memorandum for: from: rkrver-changes rather than simple dramatic differences, so that there is...

sebXCAj A- r

- ,q 9

-- ftft-dJ L/84/07/23/0

JUL 16 4

ibution

N4SS r/fCFREBrowningMJBellPAltomareLBHigginbothamMRknappLBarrettHJMillerJLinehan & r/fRRBoyle

!r SMCoplanJKennedyPDR D Rkrver-

iY DOCUMENT (EXECUTIVE OPS

MEMORANDUM FOR:

FROM:

SUBJECT:

Hubert J. Miller, ChiefRepository Projects BranchDivision of Waste Management

John J. Linehan, Section LeadeRepository Projects BranchDivision of Waste Management

COMPLETION OF MODELING STRATEGPLAN COMMITMENT: 5313112)

Enclosed is a copy of the Modeling Strategy Document recently completed by

D. Fehringer. Drafts of the document have been reviewed during preparation by

WMEG, WMGT, and RES. Any significant comments received have been resolved and

factored into the revised document. WMPC has elected not to comment on the

drafts due to higher priority work.

John J. Linehan, Section LeaderRepository Projects BranchDivision of Waste Management

Enclosure:Modeling Strategy Document WM Record File

I0'9. 9WM ProjecLt .JiniDocket No.

PDR ZLPRD

Distribution:

_Return to WM, 623SS) OW

OF : RP kg :

NAM <:8Q1 Ghan :

840803004 34ODAT 7 4 PDR WASTE_______ _____ W -1 PDR _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

clas t-ke- 7 ̂699-2 60I--4-a_ ,8

I

4

REVISED

MODELING STRATEGY DOCUMENT

FOR

HLW PERFORMANCE ASSESSMENT

I INTRODUCTION

This document describes the Nuclear Regulatory Commission (NRC) staff's

strategy for using numerical models* and computer codes for evaluating

the performance of high-level waste (HLW) repositories. The purpose of

this modeling strategy document is to establish the overall philosophy

and approach for development, evaluation, and application of numerical

models and computer codes by NRC staff and contractors in sufficient

detail to aid in planning for staffing and technical assistance needs

in this area. This document identifies the specific areas for which

the NRC staff plans to independently perform numerical analyses to

evaluate compliance with applicable standards and regulations, and

describes (in general terms) the nature of the independent analyses

planned.

*As used In this document, the terms "numerical model" and "numericalanalysis" refer to those models and analyses which are sufficientlycomplex to require computer applications. While less detailed analysesmay be important in evaluating a license application, it is anticipatedthat such analyses can be conducted with relatively little advancepreparation. In contrast, the larger resource requirements and longerlead times involved in developing computerized analysis capabilitiesrequire an overall strategy to guide the development of these analyticalcapabilities.

3111/DJF/84/02/14/0

The details of model development and application will be discussed in

other documents to be developed including technical positions related to

specific technical disciplines.

The scope of this document is intended to encompass those numerical

modeling and computer related activities needed to fulfill NRC

responsibilities during site characterization and when reviewing a

license application. These activities include (1) establishing licensing

information needs, (2) evaluating site characterization data and

providing additional guidance to the Department of Energy (DOE), (3)

preparing to review a license application, and, finally, (4) reviewing

the application. The development and evaluation of conceptual models is

recognized as an important preliminary step in the application of

numerical models, but is not the main focus of this document. Similarly,

the level of detail required for a particular application (e.g., 3-

versus 2- or 1-dimensional models) is not discussed in this document.

The details of implementing a modeling program, including such activities

as verification and documentation (ref. 1) of computer codes, are also

considered to be outside the scope of this document. This document does,

however, discuss the broad requirements of an adequate licensing

assessment methodology to set the background for identifying specific

mathematical modeling strategies.

3111/DJF/84/02/14/0- 3-

The following is an outline of this document:

Section I--The nature and form of applicable standards and regulations,

and examples of the types of numerical analyses which might be used by

the NRC staff to determine compliance with these standards and

regulations.

Section III--The roles of the Department of Energy (DOE) and the NRC in

demonstrating and evaluating repository performance.

Section IV--A discussion of the uncertainties which affect the

development of a modeling strategy.

Section V--The key assumptions on which the modeling strategy is based.

Finally, in Section VI, the modeling strategy is described.

(Sections II-IV contain information similar to that described in other

documents ncluding the staff's draft technical position on Licensing

Assessment Methodology. These sections are included in this document to

set the background for development of a modeling strategy.)

3111/DJF/84/02/14/0-4-

The Appendix presents a glossary defining a number of terms used in this

document.

II STANDARDS AND REGULATIONS

Nature of the NRC Regulation

The principal portions of Subpart E of 10 CFR 60, and examples of the

types of numerical analyses which may be used to determine compliance

with them, include:

(1) Prior to Permanent Closure

§60.111(a), limiting radiation exposures and releases of radioactive

material during operations. Numerical analyses may be used to estimate

source terms for potential releases, transport of radionuclides by the

repository ventilation system, movement of radionuclides through the

environment, and the resulting doses to members of the public.

§60.2, defining the term "important to safety" in terms of "engineered

structures, systems, and components essential to the prevention or

3111/DJF/84/02/14/0-5-

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." The use of numerical analyses for evaluating

potential accident sequences is expected to be the same as described in

the preceding paragraph.

§60.111(b), requiring that the option of waste retrieval be preserved

during operations. Numerical analyses may include heat transport in the

repository system, structural analyses for the waste packages and/or the

underground facility, and estimates of waste package degradation.

(2) After Permanent Closure

§60.112, limiting releases of radioactive materials to the accessible

environment after permanent closure to those permitted by the EPA

standard (proposed 40 CFR 191). (The nature of the EPA standard is

discussed in the following section.) Numerical analyses will include

flow of groundwater into and through the repository system and transport

of radionuclides from the waste form to the accessible environment as

illustrated in Figure 1.

DIAGRAMMATIC PLAN VIEW Inot to scale) DIAGRAMMATIC CROSS SECTION VIEW (not to scale)

ACCISSISLE tIANVICONET.-EPA STANDARD APPLIES

-.-.... a Al~~ LIAH! Of ACCESSLEBEA PoiO %%! sINVIaNCINNIN

PERFORMANCE ISSUES

P.3 When end how do"e water contact the backill?

P4 When md how doe water contact th waste packge?

P.S When and how doe water contact the waste lom?

P.S When, how, a at wht rate ate radionuclidee releaed

from the waste form?

P.7 When, how, and at what rate are radionucbde releasedftmo th waste pckge?

PAS When, how, and at what ia e radionucides releasedfrom th beafil?

P-3 When, how, and at what rate are radiond releasedfrom t disturetd zone?

P.10 When, how, and at what rate e radoid telased

from the hr field to th accesaible anvronnt?

P.t1 What is the prewate emplacement groundwatertrael Wm along th fastt path of raionucidatravel from thc diet d aone to t accsleanvionmm?

FIGURE 1 REPOSITORY SYSTEM ELEMENTS AND PERFORMANCE ISSUES RELATED TO LONG-TERM

PERFORMANCE AFTER PERMANENT CLOSURE

-

3111/DJF/84/02/14/0

§60.113(a)(1)(ii)(A), requiring a minimum waste package containment time.

Numerical analyses of waste package degradation may nclude structural

analyses, extrapolations of corrosion data obtained by accelerated

testing, and geochemical estimates of the waste package environment under

the influence of heat and radiation.

§60.113(a)(1)(il)(B), limiting the radionuclide release rate from the

engineered barrier system. Numerical analyses may be used to extrapolate

laboratory-generated leaching data, to estimate solubillty-limited

radionuclide releases, to evaluate containment by backfill materials, and

for estimates of geochemical conditions in the engineered barrier system.

§60.113(a)(2), addressing the minimum pre-emplacement groundwater travel

time from the disturbed zone to the accessible environment. Numerical

analyses may include thermal or coupled thermal-hydrologic analyses to

determine the physical extent of the disturbed zone, and groundwater flow

analyses to estimate travel times.

§60.122, addressing favorable and potentially adverse siting conditions.

Numerical analyses may include estimates of the effects of favorable or

3111/DJF/84/02/14/0-8-

potentially adverse conditions on achieving compliance with any of the

criteria discussed above.

For detailed discussions of the important terminology and points of

consideration in the regulation, the reader should consult the

Supplementary Information accompanying the publication of the Final Rule

(48 FR 120, 28194 - 28229, June 21, 1983).

Nature of the Proposed EPA Standard

As discussed above, §60.112 establishes the EPA standard as the overall

release limit for a repository system. The (draft) EPA standard is a

probability-based standard for which d formal probabilistic treatment of

releases similar to the probabilistic risk analyses used for nuclear

power plants and other applications would be required as one of the bases

for evaluating repository acceptability. (As the EPA standard has not

been finalized, the staff notes that the extent to which probabilistic

risk analysis will be formally required in the licensing process has not

been finally determined. This document assumes that a probabilistic risk

analysis will be required as one of the bases for evaluating repository

acceptability, although the results will not be the sole basis for such a

judgment.)

3111/DJF/84/02/14/0- 9

During the formal licensing process, the post-closure radionuclide

releases will be assessed, as well as the uncertainties associated with

those releases, and the results will be compared with the EPA containment

requirements. The NRC staff anticipates that the estimates of releases

and uncertainties will be used to construct a Complementary Cumulative

Distribution Function (CCDF), which is a curve displaying the probability

that releases will exceed a given value. Analyses of HLW repository

performance are subject to a number of sources of uncertainty, as

discussed later in Section IV. A CCDF s a convenient format in which

all quantifiable uncertainties, including data uncertainties and the

likelihood of disruptive scenarios, can be combined into a single

probability istribution of releases. The CCDF can then

be compared to the EPA containment requirements to determine whether the

repository will comply with the release limits established in the

standard.

The NRC staff has identified three major issues which must be addressed

to construct CCDF's and successfully defend them during licensing.

First, it is not clear that there Is a consensus in the technical

community that it is appropriate or even possible to construct meaningful

CCDF's for geologic repositories. Second, there are alternative theories

associated with geologic processes and engineered barrier failure modes

3111/DJF/84/02/14/0- 10 -

that must be reconciled to reach a consensus about the probabilities and

consequences of some significant scenarios. Third, there are many

processes and conditions which differ from others by a series of small

changes rather than simple dramatic differences, so that there is

effectively a continuum of scenarios associated with some types of

events. Selection of and differentiation between various scenarios

therefore may not be obvious, and it is appropriate to try to stimulate

discussion prior to licensing as to how this best may be done.

III ROLES OF DOE AND NRC

Prior to submitting a license application for a HLW repository, the

Department of Energy (DOE) must carry out a program of site

characterization to explore and evaluate a candidate site. DOE will

describe its plans for this program in its Site Characterization Plan

(SCP) and periodic updates. The SCP and updates are expected to include

plans for model and code development as well as the results of analyses

used to guide the development of the site characterization program. The

NRC will review and comment on the SCP and the progress of the site

characterization program.

3111/DJF/84/02/14/O- 11l -

At the time of licensing for a proposed deep geologic repository for

high-level radioactive waste, DOE has the responsibility to present and

defend a complete evaluation of the geologic repository system and its

components as required by 10 CFR Part 60. To receive authorization to

construct the repository, DOE must demonstrate through the Safety

Analysis Report (10 CFR 60.21) chat there is reasonable assurance that

the types and amounts of radioactive materials described in the

application can be received, possessed and disposed in a repository of

Lhe design proposed without unreasonable risk to the health and safety of

the public (10 CFR 60.31. (a)).

The burden of proof regarding compliance with regulatory criteria lies

with DOE. Therefore, the NRC staff does not need to take the initiative

for model or code development or application, although the NRC staff may

consider it appropriate to do so in selected areas in order to be

prepared for a thorough and timely review of a license application. NRC

staff code development will be for the purpose of independently

evaluating DOE submittals and site characterization activities, and not

to remedy perceived deficiencies in DOE's program.

The NRC staff will independently review DOE's description of the proposed

geologic repository (10 CFR 60.31 (a)(1)) and DOE's assessment of the

3111/DJF/84/02/14/0- 12 -

performance of the site and design with respect to the performance

objectives and criteria contained in Subpart E (10 CFR 60.31 (a)(2)).

The staff will critically evaluate and comment in detail on DOE's

numerical analyses, emphasizing the completeness and adequacy of (a)

models and model inputs, (b) uncertainty analyses, and (c) alternative

interpretations. In addition, the staff anticipates performing

independent assessments of compliance with the performance objectives and

criteria of Subpart E in selected areas. (The responsibility for

generation of data lies with DOE, and ndependent NRC assessments will

use DOE's data. The NRC cannot generate data to substitute for DOE's

lack of data gathering.) It is important to recognize that there Is no

statutory or regulatory responsibility for independent numerical analyses

by the NRC staff. The level of independent review is a matter of

technical judgment and policy. The minimal action which the NRC must

take to deal with an application is to pose sound technical questions to

DOE and evaluate whether the DOE responses provide reasonable assurance

that the criteria will be met. The ability to ask such questions will

not necessarily require the capability of exercising any computer codes.

It will be necessary to understand all significant phenomena

quantitatively, so that the mathematical models underlying the computer

codes can be thoroughly reviewed. It is noted that while this approach

would likely fulfill the NRC's statutory responsibility, it could also

7 .

3111/DJF/84/02/14/0- 13 -

cause a great many technical interchanges with DOE and introduce

unnecessary delays into the licensing process.

Alternatively, the NRC could develop extensive modeling capabilities

(i.e., different codes developed independently) and attempt to

independently reproduce the bulk of the DOE modeling results. Such code

development and analyses would require resources, including technology,

skilled staff and contractors, and review time on the order of the

resources used by DOE to produce the results in the first place. As

similar, but less extreme alternatives, the NRC staff and contractors

could attempt to demonstrate through benchmarking studies (conducted by

DOE, NRC or both) that the DOE codes could be relied on, and the DOE

codes could be used by the NRC staff for independent analyses, or simple,

conservative models could be used with conservative data for "bounding"

analyses. These four levels of detail for NRC staff reviews can be

summarized as follows:

1) Critically evaluate and comment in detail on DOE's work.

2) Use simple, conservative models with conservative data.

3111/DJF/84/02/14/0- 14 -

3) Review and qualify DOE (or third party) models and codes to the

extent practicable. Use these models and codes to verify some or

all of DOE's analyses.

4) Independently develop models and codes for use in independently

verifying some of all of DOE's analyses.

The NRC approach to staff use of models and codes recommended in this

document varies depending on the type of issue to be addressed and

generally consists of item 1) above combined with one or more of items 2,

3 or 4. The central issue in developing a modeling strategy is to

determine, for each finding to be reached, the appropriate level(s) of

detail for an NRC licensing review.

Figures 2 and 3 summarize briefly the responsibilities of DOE and NRC in

preparing and reviewing a license application.

3111/DJF/84/02/14/0- 15 -

O DOCUMENT FULL LICENSING/PERFORMANCE ASSESSMENTDEMONSTRATING COMPLIANCE WITH 10 CFR PART 60

O PRODUCE COMPLETE AND QUANTITATIVE IDENTIFICATION ANDCHARACTERIZATION OF UNCERTAINTIES:

- BASIC PHENOMENA AND PROCESSES

- CONSTITUTIVE RELATIONSHIPS AND SIMPLIFYING ASSUMPTIONS

- PARAMETERS AND VARIABLES-DATA GATHERING AND ANALYSES

- CALCULATIONAL UNCERTAINTIES

O DOCUMENT COMPLETE TECHNICAL DEFENSE OF INSIGNIFICANCE OFUNCERTAINTIES BASED ON:

- HARD DATA AND FACTS

- DETAILED CONSIDERATION OF ALTERNATIVE INTERPRETATIONS

O IMPLEMENT APPROPRIATE QUALITY ASSURANCE PROGRAM FOR SUPPORTINGFACTS AND DATA AND PROVIDE ADEQUATE DOCUMENTATION

Figure 2 DOE's Licensing Responsibilities

3111/DJF/84/02/14/0- 16 -

O PROPOSE FINDINGS TO ATOMIC SAFETY AND LICENSING BOARD

O BASE PROPOSED FINDINGS ON INDEPENDENT REVIEW

1. INDEPENDENT DATA REVIEW

- ESTABLISH RELIABILITY AND ACCURACY

2. REVIEW DOE PERFORMANCE ASSESSMENT -- COMPLETENESSAND ADEQUACY OF:

- MODELS, MODEL INPUTS

- UNCERTAINTY ASSESSMENTS

- ALTERNATIVE INTERPRETATIONS

3. INDEPENDENT PERFORMANCE ASSESSMENT IN SELECTED AREAS

O NRC STAFF CAN CARRY NONE OF DOE'S "WATER" IN PROVING ITS CASE--STAFF CANNOT MAKE UP FOR LICENSE APPLICATION DEFICIENCIES.

Figure 3. NRC's Licensing Responsibilities.

3111/DJF/84/02/14/0- 17 -

IV UNCERTAINTIES

Two broad categories of uncertainties exist which significantly affect

the development of a modeling strategy--programmatic and technical.

Programmatic uncertainties involve questions about how DOE will carry out

its repository development program, the development schedules DOE will

follow, and, to a lesser extent, how the NRC will conduct its review of a

DOE license application. Technical uncertainties include (1)

identification of the basic phenomena which might affect a repository

system, (2) the accuracy of the models used to describe and evaluate

these phenomena, (3) the accuracy of available data, and (4)

calculational uncertainties. Each of these is discussed below.

Programmatic Uncertainties

At the current time, DOE's repository design and site selection efforts

are still evolving, leading to uncertainties about the types and accuracy

of analyses which might be required to review a license application. For

example, it is not clear how much reliance DOE will place on individual

engineered barriers in order to achieve an acceptable level of overall

repository performance, nor is it clear what means (experimental,

3111/DJF/84/02/14/0- 18 -

analytical, or some combination of both) DOE will use to evaluate the

performance of each major barrier. Similarly, uncertainties exist

regarding the timing of DOE's development programs which directly affect

the time available for development of NRC review capabilities.

Information regarding DOE's program must be made available to the NRC as

DOE's program evolves, and a substantially complete synthesis of DOE's

program must be presented in DOE's Site Characterization Plans, if not

sooner. To the extent that information about DOE's plans is currently

lacking, the NRC will attempt to deal with the resulting programmatic

uncertainties by preparing to review all of the more likely (or the more

conservative) resolutions which DOE might pursue.

In developing this modeling strategy, the NRC staff has relied on current

knowledge and informed estimates of the approach(es) and schedules which

DOE is expected to follow in demonstrating acceptable repository

performance and, where appropriate, alternative approaches (and their

effects on a modeling strategy) are discussed. This information is

presented in a general way in the following section, "Key Assumptions,"

and more specific assumptions for individual analyses are presented under

the heading "Anticipated DOE Technical Analyses" in Section VI. This

modeling strategy is expected to evolve n step with future evolutions in

3111/DJF/84/02/14/0- 19g -

DOE's program, and the assumptions on which the document is based will be

revised as appropriate to conform to the DOE program.

Technical Uncertainties

DOE bears the responsibility for identification of sources of technical

uncertainty and for addressing uncertainties, either by reducing

uncertainties to a manageable level or by providing adequate compensation

for them. Technical uncertainties can be classified into the following

four categories.

1) Identification of basic phenomena. This type of uncertainty is, of

course, the most fundamental and has the greatest potential significance

for development of a modeling strategy. If previously unrecognized

phenomena should be discovered, major disruptions to the licensing

process could result. While the NRC staff considers the identification

of basic phenomena to be DOE's responsibility (as part of its burden of

proof to demonstrate repository safety), the staff is also working to

identify relevant phenomena by (a) closely following DOE's development

programs and (b) conducting an independent confirmatory research program

3111/DJF/84/02/14/0- 20 -

aimed at maintaining an awareness of the state-of-the-art in the relevant

technical disciplines.

2) Accuracy of models. Once basic phenomena have been identified which

might affect repository performance, it is necessary to model these

phenomena (either conceptually or mathematically) in order to evaluate

their significance. Uncertainties in the models result directly from

uncertainties in the understanding of the phenomena, and it may not be

possible to completely eliminate such uncertainties through validation

and calibration efforts. The NRC staff is addressing these uncertainties

as described above by closely following DOE's development programs and by

conducting an independent confirmatory research program, and also by

using "conservative" models where substantial uncertainties remain

unresolved.

3) Accuracy of data. Maintaining an awareness of data uncertainties is

important in developing a modeling strategy in order to avoid the GGO

(garbage in, gospel out) phenomenon whereby more credence might be

attributed to analytical results than the fundamental data will support.

In conducting licensing reviews, the NRC staff does not propose to use

computer codes which are more sophisticated than the available input data

they require, or more complex than needed to conservatively bound a

311 1/DJF/84/02/ 14/0- 21 -

phenomenon. The staff may, however, use more sophisticated codes than

will be used in licensing reviews prior to receipt of a license

application to evaluate the potential significance of missing data in

order to provide guidance to DOE for conducting its site characterization

program.

Sensitivity analyses may, of course, be used to evaluate the significance

of data uncertainties with respect to the overall uncertainty in

repository system performance.

4) Calculational uncertainties. The numerical accuracy of analytical

results can, in principle, be controlled in a straightforward manner

through a program of code verification, benchmarking and quality

assurance. The NRC staff is providing guidance to DOE in this area, and

is independently conducting a program to benchmark several codes likely

to be used for evaluating repository performance. The NRC staff does

not, however, intend its benchmarking effort to relieve DOE of its

responsibility to ensure adequate benchmarking of the codes used by DOE

to support a license application.

3111/DJF/84/02/14/0- 22 -

V KEY ASSUMPTIONS

To simplify the development of the modeling strategy, certain assumptions

need to be made regarding the content of a DOE license application, the

types of findings to be reached by the NRC, and the limitations on the

resources available to the NRC staff. The following are the key

assumptions on which this document is based:

-DOE will use codes to demonstrate that repository performance complies

with several of the performance criteria of 10 CFR 60 and 40 CFR 191.

Numerical analyses may be the primary (or sole) demonstration of

compliance for some criteria, and may be combined with other arguments

(e.g., empirical studies or expert judgment) for others. For example,

demonstration of compliance with the groundwater travel time criterion is

expected to rely heavily on the results of computerized analyses of

empirical data and determination of engineered barrier release rates may

involve a combination of empirical leach rate data and computerized

analyses of radionuclide transport through backfill materials.

-DOE will assert that these codes either (1) address all of the features

and/or processes which significantly affect repository compliance with a

3111/DJF/84/02/14/0- 23 -

particular criterion, or (2) bound features and/or processes not directly

addressed by the codes.

-The NRC staff will develop the capability to independently evaluate

compliance with each of the criteria of 10 CFR 60 and 40 CFR 191.

However, such evaluations will not necessarily involve the use of models

or codes.

-In those areas where the NRC staff does want to use models or codes for

an independent evaluation, Independent NRC development of such models or

codes is not necessarily required. It may be appropriate to use models

or codes developed by DOE or by a third party if the NRC staff's review

of the technical merits of the models or codes allows the staff to use

them Confidently.

-The NRC staff will not have sufficient resources available to it to

independently develop, operate and maintain a full suite of codes for all

facets of repository performance. Similarly, even if DOE or third party

codes are used in some areas, the NRC staff may not have sufficient

resources available to conduct detailed independent computer code

analyses n all areas without significantly affecting the timeliness of

its license application review.

3111/DJF/84/02/14/0- 24 -

-The greatest degree of controversy in a licensing review is likely to

involve the performance of those barriers for which projected performance

is least certain. The geologic setting and the engineered barriers both

provide significant contributions to the overall performance of the

repository system, but the uncertainties in projecting the performance of

the geologic setting may be more significant than the uncertainties in

engineered barrier performance, depending on what DOE attempts to claim

credit for.

-Analyses will need to consider uncertainties in one of two ways: (1)

quantification and propagation of uncertainty estimates using Monte Carlo

or similar techniques, or (2) use of "bounding" estimates and

conservative analyses. To the extent that DOE uses the "bounding"

approach, the NRC staff will need to have sufficient knowledge of the

appropriate technical disciplines to independently determine whether or

not DOE's proposal is, in fact, conservative. The NRC staff may test

certain DOE assumptions with simple calculations.

-The EPA HLW standard will require an analysis of repository performance

similar to a probabilistic risk analysis for a nuclear power plant. The

degree to which this analysis will serve as the basis for determining

repository acceptability remains to be worked out with EPA. Regardless

3111/DJF/84/02/14/0- 25 -

of the outcome, however, the requirement to perform such an analysis will

require the NRC staff to ndependently evaluate DOE's analyses of: (1)

identification of potentially disruptive events and processes, (2)

estimation of the probabilities of such events and processes, and (3) the

impacts of such events and processes on the amount of waste projected to

be released to the environment.

-Simple models and codes are preferred wherever their use can be

defended. In order to defend the use of simplified models and codes, it

will be necessary to have a good understanding of the state-of-the-art,

including familiarity with more complex models and codes, and a firm

grasp of the phenomena which underlie both simplified and more complex

models and codes. Lack of data for more complex models and codes is not,

by itself, a sufficient basis for simplification. It will be necessary

to show (for example by sensitivity analyses) that the missing data would

not lead to different conclusions even If available.

-NRC will have substantial advance notice of the codes that DOE will use

to demonstrate compliance, and these codes will reflect NRC guidance to

DOE as to how processes, parameters, and variables should be treated.

DOE codes will have been developed, documented, verified, benchmarked and

validated (to the extent practicable) in accordance with NRC guidance.

3111/DJF/84/02/14/0- 26 -

-The codes, data and results of analyses used by DOE to support the

application will be sufficiently well documented that the simulations

could be repeated independently by technically competent reviewers.

-The codes will be made available to the NRC sufficiently in advance of

the application for the staff (or contractors) to become competent in

exercising them, should they choose to do so.

-The NRC staff, through (1) access to DOE data, (2) interactions with DOE

investigators during site characterization, and (3) modeling assessments

and sensitivity studies during site characterization, will be very

familiar with sites and data at the time of license application.

3111/DJF/84/02/14/0- 27 -

VI DEVELOPMENT OF MODELING STRATEGY

The NRC staff modeling strategy is organized and presented in the

following manner:

o Identify the specific issue(s) being addressed (e.g., compliance with

the release limits of the EPA standard.)

o Describe the technical components of the issue(s), and the extent to

which DOE is expected to use numerical analyses to address the issue(s).

o Describe the specific actions which the NRC will take to arrive at a

finding which resolves the issue(s).

o Discuss the programmatic and technical uncertainties which might

affect the NRC's modeling strategy.

Much of this modeling strategy is based on key technical assumptions

regarding the performance of individual barriers, the events or processes

which might affect performance, and the means which are likely to be used

to evaluate performance. (An illustrative licensing assessment

3111/DJF/84/02/14/0- 28 -

methodology for a repository in a porous, saturated medium is illustrated

in Figure 4.) As DOE's development programs (and NRC's independent

technical programs) progress, new information may require revisions to

some of these assumptions and to the portions of the modeling strategy

developed from them.

The following sections list the major performance issues identified in

Appendix C of NUREG-0960 (Figure 1, page 6) and, for each issue,

describes the modeling strategy to be used for resolution of the issue.

P.1 HOW DO THE DESIGN CRITERIA AND CONCEPTUAL DESIGN ADDRESS RELEASES OF

RADIOACTIVE MATERIALS TO UNRESTRICTED AREAS WITHIN THE LIMITS SPECIFIED

IN 10 CFR 20?

1.1 Anticipated DOE Technical Analyses

DOE is expected to submit estimates of off-site radionuclide releases

which are well below the limits of 10 CFR 20 (and the more restrictive

limits of 40 CFR 191). Although 10 CFR 20 (and the portion of 40 CFR 191

which applies to pre-closure operations) restrict releases from normal

operations only, DOE is expected to also submit estimates of releases

Identificationof Scenarios(Event/Fault Tree)(Judgment)

L

Estimates ofScenario Probabilities(Judgment) Compliance

with EPAEffects of Scenarios Standard?on Repository(Judgment) I

Compilancewith Part 60?

Q ContainmentRelease RateGW Travel TimeRetrievability

_i -j

I

Radiation _ (lding IILevel s - IL I (ORIGEN) I

t IWaste IInventory I(ORIGEN)

- IL I'I

Radiolysis of Groundwater& Other Rad. Effects( __)

1mGehemistry

I

I

I

I

I

I

I

I

[-

Geochemistry( ___)

-

Waste PackagePerformance

oCorrosionoLeachingoStructural

(WAPPA)(Empirical)

e-."-ContaminantTransport(SWIFT)(NWFT/DVM)

ContaminantTransport(NWFT/DVM)

I

To DI

4To D

I

I

I

I

III

I.GroundwaterFlow(SWIFT)(Judgment)

I ~ T I

|Groundwater IFl ow (SWIFT) I

I 1 I

GroundwaterFlow

SWIFT)JudgmentJ

:

ThermalOutput(ORIGEN)

T IHeat

-> Transport -( j

II

Mechanical/IStructural

1'- I

ITo @

I

I

I

I

ITo @

I Underground Facility/Near FieldWaste Package Far Field Basin/Reaional

Figure 4, Illustrative Licensing Assessment Methodology

3111/DJF/84/02/14/0- 30 -

following accidents of various types, for the purpose of identifying

components, systems and structures which are "important to safety."

DOE's use of numerical analyses in addressing this issue is anticipated

to be the following:

-Source Term--DOE is expected to use a code such as ORIGEN to estimate

the inventory of radionuclides present in any waste available for

release.

-Criticality--Codes currently in use for nuclear fuel handling facilities

are expected to be used for evaluating potentially critical waste

configurations both prior to and following waste emplacement.

-Transport by Ventilation System--A code is expected to be used which can

evaluate such phenomena as dispersion, settling-out, and plate-out during

transport. No specific code has yet been identified.

-Movement Through the Environment--Codes similar to those currently used

to evaluate airborne releases from nuclear power plants and other nuclear

fuel cycle facilities are expected to be used to estimate concentrations

3111/DJF/84/02/14/0- 31 -

of radionuclides reledsed to unrestricted areas in order to demonstrate

compliance with 10 CFR 20.

-Doses to Individuals--40 CFR 191 requires DOE to estimate the maximum

doses to individual members of the public. Codes similar to those used

for analyses of nuclear power plants and other nuclear fuel cycle

facilities are expected to be applicable.

Releases in water (for example, discharges of process water used during

facility operations), if any, are expected to be evaluated by relatively

simple hand calculations or by using existing computer codes currently

used for evaluating releases from nuclear power plants and other nuclear

fuel cycle facilities.

1.2 NRC Review Actions

The codes described above are, with one exception, fairly standard codes

similar to ones which have been widely used for other applications. The

NRC staff's review will therefore be primarily limited to reviewing the

input data for the codes (especially the site-specific data) and

verifying that the conceptual models on which the codes are based are

valid for the specific DOE application. If projected releases are near

3111/DJF/84/02/14/0- 32 -

the limits of 10 CFR 20 or 40 CFR 191, the NRC staff may perform some

sensitivity analyses using DOE's codes or available third-party codes to

evaluate the significance of individual input data values.

The code used to estimate radionuclide transport in the repository

ventilation system is not known to be a standard type of code, and the

NRC staff will need to become familiar with the physical processes

occurring during such transport as well as with the input data needed for

such analyses. An independent NRC research effort is addressing this

question at the present time.

The NRC staff's review capabilities in the other areas related to this

issue are reasonably complete. Computer codes are readily available (see

ref. 2) for evaluating airborne radionuclide releases, and are

anticipated to be sufficiently accurate for the intended application. No

additional computer code development related to this issue is

anticipated, and acquisition of existing codes should not be pursued

until DOE's repository development program is more advanced.

31 11/DJF/84/02/ 14/0- 33 -

1.3 Uncertainties

While DOE's designs for repository facilities are still under

development, the means for controlling radionuclide releases (e.g., air

filtration systems) are expected to be the same as used in other nuclear

materials handling facilities. Thus, uncertainties regarding the

performance of these systems are not expected to be highly controversial.

To the extent that such systems can be shown to control accidental

releases (e.g., resulting from cask drops), uncertainties regarding

initiation and evaluation of accidents will also be of reduced

importance. In any case, identification and evaluation of potential

accidents has little effect on development of a modeling strategy since

such evaluations primarily involve development of data (e.g., hoist

system failure frequencies) and relatively simple models amenable to hand

calculations (see ref. 3) rather than use of the more detailed analyses

discussed in this document.

3111/DJF/84/02/14/0- 34 -

P.2 HOW DO THE DESIGN CRITERIA AND CONCEPTUAL DESIGN ACCOMMODATE THE

RETRIEVABILITY OPTION?


DOE is expected to show that waste can be retrieved without significant

occupational doses or environmental impacts, and on a schedule

approximately the same as the original waste emplacement schedule. This

demonstration is expected to be based on the following:

-Structural Stability of Openings--DOE is expected to show that the

underground facility will remain structurally stable for a sufficient

period of time to allow both waste emplacement and possible retrieval.

Numerical analyses of stability similar to those currently used in mining

engineering are anticipated, with appropriate modifications to account

for the longer time periods of interest and the effects of the heat

generated by the emplaced wastes. In salt, numerical analyses of creep

closure are expected.

-Non-degradation of Waste Packages--DOE is expected to argue, based in

part on the results of numerical analyses, that the emplaced waste

3111/DJF/84/02/14/0- 35 -

packages will not degrade significantly prior to retrieval. Numerical

analyses of the structural integrity of the waste package and of the

effects of corrosion on waste package materials are expected. However,

analyses of corrosion rates are expected to be primarily extrapolations

of empirical data rather than first principles" analyses.

-Availability of Technology--DOE is expected to show that the technology

for waste retrieval, including steps such as removing backfill materials,

is available. Numerical analyses are not anticipated.

The types of analyses to be submitted by DOE will depend, In part, on the

design of a specific repository and waste package.


The NRC staff review related to this issue is expected to concentrate on

the input data used by DOE and, particularly, on the theory embodied in

DOE's codes (especially the basis for extrapolation of empirical

corrosion data) . Current limitations in structural analysis

capabilities appear to be due to data deficiencies rather than

deficiencies in codes, and independent NRC code development is therefore

not anticipated. The NRC staff may perform sensitivity analyses using

3111/DJF/84/02/14/0- 36 -

DOE or third-party codes to evaluate the significance of input data

values.

2.3 Uncertainties

Substantial uncertainties exist regarding this performance issue. DOE

has not yet finalized its underground facility and waste package designs,

leading to programmatic uncertainties. Technical uncertainties involve

Identification of the basic processes involved (e.g., different corrosion

processes and various manufacturing flaws which could affect waste

package integrity), development of appropriate models of processes, and,

most importantly, development of data. As indicated above, the NRC staff

considers that data limitations represent the largest source of

uncertainty related to this performance issue (e.g., uncertainties

regarding the applicability of currently available structural analysis

codes result primarily from the data, including descriptions of failure

modes, incorporated in the codes). Independent NRC development of

computer codes is not planned unless future investigations identify a

need for more sophisticated analytical capabilities, and indicate that

appropriate data are available to support such analytical capabilities.

3111/DJF/84/02/14/0- 37 -

P.3 WHEN AND HOW DOES WATER CONTACT THE BACKFILL?

P.4 WHEN AND HOW DOES WATER CONTACT THE WASTE PACKAGE?

3.1, 4.1 Anticipated DOE Technical Analyses

DOE is expected to use one (or a combination) of three possible

approaches to estimate the time when water first contacts the backfill:

-Assume that the underground facility resaturates (or returns to its

original state of partial saturation) immediately after repository

closure.

-Use hydrologic or coupled thermal-hydrologic numerical analyses to

estimate the water infiltration rate after repository closure.

-Measure water infiltration rates into the underground facility during

operations and extrapolate, using "back-of-the-envelope" calculations, to

estimate infiltration rates after repository closure.

3111/DJF/84/02/14/0- 38 -

Similarly, DOE is expected to use one of two possible approaches in

determining when and how water contacts the waste package:

-Assume that water contacts the waste package immediately after

repository closure.

-Use numerical analyses to estimate the water flow conditions in the

backfill material.

DOE's estimate of how the contact between water and the waste package and

backfill occurs will depend strongly on the repository design, including

the nature of any backfill materials, and is expected to be based on

expert judgment, possibly supplemented by empirical results obtained from

a performance confirmation program conducted prior to repository closure.

The degree of sophistication of DOE's analyses in the near-field will

depend strongly on how much credit DOE places on engineered barriers and

on the extent to which uncertainties in near-field performance are

compensated for in the facility design (e.g., by reduced heat loading).

3111/DJF/84/02/14/0- 39 -

3.2, 4.2 NRC Review Actions

The NRC staff will review the data and theory used in any DOE numerical

analyses addressing these issues. Independent NRC staff analyses using

an unsaturated groundwater flow code (possibly coupled with a thermal

analysis capability) may be conducted. The TOUGH code may be appropriate

for analyses related to these issues. This code is available for NRC

staff use, and staff members are currently evaluating the capabilities of

this code.

3.3, 4.3 Uncertainties

Uncertainties exist regarding DOE's selection of repository media,

backfill materials and underground facility designs. More importantly,

substantial uncertainties currently exist regarding the fundamental

physical phenomena (especially under the influence of heat) which govern

groundwater flow in an unsaturated zone during resaturation and the data

necessary to evaluate unsaturated flow. (Resaturation of repositories in

salt or fractured media may involve additional complications.) The NRC

currently has independent research projects on-going to provide an

3111/DJF/84/02/14/0- 40 -

improved understanding of unsaturated flow. Additional development of

computer codes will be guided by the results of these research projects.

P.5 WHEN AND HOW DOES WATER CONTACT THE WASTE FORM?


This issue essentially involves an estimate of the waste package life.

DOE is expected to demonstrate an estimated life of at least 300 years

through a combination of. numerical analyses using WAPPA (or a similar

code) and extrapolations of empirical data. Accelerated testing results

may be used as the basis for extrapolations. Neither waste package nor

underground facility designs are sufficiently complete to identify the

dominant failure mechanisms at the current time, although corrosion is

expected to be important in all media, and structural deformation of

waste packages may be important in salt. Rapid or discrete failures

(e.g., cracking of canisters due to manufacturing defects), as contrasted

with the more continuous processes of corrosion and structural

deformation, may also be important. Geochemical analyses of the

groundwater constituents under the ambient effects of heat and radiation

will be important in evaluating corrosion rates. (To the extent that the

waste package life s dependent on the quantity of water reaching the

- -

3111/DJF/84/02/14/0- 41 -

waste package, the discussion under issues P.3, P.4 and P.8 will be

relevant.)


The NRC staff will review the data cited by DOE in its estimate of waste

package life (including information regarding the geochemical environment

of the waste packages), and will review the theory underlying any

numerical analyses. Any independent NRC staff analyses are expected to

rely primarily on simple analytical models and empirical relationships,

and any more detailed analyses will use DOE or third-party codes. The

NRC staff and contractors are currently reviewing potential DOE designs

and are studying possible failure mechanisms in preparation for reviewing

a license application. The NRC staff has concluded that the code WAPPA

will require additional work by DOE (to incorporate more realistic models

of the processes treated by the code) before it will be suitable for use

in licensing. Development of waste package codes by the NRC is not

anticipated.

The NRC staff is also reviewing available information and codes which

might be useful for estimating the geochemical environment of the waste

packages. Available geochemical codes may provide some guidance in

3111/DJF/84/02/14/0 - 42 -

predicting geochemical conditions, but such evaluations are expected to

be based primarily on expert judgment derived from empirical information.

5.3 Uncertainties

DOE has not yet selected a waste package design, nor has DOE determined

how much reliance will be placed on waste package containment in

achieving acceptable levels of overall repository performance.

Substantial uncertainties result related to (1) the types of materials to

be used (and, therefore, the physical phenomena to be evaluated), (2) the

types and precision of data needed, and (3) the time periods over which

empirical results are to be extrapolated and the means (e.g., accelerated

testing) by which such extrapolations can be made with confidence.

However, as mentioned above, the NRC staff anticipates that DOE's

analyses will consist principally of extrapolations of empirical data

(possibly using codes such as WAPPA), and there should be no significant

impacts on the development of a modeling strategy.

3111/DJF/84/02/14/0- 43 -

P.6 WHEN, HOW, AND AT WHAT RATE ARE RADIONUCLIDES RELEASED FROM THE

WASTE FORM?


DOE is expected to demonstrate an estimated release rate from the

engineered barrier system of less than 10 5/year. It is assumed that DOE

will rely strongly on the waste form to achieve this release rate, with

possible assistance from backfill materials (see issue P.8). DOE's

evaluations of waste form release rate are expected to be based on a

combination of numerical analyses using WAPPA (or a similar code) and

extrapolations of empirical data. Accelerated testing results may be

used as the basis for extrapolations. While waste package and

underground facility designs are not yet complete, both the leach

resistance of the waste form and the solubility limits of individual

radionuclides in the geochemical environment of the underground facility

are expected to be important factors affecting the release rate. (To the

extent that the release of radionuclides is solubility-limited, the

quantity of water available for radionuclide release will be important as

discussed under issue P.8.) Geochemical analyses of the groundwater

311 1/DJF/84/02/14/0- 44 -

constituents under the ambient effects of heat and radiation will be

important in evaluating release rates.


The NRC staff will review the data cited by DOE in its estimate of

release rates, and will review the theory underlying any numerical

analyses. Any independent NRC staff analyses are expected to rely

primarily on simple analytical models and empirical relationships, and

any more detailed analyses will use DOE or third-party codes. The NRC

staff and contractors are currently reviewing potential DOE designs and

are studying possible release mechanisms in preparation for reviewing a

license application. The NRC staff has concluded that the code WAPPA

will require additional work by DOE (to incorporate more realistic models

of the processes treated by the code) before it will be suitable for use

in licensing. Development of codes by the NRC is not anticipated.

6.3 Uncertainties

Programmatic uncertainties result from the current lack of information

regarding DOE's selection of repository media and waste forms, and from

3111/ DJF/84/02/14/0- 45 -

uncertainty regarding the degree of reliance which DOE will place on the

waste form in achieving an acceptable level of overall repository

performance. Technical uncertainties involve Identification of the

fundamental phenomena of importance, appropriate mathematical

descriptions of these phenomena, and measurement of supporting data. The

NRC staff anticipates that analyses of waste form release rates will rely

primarily on extrapolations of empirical data (possibly using codes such

as WAPPA), and that there will be no need for the NRC to independently

develop codes related to this issue. The NRC staff does anticipate

closely monitoring DOE's work with WAPPA or similar codes as these codes

are adapted for use at specific sites.


WASTE PACKAGE?


Consistent with paragraph 6.1 above, DOE is expected to argue that the

release rate performance objective of 10 CFR 60 can be achieved by the

waste form and backfill (or possibly by the waste form alone) and that

other components of the waste package need not be relied on to meet this

criterion. Therefore, in the absence of information to the contrary from

- -

3111/DJF/84/02/14/0- 46 -

DOE, issue P.7 is considered to be adequately addressed by issues P.6 and

P.8.


BACKFILL?


Backfill materials may serve two general functions:

-Prolong waste package life by delaying the time of contact between water

dnd waste package, reducing the amount of water reaching the waste

packages, or altering the groundwater chemistry.

-Reduce the rate of radionuclide release by reducing the quantity of

water reaching the waste, altering the groundwater chemistry, or sorbing

radionuclides which have been released from the waste form.

3111/DJF/84/02/14/O- 47 -

In the absence of specific repository designs, it is not clear to what

extent DOE may rely on the backfill functions described above. It is

possible, however, to identify the types of numerical analyses which

might be used to evaluate backfill performance. These analyses are:

-Unsaturated groundwater flow (possibly coupled with a thermal analysis

capability) to estimate the time when water contacts the waste packages

and the quantities of water available.

-Coupled thermal and saturated groundwater flow to estimate the

quantities of groundwater reaching the waste packages after the

underground facility has been resaturated with water (for repositories in

saturated media only).

-Geochemistry to evaluate the chemistry of the groundwater reaching the

waste packages and the sorbing capabilities of the backfill (under the

influence of heat and radiation).

-Contaminant transport to predict migration of radionuclides through the

backfill.

3111/DJF/84/02/14/0_ 48 -

Inclusion of the results of such analyses in DOE's license application

will depend, in part, on the extent to which DOE can demonstrate

compliance with the release rate criterion of 10 CFR 60 using simpler

arguments (e.g., leaching or solubility data).


The NRC staff's review will concentrate primarily on the data used by DOE

and on the theory underlying DOE's numerical analyses. Independent NRC

staff analyses of groundwater flow and contaminant transport may be

conducted on a limited basis, as a check on any DOE analytical results

submitted. The codes TOUGH and SWIFT may be appropriate for analyses of

groundwater flow in unsaturated and saturated media, respectively. SWIFT

may also be appropriate for contaminant transport analyses in saturated

media, and a DOE code, TRACER30, may be useful for contaminant transport

calculations in unsaturated media.

8.3 Uncertainties

The major uncertainties related to this performance issue have been

alluded to above, i.e., programmatic uncertainties regarding the degree

of reliance which DOE will place on backfill materials, and technical

3111/DJF/84/02/ 14/0_ 49 -

uncertainties regarding the fundamental physical phenomena, descriptions

of these phenomena and availability of required data. The physical

phenomena occurring in this "very near field" region will be quite

complex due to the interacting effects of heat, radiation and "foreign

materials" (.e., waste form, canister, and backfill materials) on

groundwater flow and geochemistry. Furthermore, the fundamental physical

phenomena which govern unsaturated flow and transport are not well

understood. Because of this complexity, the NRC staff assumes that DOE

may not attempt to demonstrate any significant contribution to overall

repository performance by backfill materials, except possibly to limit

the quantities of groundwater available to the waste packages.

Therefore, only coupled thermal-hydrologic analyses (e.g., using TOUGH or

SWIFT) are currently planned by the NRC staff. Should DOE decide to try

to demonstrate a more substantial contribution by the backfill (e.g.,

sorption of radionuclides), significant changes to this modeling strategy

would be necessary. The NRC staff is closely nteracting with DOE and is

conducting ndependent research studies to dentify any possible changes

as early as possible.

3111/DJF/84/02/14/0- 50 -


DISTURBED ZONE?


DOE is expected to address this ssue in two separate steps:

-Determination of Disturbed Zone--DOE is expected to submit the results

of coupled thermal-hydrologic analyses dentifying the physical

boundaries of the disturbed zone as defined in 10 CFR 60. These analyses

will estimate the extent to which the properties of the emplacement

medium In the vicinity of the repository will be altered by the presence

of the repository, and the effects of these alterations on the

performance of the repository. Limited geochemical analyses may be

conducted by DOE, but DOE's estimates of repository effects on the

chemical properties of the emplacement medium are expected to be based

primarily on expert judgment supported (or confirmed) by empirical

measurements.

3111/DJF/84/02/14/0- 51 -

-Waste Isolation Capability--The results of groundwater flow and

contaminant transport analyses to evaluate the transport of radionuclides

through the disturbed zone are expected to be submitted by DOE.


The NRC staff's review will concentrate primarily on the data used by DOE

and on the theory underlying DOE's numerical analyses. Independent NRC

staff analyses of groundwater flow and contaminant transport may be

conducted on a limited basis, particularly to determine the extent to

which repository performance is affected by phenomena such as thermal

buoyancy effects in groundwater. The codes TOUGH and SWIFT are available

to the NRC staff for analyses of groundwater flow in unsaturated and

saturated media, respectively.

9.3 Uncertainties

Technical uncertainties related to this performance issue include: (1)

identification of how physical phenomena (e.g., groundwater flow,

geochemistry) might be affected by the presence of the repository, (2)

development of realistic models of the effects of changes in physical

properties on repository performance, and (3) collection of data

3111/DJF/84/02/14/0- 52 -

necessary to evaluate the significance of changes in repository

performance. The NRC staff considers this performance issue to be a

major factor In developing a modeling strategy. While the NRC staff does

not expect DOE to claim substantial credit for Isolation of wastes within

the disturbed zone, determination of the physical extent of this zone is

expected to be difficult due to the complex nature of the physical

processes of interest. The NRC staff is working on a technical position

which will provide guidance to DOE on the disturbed zone, and is

conducting independent research efforts, the results of which will guide

future computer code development and application.


FAR-FIELD TO THE ACCESSIBLE ENVIRONMENT?


Because of the nature of the EPA high-level waste standard (described in

section II), DOE is expected to submit estimates of the cumulative

amounts of waste released from the far-field (region outside the

disturbed zone) to the accessible environment for 10,000 years after

repository closure. The estimates will address both "normal" conditions

and unlikely events and processes, and will reflect all reasonably

3111/DJF/84/02/14/0 -

quantifiable uncertainties in the performance of the overall repository

system. The release estimates are expected to be displayed in the form

of a "complementary cumulative distribution function" to facilitate

comparison with the release limits of the EPA standard. The principal

numerical analyses expected to be used by DOE for the release estimates

include:

-The analyses of engineered barrier and disturbed zone performance

discussed above.

-Groundwater flow, contaminant transport and geochemical analyses for the

geologic media between the disturbed zone and the accessible environment.

-Estimates of the likelihood that potentially disruptive events and

processes will occur, nd evaluations of the effects of such events and

processes on repository performance.


The NRC staff will review the data and theory used by DOE for its

analyses of far-field performance. The NRC staff will independently

perform numerical analyses of groundwater flow and contaminant transport

3111/DJF/84/02/14/0- 54 -

to confirm the results of DOE's analyses, but will perform Independent

geochemical analyses only on a limited basis. The NRC staff's review of

the likelihood of potentially disruptive events and processes will

primarily be limited to reviews of data and theory, and evaluations and

interpretations of DOE's data and analyses.

The NRC staff (through contractors) has ndependently developed

groundwater flow and contaminant transport codes, including SWIFT and

NWFT/DVM, which are expected to be applicable for far-field analyses in

porous, saturated media (including the far-field, saturated transport

analyses at unsaturated sites). Modifications to SWIFT and NWFT/DVM to

adapt these codes for analyses of fractured, saturated media are

underway. The physical phenomena involved n unsaturated flow and

transport are less well understood, and code development is consequently

less advanced. TOUGH may be applicable for unsaturated flow analyses,

and FEMWASTE (or similar codes) may, if used with conservative data,

provide bounding estimates of radionuclide transport in unsaturated

media. The staff is continuing to study the physical and chemical

phenomena likely to be present at repository sites, and will either

modify existing codes or develop new codes as necessary in order to

maintain an independent capability for far-field analyses.

3111/DJF/84/02/14/0- 55 -

10.3 Uncertainties

While significant uncertainties remain, the NRC staff considers that the

current level of understanding of relevant phenomena (e.g., groundwater

flow and contaminant transport) in saturated regions of the far feld

(including the saturated zone below an unsaturated site) is substantially

better than in the regions affected by the presence of a repository. In

particular, the nature of saturated groundwater flow in porous media is

quite well understood, and current research promises to substantially

advance our understanding of flow in fractured media. The data required

to confidently evaluate saturated groundwater flow on a regional scale

may require a significant number of data measurements, but are considered

to be more readily measurable (and with better accuracy) than data for

dreas affected by the presence of a repository. (Appropriate

measurements may also provide data which will assist in code validation.)

Substantial uncertainties remain regarding groundwater flow and

contaminant transport in the unsaturated zone. However, analogous to the

saturated zone, the nature of the fundamental phenomena of importance are

likely to be more easily and thoroughly understood in the far-field

regions which are unaffected by the presence of a repository.

3111/DJF/84/02/14/0- 56 -

For these reasons, the NRC staff has concentrated Its modeling efforts on

far-field groundwater flow and contaminant transport, and the staff

continues to believe that these areas are the most suitable for

analytical (as opposed to empirical) treatment in a licensing review.

P.11 WHAT IS THE PRE-WASTE EMPLACEMENT GROUNDWATER TRAVEL TIME ALONG THE

FASTEST PATH OF RADIONUCLIDE TRAVEL FROM THE DISTURBED ZONE TO THE

ACCESSIBLE ENVIRONMENT?


DOE is expected to submit the results of numerical analyses of far-field

groundwater flow conditions which estimate the groundwater travel time

between the disturbed zone and the accessible environment under pre-waste

emplacement conditions. These analyses are essentially a subset of those

discussed under issue P.10 above, but without considering the

perturbations caused by the presence of the repository.

DOE may also submit empirical data (such as age-dating information for

groundwaters) to support analyses of compliance with the travel time

performance objective of 10 CFR 60.

3111/DJF/84/02/14/0- 57 -


The NRC staff will independently perform numerical analyses of

groundwater travel times in saturated media between the disturbed zone

and the accessible environment under pre-waste emplacement conditions.

The groundwater flow code SWIFT is expected to be appropriate for the NRC

staff's independent analyses. Modifications to the code for particular

media may be necessary.

The NRC has proposed an amendment

groundwater travel time criterion

Should this amendment be adopted,

evaluate groundwater travel times

disturbed zone and the accessible

appropriate for such analyses.

to Part 60 which would apply the

to repositories in unsaturated media.

the NRC staff will independently

In unsaturated regions between the

environment. The code TOUGH may be

11.3 Uncertainties

As discussed in Section 10.3, the NRC staff considers that the far-field

geologic barrier (including groundwater flow conditions) is the component

of the repository system most amenable to analytical treatment, and the

3111/DJF/84/02/ 14/0- 58 -

staff anticipates that the bulk of its computer code development and

application efforts will continue to concentrate on this barrier.

P.12 HAVE THE NEPA ENVIRONMENTAL/INSTITUTIONAL/SITING REQUIREMENTS FOR

NUCLEAR FACILITIES BEEN MET?


Results of numerical analyses are not anticipated for this issue.

VII SUMMARY

Figure 5 lists four ncreasingly detailed levels of review ranging from 1

(least detailed) to 4 (most detailed). Figure 6 displays an illustrative

licensing assessment methodology and, for each technical discipline,

indicates the level(s) of detail which the NRC staff currently

anticipates will be appropriate for an NRC licensing review.

3111/DJF/84/02/14/0_ 59 _

1. CRITICALLY EVALUATE AND COMMENT IN DETAIL ON DOE WORK

2. USE SIMPLE, CONSERVATIVE (BOUNDING) MODELS WITHCONSERVATIVE DATA

3. REVIEW AND QUALIFY DOE (OR THIRD PARTY) MODELS AND CODES TOTHE EXTENT PRACTICABLE. USE DOE (OR THIRD PARTY) MODELS ANDCODES TO VERIFY SOME OR ALL OF DOE'S ANALYSES

4. INDEPENDENTLY DEVELOP MODELS AND CODES FOR USE IN INDEPENDENTLYVERIFYING SOME OR ALL OF DOE'S ANALYSES

Figure 5. Levels of Detail for Licensing Reviews

Estimates ofScenario Probabilities

J _

Identificationof Scenarios(Event/Fault Tree)Level 1 I ILevel 1

Effects of Scion RepositoryLevel 1

Compliance .with Part 60?

0 ContainmentRelease RateGW Travel TimeRetrievability

Radiolysis of Groundwater& Other Rad. EffectsLevels 1,2

I

I

II

IWaste PackagePerformance

oCorrosionoLeachingoStructural

Levels 1,2,3

To D

To W

I

IIIII

I

1 IHeatITransport -Levels1,2,3 -

III

Mechanical/StructuralLevels

-1,2,3

To @

III

II

To D

II

Waste Package I Underground Facility/Near Field I Far Field I Basin/Regional

Figure 6. Illustrative Licensing Assessment Methodology

and Levels of Review for odel/Code Application

3111/DJF/84/02/14/0- 61 -

REFERENCES

1) Slling, S. A., "Final Technical Position on Documentation of

Computer Codes for High-Level Waste Management," NUREG-0856, U. S.

Nuclear Regulatory Commission, 1983.

2) Hoffman, F. 0., et al., "Computer Codes for the Assessment of

Radionuclides released to the Environment," Nuclear Safety, Vol. 18, pp.

343-354, 1977.

3) Heckman, R. A., and T. Holdsworth, "A Probabilistic Safety Analysis

for Solidified High-Level Nuclear Waste Management Systems: A Status

Report," NUREG/CR-0577, U. S. Nuclear Regulatory Commission, 1979.

3111/DJF/84/02/ 14/0- 62 -

APPENDIX--Definitions

Accessible environment. (1) the atmosphere, (2) land surfaces, (3)

surface water, (4) oceans, and (5) the portion of the lithosphere that is

outside the controlled area. The overall system performance for the

geologic repository s calculated at this boundary (§60.2).

Computer code. A set of computer instructions for performing the

operations specified in a numerical model. Syn: Computer Program.

Conceptual Model. A pictorial and/or narrative description of a

repository system or subsystem which represents all relevant components

and structures contained within the system or subsystem, the interactions

between the components and structures, and any internal or external

processes which affect the overall performance of the system or

subsystem.

Consequence analysis. A method by which the consequences of an event are

calculated and expressed in some quantitative way, e.g., money loss,

deaths, or quantities of radionuclides released to the accessible

environment.

3111/DJF/84/02/14/0- 63 -

Controlled area. A surface location, to be marked by suitable monuments

extending horizontally no more than 10 km in any direction from the

underground facility, and the underlying subsurface, which area has been

committed to use as a geologic repository and from which incompatible

activities would be restricted following permanent closure (§60.2).

Deterministic code. A code that s based solely on physical

relationships and that does not consider ranges and distributions of

input parameters. For a given set of input parameters, the code always

produces the same result.

Disturbed zone. That portion of the controlled area whose physical or

chemical properties have changed as a result of underground facility

construction or from heat generated by the mplaced radioactive wastes

such that the resultant change of properties may have a significant

effect on the performance of the geologic repository. The minimum

groundwater travel time is calculated between this boundary and the

accessible environment (§60.133(a)(2)).

Engineered barrier system. The waste packages and the underground

facility. The maximum radionuclide release rate is measured at this

boundary (§60.113(a)(1)(ii)(B)).

3111/DJF/84/02/ 14/0- 64 -

Finding. A determination of compliance or non-compliance 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 on which it is based, expert

opinion, and empirical studies.

Flow Path. The model trajectory of an hypothetical groundwater particle

from a release point at the underground facility to the boundary of the

modeled system. This general term can be applied to laminar or

turbulent, steady-state or transient groundwater flow.

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 60.

Mathematical model. A mathematical representation of a process,

component, or system.

Model. A representation of a process, component, or system.

3111/DJF/84/02/14/0- 65 -

Numerical method. A procedure for solving a problem primarily by a

sequence of arithmetic operations.

Numerical model. A representation of a process, component, or system

using numerical methods.

Performance assessment. The process of quantitatively evaluating

component and system behavior, relative to containment and isolation of

radioactive wastes, to support development of a high-level waste

repository and to determine compliance with the numerical criteria

associated with the regulation (10 CFR 60).

Performance confirmation. The program of tests, experiments, and

analyses that is conducted to evaluate the accuracy and adequacy of the

information used to determine reasonable assurance that the performance

objectives for the period after permanent closure can be met.

Quality assurance. Those planned and systematic actions necessary to

provide adequate confidence that a structure, system, or component will

perform satisfactorily in service, or that a product such as a

3111/DJF/84/02/14/0- 66 -

mathematical analysis or a data measurement will be sufficiently free

from error to serve its intended purpose.

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.

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 etriment per unit time to a person or a

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

events.

3111/DJF/84/02/14/0- 67 -

Scenario analysis. The process of dentifying scenarios and estimating

the probability of their occurrence.

Sensitivity analysis. An analysis in which one or more parameters are

varied to observe their effects on the performance of a system or some

part of it. Such an analysis requires definition of a system, the ranges

of parameters over which the system is to be investigated, and the

characteristics of the system which is to be observed.

Simulation. The application of an operating computer code.

Streamline. A groundwater flow path for which each particle passing

through a given point follows the same path as the preceeding particle.

Streamline flow, strictly speaking, applies only to laminar, steady-state

flow regimes.

Uncertainty analysis. An analysis that estimates the uncertainty in a

system's performance resulting from the uncertainty of one or more

factors associated with the system. Such an analysis requires definition

of a system, description of the uncertainties in the factors that are to

be investigated, and the characteristics of the system that is to be

observed.

-

3111/DJF/84/02/14/0- 68 -

Underground facility. The underground structure, ncluding openings and

backfill materials, but excluding shafts, boreholes, and their seals.

Validation. Assurance that a model as embodied in a computer code is a

correct representation of the process or system for which it is intended.

Verification. Assurance that a computer code correctly performs the

operations specified in a numerical model.

Waste form. The radioactive waste materials and any encapsulating or

stabilizing matrix.

Waste package. The waste form and any containers, shielding, packing and

other components surrounding the waste form. The minimum waste package

containment time is calculated at this boundary (60.113(a)(1)(li)(A)).

memorandum for: from: rkrver-changes rather than simple dramatic differences, so that there is...

Documents