 60. 00E is expected to allocate performance'among barriers based on available data before site characterization begins. These initial allocaticns of performance will provide a basis for determining how much site characterizaton testing will be needed. For barriers contributing substantially to the performance objectives, niore cata will be needed than for those with little or no contribution. The initial allocation of performance among the barriers is likely to char.ge based on new data and the results of performance assessments.

Prior to and during the early phases of site characterization, when the relative importance of items and activities to waste isolation is not known, all activities DOE plans for characterization of barriers as well as those barriers themselves should be covered by a 10 CFR 60 Subpart G QA program. At the time of the license application, specific barriers may not need to be on the Q-list if they will not be relied on to meet the system performance objectives and if they will not detract from the performance of those barriers which will be relied on. This will only be evident after extensive data are collected and analyzed in performance assessments. Until that time, it is prudent to censider all barriers for which characterization activities are planned to be on the Q-list. Failure to apply QA in a manner comensurate with the potential importance of an item or activity may adversely affect the credibility of information used in the license application.

In determining which barriers are important to waste isolation and how much each contributes to performance, the DOE should utilize as a basic reference the guidance given in the NRC staff's " Draft Generic Technical Position on Licensing Assessment Methodology for High-Level Waste Geologic Repositories" (USNRC, 1984). This document presents information on the identification of credible scenarios, determination of the likelihood of these scenarios, development of conceptual models that describe the scenarios, formulation of mathematical models that are consistent with the conceptual models, incorporation of data and associated uncertainties into the numerical mcdels, assessment of the consequences of the scenarios, and the comparison of the results with numerical performance objectives.

19

6.4 Staff Information Needs Q-list items and activities should be identified prior to conducting related work or purchasing related items and services. Certain items and activities represent obvious or likely candidates for the Q-list.

During operation, systems important to safety may include the ventilation system, waste transport system, instrumentation, the electrical system, and the rock support system. After emplacement of the wastes there are various components or systems that are likely to be important to isolation, such as the waste form, backfill, waste package, and natural barriers. Major site characterization activities to be on the Q-list include waste package and backfill testing, borings, and in-situ testing at depth.

(a) License Application DOE should submit with the license application a description of the QA plans to be applied to items and activities during

! construction, the period of operations, and performance

! confirmation. A final list of items and activities important to safety and waste isolation should be provided with the description of the 10 CFR 60 Subpart G QA program to be provided i in the license application [10 CFR 60.21(c)(4)]. Analyses to identify systems, structures, and ccmponents important to safety shall also be provided in the license application [10 CFR 60.21(c)(1)(ii)(E)and60.21(c)(3)]. DOE should furnish information supporting development of the final Q-list and the applicable 10 CFR 60 Subpart G QA program in the periodic SCP updates so that the staff will be thoroughly familiar with and in close agreement on the content of the list prior to the license application.

l (b) Site Characterization Plans Prior to submission of the license application, 00E is required to submit a site characterization plan. In order for the NRC staff to identify licensing issues early so that they may be resolved by the license application, the methodology for determining the scope of items and activities important to safety and waste isolation, a preliminary Q-list, and a description of the 10 CFR 60 Subpart G QA Program applicable to items and activities on the Q-list for the site characterization i phase should be provided. While changes in the level of detail i

i 20 1

1

and content of the Q-list are likely to occur between the SCP and the license application based on an increased level of knowledge and maturity of design, DOE should provide a provisional Q-list in the SCP based on available information.

This provisional list should include items and activities important to safety and waste isolation and should be supported by conservative analyses to assure all potential items and activities are identified at least at the system and major component level. As the design matures and more information is collected, items may be removed from or added to the Q-list.

Changes of this type are expected and should be documented with supporting analysis and rationale. The staff will periodically review the process for adding and removing items from the Q-list to assure that it is adequate and reliable. It is important to cephasize the need for conservativism in ceveloping the provisional Q-list as information on items and activities which are added to the Q-list after site characterization activities begin that was not collected under a 10 CFR 60 Subpart G QA program may not be aoequate to support licensing.

The staff acknowledges that it is difficult to determine what items will be important to safety prior to site characterization due to limited site data and design detail. With the limited data base available at this time, application of a rigorous probabilistic analysis may not be practicable. Accident scenarios including initiating events as well as dose consequences for accidents will therefore need to be identified and estimated using conservative engineering judgment. The available information base may include data collected and analyzed for other similar activities, such as external events for reactor facilities and design basis accidents for independent spent fuel storage installations (ISFSIs) and refueling operations at nuclear power plants where these can be shown to apply directly to the high-level waste facility.

Although the repository operational system represents a new and unique nuclear facility, comparisons can be made with similar nuclear facilities in order to facilitate kncwledgeable decisions and avoid duplication of effort. Use of analyses for similar facilities should be carefully conducted and the information obtained should be rigorously examined to assure that key differences in facilities have not been overlooked.

Activities will be conducted during the site characterization phase that may not relate to items on the Q-list but will be 21

used to support findings on other licensing requirements.

Therefore in addition to describing the QA program for items and activities on the Q-list, plans for development and implementation of a program to assure and. document that ncn-Q-list requirements are met should be described in the SCPs.

6.5 Graded Application of QA Measures The amount of QA necessary to provide adequate confidence in the quality of the items and activities within the scope of the Q-list may vary. This flexibility is provided in 10 CFR 60 Subpart G through reference to 10 CFR 50 Appendix B. Criterion II of Appendix B states that the QA program shall provide for control over activities affecting the quality of structures, systems, and components to an extent consistent with their importance to safety.

It is expected that probabilistic and performance assessment analyses will provide qualitative and quantitative evaluations of the importance to safety or waste isolation of particular structures, systems, or components for developing the scope of the QA program.

These evaluations can then provide a logical framework for application of graded QA measures.

NQA-1 (ANSI /ASME, 1983) is a consensus standard which provides more detailed guidance on grading QA for nuclear facilities. NQA-1 Appendix 4A-1 contains these criteria for grading QA measures, which i have been slightly modified for repository work:

1 o The impact of malfunction or failure of the item on safety or waste isolation. For example, components may play a major role in safety, perfonn supporting functions for primary equipment, perform redundant functions (i.e., two items may perform identical functions but only one may be needed to prevent or

mitigate an accident), or perform functions for low consequence events or accidents with very low probabilities of occurrence.

l QA measures should be applied to a degree consistent with the l

importance to safety or isolation of a specific item or t activity. Likewise, data will vary in degrees of importance to i

safety or waste isolation.

o The ccmplexity of design and fabrication of an item or design and implementation of a test, or uniqueness of the item or test.

First-of-a-kind items or tests or complex items or tests may 22 l

t

(

require extensive design efforts or extensive inspection or peer review during their development to assure satisfactory results.

o The special controls and surveillance needed over processes and equipment. Processes and equipment which affect the quality of components, data or analyses and whose effects on the components, data or analyses cannot be easily measured or evaluated in the final product, such as welding and heat treatment, should be controlled as prescribed by Criterion IX of 10 CFR 50 Appendix B.

o The degree to which functional compliance can be demonstrated by inspection or test. Proof of the quality of a component can sometimes be demonstrated by inspection and/or testing of a final product. In such cases, the in-process control program' may be reduced. The limiting case is whether an end' product test can properly assess the degree of compliance to quality requirements and thereby eliminate the need for in-process control.

o The quality history and degree of standardization of the item or test. If a manufacturer or organization has been producing a particular standard item or conducting a standard test for a long period and if the quality history of the item or test-indicates acceptable performance, QA measures may be applied to that item or test to reflect the demonstrated performance.

Conversely, if certain characteristics are determined to be unsatisfactory based on operational data, additional QA measures may be required to assure that experienced deficiencies are identified and corrected or controlled.

It is expected that probabilistic and performance assessment analyses can provide qualitative and quantitative measures of the importance to safety or waste isolation of particular structures, systems, or ccmponents for developing the scope of the QA program. These measures should then provide a logical starting point for grading quality assurance elements.

In implementing the above guidance for items important to safety, the amount and types of inspection, testing, and record keeping will be the most variable measures in the grading program. The following examples of how grading can be accomplished are helpful in interpreting the guidance given above. Assume that a standard off-the-shelf radiation detector is used in the pre-closure phase as 23

T 1

I l one of several indicators that an accident has occurred and that safety systems, such as the ventilation system, need to be activated.

l If this detector is on the Q-list due to the safety functions it i performs, the designer working under the 10 CFR 60 Subpart G QA

program would be expected to review the information supplied by the manufacturer on the range, accuracy, power requirements, i environmental conditions under which the specific radiation detector l will operate and other pertinent information about the device and its use. Following this review, the designer would confirm that the g available instrument would be suitable for its application in the facility,. or make revisions to the design to accomodate the instrument. After specifying or referencing the necessary requirements in the purchase order, a receipt inspection would be conducted to assure that the instruments received were in accordance l with the purchase order. After receipt, the instrument would be -

! installed, calibrated, tested, and maintained in accordance with the requirements of the manufacturer and/or designer. The repository designer and constructor-should have the required Subpart G QA programs in place for the work that they performed on the-device and for selecting and auditing the supplier, while the manufacturer of 1

the device may only implement a portion of a Subpart G QA program.

The specific controls selected by the designer's and constructor's.

. organizations, along with the manufacturer's standard controls, could suffice.

As a contrasting example, the hoist used to transport waste packages can be considered. If failure of the hoist had the potential for

! contributing to an off-site dose of 0.5 rem or greater and assurance were needed that such a failure were not credible, essentially all l elements of the Subpart G QA program might be' applied to the design, fabrication, installation, and inspection of the hoist. The supplier 2

of the hoist might be required to establish its own Subpart G QA program including all elements of Appendix P, with in-process inspections, design reviews, training programs, etc. Also the

purchaser would probably conduct surveillances, source inspections, l and audits of the hoist supplier to assure that the QA program was t being carried out properly. Numerous detailed and complex inspections might be required to assure that the device were safe and reliable. The records associated with these activities would be extensive.

For items and activities important to waste isolation, QA measures

! should not be significantly graded in the early phases of site l

characterization. The characteristics of individual components or l'

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phenomena of the natural or engineered systems are not well known prior to extensive data collection and analysis and their contribution to meeting the numerical performance objectives of 10 CFR Part 60 cannot be established with certainty at that time. Until a defensible basis for grading QA measures can be established, a ,

conservative level of QA should be applied to testing and design of barriers in the event that subsequent data analyses show them to be important in meeting the isciation and containment requirements of 10 CFR Part 60. Scre flexibility in this approach may be permitted if conservative performance goals are established for individual components or phenomena and for routine or simple tests or components.

r During the field investigations, the amount of inspection and control that is placed on the various activities may vary due to the complexity of the tests and the amount and importance of the information to be collected. If, for example, a boring is planned for the sole purpose of obtaining an additional ground water level measurerent to confirm existing gradients, it may not be necessary to

} place a large number of controls on the drilling of this boring; for exemple, a full-time inspector may not be required to observe the drilling. If, on the other hand, the purpose of the boring was to obtain accurate measurements of the orientation of joints and fractures, and to obtain ground water samples for chemical analysis, the controls which should be imposed would be more stringent. The drilling rig may need to be inspected prior to operation to assure that the appropriate equipment and procedures were in place to achieve the desired objective. Any drilling fluid introduced into

- the hole would likely be analyzed and have a tracer added, and controls might be placed on the types of additives which could be used in the hole. In this example, it would be likely that a full-time inspector would be on the drill site to assure that all procedures were followed and documentation completed. Boring completion and testing of this hole should be dccumented in a detailed series of field reports. In both cases, the NRC staff would expect documentation showing that information was obtained correctly.

However the level of effort would be considerably different for these two situations.

4 25 4

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7.0

SUMMARY

This GTP provides guidance on approaches the staff considers acceptable for identifying . items and activities within the scope of the 10 CFR 60 Subpart G QA program. It also gives guidance on how to apply QA to these and other items and activities in order to demonstrate compliance with the requirements of 10 CFR Part 60. DOE should use PRA techniques, to the extent practicable, to identify items and activities important to safety. DOE should identify barriers it..portant to waste isolation based en performance allocation and performance assessments. The final Q-list should be contained in the description of the 10 CFR 60 Subpart G QA program included in the license application and a provisional Q-list should be presented in the SCP. The provisional Q-list should be based on available data, engineering judgment, and conservative assumptions. In addition to addressing items and activities on the Q-list, DOE should apply an appropriate level of QA to non-Q-list items and activities that will be needed to support other licensing requirements, such as those in 10 CFR 60.131 for radiological protection of workers. DOE may apply graded QA measures to items and activities on the Q-list based on the criteria in NOA-1 Appendix 4A-1 (ANSI /ASME, 1983) when information is available to support such grading.

1 26

4

8.0 REFERENCES

T d

American Nuclear Society (ANS) and Institute of Electrical and Electronics Engineers (IEEE). 1983. PRA Procedures Guide, NUREG/CR-2300, Washington, DC.

American Society of Mechanical Engineers (ASME).1983.

Ouality Assurance Program Requirements for Nuclear Facilities, ANSI /ASME NOA-1, Bew York, NY.

Chang, W. Y. 1986. Evaluation of Regulatory Guides Potentially Useful to Geologic Repository-Development, BMI/0NWI-588, Ebasco Services, Columbus, OH.

Harris, P.A., D.M. Ligon, M.G. Stamatelatos. 1985. High-Level Waste Preclosure Systems Safety Analysis Phase 1, Final Report, NUREG/CR-4303 (SAND 85-7192), GA Technologies, Inc., San Diego, CA.

U. S. Nuclear Regulatory Commission. 1979. Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment, Regulatory Guide.4.15, Washington, DC.

U. S. Nuclear Regulatory Commission. 1984a. NRC Review Plan: Quality Assurance

! Programs for Site Characterization of High-Level Nuclear Waste Repositories, 4 Washington, DC.

V. S. Nuclear Regulatory Commission. 1984b. Draft Generic Technical Position on Licensing Assessment Methodology for High-Level Waste Geologic Repositories, Washington, DC.

U. S. Nuclear Regulatory Commission. 1985. Draft Generic Technical Position on Waste Package Reliability Analysis, Washington, DC.

i U. S. Nuclear Regulatory Commission. 1986. Draft Generic Technical Position on Qualification of Existing Data for High-level Nuclear Waste Repositories,

. Washington, DC.

1 27

l 9.0 BIBLIOGRAPHY Tenera Corporation.1985. Evaluation of the Waste Isolation Pilot Plant Classification of Systems, Structures and Components, EEG-30, Contract Report for the State of New Mexico, Berkeley, CA.

Nathwani, J.S. 1986. Safety Assessment of a Conceptual Design of a Facility for Geological Disposal of Candu Nuclear Fuel in the Canadian Shield, Paper presented at the International Atomic Energy Agency Symposium on the Siting, Design and Construction of Underground Repositories for Radiological Wastes, March 3-7, 1986, Hanover, Germany.

o

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28

l APPENDIX A GLOSSARY 9

1 29

Accessible environment: (1) The atmosphere, (2) the land surfaces, (3) surface water, (4) oceans, and (5) the portion of the lithosphere that is outside the controlled area (10 CFR 60.2). The overall system performance for the geologic repository is calculated at this boundary.

Backfill: Material used to fill access tunnels, shafts, and other openings, excluding waste emplacement holes, and forming part of the underground facility. (An example is bentonite clay mixed with crushed rock and secondary minerals used to fill repository drifts).

Consequence analysis: A method by which the consequences of an event are calculated and expressed in scme quantitative way, e.g., money loss, deaths, or quantities of radionuclides released to the accessible environment.

Design basis: Information which identifies: a) the specific functions to be performed by the structures, systems, or components of a geologic repository; b) assumptions regarding design controlling parameters; c) the specific parameter values selected as a basis for the design; and d) the supporting rationale for assumptions and parameter value selection.

Design process: An iterative process of developing a geologic repository design from the preliminary stages where the level of uncertainty in design inputs is high, to a final stage where the level of uncertainty is icw enough to meet established performance criteria.

Engineered barrier system: The waste packages and the underground facility (10 tFR 60.2). The maximum radionuclide release rate is measured at this boundary (10 CFR 60.113(a)(1)(ii)(B)).

Finding: An assessment of compliance or noncompliance with a specific requirement. A finding addressing a numerical performance objective will be reached after the following are weighed; the results of a reliability analysis and the laboratory and field tests upon which it is based, expert opinion, and empirical studies.

Geologic repository: A system which is intended to be used for, or may be used for, the disposal of radioactive wastes in excavated geologic media. A geologic repository includes: (1) the geologic repository operations area, and (2) the portion of the geologic setting that provides isolation of the radioactive waste (10 CFR 60.2).

Geologic repository operations area: A high-level radioactive waste facility that is part of a geologic repository, including both surface and subsurface areas, where waste handling activities are conducted (10 CFR 60.2).

30

High-level radioactive waste (HLW): (1) Irradiated reactor fuel, (2) liquid wastes resulting from the operation of the first_ cycle solvent extraction 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 (10 CFR 60.2).

Licensing assessment: An assessment of whether a license application complies with all of the requirements that it purports to meet. For this program it is the sum of the individual findings for each of the requirements of 10 CFR Part 60.

Packing: The material that is placed in the waste emplacement hole in the annular space between a canister or overpack (if one is present) and the host rock. The packing is a component of the waste package which serves to control the release of radionuclides from the waste package by sealing against water, modifying the water chemistry, sorbing or retarding the transport of radionuclides or by establishing other improvements in environmental parameters. (An example is a mixture of bentonite clay and crushed rock placed in the annulus between the overpack and host rock).

Performance assessment: The process of quantitatively evaluating component and system behavior, relative to containment and isolation of radicactive waste, to support development of a high-level waste repository and to determine compliance with the numerical criteria associated with 10 CFR Part 60.

Performance confirmation: The program of tests, experiments, and analyses which is conducted to evaluate the accuracy and adequacy of the information used to determine with reasonable assurance that the performance objectives for the period after permanent closure will be met (10 CFR 60.2).

Quality assurance: Those planned and systematic actions necessary to provide adequate confidence that the geologic repository and its subsystems or components will perform satisfactorily in service. Quality assurance includes quality control, as defined below.

Quality control: Those quality assurance actions related to the physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, system, structure, or component to predetermined requirements (10 CFR 60.150).

Reliability: The probability that a system or component, when operating under stated environmental conditions, will perform its intended function adequately for a specified interval of time.

31

Reliability analysis: An analysis that estimates the reliability of a system or component.

Risk: A measure of the probability and severity of adverse effects (consequences); the expected detriment per unit time to a person or population from a given cause.

Risk analysis: An analysis that combines estimates of the probabilities of scenarios with estimates of the consequences of those scenarios, while considering the uncertainties associated with both.

Scenario: An account or sequence of a projected course of action or event.

Scenario analysis: The process of identifying scenarios and estimating the probability of their occurrences.

Site: The location of the controlled area (10 CFR 60.2).

Site characterization: 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 particular site relevant to the procedures under this part. Site characterization includes bcrings, surface excavations, excavation of exploratory shafts, limited subsurface lateral excavations and borings, and in-situ testing at depth needed to determine the suitability of the site for a geologic repository, but does not include preliminary borings and geophysical testing needed to decide whether site characterizaticn should be undertaken (10 CFR 60.2). .

Site characterization plan: A general plan for site characterization activities for a candidate site for a high-level waste repository, as required in the Nuclear Waste Policy Act and 10 CFR Part 60.

System or cornonent performance: How each element or a combination of all elements of t1e engineered barrier system of the geologic repository contributes to meeting the numerical performance objectives set forth in 10 CFR 60.113. ,

Waste package: The waste form and any containers, shielding, packing and other absorbent materials imediately surrounding an individual waste container (10 CFR 60). The minimum waste packa boundary [10 CFR 60.113(a)(1)(II)ge (A)]. containment time is calculated at this 32

i

Underground facility

The underground' structure, including openings and j backfill naterials, but excluding shafts, boreholes, and their seals (10 CFR 60.20).  :

i Waste form: The radioactive waste materials and any encapsulating or stabilizing matrix (10 CFR 60.2).

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