UNITED STATES

NUCLEAR WASTE TECHNICAL REVIEW BOARD

WINTER BOARD MEETING

February 1, 2006

Desert Research Institute

755 East Flamingo Road

Las Vegas, Nevada 89119

[NWTRB BOARD MEMBERS PRESENT]{.underline}

Dr. Mark Abkowitz

Dr. William Howard Arnold

Dr. Thure Cerling

Dr. David Duquette

Dr. B. John Garrick, Chair, NWTRB

Dr. George M. Hornberger

Dr. Andrew Kadak

Dr. Ronald Latanision

Dr. Ali Mosleh

Dr. Henry Petroski

[SENIOR PROFESSIONAL STAFF]{.underline}

Dr. Carlos A.W. Di Bella

Dr. Daniel Fehringer

Dr. Daniel Metlay

Dr. David Diodato

Dr. John Pye

[CONSULTANTS]{.underline}

Dr. William Murphy

Dr. Leon Reiter

[NWTRB STAFF]{.underline}

Dr. William D. Barnard, Executive Director

Joyce Dory, Director of Administration

Karyn Severson, Director, External Affairs

Linda Coultry, Program Support Specialist

Davonya Barnes, Office Assistant

[I N D E X]{.underline}

[PAGE NO]{.underline}.

**Opening Remarks**

B. John Garrick, Chairman,

U.S. Nuclear Waste Technical Review Board. . . . . . 6

**OCRWM Program and Project Overviews**

**Yucca Mountain**

J.R. (Russ) Dyer, U.S. Department of Energy,

OCRWM Chief Scientist . . . . . . . . . . . . . . . 18

**Conservatism, Non-conservatism, and Uncertainty in**

**Dose Calculations - Risk Informed Dose Calculations**

Michael Ryan, Editor-in-Chief,

Health Physics Journal

Adjunct Professor,

Medical University of South Carolina . . . . . . . . 42

**Implementation of a Dose Standard after 10,000 years**

Tim McCartin, U.S. Nuclear Regulatory Commission . . 63

**Mass of Water Seeping into and out of**

**Drifts over Time**

Jens Birkholzer, Lawrence Berkeley

National Laboratory . . . . . . . . . . . . . . . . 97

Ernest L. (Ernie) Hardin

Bechtel SAIC Corporation (BSC) . . . . . . . . . . . 127

What is the mass flux of water seeping into and out

of the drifts over time? What key processes affect

these estimates? What are the assumptions and

uncertainties associated with these estimates? What

is the technical basis for the models used to

estimate water mass flux into and out of the drifts

over time? How does the mass of water seeping into

and out of the drifts vary temporally and spatially?

**Lunch** . . . . . . . . . . . . . . . . . . . . . . . 152

**Mass and Activity of Key Radionuclides Potentially**

**Released from Waste Forms, Waste Packages,**

**and Drifts over Time**

David C. (Dave) Sassani,

Management and Technical Support (MTS) . . . . . . . 153

Robert L. (Rob) Howard, BSC . . . . . . . . . . . . 189

[I N D E X]{.underline}

(Continued)

[PAGE NO.]{.underline}

What are the most significant physical and

chemical controls on release of radionuclides

from the waste package? What is the qualitative

and quantitative level of understanding of those

controls? What is being done at present to

develop a better understanding of the most

significant controls on the release of radio-

nuclides from the waste package? What key

processes and events affect the estimate of

mass/activity flux from the waste package and

engineered barrier system? What is the technical

basis for the models used to evaluate those key

processes and events? What are the key assumptions

and uncertainties associated with these models?

**Mass and Activity of Key Radionuclides Potentially**

**Released from the Unsaturated and Saturated**

**Zones over Time**

Bruce Robinson, Los Alamos National Laboratory . . . 219

Bill W. Arnold, Sandia National Laboratory . . . . . 255

For a potential unit release of radionuclide mass

or activity from the drifts, what is the rate of

release from the unsaturated and saturated zones,

respectively? What are the key processes that

affect these estimates? What is the technical

basis for models used to estimate the mass or

activity flux out of the unsaturated and saturated

zones? What are the key assumptions and uncertain-

ties associated with this estimate? Given a unit

release from the drift tunnels or unsaturated zone,

respectively, how do releases from the unsaturated

and saturated zone vary temporally and spatially?

**Public Comments** . . . . . . . . . . . . . . . . . . 281

**Management and Technical Support Peak Dose**

**Sensitivity Analysis**

Mark Nutt, MTS . . . . . . . . . . . . . . . . . . . 288

What key assumptions were used in the peak dose

sensitivity analysis? What are the most

[I N D E X]{.underline}

(Continued)

[PAGE NO.]{.underline}

significant processes in the peak dose

sensitivity analysis? Which processes have the

greatest impact on the magnitude of the peak

dose? Which processes have the greatest impact

on the timing of the peak dose?

**Adjourn Public Meeting** . . . . . . . . . . . . . . . 319

[P R O C E E D I N G S]{.underline}

8:07 a.m.

GARRICK: Good morning and welcome.

My name is John Garrick. I\'m Chairman of the Nuclear Waste Technical
Review Board. And, on behalf of the Board, I\'d like to say we\'re very
pleased to be back in Las Vegas, a city of frequent visits by the Board,
and the primary location of most of our public meetings.

As you know, we, at each of our meetings, introduce the Board members,
and I want to do that first. And, in my case, I am in the role of a
consultant these days, primarily on the application of the risk sciences
to a variety of industries, space, chemical, nuclear, transportation.
And, my background and areas of interest are risk assessment and nuclear
science and engineering.

Now, as I introduce the Board members, I want them to raise their hands
when their name is called. And, the first one I want to introduce is
Mark Abkowitz. Mark is Professor of Civil Engineering and Management
Technology at Vanderbilt University, and Director of the Vanderbilt
Center for Environmental Management Services. He chairs the Board\'s
Panel on the Waste Management System.

Howard Arnold. Howard is a consultant to the nuclear industry, having
previously served in a number of senior management positions, including
vice-president of the Westinghouse Hanford Company, and president of
Louisiana Energy Services.

Thure Cerling. Thure is a Distinguished Professor of Geology and
Geophysics and a Distinguished Professor of Biology at the University of
Utah. He is a geochemist, with particular expertise in applying
geochemistry to a wide range of geological, climatological and
anthropological studies.

David Duquette. David is Department Head and Professor of Materials
Engineering at Rensselaer Polytechnic Institute in Troy, New York. His
areas of expertise include physical, chemical, and mechanical properties
of metals and alloys, with special emphasis on environmental
interactions. His current research interests include studies of cyclic
deformation behavior as affected by environment and temperatures, basic
corrosion studies, and stress-corrosion cracking.

George Hornberger. George is the Ernest H. Ern Professor of
Environmental Sciences and Associate Dean for Sciences at the University
of Virginia. His research interests include catchment hydrology,
hydrochemistry, and transportation of colloids in geological media. He
chairs the Board\'s Panel on the Natural System.

Andrew Kadak. Andy is Professor of the Practice in the Nuclear
Engineering Department of the Massachusetts Institute of Technology. His
research interests include the development of advanced reactors, space
nuclear power systems, improved technology-neutral licensing standards
for advanced reactors, and operations and management issues associated
with existing nuclear power plants.

Ron Latanision. Ron is an Emeritus Professor at MIT and a principal and
Director of Mechanics and Materials with the engineering and scientific
consulting firm, Exponent. His areas of expertise include materials
processing and corrosion of metals and other materials in different
aqueous environments. He chairs the Board\'s Panel on the Engineered
System.

Ali Mosleh. Ali is the Nicole J. Kim Professor of Engineering and
Director of the Center for Risk and Reliability at the University of
Maryland. He has performed risk and safety assessments, reliability
analyses, and decision analyses for the nuclear, chemical and aerospace
industries. He chairs the Board\'s Panel on Repository System
Performance and Integration.

Henry Petroski. Henry is the Aleksandar S. Vesic Professor of Civil
Engineering and Professor of History at Duke University. His current
research interests are in the areas of failure analysis and design
theory. Ongoing projects include the use of case histories to understand
the role of human error and failure in engineering design, as well as
models for invention and evolution in the design process.

We also have two consultants with us today. Dr. William Murphy. Bill is
an Associate Professor in the Department of Geological and Environmental
Sciences at California State University-Chico. His areas of expertise
are geology, hydrogeology, and geochemistry. Dr. Murphy has worked with
the Board many times in the past, and we are pleased to have him with us
today.

And, recently retired from more than 31 years of federal government
service, and about half of that on the Board staff, we are pleased to
have with us today Dr. Leon Reiter. Leon\'s areas of expertise include
probabilistic seismic hazard assessment, probabilistic volcanic hazard
assessment, and total system performance assessment. And, on behalf of
the Board, Leon, I want to thank you for a career of distinguished
service, and we look forward to having access to your expertise.

There is one other group I would like to introduce as a group that\'s
with us today, distinguished group. The Board, some three years ago,
started participating, at least having representatives participate with
other similar boards around the globe that advise their governments on
matters pertaining to nuclear waste, and in some cases other nuclear
matters. And, we have now had three meetings. We had a meeting initially
in Paris, and our second meeting was in Berlin, and our third meeting,
which I have had the privilege of chairing the last two days here in Las
Vegas, for the purpose of seeing how we, and each of our advisory
boards, can maximize our effectiveness in advising our respective
governments on matters, in this case, pertaining to nuclear waste.

And, we have representatives from Germany, from France, from Japan, from
Sweden, from the Nuclear Energy Agency, and, of course, from the U.S.
And, I\'d like to ask these distinguished members of the International
Advisory Board Group to stand at this time, and I encourage those of you
to interact with them and question them and help us continue this
process of an effective exchange with the international community, such
that we can all adequately address this international issue.

Thank you very much.

Now, at the beginning of each meeting, there\'s a few routine things
that we do. One is we read the following statement for the record, so
that everybody is clear about our behavior as Board members.

Board meetings are spontaneous by design, even though it\'s kind of a
daunting, stuffy environment we have with reporters and what-have-you,
we nevertheless express ourselves pretty freely, and we want to be able
to continue that. But, when the Board members speak extemporaneously, it
is important to realize that we are speaking on our own behalf, not on
behalf of the Board. When a Board position is articulated, we will do
our best to make that known to you.

In December last year, the Board sent a letter report to Congress, last
year, that was just a couple months ago, we sent a letter report to
Congress and Secretary of Energy highlighting some Board findings from
the calendar year 2005. The findings contained in that report are based
on information obtained during 2005 at various fact-finding meetings,
public meetings, laboratory visits, and field excursions. And, there are
copies of that letter report available to you out front. The findings
that we made in that report are grouped into four areas, and I will
briefly highlight some of them now.

In the area of the waste management system, DOE announced a decision to
evaluate a canister system for transportation, aging, and disposal, thus
gave birth to another acronym known as TAD. The Board believes that this
system warrants examination, and recommends that DOE determine
first-hand the compatibility of possible TAD canister designs with the
capabilities for storage, handling, and transportation options at each
reactor site. We are also following the development and use of DOE\'s
Total System Model, given the model\'s potential to increase
understanding of how everything fits together, of the interactive
effects of various waste management system design and operational
components.

DOE\'s analyses and understanding of the engineered system continue to
evolve in a number of areas. However, the Board reported that it did not
find compelling DOE\'s arguments for screening out localized corrosion
of the waste package from deliquescent salts at high temperatures. And,
the main reason we reached this finding was that the data that we saw
that was presented to the Board were not for temperatures above 150
degrees centigrade.

DOE\'s studies of the natural system and natural processes continue to
enhance fundamental understanding of their potential behavior. For
example, DOE presented experimental data indicating that the transport
of neptunium may not be significantly delayed by co-precipitation,
unlike uranium. The Board is encouraged that the Office of Science and
Technology continues to support fundamental investigations regarding
source term phenomena and the radionuclide transport characteristics of
the natural system.

Finally, with regard to repository performance assessment, the Board
recommended that, in addition to and parallel with the repository
compliance case, the performance assessment being the primary document
against which the technical issues are resolved in the licensing of the
repository, that the DOE should also develop a realistic analysis of
repository performance. The Board believes that such an analysis would
be invaluable for fundamental understanding, for informing key
constituencies, and for building confidence in the DOE\'s estimate of
repository performance.

Now, let me turn to today\'s meeting, and set the stage, if I may.

In evaluating the technical validity of DOE activities, the Board
considers a number of factors, including (1) the relevance of the
activities to the performance of the total waste management system and
repository systems. By performance, we mean safety and throughput. (2)
the extent to which the activities are integrated and compatible with
the total waste management and repository systems, and (3) the extent of
fundamental understanding of the technical and scientific issues
involved. To make these determinations, the Board assesses project
priorities and issues on a regular basis. And, we have such a session
scheduled later this month, as a matter of fact.

An example of this is today\'s meeting. The agenda focuses on specific
factors affecting the radiation dose at the accessible environment of
the proposed repository. Key parameters for assessing radionuclide
processes include the mass flux of water and relevant radionuclides
entering and exiting the different barriers and zones of the repository.
Barriers to radionuclide mobilization and transport into the natural
system include the surface geology above the repository horizon, the
engineered barriers consisting of the drip shield, the waste package,
and the invert; and the byproducts of the degradation process involving
different chemical and mineral phases of the mobilized waste. The
quantity and chemical form of the radionuclides in the source term have
a major impact on radionuclide transport, retardation, and retention in
the natural system.

Our objective is to evaluate DOE\'s fundamental understanding of
radionuclide containment and transport processes that dominate the
radiation dose at the accessible environment. That\'s what we want to do
today. Such understanding includes knowledge about uncertainties in dose
calculations over times that include the peak dose and the likely
behavior of the repository system and its individual components.

Fundamental understanding of radiation dose calculations implies
knowledge of the processes that affect radionuclide behavior and how
such processes might be perturbed by episodic events such as earthquakes
and volcanic activity, and knowledge of the behavior of different waste
forms and how they contribute to radionuclide transport. It also implies
a basic understanding of the methods of analysis and their connection
with the supporting evidence, be it from the site characterization
program, analog investigations, or general scientific knowledge.

Of particular interest to the Board are DOE estimates of the amounts and
chemical form of the dose-contributing radionuclides entering and
exiting different hydrogeologic units within the unsaturated and
saturated zones of the repository. Increased understanding in this area
could provide a better accounting of the contribution to radionuclide
retention and transport of specific elements of the repository system
while providing greater visibility into the difference between the flow
of water and the transport of radionuclides, a difference that could
prove to be important in demonstrating the waste isolation capability of
the repository.

One of the complications of modeling the disposition of radionuclides in
the proposed Yucca Mountain repository has to do with scale. Because of
the relatively low water seepage rates and small quantities of
radioactive material (mass and curies) exiting the engineered barrier
system in comparison to the very large scale natural hydrogeological
system, it is extremely difficult to take full credit for the
geochemical processes that may be important to quantifying the isolation
capability of the natural system. It is a matter of grams and curies of
a few radionuclides interacting with megatons of material. In part, this
is why the Board has asked DOE to present predictions of mass and
radionuclide activity fluxes over time and space at the interfaces of
the repository subsystems. The Board believes that a clearer
understanding of such scale issues and the processes involved could lead
to greater confidence in the performance assessments of the repository.

Now, a few words about the agenda. Today, DOE and its contractors will
be making four presentations, beginning with an introduction by the
almost new Chief Scientist of the Office of Civilian Radioactive Waste
Management, Dr. Russ Dyer. And, Russ will explain what I mean by almost.
Russ has many years of valuable experience with the program, and we look
forward to hearing his thoughts. And, he\'s certainly no stranger to us.

Next, we\'re going to hear from Mike Ryan. Mike is chairman of the
Nuclear Regulatory Commission Advisory Committee on Nuclear Waste, and
will present his views on conservatisms, non-conservatisms, and
uncertainty in dose calculations and the risk-informed approach to dose
calculations. As a health physicist, an editor of the official journal
of the Health Physics Society, and Chairman of the Advisory Committee on
Nuclear Waste, Mike has substantial expertise in the human health
effects of radioactivity.

Next up following Mike will be Tim McCartin of the U.S. Nuclear
Regulatory Commission staff describing the implementation of a dose
standard beyond 10,000 years.

Following Tim\'s presentation, the agenda will focus on mass and
activity fluxes through repository subsystems, first of water seeping
into and out of the drift tunnels. And, after lunch, we will discuss
radionuclide releases from the waste package, waste form, and drift
tunnels over time. We will then have a presentation describing the mass
and activity of key radionuclides potentially released from the
unsaturated and saturated zones over time. The last presentation will be
a study by the DOE Management and Technical Support describing their
peak dose sensitivity analysis over a one-million-year time frame.

And, as usual, following the presentations, we have scheduled time for
public comment, which is an aspect of our meetings that is extremely
important to us. If you would like to comment at that time, please enter
your name on the sign-up sheet at the table near the entrance of the
room. Of course, written copies of any extended remarks can be
submitted, and will be made part of the meeting record. Some of you have
asked about questioning during the course of the presentations. Our
preference for that would be for you to write down your questions, and
submit them to Linda Coultry, she will be visible in the back, and leave
this information at the sign-in table. And, we will cover as many
questions as we can, time permitting.

And, then, finally, I would like to ask all of you to turn your cell
phones and pagers to the silent mode, including myself. I was guilty of
not doing this yesterday.

Without further ado, I am pleased to introduce Dr. Russ Dyer. Russ has
recently assumed the role of Chief Scientist for the Program, a position
that the Board has long recommended. He brings a wealth of experience to
the job, and we look forward to hearing his views.

Russ?

DYER: Thank you, Dr. Garrick. I would certainly like to welcome the
NWTRB back to Las Vegas, and for visitors for whom this may be the first
time here, welcome.

Let me take a minute and talk about this facility that we\'re in,
because it\'s relatively new here in Las Vegas, and it is unique. Across
the way is the Atomic Testing Museum, and if you have a chance to visit
the Atomic Testing Museum, I certainly urge you to do it, because it
captures the history of a program, a facility and a group of people that
had far-ranging impact on science and technology, not just in Nevada,
but in this nation and throughout the world.

My own experience in there, an hour in that museum will just whet your
appetite. It takes about three hours to really do it justice if you\'re
interested in the progress of technology and the role that atomic
testing had in that.

With that, let me get started. For the record, I\'m Russ Dyer. I am
currently the Assistant Deputy Director for Science and Technology of
the Office of Repository Development. But, I\'ll tell you a little bit
later what Dr. Garrick was alluding to. We have a change underway within
the program.

I\'m pleased to be here today to address the Board. Since our last
meeting in November, there have been a number of changes, organizational
changes, in the Department, and progress has been made in engineering
and science and we\'ll detail some of that progress for you today.

Today, I\'ll provide the Department of Energy\'s Office of Civilian
Radioactive Waste Management program and project overview. The overview
will focus on organizational changes, the funding for fiscal year \'06,
engineering and science. Afterwards, I\'ll take questions from the Board
on the points I present.

In organization\--let\'s go ahead and have that slide up, if we could,
please\--there have been a few changes to the OCRWM organization since
the November meeting, and more are forthcoming.

The announcement of Edward Sproke as the nominee for the Director of the
Office of Civilian Radioactive Waste Management was made some time ago.
Mr. Sproke is awaiting confirmation. Until his confirmation, Paul Golan
is the acting director. Whenever Mr. Sproke is confirmed, then Paul will
drop back to his permanent position as the Principal Deputy of OCRWM.

Now, what we have up on the screen here is an organization chart that
was just approved a few weeks ago, but is not yet in effect. We need a
transition period to move from the existing organization to this new
organization, and I\'m going to talk about some of the key points on
this organization.

This is, let me first just talk about some of the general philosophy
behind this organizational construct. It\'s a flat organization. All of
these managers, managerial offices, report directly to the Principal
Deputy here. It does away with the distinction between east and west
organizations. You will notice that there is not\--the Office of
Repository Development is not on this. So, we have organizational
structure that will employ people both in the east and in the west.

It shifts several responsibilities and functions. I\'m going to walk
through some of those. There are 13 direct reports to Paul Golan. And,
let me talk about two boxes on here to start with. As Dr. Garrick
mentioned, I\'ve been named the Chief Scientist, so the Office of Chief
Scientist is here on the left, and there\'s a progression here, which
goes from, stepping across from Study, Design, License, Build, and
Operate, from left to right on here. So, we have the Office of the Chief
Scientist\--for those of you in the back of the room, I\'ll leave these
out, so it will take a while to get through this. The Office of the
Chief Engineer here, Paul Harrington is currently acting in that
position. That\'s one of the positions. We\'ve had two positions that we
have active recruitments on. That\'s one that we hope to fill very soon.

The other that we have an active recruitment on is in the Office of
Quality Assurance. Vinnie Brown vacated that office, and Mike Ulshafer
is acting in that office there. Continuing on across in Licensing, we
have a new individual in the Department, Mark Williams. In the new
construct, he will head up the Regulatory Authority Office. Currently,
he\'s taking the place of Joe Ziegler in the Office of License
Application and Science.

Let me tell you a little bit about Mark Williams. Mark joined us in
November as the Director of the Office of License Application and
Strategy, soon to be the Regulatory Authority Office. Mr. Williams
previously worked for the Department of Energy\'s Office of Environment,
Safety and Health, where he was involved in the safety of DOE nuclear
facilities for 15 years. He also has more than 25 years of nuclear
related experience, including licensing and reactor regulations at the
Nuclear Regular Commission, process control and engineering associated
with the light water reactor at Bettis Atomic Power Laboratory and
Testing on Naval reactors at the Pugette Sound Naval Shipyard.

If I continue on on the slide, down at the next level, we have an Office
of Waste Management, Chris Kouts will be the manager of that office, and
Office of Logistics Management, Gary Lanthrum currently of the National
Transportation Program, the Office of National Transportation, will be
heading up that office. And, there is an office on here that is vacant,
and that\'s Disposal Operations, and that is a construct that we\'re
putting in for the future. We don\'t intend filling that position for
the foreseeable future.

Down at the bottom level, we have what I\'ll call the support functions,
the Office of Project Controls, Ken Powers currently the associate
deputy in the Office of Repository Development, will be heading up the
Office of Project Controls. The Office of Procurement will be Suzanne
Mellington. The Office of Government Services, Rich Minning, and
finally, the Office of External Affairs, Allen Benson. So, those are the
13 boxes in the new organization that we\'ll be standing up over the
next several months.

Let me move now to the next topic, which is Lead Laboratory. Several
weeks ago, in January of 2006, the Department announced the designation
of Sandia National Laboratory as the lead laboratory to support the
Office of Civilian Radioactive Waste Management. As Lead Laboratory,
Sandia will provide management and integration services for all Yucca
Mountain scientific programs, including the science that supports
OCRWM\'s license application, and its defense in the Nuclear Regulatory
Commission\'s review process. Sandia will perform this work in
collaboration with supporting organizations, such as other national
laboratories, subcontractors, federal agencies, universities, and expert
panels.

Lead Laboratory will fall under the purview of the Chief Scientist. Let
me go back a bit and talk about some of the other things that
fall\--that have been listed as expectations within the Office of the
Chief Scientist responsible for coordinating scientific work on the
program, that\'s all scientific work throughout the Office of Civilian
Radioactive Waste Management, including post-closure science, the Lead
Laboratory and the Science and Technology Program.

The Chief Scientist will be the program\'s primary point of contact with
international groups or agencies, such as the International Atomic
Energy Agency, and the Nuclear Waste Technical Review Board.

Let me move now to the fiscal year \'06 budget. As discussed in the
November meeting, Congress approved \$500 million in OCRWM funding in
the fiscal year \'06 appropriation, of which \$50 million was reserved
for reprocessing. OCRWM is currently seeking a legal interpretation to
determine if OCRWM or another DOE program, for instance, the Office of
Environmental Management, is authorized to spend the \$50 million. As
you are aware, there is a legal foundation for the missions and
functions of the Office of Civilian Radioactive waste Management.

The \$450 million appropriated is \$172 million below the fiscal year
\'05 funding levels, and \$201 million below the fiscal year \'06
request. Recently, the federal government announced a 1 per cent
rescission on top of other reductions in the federal budget. That\'s
across the board. This rescission will result in an additional \$5
million reduction to the OCRWM fiscal year \'06 budget.

The Department is in the process of developing its fiscal year \'06
annual work plan to establish program priorities in light of this budget
reduction. Critical Decision 1, which we talked about previously,
that\'s the corporate decision making process within the Department of
Energy for major systems. We have a proposed change that we would be
submitting to the DOE decision hierarchy associated with the
incorporation of the transportation, aging and disposal of canisters
into the system concept. This is the TAD concept that Dr. Garrick talked
about. And, requirements management are some of our top priorities. So,
CD-1 and requirements management.

In addition, the Department is continuing processing of refinements to
the Total System Performance Assessment. However, until the annual
planning is complete and impact of the Critical Decision 1 decision and
the overall program is determined, detailed priorities and science are
not yet set.

Let me move now to engineering. And, most of this is associated with
work being done to support this Critical Decision 1 milestone. As stated
in the November meeting, the Department announced its plans to simplify
the design and operations of surface facilities. This simplified
approach involves changing from a bare fuel model to one based on the
use of TADs, the transportable, aging, disposable capable containers.
The Department directed its contractor, Bechtel SAIC Corporation to,
one, stop work associated with activities that supported primarily their
fuel handling at the repository, other than that required for limited
bare fuel and off-normal operations, and, two, to develop this package
of supporting information to support the Critical Decision 1. And, that
would be a package that includes conceptual design that addresses a
simpler surface facility and canister operations, including, as I said,
the conceptual design and associated planning information.

The Department currently plans to submit the CD-1 package to the Energy
Secretary\'s Acquisition Advisory Board for review and approval in the
spring of this year. If the Board gives its approval, the Department
will proceed to implement the TAD concept.

The Department agrees with the Nuclear Waste Technical Review Board that
the TAD design will require close coordination among DOE, utilities and
the regulator, and this will be very important in the selection process
to ensure compatibility with repository operations. The Department
expects to pursue the TADs concept, consistent with this principal.

Let me move now to requirements management. In December of 2005, the
Department issued a letter to Bechtel SAIC stating effective
immediately, no engineering and preclosure safety analysis technical
work products, subject to the quality assurance requirements and
description document, may be approved until DOE re-validates the Yucca
Mountain Project technical requirements baseline, and contractually
conveys applicable requirements.

The Department\'s action was the result of a recent OCRWM concerns
program investigation into allegations that the project had not
maintained or properly implemented its requirements management system,
resulting in potential inadequacies in the design and control process.
Since these findings raised uncertainty regarding the adequacy of design
and products developed under the YMP design and control process, the
Department has made resolution of these issues a top priority.

To resolve this issue, the Department has taken immediate actions, has
plans to take remedial actions, and actions to prevent recurrence. And,
some of these actions, I\'m going to list now.

First. The Department of Energy re-validates the project\'s technical
requirements baseline, and contractually conveys these applicable
current requirements to BSC.

Secondly. BSC procedurally maintains and flows requirements down to
implementing mechanisms for programmatic requirements or products for
technical and procurements requirements.

Third. BSC internally validates that processes exist and are being
followed, and requirements are appropriately traced implementing
mechanisms and products.

Fourth. BSC provides DOE documentation that their actions are complete.
And,

Fifth. DOE verifies that the BSC processes exist and are effective.

Let me now move to science. The DOE presentations on today\'s agenda
will cover a range of post-closure topics that affect the assessment of
dose. The primary emphasis and focus will be on processes and models
developed for 10,000 to 20,000 year time frame of performance.

For today\'s topics, the presenters will provide their best estimate of
performance critical parameters under the constraint that assumptions
made need to be well justified. This will allow the Board the
opportunity to more fully explore the appropriateness of these
assumptions in these technical areas. We have an aggressive agenda
today. With the limited time allotted for each topical area, each DOE
presenter will provide background information, then summarize the basis
and significance of the key processes. Significance described with
respect to individual components will not be discussed, since the TSPA
is still in draft.

In addition to processes and models, the Department will also present
scoping peak dose analysis, conducted to provide insights into the load
of different processes and events on peak dose, and additional
information on other features, events and processes not included in the
analysis.

I\'d like to caution the Board and the audience that the peak dose
results presented today are informative in nature only. The results are
not intended to demonstrate compliance with any standard, and, thus,
should not be compared to any proposed or final regulation.

In addition, the information presented on processes and models represent
what is in the current project baseline, which, as I said, is subject to
change if the baseline changes as a result of the fiscal year \'06
planning process, namely, the incorporation of TADs into the baseline.

In summary, the Department is committed to examining canister-based
design and operations as a top priority. The Department believes that
the best operations are those that are simplest and most
straightforward. The canisterized approach incorporates this philosophy.
The Department understands that the selection of TADs is important, and
the coordination among DOE, utilities and the regulator will be crucial
to the selection process to ensure compatibility with repository
operations.

Annual planning is underway. Upon completion of the annual planning and
the CD-1 process, the program\'s priorities and a new baseline schedule
for license application submittal will be established. In addition to
Critical Decision 1, the Department anticipates that requirements
management improvements to the Total System Performance Assessment will
continue to be top priorities on the program.

That\'s the end of my prepared notes. I have one personal note that I
would like to add here. About 16 years ago, I started interacting with a
brash young staffer with the Nuclear Waste Technical Review Board, and
we were struggling at that time with developing a way to estimate
performance of a geologic repository system over a very long time
period. And, Leon was involved in some of those early very chaotic
meetings, and I have fond remembrances, and we will miss Leon. So, thank
you, Leon.

With that, I would like to take any questions from the Board.

GARRICK: Thanks, Russ.

Okay, let\'s see what the Board has to say. Henry, and then Andy.

PETROSKI: Petroski. Who indicated that under the rubric of engineering,
that the design was going to be looked at with the view towards
simplifying it, going to a simple design? Does that mean you\'re going
to look at the existing design and try to simplify it, or is there going
to be an effort to go back, say, to square one and\--is it a simple
design from scratch?

DYER: I think it depends on what opportunities were given by the TADs
concept, incorporating the TADs concept. There\'s some of that that can
be done with I think relatively minor changes to some of the existing
components. There may be great benefits from introducing a whole new
approach, and I haven\'t been that close to the design process, so I
can\'t tell you for sure where we\'re headed for that.

GARRICK: Andy?

KADAK: Kadak, Board.

I have a number of questions, but I\'m particularly interested in what
you think the schedule is for this project. It didn\'t appear that you
said anything about that. Could you kind of fill us in about the
licensing submittal? It sounds like you\'re going back to the drawing
boards, as Henry just mentioned. Give us a flavor as to where you are
relative to timeline.

DYER: We are trying to develop a timeline. We\'re trying to figure out
exactly what we need to do, what the work that needs to be done, and
then we\'ll work out the schedule of how to accommodate that work. We do
not have a schedule yet.

KADAK: And, just relative to your role, I was also a bit confused, and
in one statement you said was surprising, you haven\'t established
detailed priorities in science yet. I mean, explain that, please. I\'m
surprised.

DYER: We have some alternatives that are on the table. Bechtel SAIC
prepared a proposed work package that we are in the process of reviewing
now. It accommodated earlier priorities that we had in the program, and
they believe that it accommodates the priorities we\'ve placed on
incorporation of TADs into both the design and into the performance
assessment.

We\'re looking at ways that we might enhance that. I was involved in
meetings fairly late last night, looking at what we really need to do to
put the best program before us. So, I would say that we have maybe 80
per cent of the program pretty well understood, but there\'s quite a bit
that we\'re going to make sure that if there are important things that
need to be done now, that we bring them into the program now.

KADAK: Thank you.

GARRICK: Mark?

ABKOWITZ: Abkowitz, Board.

I took your comments, Russ, to imply that the Department of Energy is
pretty much adopting the strategy of full utilization of TADs to the
maximum extent possible. And, I was just curious whether the decision
has been made up until this point in time on studies that DOE has been
conducting alone, or whether there have been discussions with some of
the other stakeholders, particularly the utilities, about the transfer
of risk to utility workers, because more of the bleming activity will be
going on at the sites, the willingness to take fuel out of dry storage
and put it back into pools for repackaging, and a number of issues that
if I were a utility, I would want to have a seat at the table. Could you
please comment?

DYER: There have been, and there will be discussions with the utilities
that have been a fairly informal level to date. I expect that that will
kick up with time.

GARRICK: Ron?

LATANISION: Latanision, Board.

Russ, I\'m curious if at this stage, you could give us an indication of
the organization of the office that you will lead in terms of the
staffing? Is it premature to do that, or\--

DYER: It\'s premature. Right now, I have a staff of one, me.

LATANISION: May I make a suggestion? I\'ve always felt the absence of
someone who would take responsibility or would be the visible leader for
such issues as materials, for example. And, I think that senior
leadership has always been a concern to the Board, not only in terms of
materials, but joining processes, and so on and so forth. I would just
recommend that you should consider appointing to your staff some people
who provide that senior leadership, because I\'ve always felt that
absence.

DYER: Well, one thing that I\'ll inherit is the current Science and
Technology Program, which, as you\'re aware, has the thrust areas, of
which materials performance is one of the thrust areas, and there are
other thrust areas. And, there\'s pros and cons of bringing some kind of
an expert body such as that into the mix. I\'ll be looking at that.

LATANISION: Latanision, Board.

Just to follow up. I mean, to be really blunt, I mean, if I were asked
who is in charge of materials on this project, I think there should be
an answer to that question. And, right now, I can\'t tell you, and I\'m
not sure that anyone can, who is responsible for materials, and it\'s a
critical issue in terms of the waste package, and so on.

DYER: I don\'t disagree. This will give us an opportunity to do that,
because everything will be under the Chief Scientist.

LATANISION: Thank you.

GARRICK: David?

DUQUETTE: Duquette, Board.

As you know, I was privileged to attend a recent meeting of materials
issues relative to possible changes in the canister design, using
amorphous alloys instead of C-22 for the outer layer. With the TAD
concept being reexamined, or examined, you can look at it either way,
there was a TAD concept, I think, some years ago, will the Science and
Technology Program impact on the license application, on the TAD design,
will it change the design of the repository? What\'s going to happen to
the new concepts that are being introduced on the materials side for the
canisters?

DYER: Right now, we still have that program, as well as other
structurally amorphous materials, the specific program Dr. Duquette is
talking about. We\'re still in an R&D program on that, so we still have
some things to prove out to ourselves, and one of the things that we
need to understand is when and how to bring advancements such as
structurally amorphous materials into the program. When is the
appropriate time to do it? Right now, we still haven\'t convinced
ourselves that it\'s at a stage that we\'re ready to bring it into the
program. As you\'re aware, there\'s some pretty aggressive testing
that\'s due this year, and into next year, and I think that will\--that
gives us the opportunity to improve our confidence that those materials
have potential advantages associated with the use of those materials can
be brought into the program. I think, actually, we\'ve got a meeting
next week between the S&T Program and the Design Program, here to talk
about opportunities and talk about timing, start working the details
that you\'re talking about here.

DUQUETTE: Duquette, Board.

The first one, I want to congratulate you on supporting that program,
even if it doesn\'t work out, it\'s what the, I think, the S&T Program
was supposed to do.

But, my question really had more to do with if you\'re going to be
redesigning the containers as TADs, will the new materials aspects
impact that design process, and will that, in turn, impact the design of
the vault itself?

DYER: In the near term, I would say no, because I don\'t think that
program is mature enough for us to bring it into licensing at this point
in time. In the future, it may well bring about a change to the
licensing basis.

GARRICK: Okay, I have questions from Bill Murphy and Thure, Andy, Howard
and myself, and then we\'ll probably cut it off and go to our next
speaker. So, Bill?

MURPHY: Bill Murphy. I\'m the consultant to the Board.

You mentioned that we\'ll see results that are relevant to a 10,000 or
20,000 year time period, and I recognize that a great deal of work has
been done to evaluate how the system will work on that time scale, do
you perceive a different set of issues or a different set of priorities
to the scientific problems associated with a million year time scale?

DYER: To a large part, that depends on the final form of the EPA
standard, and NRC\'s implementation of the EPA standard. If you can use
and justify the use of the models for the 10,000 to 20,000 year period
and extrapolate those out in time, if there were some models that can\'t
be legitimately justified, you\'ve got to identify what those are and
how you\'re going to address them. And, I don\'t think there is an easy
answer to that.

GARRICK: Thure?

CERLING: Cerling, Board.

It\'s always unpleasant to find out you only have 80 or maybe 70 per
cent of the money that you anticipated. And, so, my question has to do
with does this mean that you will cut things across the board at 70 per
cent? Are there key pieces of the puzzle that cannot be funded at the
appropriate level and will cause a significant time lag, and what sort
of time lag does this sort of funding inequity result in, and how are
you going to make your case in the next two to three years that this has
to be changed?

DYER: Well, identifying what the critical pieces of work are that need
to be done to support a path forward, either in science or in the design
and engineering arena is where we\'re putting our priorities now. Much
of that depends on what comes out of the conceptual design associated
with this Critical Decision 1 package, and how we see the impacts of
that propagating through the system. And, then, responding to that to
make sure that we have the work in place to support that. So, I see that
being a key priority.

As you\'re well aware, we\'ve got a lot of relic issues also that we
have been and will be working on, and it\'s going to be a real challenge
to make sure that those don\'t get pushed off the table as we focus on
emerging priorities.

GARRICK: Howard?

ARNOLD: Arnold, Board.

Dr. Duquette alluded to the fact that TADs had been considered in the
past and rejected, and we still have some uneasiness about the
implementation. I think it\'s easy to see the advantages of TADs, but
the disadvantages I think need to be addressed and put to rest. as you
alluded to, our uneasiness over some aspects of it, and I think it would
be really worthwhile to make a connected story out of why the previous
decision was wrong.

DYER: I don\'t disagree with you. I\'m not prepared to do that right
now. I can offer something to the Board. There was kind of a lessons
learned that we put together in a multi-purpose canister, as it was
called at that time, which we can make available to the Board. I think
it\'s somewhere in the archives, I\'m sure. But, I would suggest that
might be a good topic for a future meeting also.

GARRICK: Andy?

KADAK: Yes. I have to prioritize my questions, because I had so many.

GARRICK: Please do.

KADAK: I will try. All right, I\'ll make two brief ones. One is I\'d
like to really understand the status of the Total System Performance
Assessment. I understand there\'s been a review made, quite critical,
wondering what you\'re doing to resolve that. That\'s question number
one.

And, two, there\'s been a lot of discussion on the Global Nuclear Energy
partnership, which talks about solving the world\'s nuclear waste
problems, and I\'d like to know what the impact of that program might be
on Yucca Mountain.

DYER: Okay. Let me take the second one first. It\'s hard to say, because
I haven\'t seen a coherent one. We, the nation, hasn\'t seen a coherent
policy or any legislative proposal associated with that yet. We see
inklings of it, for instance, the \$50 million in the appropriation for
this year that was associated with reprocessing suggests that there is a
sense of urgency and importance in bringing back reprocessing into the
nation. But, I haven\'t seen, or we haven\'t seen an overall context
that that fits in. So, it would be total speculation on my part to say
how that\'s going to all fit together.

As far as the Total System Performance Assessment, Bechtel SAIC had an
independent review, looking at the TSPA. They\'ve been looking at it
since sometime in \'04, I believe. There are a number of comments, some
are quite critical, some we agree with and they need to be addressed and
fixed, some we disagree with, and trying to figure out how to move
forward in that process is one of the challenges that we have.

KADAK: But, do you have any estimate as to when that will be resolved?

DYER: I would hope that we would be able to put that behind us by this
summer. But, it\'s clear that the TSPA that was being reviewed by that
team is not the TSPA that we\'ll be able to take into licensing. So,
we\'ll have to start work on a follow-on TSPA that accommodates the new
EPA performance requirements.

GARRICK: Two quick questions, Russ. Garrick, Board.

Number one, is it intended to have your counterpart in the Lead Lab, in
other words, is there going to be in the Lead Lab a Chief Scientist for
Yucca Mountain specifically?

DYER: I\'m not sure how the Lead Lab proposes to construct things.
We\'re starting discussions actually tomorrow about the transition.
It\'s going to take a while for the Lead Lab to stand up, just the
administrative framework of how to do project control, quality assurance
program, how we are actually going to flow money to and from Sandia to
the other organizations. So, I don\'t know the answer to that.

GARRICK: Yeah, I\'m thinking of kind of the model of the Waste Isolation
Pilot Plant, where there was such an approach.

DYER: Since one of the reasons for selecting Sandia was because of their
experience in the Waste Isolation Pilot Plant, they may well choose to
bring some of the organizational structure and approaches that they use
on WIPP to this program. But, so far, we haven\'t talked about that
level of detail.

GARRICK: A final question. You only have to read the paper to learn that
DOE, or the Department, is considering a number of initiatives that
could impact the management of radioactive waste. And, one of the issues
of some concern to the Board is how these initiatives might impact the
course forward and the focus relative to Yucca Mountain. I\'m thinking
of initiatives having to do with reprocessing and spin-offs from the
Generation Four studies, and what have you.

Can you say anything about what\--and maybe this gets into the Critical
Decision business that you mentioned earlier\--can you say anything that
would give us some confidence that the Yucca Mountain project is not
going to suffer from these initiatives as a result of any possible
distractions, or the feeling that now there are alternatives are
developing, and that\'s a basis for delays, et cetera, et cetera?

DYER: If you go back in time five to eight years, some of the same
issues were being brought up\--this is when partitioning and
transmutation was a big thing\--I think there\'s recognition on all
sides that no matter what system you go with, reprocessing, if you are
able to get a transmutation program eventually in place, all of these
approaches require a geologic repository of some kind. And, I think
there is a realization that under any approach, you need a geologic
repository. I haven\'t seen anything that takes away from the need for a
geologic repository. Some of the approaches, some of the proposals may
balance the urgency of it, but the ultimate\--

GARRICK: Well, it\'s the urgency part that I\'m most interested in. I
agree with you that a repository is needed, no matter what alternative
is considered. But, the question really focuses on the impact that these
initiatives have on the Yucca Mountain Project. And, that\'s all I was
trying to address.

Okay, thank you very much. It\'s a good start. We\'re only a couple of
minutes in time, so now we\'ll hear from our distinguished visitor from
the Nuclear Waste Group in the Regulatory Agency.

RYAN: Mr. Chairman, members of the Board, thank you very much for your
invitation to be with you today. I had planned to be with you a few
months back, but took ill and was in the sad position of not being able
to speak, which is always troublesome. But, I appreciate the invitation
to come again today.

What I would like to do today is to share with you some ideas on
conservatism, non-conservatism, and uncertainty in dose calculations,
and how to risk inform our thinking perhaps on dose calculations. We
very often hear about dose conversion factors in whatever country,
whatever venue, and there are lists and tables of dose conversion
factors in one form or fashion, and what I want to do is highlight for
you today where they\'re certain, where they\'re uncertain, and how we
can improve our thinking about such dose conversion factors.

Next slide, please. The views that I present are my own, so I want to
share with you that thought.

Back in history, 1999, the U.S. Nuclear Regulatory Commission issued a
policy paper that described its risk-informed and performance-based
concepts and how it should apply to NRC\'s regulatory work. It\'s an
important benchmark for radiological performance assessment, and I think
it should be extended to our thinking about dose conversion factors as
well.

I\'d be remiss if I didn\'t take the opportunity to put forth the risk
triplet of Kaplan and Garrick of what can go wrong, how likely is it,
and what are the consequences. I promised John when I was on the ACNW
I\'d use this slide in almost every talk I give somewhere. So, I\'m
happy to give it. But, it\'s not a set of questions, it\'s not
inappropriate to ask about dose factors.

Do the metabolic models that we ascribe to them in the reference
calculation fit the circumstances at hand? We now have, for example,
dose conversion factors for young adults, for children, and for adults.
Do the chemistry and physics of the setting in which we assumed various
physiologic parameters represent the setting in a particular case at
hand, and so forth. So, there are things that can be different in dose
conversion factors, and there can be consequences in what we calculate
for dose.

If we use some risk insights from assessments using these principles, we
can highlight important radionuclide contributors to risk and their
significance. We may, in fact, find some radionuclides that we don\'t
need to worry about that we classically have worried about, and we might
find some radionuclides that we have to pay a little bit more attention
to the details of what one calculates for a dose.

Traditionally, there are a range of tools that yield a range of results
in our understanding of risks and performance assessments. We have gone
from extreme bounding cases, the case of assuming everything that can go
wrong does go wrong, and we calculated dose, and if that numerically
satisfies us against some standard or against some way of thinking about
it, we say good enough.

Next is bounding analysis, which is a little less extreme in the
assumptions, followed by sensitivity studies where we examine various
issues, parameters, whatever it might be to say what happens if, and
then one-off calculations and comparisons where we take these kinds of
analyses and gain insight and understanding from comparison. And, then,
finally, probabilistic risk assessment, which is a more fundamental and
perhaps better way to do some of these analyses.

The applications for risk-informed approaches, of course, apply to all
aspects of a performance assessment. Inventory. People often assume
that\'s a trivial matter, but it\'s something that needs attention. What
is the inventory? What will the inventory be over time? And, what
physical and chemical form might it be in?

The source term. What fraction of the inventory leaves and goes
somewhere else, and under what process? What are the released fraction
of radionuclides, what gets released from the engineered system? Does
that change it in some way physically, chemically, or otherwise? What
are the interactions in the near field, that is, where is the source and
the packaging and the system influence still exist, and what are the
interactions in the far field? Typically example of a far field
parameter of risk significance is Kd. What is a Kd? What are the
geochemical processes that can impact? And, then, finally, uptake and
dosimetry estimates.

We spent an awful lot of time over the years talking about Kd and
transport, but we haven\'t talked much about the equivalent parameters
in the intake and dosimetry calculations, and there are some of
interest.

Internal dosimetry 101 is there are inhalation exposures typically
involving, but not exclusively involving, the respiratory tract. It\'s
true that you cannot have an inhalation exposure without having an
ingestion exposure. Part of what you breathe in ends up in your stomach.
So, it\'s a complicated situation. There are ingestion exposures
directly in food and other stuff. And, then inunction for the completion
of the three I\'s is skill absorption, which is mainly an issue for
tritium and radionuclides that enter in the workplace through wounds or
other kinds of exposures where the skin is broken.

In the inhalation case, aerosol science is a key part of assessing, for
example, workplace exposures that have little substrate for the
radioactive material, a much different circumstance than inhaling dust
where the actual physical dust itself can be a big part of the exposure.

Environmental exposures can involve significant amounts of substrate
mixing and interacting with the radioactive material. So, does that
infinitely dilute solution of radioactive material behave the same when
it\'s interacted for decades or hundreds of years with earth materials,
and does the exposure result in the same dose?

Biological sciences are also important with the aerosol science where we
deal with solubility, deposition, absorption and clearance kinetics. All
of these factors can significantly influence dose.

An interesting reference point in this arena is physiology. My typical
dose factors assume that for light work, we inhale 0.54 cubic meters of
air per hour, with a respiration rate of 12 per minute. Light exercise
is 1.5 cubic meters of air per hour, and a respiration rate is 20 per
minute. Well, obviously, this is a direct scaling factor to what\'s
inhaled. I guess I\'m not in very good physical shape, because my
respiration rate when I\'m doing work is much higher than 20 per minute.

And, the other assumption is that all the air breathed in is breathed in
through the nose. I don\'t think that\'s the case for most strenuous
activities. I actually did a little study, I look around the room and
see who has their mouth open and who has just\--and, believe me, we\'re
not all nose breathers.

GARRICK: Actually, the ones that are asleep.

RYAN: So, I find this interesting in that it is an interesting reference
point, but I think we owe it to ourselves to examine and risk inform our
calculations, our particular setting and our circumstance to determine
whether or not this is useful and applies and is accurate for what
we\'re trying to accomplish.

And, I\'ll pick on another one. The GI tract uptake fraction. The
selection of a GI tract uptake fraction is a scaling factor to dose. In
ICRP reference documents, you will find 5 times 10^-4^ is the moderate
case where absorption is moderate for unspecified compounds. And, for
slower insoluble oxides, it\'s 1 times 10^-5^. In 1983, I\'m sad to say,
23 years ago, Dave Kocher and I did a review of GI tract uptake fraction
estimates, and at that time, came up with the number of around 10^-3^
might be more appropriate for environmental species.

Again, when the doses are fractions of a micromillirem per year, this
may or may not be important. But, in terms of demonstrating
understanding, I think it could be very important, just to cover the
base and make sure that we do understand the basis for these dose
factors, particularly in the plutonium case, and understand whether or
not that\'s helpful.

Very often, in the absorption case, tritium, about 50 per cent of the
tritium that\'s inhaled is assumed to be absorbed through the skin. If
somebody is in protective clothing, for example, that may or may not be
true. But, in the worker case, we always rely on bioassay. If you take
bioassay measurements of all sorts, if there has been as exposure, and
what do we do for that individual? We develop an individual specific
biokinetic model. We then kind of set aside the dose conversion factors
and do it through the individual. So, it\'s interesting to see the range
of values. Very often, the individual model doesn\'t look too much like
the ICRP model.

HORNBERGER: Is that tritium as hydrogen?

RYAN: Usually, there\'s a water vapor. It\'s very rare to see a
workplace exposure with anything other than water vapor, and hydrogen,
of course, exchanges very rapidly with the moisture in the air. So,
it\'s almost immediate. In fact, an intake of tritium uniformly
distributes into hydrogen pool in the body in about two hours.

Another example. I-129, there\'s lots of interest in I-129, but in a
recent paper that Dave Moeller and I published, we looked very carefully
at this, and it is very dramatically dependent on the stable iodine
daily intake rate. If, for example, the daily intake rate is around 400
milligrams per day, which is one of the key references we found, you
over estimate the dose using the reference dose factor by a significant
amount.

The story continues however. One of the interesting comments we received
out of the publication of that paper was that recent data by Hollowell
and others have suggested 150 micrograms per day. And, guess what? That
would under estimate the dose, if that\'s the intake rate and you use a
reference dose factor.

The important message here is that in this case, as in the case for
Carbon 14, the stable element of carbon, the dietary intake pattern
directly influences the dose conversion factor. And, that should be
accounted for. Interesting to think about. The reference factor, by the
way, in ICRP Publication 72 is based on an intake rate of 200. So, if
you\'re 400 intake rate, you\'re over estimating the dose by a factor of
2 just on that one parameter.

Carbon 14, the same way. It\'s interesting that there are variations by
a factor of about 1000 on intake assumptions for food, water and models.
The key assumptions count, and we actually summarized, and look at the
estimated dose based on various assumptions, from 10^-2^ microsieverts
per unit intake, to 10^-5^. All legitimate assumptions in the various
settings in which they were assumed, but it\'s relatively important to
make sure you understand how the stable element dilutes carbon in the
dose factor.

Let me talk about some interesting extremes. 10 CFR 61 is based on an
extreme bounding case. The probability of intrusion into a low level
waste site is 1. That\'s the assumption. The probability of intrusion
into Class C waste is 1. Exposure is maximized via all pathways to the
resident farmer. In a way, the resident farmer has to be unemployed
because he gets external exposure for 18 hours a day. He has to be an
expert agriculturist because he has to grow his entire inventory of food
in exhumed waste. I don\'t know how you grow food in ground up mop heads
and coveralls and shoe covers and ground up metal. But, nonetheless, all
of these assumptions were made to set the concentration values that are
in 61, 54. It\'s a bounding case, and it\'s an extreme one. But, it
certainly can mask overall risk, and certainly can mask our insights
into what contributes to risk. That\'s the message.

Interestingly enough, in 61, 58, there is a provision to allow for
alternative concentrations, calculations from alternative scenarios of
exposure as long as the principal requirements for protecting the public
health and safety are met.

RESRAD, commonly used calculation to analyze for disposition of solid
material. Is it a bounding analysis? It\'s specific to groundwater, and
it ends up basically assuming that one is using leachate as drinking
water. Water is taken through some disposal unit, through a theoretical
drinking water source, and there\'s no dilution and no dispersion in
that water from the leachate. So, in essence, you\'re drinking the
leachate. It depends on analysts choices for parameters in the models.
Infiltration rate, of course, being a key one, aerial projection of the
waste being another. And, again, risks may be masked or overstated.
Important processes can be missed, channeling, for example, rather than
sheet flow.

MARSSIM and D&D applications are a step forward. It\'s a statistical
approach, using, of course, the Wilcoxon rank sum statistical test to
assess residual contamination. The good news here is different analysts
will get similar results if they follow the method. It\'s appropriate in
that way. It\'s a more rigorous approach to uncertainty analysis. The
next step, of course, will be to tie those residuals to risk. How do we
understand what they mean?

10 CFR 63 is a stylized and prescribed approach, a representative
volume. It\'s required by law. Is it representative or conservative or
not conservative? Does it address all the sources of variability? And,
it helps to understand risks by exploring these conservatisms, both
positive and negative.

In these short few minutes, I hope to introduce you to the idea that
dose conversion factors and dose calculations shouldn\'t be left
unexamined from the standpoint of how we assess risk and what\'s
important. Should it be done for all radionuclides? Perhaps not. But,
for those radionuclides that are long-lived, mobile in the environment,
and give an inherently high dose per unit intake from any reckoning, it
probably ought to flow to the top of the list.

We often hear about Plutonium 239, Neptunium 237, two I\'ve mentioned,
Carbon 14, I-129, Technetium 99, and perhaps a handful of others, but
these kinds of detailed assessments may not change your view of their
relative importance, or of their potential dose consequence. But, I
think informing those calculations by understanding what variability
there might be in these radionuclides and their behavior can help
certainly solidify our understanding of risk.

There\'s are broad spectrum of approaches to risk informing decision
making, particularly in the performance assessment part, and
particularly in dosimetry. The goal should be that there should be
realism that presents best estimates and transparent assessments of
risk.

I think we need to challenge some of this old reference wisdom and dose
factors, and understand the foundations for all parameters. I think more
importantly, we need to know the limits of our established scenarios.

And, finally, we need to really understand when we\'re using extreme
bounding analysis and when various other tools, like sensitivity
studies, one-off calculations and probabilistic risk assessment are more
appropriate. All of these can play a role in dose calculations, and all
have strengths and weaknesses, and some are better than others.

So, with that, that\'s it. Thank you very much.

GARRICK: Thanks, Mike. Let me lead off with a kind of a question. As you
know from your ACNW experience, and from the proceedings of the Board,
there has always been an interest in getting to the fundamentals of
what\'s really happening. We\'ve talked extensively about it\'s
important for the experts to communicate what they indeed believe is the
performance capability of the repository.

This is an interesting example of being able to separate that kind of
analysis from a compliance analysis. And, given that the Yucca Mountain
doses are driven by a relatively few radionuclides in particular,
Iodine, Technetium, Plutonium, Americium, and Neptunium, are you aware
of anybody turning up the microscope on those specific radionuclides and
taking into account some of the things you have mentioned here that
might have an impact on, if we had such a thing, the experts\' realistic
assessment of the performance of the repository?

RYAN: I think, as I mentioned Dave Moeller and I have actually published
now four papers on this topic. We looked specifically at I-129. We took
up Carbon 14 next, and I think the other key radionuclides you mentioned
also deserve scrutiny. To my knowledge, at this point, I don\'t know of
anybody else that\'s done these kind of detailed probes into, you know,
what the risk informed results might be for dose factors. But, it\'s
certainly worthy of study.

GARRICK: Yes. Yes, Ali?

MOSLEH: Mosleh, Board.

Dr. Ryan, I\'m looking at your conclusion and was wondering if you have
any preference on the methods, approaches that you would list, and
particularly, I was interested in learning your opinion about the more
comprehensive use of PRA type approach to capturing all these
uncertainties that you had mentioned? Would that not be a
comprehensive\--for addressing all these?

RYAN: You know, it sure could be. Instead of assuming a dose factor as a
rigid number, you could, after studying a particular radionuclide\'s
dose factor, make a decision as to how you might treat that in a PRA. I
think that would be an interesting thing to do. Of course, it could be
as simple as the dose you calculate, even with an extreme dose
conversion factor, is so unimportant that it tends to fall off in
importance. But, then, you have a risk insight, don\'t you?

My point is that by using reference factors without challenging them,
and their fundamentals, that we miss that opportunity to better risk
inform.

GARRICK: Mark?

ABKOWITZ: Abkowitz, Board.

Mike, I found your primmer interesting, and I couldn\'t help but think,
as I looked at the different forms of analyses, whether or not you or
someone else has actually looked at TSPA, and the various models and
assumptions that are in TSPA, and tried to classify each element of TSPA
in terms of what type of analysis was used to establish that particular
set of parametric relationships. And, I\'d be kind of curious if you\'re
aware of, or whether you think there would be value to having a
scorecard where you could break down TSPA into various components and
associate with each of them the style of analysis that was used. I think
that would help the Board with its understanding, probably the DOE as
well, and also give us a better appreciation for what kinds of issues
are propagating in terms of the biases in the models.

RYAN: Again, I\'m not expert enough on TSPA to give you a full answer.
Much of that earlier work was done at ACNW, and in its reviews prior to
my watch. I do think that risk informing whatever performance assessment
code you might be using, and the way you\'ve described is often helpful,
because it tends to allow you to focus on issues that tend to rise up in
risks, rather than those that do not. Whether it\'s adjusted dose issues
or whether it\'s geosphere or near field or far field or whatever it
might be, I think that kind of thinking helps you to organize your
priorities a little bit better.

GARRICK: Ron?

LATANISION: Latanision, Board.

Mike, could we turn to your last slide, last conclusion slide, Number
23? It seems as though there has to be a lot of judgment made on which
of these approaches to use under what circumstance, and I wonder for
someone who is relatively uninitiated, mainly myself, could you give us
a couple of concrete examples of under what circumstance you would, as
you say, they all have strengths and weaknesses, but under what
circumstances would you use PRA, for example, as opposed to a
sensitivity study?

RYAN: I think in the absence of other information where I don\'t have a
foundation to make those judgments, PRA is a helpful tool that can begin
to help develop those insights.

LATANISION: So, that would be the way you\'d lead off. And, then, you
perhaps\--

RYAN: You could. On the other hand, for example, in radiation production
in the workplace, I think there\'s such a wide body of experience that
you can very often come up with sensitivity studies or bounding analyses
that will tell you for a work activity, you\'ve accomplished your
radiation protection objectives. I think they all have a role. But, you
know, I think we have to get a little bit more specific in when to
decide. And, I think it\'s hard to, up front, say let\'s always use PRA,
because that\'s the gold standard, and we start with nothing and end up
with risk insights, I don\'t think that\'s well schooled. I think you
have to look at the individual case and make that decision.

I think very often in the environmental performance assessment area,
you\'re often in a situation where one-off calculations and comparisons,
and sensitivity studies get you a long way down the road to risk
informing. They can also help you design a better PRA. That\'s where you
really want to end up.

GARRICK: The main plus of a comprehensive PRA, that\'s what it was
invented for, is to provide you a mechanism of developing realistic
models. That\'s when the whole concept evolved out of frustration of not
having realism in the investigations of what could happen, an accident,
for example, and it was the intention of PRA to fill that need, and
that\'s why conservative or bounding PRAs are nonsensical. They don\'t
make sense. PRA, by definition, is a realistic model.

Yes, Leon?

REITER: Leon Reiter, Consultant.

Mike, I understand a few months ago, a committee of the National Academy
relooked at the linea extrapolation hypothesis, and concluded that it\'s
still operable. On the other hand, there\'s a French group came out that
said that it was not appropriate. I wonder if you\'d talk about that,
and what implication this might have for the TSPA?

RYAN: That\'s a great question, Leon, thank you.

I think you\'ve characterized the Academy study that not much has
changed from year one through six, that they\'ve really recognized
updates in the epidemiologic studies and a few other things, but they
could not step away from the LNT as a basic model for radiation risk. I
was fortunate enough to be in France a couple of weeks ago and heard a
talk on their study, and, in fact, the Commission at our last briefing
asked us to further investigate the French report, and we\'re planning
now with the French to have them come and give a presentation on their
report. So, it\'s a little premature for me to comment on their report,
other than to say I believe they\'re recognizing a level of 1
millisievert or 10 rad. as a dose below which they\'re having a hard
time demonstrating risk.

REITER: Let\'s say that they\'re accepted, I mean, the community accepts
that maybe it\'s like 100 millirem, I guess\--

RYAN: It\'s e millisievert. 10 rads. I\'m bilingual, so I can help you
with that. It\'s 10 rad.

REITER: So, what impact would this have on\--

RYAN: It\'s a great question. You know, it depends on how you apply the
idea. If you apply it to all radiation exposure, well, let\'s include
background, let\'s include medical exposure that folks get as part of
the radiation environment we live in, I don\'t know. If you also look at
performance in the workplace, you know, the nuclear industry has a very
clear record of decreasing occupational exposure. So, it\'s one of those
issues that crosses from a science question, one of the risks at low
dose for various kind of effects, whatever they might be, versus what
kind of policy judgment do we make about it. What I can tell you a
little bit about at the moment is science aspects of it. The policy
aspects of it may be above my pay grade.

GARRICK: Okay. Andy, then Ali, and then David.

KADAK: Kadak, Board.

Based on your comments, do you have any feel, and your knowledge of the
licensing of Yucca Mountain, do you have any feel for if one were to
apply this kind of analysis, in terms of really understanding impacts on
humans, what do you think the numbers would do relative to public health
risk?

RYAN: I guess when you accept the idea that all the doses whoever has
calculated them to date are very very low, my guess, and it\'s strictly
a guess, because we haven\'t done all the analyses to firm that guess
up, is that I don\'t think impacts would end up changing very much
relative to any kind of comparison to a standard. But, I think it\'s a
weakness to not have that analysis in hand.

KADAK: Let me try to interpret what you just said. Are you saying that
if you apply a realistic model, as we\'re trying to apply realism in the
performance of the natural barriers, and other things, and if you
consider the fact that humans don\'t breathe like, you know, through
their mouth and through their nose, and other things relative to the
leachate, as you call it, the doses\--and I\'m assuming this is what is
assumed in the DOE analysis\--that the dose numbers could be
significantly lower?

RYAN: Well, I\'m by no means an expert on DOE\'s analysis, so I can\'t
speak to it. Let me speak in a more general way, if I may.

KADAK: All right.

RYAN: If you accept reference dose factors, you could be over estimating
or under estimating dose in whatever setting you might want to apply it.
My simple point is I think we need to understand it in the setting in
which we apply it, what error, uncertainty, conservatism,
non-conservatism, might be implied. And, I think it\'s a little bit of
all of those.

GARRICK: Ali?

MOSLEH: In your opinion, these uncertainties are mostly uncertainties
due to lack of understanding of the models and the processes, or is it
more like the parameters and the data that you need to inform those
models?

RYAN: I think our understanding of the basic physiology is certainly
robust. I think it\'s often a combination of a lack of how we can assess
for a specific exposure, a specific physical form, a specific chemical
form, and then make a dose calculation on that basis. If you look at any
occupational internal exposure, we end up with calculating for the
purposes of reference using reference models. But, then, we very
specifically say all right, we\'re going to do bioassay measurements,
whole body counting, body fluids analysis, whatever it might be, and we
determine a specific model for that individual.

A simple case is tritium. The reference excretion half-time is 12 days.
Well, what happens if it\'s eight days? The dose goes up by that
relative ratio, or down, rather. If it\'s 16 days, it goes up. The point
is without knowing what that individual\'s biological retention
half-time is for tritiated water, you really are estimating the dose.
You\'re not determining it as precisely as one could. So, I think to
answer your question, it\'s those metabolic parameters that describe,
you know, that come from the physiology for an individual, and then the
specifics of the physics and chemistry of what\'s inhaled or ingested.

GARRICK: David?

DUQUETTE: Duquette, Board.

This is probably more a comment than a question, but could you go to
Slide Number 4? I spent most of my career looking at the first two as a
corrosion person, because normally, you don\'t get called in unless you
already know what the consequences are, and usually, they\'re pretty
dire. The first two, of course, are the things that PRA is somewhat
based on.

RYAN: Yes.

DUQUETTE: In this particular project, there\'s been an attempt by the
project to eliminate corrosion as a possible source of radionuclides,
because the probability of it occurring are not very high. At least two
members of this Board disagree with that particular conclusion. But, the
fact of the matter is that I think you can take the wrong path if you
don\'t take a look at what the consequences are. And, I think the first
two address PRA, and I personally think that quite often, when people
assess probabilities, they do it incorrectly because they don\'t always
understand the mechanisms of what\'s going to happen.

RYAN: Fair enough.

GARRICK: Any other questions? Any questions from the staff?

(No response.)

GARRICK: Thank you very much.

RYAN: Thank you, Mr. Chairman. I appreciate the time to be with you.

GARRICK: Our next speaker, Tim? Tim McCartin from the NRC. You might
tell us what your job is there, Tim.

MC CARTIN: I work for the Nuclear Regulatory Commission in the Division
of High Level Waste Repository Safety. And, in that capacity, I really
have two roles. I\'m the Senior Advisor for performance assessment, but
I also have the lead role for developing the regulations for Yucca
Mountain.

And, today, in truth in advertising, although the title says
Implementation of a Dose Standard after 10,000 Years, if someone was
expecting I was going to provide the entire context of our implementing
the dose standard, I apologize. I am going to focus on two aspects of
our implementation in the time I\'m allotted. And, those two aspects
are, one will be the kinds of quantitative analyses, and our
understanding to date with respect to implementing a peak dose standard
with respect to dosimetry and performance assessment. The other is, and
I\'ll get into a little more detail on that, is our basis for specifying
how to represent climate change.

And, you will notice Dr. Gordon Wittmeyer, who my colleague, is much
more informed with respect to the climate change, and if I get questions
on that, I will most likely defer the hard ones to Gordon and take the
easy ones for myself.

Next slide? I\'ll give a slight introduction on the purpose of proposed
Part 63 to set up those other discussions that will get into our
understanding from an inventory standpoint, from a dosimetry standpoint,
and then for the most part, talk about the representation of climate
change.

In terms of our regulation, EPA has proposed a standard for a peak dose
limit beyond 10,000 years. NRC is implementing that by adopting that
standard. There are some other aspects with respect to how the
performance assessment is to be done, et cetera. And, I\'m not going to
get into the details of that, but the dose limit is probably the most
noticed aspect of the standard. But, there are other aspects to that.

Secondly, EPA did specify to update the dosimetry with respect to the
public dose assessment. For people who don\'t know, there\'s a slight
distinction between the EPA standards and NRC regulations. EPA\'s
standards are for the public exposures. NRC regulations are for\--we
adopt that public exposure limit, but we also specify our own limits for
worker exposures. And, so, when EPA specified use for current dosimetry
for public dose estimates, we have also taken the initiative to, rather
than having two different types of dose calculations, that use that same
newer dosimetry for the workers.

And, lastly, the specification of climate change at Yucca Mountain after
10,000 years, EPA specifically identified the NRC to come up with a
value for representing that climate change after 10,000 years.

Next. Okay, and, with that, I just want to go into first, our
perspective from an inventory standpoint. And, this is just looking at
the inventory going out in time, and you can see just using the 1,000
year inventory, and be aware there certainly is a fair amount of very
short-lived radionuclides that have decayed by 1,000 years, but using
1,000 year inventory in terms of curies as a metric, you can see at
10,000 years, it\'s approximately a quarter of what it was, going down
to 100,000 years, it\'s approximately 2 per cent at 100,000 years.

GARRICK: This is strictly based on decay?

MC CARTIN: Strictly decay. Just based on curies.

GARRICK: Are you going to show us a similar curve based on performance?

MC CARTIN: I have some information on performance. I don\'t know if it
will be as much as you would like, but I do have some, and happy to talk
to the basis for our calculations.

GARRICK: That\'s the one we\'re really interested in.

MC CARTIN: Okay. And, if one went out to a million years, my
recollection is that the 2 per cent drops by about an order of
magnitude. It would be about .2 per cent if you carried it out to a
million years.

Next slide? With respect to those time frames, one of the important
radionuclides, one of the things that the Commission is interested in,
and the Staff is interested in, is we have a perspective in the 10,000
years. Now, if you went out to a million years, how does your
understanding of what nuclides are important change over that million
year time period. And, you can see the thousand years were dominated by
americium and plutonium. Going out to 10,000 years, we still have
Plutonium 239 getting a little more important, as the americium and the
Plutonium 240 decays away. Going further, if you go out, you can see you
still have Plutonium dominating at 100,000 years.

I would point to one of the interesting things to the far right of that
100,000 year curve, you have neptunium starting to show up as a
contributor of that. Clearly at 1,000 years, in terms of curie amount,
it is not a significant amount of the inventory.

Next slide? Carrying it out further, you can see there is Thorium 230
has some significant in growth, and is the dominant radionuclide at
300,000 years with respect to curies. And, if you go all the way out to
a million years, you can see neptunium.

One thing I would like to point out in our analyses over the last two
decades, we have always had neptunium as a significant contributor, and
a significant nuclide to look at, even though it\'s only recently, the
idea of a million year calculation has come into vogue. The concept, you
know, you don\'t need a million year requirement to know that neptunium
was an important radionuclide, and you can see at the million years, it
is the single largest inventory. But, we had that with our 10,000 year
calculation. We were still interested in neptunium.

Next slide? That\'s some perspective at least on how the inventory
changes over that million year period. Now, as I stated, the EPA
standard has suggested using newer dosimetry, and if I look at Federal
Guidance Report Number 11, which is September of 1988, that is the
current basis for doing dose calculations. What EPA is proposing is
really moving to Federal Guidance Report 13, which is September of 1999,
newer dosimetry.

And, the next slide will show the changes for some of the key
radionuclides. And, you can see for some, in terms of the ingestion dose
conversion factor, for some of the nuclides, they increase, most notably
the first two, technetium and iodine increase by about a factor of 2.
Those typically are nuclides that show up in 10,000 year assessments,
because they\'re highly mobile, relatively soluble.

Neptunium dropped a little bit more than an order of magnitude. Thorium
230 increased some, and the americium, plutonium and uranium dropped
somewhat.

Next slide? In terms of the relative dose conversion factor, if I look
at FGR 11, the older dosimetry, you can see neptunium was the most
dominant radionuclide from a dose conversion factor. Technetium may look
like there\'s nothing there, but technetium, although there\'s a fair
amount of technetium in the repository, very long-lived, it is an
extremely low dose conversion factor, and there is a value there, it
just doesn\'t show up on this scale it\'s so small.

Next slide? Going to Federal Guidance Report 13, the newer dosimetry,
you can see sort of what\'s happened with neptunium. Everything else has
sort of come down to similar contributions. Neptunium is no longer the
dominant radionuclide, and other nuclides are comparable to the dose
conversion factor for neptunium. And, that\'s in terms of, as you will
see for our analyses, we were interested in prior to the newer
dosimetry, neptunium has always been the dominant radionuclide in our
dose calculations. And, so, we\'ve implemented new dosimetry and done
some calculations.

Next slide? Everyone is always interested in dose estimates. I will say
I debated long with myself on this one in terms of what\'s the best
title to give this slide. And, I chose the word \"illustrative.\" I want
to say that we are in the process of revising our models. It is a work
in progress. This curve is for us, we were using it primarily to see
with the newer dosimetry\--

GARRICK: But, it is a Yucca Mountain illustrative?

MC CARTIN: Oh, absolutely. Yes. We\'re in the middle of revising our
models, and I wouldn\'t want to say this is where we will end up with
our new model. But, we were interested in would the newer dosimetry
change the relative importance of neptunium, and you can see the red
line is neptunium, it still dominates the overall dose curve in our
analyses. That peak is, I think, somewhere on the order of 150,000
years, and around 9 millirem, and it tapers off.

GARRICK: You\'ve already alluded to some of the reasons why maybe this
is a different curve than we\'ve seen in the past. Are you going to
expand on that a little? This is very different in both the time of peak
and in the magnitude.

MC CARTIN: Well, I can give you some perspectives on why it looks the
way it is.

GARRICK: Okay.

MC CARTIN: And, what changes could alter things in the future as we
proceed. Number one, we have a waste package that fails primarily
between the period of 40,000 to 80,000 years. Those models continue to
be improved, enhanced, based on new data, new information, and I can\'t
say what that change might end up, but we are revising the corrosion
models. That\'s primarily why you see the peak around 150,000 years,
it\'s related to a waste package that on average, is failing at around
60,000 years, and that\'s the transport time to get neptunium to the
accessible environment, for the most part.

KADAK: The failure mode is what did you say, corrosion?

MC CARTIN: Corrosion, yes.

KADAK: Localized?

MC CARTIN: General.

KADAK: General?

MC CARTIN: Yes.

HORNBERGER: This one indicates that neptunium is transport limited and
not solubility limited?

MC CARTIN: Neptunium can oscillate between solubility limited and
release rate limited, depending on the particular selection of
parameters and the time period. Time period is affected by how much
water is infiltrating at that particular time. But, also, for our
calculations, we don\'t use a strict release rate. We have a calculation
that it goes through, but on average, our release rates from the spent
fuel are on the order of 10^-4^ to 10^-5^, depending on where you are in
that curve. When you tend to be at the higher end, and you have a lot of
water, there could be a solubility limit. When you\'re at the lower end,
and littler water, you won\'t have a solubility limit. So, it\'s not
strictly controlled by solubility universally, but it can be, depending
on the particular selection parameters.

And, we continue to look at the release rate. I would say based upon
previous TSPAs, that is a large difference between ourselves and the DOE
models, that DOE tends to have, my understanding, and I will say I could
be wrong, but my understanding, tends to have a much higher release
rate, and they take credit for cladding. We, on the other hand, for the
source term, we have a much lower release rate, but we take no credit
for cladding. And, so, that can be a significant difference, depending
on particular parameters. But, that\'s one reason you\'re seeing those
doses, and this is using the ICRP, the new dosimetry, ICRP, as
implemented in Federal Guidance Report 13, that an order of magnitude
higher would take a neptunium dose of 8 millirem, and make it 80.

There are other differences. One thing we do not, in this particular
calculation, we do not have Carbon 14 in our calculation. We, due to
isotopic dilution reasons, we do not transport Carbon 14. We are
relooking at those assumptions and analyses to further confirm that,
partly because in some of the DOE analyses, Carbon 14 is not a dominate
contributor, but is a non-trivial contributor dose. But, we don\'t have
Carbon 14. We think isotopic dilution would render it rather small. But,
we\'re reconfirming that.

This particular calculation also does not have colloids. Depending on
your assumptions for the generational colloids, and for the transport of
colloids, that could be different. We have done analyses in the past
where we felt the colloidal contribution was limited. The next version
of what we\'re working on today is looking at an appropriate way to look
at the generational colloids within the waste package, given the waste
package materials and the internals, and then also the filtration and
transport of colloids in the far field. And, that\'s something that
could change the results to date. Those don\'t result in that.

Other than that, you know, we continue to, as information comes in, be
it the solubility of neptunium, be it the retardation factors in the
unsaturated/saturated zone, we continue to update the model, and
traditionally, we have modified the parameter ranges somewhat based on
new information. I don\'t know if I would expect big differences there,
but that work still is being done.

PETROSKI: Could you clarify? What does that low curve include?

MC CARTIN: Everything. All the radionuclides in our inventory. There\'s
approximately 21 radionuclides.

DUQUETTE: You assume that all of the packages fail at the same time?

MC CARTIN: No. There is a distribution of packages, and, like I said,
generally, the waste package failure times are varying from 40,000 years
to 80,000 years. We have eight sub-areas, eight regions of the
repository where we represent corrosion of the waste package. It partly
depends on the temperature in terms of when they fail, however, within a
sub-area, one of those eight sub-areas, when we use a representative
package. So, when a representative package fails, they are all assumed
to fail.

KADAK: How sophisticated is this model relative to what DOE has
prepared? And, is this the best estimate? I mean, are you just making
gross assumptions trying to just get a feel for what\'s going on, or are
you trying to really predict peak dose?

MC CARTIN: I would like to think we\'re trying to give our best shot at
representing the performance. However, there is\--and, I think this is a
strength of the system\--there are at least two independent groups,
ourselves and DOE, EPRI also is out there doing calculations, but,
looking at ways to represent this. There are some areas where I believe
we have a more sophisticated approach. There are some areas where DOE
has a more sophisticated approach. And, it would take, I mean, in the
limited time I have, I really would be pressed to go into all the
examples. But, certainly, I\'m willing to talk to you off line somewhat.

But, it\'s a variation, and it really, I mean, probably the best example
I can give, in my own opinion, is the source strength. And, we looked at
the degradation rates of spent fuel, and cladding. We felt that there
was a better basis for the degradation of spent fuel for very long-term
behavior than credit for cladding, so we did not look a lot at credit
for cladding. We spent most of our time looking at the basis for the
release rates from the spent fuel. You know, whereas, DOE is taking at
early times, certainly a substantial amount of credit for cladding, and
has a very fast release rate from the spent fuel. And, I think that
should be looked on as a strength to the system. Different analysts look
at the system and see where they are more comfortable ascribing
performance to, and go down one path.

GARRICK: Tim, I knew this slide would create a great deal of interest,
and I applaud you for showing it. We are very grateful to you for doing
that. But, I think we\'d better go on. I will give Ron\--

LATANISION: Mr. Chairman, I do have to ask this question. My friends on
the other side of the table here just provoked me, so I will do so.

I just want to make sure I\'m clear on the point of the mode of
corrosion failure that you\'re looking at. My understanding is that if
you look at the data that\'s been generated, and if you assume that the
waste packaged mode of degradation is by uniform corrosion, by general
corrosion, and the system is operating in a passive regime, then this
system ought to have about a million years of life. So, I\'m wondering
how you then translate that into a 40,000 or 50,000 year failure, if
it\'s not something like localized corrosion.

MC CARTIN: Certainly. That part of the model is being revised currently,
and there is certainly with respect to the types of chemistries that
we\'ll be seeing, I would say that is a correct assumption, that the
container lifetimes from uniform corrosion are going longer. But, you
know, it\'s still, we\'re in the middle of that.

LATANISION: But, just to follow that up, if that were the case, then
presumably those curves are going to move around quite a lot. Am I
correct? I mean, that peak that you\'re showing\--

MC CARTIN: It could move out further in time.

LATANISION: Right.

MC CARTIN: But, it wouldn\'t be a spread, say, from 40,000 years to a
million years.

LATANISION: No, I know. I just meant that it would move from\--

MC CARTIN: Oh, absolutely.

LATANISION: It could be significant.

MC CARTIN: Yes. However, I will say we have looked, just the way of
calculations, what if I just move that out 500,000 years, and have all
the waste packages fail within, say, 500,000 to 600,000 years. The dose
curve goes down some. It doesn\'t really drop that much, because it\'s
dominated, in our calculations, dominated by neptunium that essentially
has a very long half-life, and is still pretty much there.

GARRICK: Are you saying that\--I don\'t want to prolong this
unduly\--are you saying that the impact is principally on the time of
peak dose, but you don\'t expect a major change with respect to the
magnitude of the dose?

MC CARTIN: To date, that is correct.

GARRICK: Okay.

MC CARTIN: Given all the other assumptions, as other things change, you
know, I will say I don\'t want to\--you\'re getting a snap shot. It\'s
not a final, you haven\'t gotten to the end of the movie yet. So, it\'s
just one part of it. But, yes, I mean, I guess I wasn\'t overly
surprised by that.

GARRICK: Yes. Carry on.

MC CARTIN: Yes, next slide, please.

Getting to the, as I said, the EPA, they suggested that the assessment
for climate change could be limited to the effects of increased water
flow through the repository. And, they said the nature and degree of
climate change could be represented by constant conditions after 10,000
years.

Next slide. And, in that, they said NRC should specify in regulation the
values to be used to represent that constant climate after 10,000 years.

Next. And, in looking at that, clearly, temperature and rainfall are the
most straightforward things when you think of climate change. But,
looking at the problem, really, the part of that that affects the
performance of a Yucca Mountain repository is deep percolation, how much
water gets to the repository horizon. And, so, rather than specifying a
temperature and a rainfall, we felt a more direct way was to go directly
to deep percolation, and have a representation of that.

Certainly, we recognize that that deep percolation rate, whatever it is,
is controlled by a variety of processes, such as the precipitation, the
temperature, evaporation, plant transpiration, those things have to be
considered in getting to the deep percolation, but we thought that it
would be a more direct way to represent what should be used for the
climate change, rather than, say, going to the first factors of
temperature and rainfall.

Next slide. In estimating that deep percolation, clearly, one has to
look at some range for the mean annual precipitation over that million
years. There was an understanding in the literature we looked at that
the glacial-transition/monsoon states tended to dominate the long-term
climate states. These were wetter conditions. On average, we would
expect it to be wetter during the next million years than it is today.

Next. A harder thing is what\'s arranged for the fraction of
precipitation that ends up as deep percolation. And, I\'ll talk about
those in a little bit more detail.

Next slide. In terms of precipitation, looking at analog sites based on
vegetation related to packrat middens. There was analog sites that the
estimate for the rainfall was on the order of 266 to 321 millimeters per
year, representative of the last glacial maximum. I\'ve been criticized
before for this for showing the three significant figures to that, and
how could I possibly do that. I will say that was the number that was
reported in the reference that we used. And, so, I felt we should not
change the numbers in the report, just reporting what was used, but we
do recognize the limitations in any type of calculation of this nature
to three significant figures. But, there is a rainfall rate.

Next slide. In terms of the fraction of precipitation that results in
deep percolation, there, we had to draw upon calculations. Some of the
models in our TPA code that account for and support the values we have
in the performance assessment that consider the temperature, soil depth,
evapotranspiration, and what we found for the two ranges of
precipitations, 250 to 420 millimeters per year, that approximately 5 to
20 per cent of that precipitation ended up as deep percolation. And, so,
that\'s our fraction for the amount of rainfall that ends up at the
repository horizon, giving the modeling that takes into account such
things as the evapotranspiration, which is highly dependent on the
temperature.

KADAK: So, you used 20 per cent, so that\'s fine.

MC CARTIN: Well, here\'s what we did. And, so, what we took was we
took\--we had that range for precipitation of 266 to 321, and then we
had a range of 5 per cent to 20 per cent for the fraction that would
remain, and we got a range to be used for deep percolation by just doing
a straight arithmetic multiplication. And, so, we ended up with a range
for that long-term average of 13 to 64.

Now, some might ask, well, if you have to specify an average, why don\'t
you just give a single value. And the reason we didn\'t was that we
think it\'s useful that you have a range that in the performance
assessment, then you\'re going to get an understanding, well, how
important is it when you get towards the higher end to the lower end.
And, it\'s not a huge variation, but you will be able to, in the
results, see some sensitivity to what that specification was. And, so,
we thought that was valuable information that we think would assist any
safety decision the NRC might make in the future.

In terms of the distribution for that range, we looked at it that really
deep percolation is a result of a lot of multiplicative processes. So,
that would suggest a logarithmic distribution. And, when we look at
those numbers, there\'s limited basis for favoring either end\--well, in
actuality, really, for favoring any number within that range. And, so,
we said a log uniform distribution would seem to not be unreasonable for
a range.

Next slide. With that, you end up with if you did that range, the mean
value from that range and that distribution is a value of 32 millimeters
per year. That\'s approximately six times larger than the infiltration
rate, deep percolation rate estimated by some for the site today,
although that\'s assuming the estimate is on the order of 5 millimeters
per year. There are ranges for that. So, we do end up with a value
that\'s deeper, that is more infiltration to the repository, we believe,
and in a qualitative sense that if you are assuming that a lot of the
time it\'s going to be wetter and cooler, and certainly the cooler
conditions really enhance the infiltration, that it was a reasonable
proposal to go out and seek public comment.

Last slide. And, with that, where is the process today? The EPA comment
period ended on November 21st. The NRC comment period ended on December
7th. We did get comments on the approach for estimating the deep
percolation, as well as other items. We are in the process of
considering those comments. We would expect to finalize our regulation
shortly after EPA finalizes theirs. So, we will not be finalizing\--we
can\'t finalize ours until we see what the final EPA standard is.

And, with that\--sorry about that mid-talk delay, but\--

GARRICK: It was a highlight. Okay, David, Thure, and then Bill?

DUQUETTE: Duquette, Board.

Given that NRC is sort of a junkyard dog on this whole thing, and as
you\'re the regulator and have to pass through whatever DOE gives you,
and at the present time, at least some of the numbers that I seem to
recall from the DOE data are larger than your dose numbers, do you think
the NRC as a regulating body would lean towards accepting, assuming that
the DOE standard meets the EPA standard, how do you think you\'re going
to compare your data with their data?

MC CARTIN: Well, and I probably should have qualified this, the
calculations we\'ve done for the past 20 years or so and the performance
assessment tools, and techniques, we have are done in a way to assist
our review of the DOE license application. And, the fact that we have a
particular number doesn\'t really help DOE one way or another. It\'s a
way for us to get insights. We obviously understand our code very well,
and we can do changes to it, and things of that nature.

The bottom line is we need to understand the assumptions and the bases
for DOE\'s performance assessment, and look at their defense for what
they have proposed. And, there are cases where I think we have an
approach in our PA code that helps us better understand what some of the
assumptions might mean, and allow us to ask more informed questions of
the Department. But, basically, it\'s really the Department\'s
performance assessment.

As I said, an example, in our code, we currently take no credit for
cladding. Well, we do have a parameter where we can do\--take credit for
cladding. We haven\'t invoked it. But, that\'s an example of things
we\'re aware of, some of the things the Department has in their code.
This is a way for us to better understand the problem. But, ultimately,
it is based on the Department\'s. I don\'t know if that\'s helpful.

GARRICK: Thure?

CERLING: Cerling, Board.

Two questions. I didn\'t ask any questions on Slide 11, so I get to now.

GARRICK: So, look out.

CERLING: One of the things that\'s of particular interest to me is why
the Neptunium 237 goes down instead of perhaps even continuing to rise,
or reaching a plateau, and I was just wondering why that is in your
model?

MC CARTIN: Well, part of that is, as I said, you know, we\'re on a
10^-4^, 10^-5^ release rate, so the spread of that\--the larger part of
the release, you\'re looking at around 100,000 years, so neptunium is
depleted. I mean, if you think of a release rate on the order of 10^-5^,
it\'s gone. And, there\'s some lingering\--some of the parts that
you\'re seeing there is uncertain\--

CERLING: Well, then does this mean that all of the waste packages have
failed by 80,000 years, or what?

MC CARTIN: Oh, yes. In our performance assessment\--

CERLING: So 100 per cent have failed between 40,000 and 80,000?

MC CARTIN: Yes.

CERLING: Okay. The second question that I had had to do with Slides 9
and 10, which were the difference in the dose conversion factors. And,
everything is normalized to 100 per cent of the maximum, and, so, could
you just help us along on what are the\--what happened to the absolute
values? MC CARTIN: Well, I mean, I could put them in. I don\'t know if
it would be\--I mean, where you\'re at about the 50 per cent point for
neptunium versus Plutonium 239, it\'s a factor of 2 difference, I mean,
in the absolute values.

CERLING: No, I know that. But, if you compare this to Slide 9, Neptunium
237 I think has 100 per cent value, so we don\'t really know\--if
plutonium stays the same, and suddenly is now 100 per cent, then we can
work that out. So, what has happened\--

MC CARTIN: Go to the previous slide. This is the slide that is
comparing, when I switched from FGR 11, the older dosimetry, to FGR 13.

CERLING: Right. Well, there\'s still no absolute values.

MC CARTIN: Well, neptunium dropped by a little more than a factor of 10.
So, it would be ten times less\--the absolute value would be ten times
less. The dose would be ten times less for a unit intake of neptunium.
Likewise, with iodine and technetium, it\'s approximately twice. Those
absolute values of the dose\--

CERLING: No, I see. I guess with an Excel spreadsheet, we could figure
this out, but it\'s kind of hard to put it all together with the three
slides. I guess one could put it together.

MC CARTIN: Well, the only reason for the other slides was this showed
you neptunium dropped a lot. But, if you go to the next slide, one of
the things that\'s helpful, in my mind, was that neptunium was a very
large contributor. Technetium might change some, but you can see
overall, that technetium dose co-efficient is extremely small to begin
with. So, a factor of 2 change in technetium is taking a very small
number and making it a little bit larger. It\'s still a very small
number. That was the only perspective that these slides were trying to
provide.

GARRICK: Let me go to Bill, Bill Murphy.

MURPHY: Bill Murphy, consultant.

My first question was Thure\'s first question, why did neptunium drop
between 100,000 and 200,000 years. And, I think I understand the answer.
It provokes, in my mind, another, where does it go? Have you devoted
attention to considering where this neptunium that\'s been released and
transported out of the system ultimately accumulates, and is that a
potential hazard?

MC CARTIN: In terms of if it wasn\'t withdrawn from the aquifer and it
stayed in the aquifer?

MURPHY: Or if it became concentrated at a discharge point, for example.

MC CARTIN: Oh, boy, about five or ten years ago, we did look at what if
it was never taken out, in which case generally, one might say, well,
there\'s no dose if it remains in the aquifer. But, if it continued
onward, and recognizing there could be some spring deposits, where there
could be some seepage, and at the time we looked at it, there was\--I
mean, obviously, this is going much further in time because the
potential delay mechanisms are extremely long because you\'re in the
alluvium, which tends to be fairly absorptive. But, you get to a point
where if you had spring deposits, you have more of a inhalation dose
from things that get up in the air and the dust and things, and it
doesn\'t tend to be nearly as much in our calculations than the current
assumptions, where you\'re drinking two liters of water a day of
contaminated water, in addition to irrigating crops and consuming crops,
that the inhalation dose was significantly less.

MURPHY: Thank you. I have one other question on a different subject. You
said that the nature and degree of climate change can be reasonably
represented by constant conditions after 10,000 years. And, I wonder
about the affects of transients in flow systems on dose effects. For
example, in environmental systems, the first flush of the season is the
big effect. And, in fact, a variable climatic regime where fuel oxidizes
and material become labile for 10,000 or 50,000 years, and then there\'s
a flush, could lead to substantially, hypothetically substantially
greater peak doses than a continuous process even at a higher flow rate.

MC CARTIN: We have gotten comment in terms of the issue of, gee, maybe a
variable climate still retaining that average would be worse, if you
will, quote, unquote, worse than say a constant. And, it\'s something
we\'re looking at.

There are a number of assumptions that I think if you\'re going to do
that performance calculation, that you need to think about with respect
to how the flushing occurs. There\'s a part of that that the timing is
probabilistic of when those, say, a large climate change might occur is
not known, so that would be a variable. There\'s other things. It
certainly depends on the timing and extent of waste package degradation
in terms of when did they fail relative to that time. And, so, there\'s
a lot of subtleties to the calculation that we are looking at. There are
other things. I know, once again, my understanding of the DOE analysis
that I will say, I won\'t say I know exactly this is what happened, but
I know when they did the water table rise for climate change, that was
done instantaneously. And, if you raise the water table up
instantaneously 100 meters, and now you have all the radionuclides that
are in 100 meters of the unsaturated zone are now instantly available
for transport in the saturated zone, that certainly will create a large
pulse.

The question one has to step back and say, well, this instantaneous rise
of the water table 100 meters, is that actually the way it\'s going to
occur. And, so, there\'s a lot of things I would say it\'s a\--you\'re
right, there are subtleties to how climate changes, but one needs to
think through the problem. We\'re in the process of doing that, because
we certainly got comments to that effect in this approach.

GARRICK: Okay, Ali, Leon, Howard, and then David Diodato.

MOSLEH: Mosleh, Board.

It\'s back to your peak slide. I understand these are illustrative, but
are they illustrative mean values, or something else?

MC CARTIN: Yes, this was a mean value calculation. It was a
probabilistic calculation, and I\'m representing the mean dose curve.

MOSLEH: And, how do you think the median curves would fall?

MC CARTIN: I debated, because I knew someone was going to ask that
question about whether to add other percentiles, including the median.
Given the nature of the calculation and our continued revisions to the
code, I felt comfortable in presenting the mean curve. The median, in
general, will be less, but how much less, I don\'t know. But, in
general, it\'s less.

GARRICK: Leon?

REITER: Leon Reiter, Board.

Tim, is it correct to assume that disruptive events are not included in
this?

MC CARTIN: Yes. I should have said that. Yes, this does not include
igneous activity.

REITER: Could you give us\--or seismic activity, or are seismic
activities included?

MC CARTIN: There was some inclusion of seismic activity, but it did not
have a significant effect in these analyses. That is an area of our
model that we are looking at.

REITER: Could you give us any feeling as to if you included igneous
activity, and other parts of seismic activity, how might this curve be
affected? Which parts might be affected?

MC CARTIN: Well, a failed waste package is a failed waste package. So,
the fact that we don\'t, if we were taking credit for cladding, and
seismicity was affecting the cladding, it would have potentially a
larger effect. In our case, we are not taking credit for cladding. In
general, I would not expect a dramatic difference, but I will say, you
know, all of this please take with a grain of salt, that the way to
project things forward in time is something that with respect to
seismicity, and these other processes, we\'re looking at. But, you know,
like I said, all the waste containers were failed at around 60,000
years.

REITER: You don\'t expect igneous to have much effect?

MC CARTIN: Pardon?

REITER: What about igneous? You don\'t expect igneous\--

MC CARTIN: I personally would be somewhat surprised if igneous was
changed dramatically, for the simple reason the one aspect of the
repository system when represented, igneous activity occurring early on,
early on I will define in the first 1,000 years, has the effect of
releasing short-lived radionuclides that can\'t make it any other way
except this release. And, as you go out in time, our 1,000 year doses
are dominated by the short-lived radionuclides. At 10,000 years,
they\'re not there anymore. And, so, the things that were causing the
larger doses were the short-lived radionuclides that dot out to an air
pathway.

So, assuming, once again, we continue to revise and understand, but
given, if you look at that inventory slide, you can see Americium 241
and Plutonium, one of the plutoniums, maybe 239, that were very large
sources of the inventory at 1,000 years, at 10,000 years, they\'re
essentially gone.

REITER: But, if you assume an intrusive dose, you might have releases
later on.

MC CARTIN: Oh, I was talking about extrusive. Certainly you could. Once
again, I think this curve is representing all waste packages fail, with
the probability of one, essentially. The igneous activity, it might fail
some waste packages at some later time, but it has a lower probability.
In the groundwater pathway, we would still expect to see neptunium
dominate the dose curve. But, once again, those questions need to be
evaluated, and I\'m giving you my, at least, sense of the calculation.

GARRICK: One of the reasons you don\'t think it would have much of an
effect is things like the probability of an intrusive event or an
extrusive event is very small?

MC CARTIN: Yes.

GARRICK: Okay. Let\'s hear from Howard, and then David Diodato.

ARNOLD: Arnold, Board.

I\'m interested in the cladding. Is the fact that, you\'re assuming no
credit, is that based on good analysis, or are you working to see
whether it\'s due some credit?

MC CARTIN: Well, I mean, I\'m struggling with the word \"analysis.\" I
would say there isn\'t a quantitative calculation that we\'ve done, but
it was a recognition that if you had some slight perforations, or pits
in the cladding that allowed oxidation, unzipping of the cladding
relatively quickly, the cladding is extremely thin, that over thousands
of years, some seismic events, other things that we felt that we would
prefer to, as Dr. Garrick would say, we turned our microscope up on the
release rate from the spent fuel, and we recognized that possibly you
can come up with some basis for that, we felt, for our effort, we want
to focus more on the looking at issues with respect to the degradation
rate of the spent fuel. That\'s not to say that, you know, the
Department will have a basis. They will review it, but like I said, we
have not included it in our calculation.

As with every performance assessment, and I would say\--one thing that I
would like to say, I don\'t think anyone tries to be overly conservative
in the performance assessment. You\'re always trying to do what you
think is a reasonable estimate of performance. But, you generally have
limited resources, and, so, you look at areas where I think I\'ll
improve that part, and this part, I\'ll let go. And, in our case, we
felt that the cladding credit was not as important in our mind to the
degradation of spent fuel. But, you have many of those questions of
where do you devote your resources in developing a model.

GARRICK: Okay, David?

DIODATO: This is Dave Diodato, Staff. I\'ll be brief.

Tim, I appreciate you doing this presentation. On Slide 16, I\'m going
to focus just on the precipitation question, with the assumption of
long-term average climate specification that EPA gave you the
opportunity to decide to choose to put in the regulation.

This 266 to 321 millimeter per year number, when you gave this
presentation at the public meeting of the Advisory Committee on Nuclear
Waste at the NRC, some questions came up about that because that was
based on the USGS open file report, where the analyst found the
regression coefficient in the analysis was only about .75, and that was
based on analog sites and other information. So, there were questions
raised about that in that meeting, and I just wonder since that time,
you have some confidence in these numbers so far, but are there any new
analyses to date that have increased your confidence in this range of
precipitation?

MC CARTIN: Well, we got comments on our specification, and we are in the
process of looking at it from a variety of different approaches. With
that particular range, we don\'t have anything right today. I mean, I
will say part of it that I was somewhat, you know, currently, the
precipitation at Yucca Mountain is on the order of 125 millimeters per
year. If you double it, you get 250. You\'re sort of at the low end of
that range, and, you know, doubling of rainfall, we are a little higher.
I know the ACNW expressed an interest that it\'s too high. We are
looking at that.

The only thing that I\'ve done, as much out of curiosity to see what it
was, I took in our performance assessment, we use 100,000 year
maloncovich cycle to represent climate change over a million years,
where we did a gradual up and down. Using average conditions there for
that cycle and rainfall, I just did a time average over, you know,
accounting for the wet and dry periods, just over time, and got a value,
and I know people will think I cooked the books on this one, but it
ended up being 33 millimeters per year, on average, accounting for the
cycle. And, our average came out to 32, and that\'s maybe pure chance.

I also went to, I believe it was the DOE site recommendation where they
had an alternative model for climate change, where I think they had a
400,000 year cycle, with these spikes. And, I did the same thing, did a
time average of that representation of a million year period, and ended
up with 26 millimeters per year. And, actually, I was stunned at how
close those were. We continue to look at additional items, different
ways to represent climate, and, you know, to date, that\'s pretty much
what we have. There will be a report in the not too distant future that
the Center could be publishing, looking at some other alternatives.

DIODATO: Well, I just am a little confused because, if I heard you
right, you\'re arguing for numbers that are an order of magnitude lower
than these numbers.

MC CARTIN: I\'m sorry. The 26 and 33 were for deep percolation. That is
precipitation.

DIODATO: Okay.

MC CARTIN: Using that range, our ranges, on average, when you translate
that on the next slide to\--but when you translate that to the deep
percolation, you end up with a range of 13 to 64. That is, on average,
32. So, that precipitation amount gets you a\--yeah, that would be
extremely large if you had 266 millimeters of deep percolation. But,
that\'s the rainfall.

DIODATO: Yes, and that\'s what we\'re looking for, is the technical
basis on the rainfall, because 190 millimeters per year at present;
correct?

MC CARTIN: Estimate is around 125. But, you know, no one has a\--it\'s
going to be variable. I mean, that\'s why we were more comfortable with
a range. You can\'t really, a single value is pretty hard to defend.

DIODATO: But, still, no new numbers that would increase your confidence
in that 266 to 321?

MC CARTIN: Not yet. And, we\'re just in\--the comment period hasn\'t
been closed that long. You know, we\'ll have more to say when we
finalize the regulation.

DIODATO: Thank you. I appreciate that.

GARRICK: I wanted to cut it off at this point for our break, but I have
a colleague here that has a pressing question. And, if he can keep it
brief, we will handle it.

KADAK: Kadak. It appears that you\'ve taken much more credit for the
natural barrier in terms of time to get to the environment than DOE has.
And, again, this gets back to my question about modeling sophistication.
What is it that you\'re doing that\'s different than what DOE is doing?

MC CARTIN: In terms of the natural system, very briefly, and I\'ll break
it into two components, the unsaturated zone and the saturated zone. In
terms of the unsaturated zone, we probably take less credit than the
Department in that we have on the order of 50 per cent of the repository
footprint lies above Calico Hills vitric unit, where there\'s porous
flow sorption. The other 50 per cent is fracture flow. And, so, there is
essentially very little\--well, in our model, there is no retardation,
so for half of the footprint, we have fairly rapid transport. The
Department has slower rates of transport in the unsaturated zone than we
do, based on previous models. I do not know what current models is.
It\'s primarily based on the FEIS model, final environmental impact
statement model of the Department.

They have traditionally shown more there, more matrix diffusion, et
cetera. We take no credit for matrix diffusion in the unsaturated zone.

In the saturated zone, I believe the primary difference is we have a
longer saturated zone flow path in alluvium, where most of the
retardation is. With respect to retardation factors, we\'re really, I
don\'t think we\'re that different, but I don\'t believe there\'s a
dramatic difference, but there could be some small differences, but I
think it\'s the distance in the alluvium. That would be the primary ones
that I\'m aware of.

GARRICK: Okay, thank you very much, Tim.

We\'ll now take a 15 minute break.

(Whereupon, a brief recess was taken.)

GARRICK: Jens Birkholzer, go ahead.

BIRKHOLZER: Okay, thanks for the invitation to speak here. I feel
honored, and I hope I can give you the information that you\'re seeking.

My part of the talk, it\'s a split talk, it\'s on the mass of what is
seeping into drifts. The second part is Dr. Hardin\'s topic, and he will
talk about seepage and water coming out of drifts.

Next slide, please. This is a brief outline of my talk. It will first
present to you the technical basis and key processes affecting seepage
into drifts. We\'ll give a brief introduction on how we propose that
TSPA should calculate seepage in the performance assessment. We\'ll then
go into detail on how we predict seepage, both at ambient conditions,
long-term conditions, and the effects that the thermal period will have
on seepage, give you a basis for technical assessments, some
assumptions, some uncertainties. And, finally, discuss some seepage
calculation results. Those are results of the probabilistic analysis for
the 10,000 year or 20,000 year time frame, if you want, an analysis that
was done in Berkeley, which is sort of similar to what TSPA is expected
to do, but not identical, there are some simplifications.

Next slide. First, a few definitions. What we define as seepage is the
dripping of liquid water, so we\'re not talking vapor, we\'re talking
liquid water from the formation into an underground opening, because of
the forces that amount of water is much less than what is percolating
down to the tunnel drifts.

When we talk about the seepage rate, we talk about the mass of water
seeping per time, and we give that rate for a drift section that
contains one waste package. So, you kind of know what the seepage rate
is that\'s going to be per waste package.

Seepage percentage is the ratio of seepage rate divided by the flux
getting to a tunnel drift. In other words if you have, say, a seepage
percentage of 5, that would mean that 95 per cent of the water that gets
down to a drift will actually be diverted around it and not seep.

Finally, the seepage fraction is the relative number of waste packages
that could be affected by seepage.

Next slide. This slide gives you a brief introduction to the processes
and factors that we believe are important for ambient seepage. First of
all, starting on the right side here, we really have a small scale, or
drift scale problem to solve when we look at seepage. The main
parameters, sort of in theory, are the amount of water getting to a
drift, the shape of such an opening, the capillarity of the formation,
and also the permeability that would allow water to divert sideways.

Now, in practice, there are more issues that we need to look at, flows
and the fraction where it could be episodic or channelized. We have some
excavation disturbance here, so properties change. We may have some film
flow along the wall. We may have, you know, certain surface roughness to
look at. Also, there could be drift degradation with time or with
seismicity, all the way up to a full collapse of drift. We\'ll talk
about that later.

Now, in terms of the flow, it really is dependent on the overall flow
patterns in the unsaturated zone at Yucca Mountain, sort of shown here,
depending on infiltration patterns, on the flows and the different
geological horizons. One important one for seepage purposes is the
so-called Paintbrush unit, which is a porous non-fractured unit that
tends to dampen seasonal changes. So, below that, we kind of assume we
have some sort of a steady state flow behavior that would change with
sort of climate changes over long time scales, but not rapidly.

Next slide. Now, if heat comes into play, there are more issues to look
at. The most important in terms of seepages, that for a few hundred to a
thousand years, maybe, you will have above boiling temperatures right
next to your drifts. They will dry out and they will tend to vaporize
any water that tries to get next to the\--towards the drift, and, so,
really seepage is not even an issue in that time frame.

On the other hand, there are changes in properties, maybe because of
mechanical stresses that would affect them, or geochemical changes.

Next slide. All right, so we have a bunch of complex processes to look
at, and they occur on different scales. The project in the past five,
ten years has done a very elaborate experimental and modeling analysis
in order to understand and predict seepage. And, the results of that
analysis must obviously be moved and propagated into RSPA as reasonable
as possible, accounting for spatial variability and uncertainty.
Simplifications have to be made, on the other hand, because it is a
performance assessment, and you can\'t translate process model results
one on one usually.

The so-called seepage abstraction model does those simplifications and
integrates the inputs from various sources, defines the methodology for
TSPA to handle seepage, and also provides parameter distributions and
look-up tables, which are being used.

Next slide. This is really what will be done in TSPA. There\'s a two
step approach. The first one is a probabilistic analysis of ambient
seepage. TSPA loops over time, over locations in repository, and over
realizations of uncertainty, Monte Carlo type uncertainty, sampling. It
uses look-up tables as a function of some key parameters. I\'ll talk
about those in a bit. Then, it calculates ambient seepage rate and also
the seepage fraction over the entire repository, and parts of the
repository.

The second step is a simplified bounding treatment of the coupled
thermoprocesses. For example, accounting for the vaporization barrier
for early time periods, and also, if needed, accounting for changes in
properties as a result of mechanical geochemical effects. Currently,
there is no adjustment made because the effects in the past have been
shown to be small.

So, in short, seepage is a function of location, and will be a function
of location in TSPA, and is also a function of time.

Next slide. I\'m talking a bit about the technical basis, starting with
ambient seepage. Ambient seepage has been extensively tested by about
100 liquid release tests. The water was released above drifts at Yucca
Mountain, and then water is captured and the amount of seepage was
measured.

In parallel, a pretty sophisticated heterogeneous fracture model was
developed on that drift scale in order to understand those tests, in
order to calibrate parameters, in order to provide a conceptual model to
come up with seepage predictions.

Next slide. Overall, all the tests have demonstrated that there is a
significant capillary barrier behavior and significant flow diversion,
so seepage is always much less than the water that was injected. the
model was shown to be able to accurately capture the data, and it was
also shown that seepage can be described as a function of three key
parameters. One is the local permeability, and actually it\'s
variability, it small scale variability as measured in air permeability
testing. One is an effective capillary strength that includes both the
physical capillarity, as well as some small-scale effects that cannot be
explicitly handled by the model. And, the third one is the percolation
flux.

Next slide. Now, the same model was taken, the same validated calibrated
model was taken and was used to do a systematic prediction of seepage
over those three parameters, and essentially varying permeability,
percolation flux, and capillarity over the ranges that we expect to be
the ones to be used over the repository locations we\'re looking at.
And, results of that analysis was a look-up table for TSPA. The blue
arrow shows you where seepage would tend to increase in that look-up
table, say permeability goes down, capillarity goes down, percolation
flux goes up. You have pretty high seepage rates. So, depending on the
properties you have, you feed into that look-up table, you get seepage
results. We have separate seepage look-up tables for intact and
moderately degraded drifts, as well as fully collapsed ones.

Next slide. Why is that? Drift degradation analysis has shown that
drifts in the lower lithophysal unit, it\'s sort of a softer rock unit,
may collapse in extreme seismic cases, and what that collapse does is
that essentially the diameter of drifts would double. They would be
filled with fragmented rock material. There would still be a capillary
barrier up here because that material has a very low capillarity, but
the seepage overall would be larger because (a) more water has to be
diverted sideways, and secondly, because it\'s harder to divert if your
opening is large.

So, for the same parameters, you would expect more seepage into a
collapsed drift compared to an intact drift. And, that\'s an important
thing to understand.

Next slide. Now, I said that we have look-up tables are developed for
TSPA to work with. They can obviously not work with those tables if they
do not have the input parameters described to feed into those tables.
And, we have provided those parameter distributions accounting for both
spatial variability within the repository, and also uncertainties that
we see. I will briefly walk you through those.

Next slide. Permeability effects. What we have to do is we have to
describe the variation of mean permeability on the drift scale at
different locations in the repository. And, we do that using all the
data we have at hand from various air permeability measurements. Some of
those have been done on the scale in the niches where the liquid release
was conducted. And, some are done with larger packer lengths, or some
scaling analysis has to be done.

Next slide. Anyway, we use all that information. We perform scaling
analysis to make results comparable. We also adjust for measurements
that have been conducted in intact rock to include the impact of drift
excavation on properties. We distinguish between geological units, and
then we develop the spatial variability distributions.

Now, we consider those to be uncertain because our sample size is not
all that large. And, so, on top of using those distributions, we develop
also uncertainty distributions. And, I show those in the next slide,
please.

This log-normal distribution is a spatial variability, using best data
for the lower lith unit, about a 10^-12^ mean permeability. And, then, a
triangular distribution was developed that shifts that variability to
upper and lower values, essentially to account for the uncertainty in
that distribution to be representative for the entire repository.

Next slide. Some of the same procedures done for the effective
capillarity, that\'s the parameter that was calibrated in liquid release
tests. We have ten locations, ten calibration values, and then we
developed appropriate distributions again following that same
methodology.

Next slide. A little bit different is the understanding of the local
percolation flux. Here, we use actual modeling results of the three
dimensional model scale, unsaturated zone flow model. That is a model
that evaluates the steady-state flow fields for the current three
climate states expected to occur within the 10,000 year compliance
period, the present day, from zero to 600 years, monsoon climate, from
600 to 2000, and a glacial transition climate afterwards, basically
getting wetter, as Tim alluded to with time because of those changes.

And, also, there are three alternative infiltration scenarios to include
some of the uncertainties that are related to those infiltration
predictions. Anyway, those distributions of percolation flux at the
repository are being used and fed into those look-up tables.

Next slide. There\'s one issue with those predictions. The model has a
grid resolution of about 100 to 200 meter. That resolution is not fine
enough to account for heterogeneity below that scale, and we need to
look at the heterogeneity, because our drift scale is five meters, or
so. So, we have done some sub-grid modeling to understand what the
sub-grid heterogeneity could be, and we actually developed a we\'ll call
it a flow focusing regression curve that is used to broaden the
distribution of fluxes from the site scale model.

Next slide. Okay, moving on to the impact that coupled processes may
have on seepage. A so-called thermal seepage model was developed that
solves for the evolution of seepage over time for the thermal period of
the repository. It\'s based on both the methodology of ambient seepage
and a methodology developed for understanding the Yucca Mountain drift
scale test, the heater test. That model is actually complemented by
pretty similar geomechanical/geochemical models, which I\'m not talking
about today, to understand what the additional impact of property
changes from geomechanics and geochemistry might be.

Next slide. Typical model results are, and those are basically supported
by all the heater testing validation we\'ve done, and other confirmation
studies, that the barrier for flow to reach a drift during the
above-boiling phase is essentially a perfect one. There is no water that
could get down that could seep during a period that temperatures are
above 100 degrees.

Later, there is a period of resaturation, and that would lead to some
delayed seepage initiation if the properties would allow for seepage.
Always, that thermal seepage would be less than or approaching ambient
seepage would not be above that. So, there isn\'t really a puddle of
water that would all of a sudden flow down and would create a seepage
that is much larger than any ambient seepage would be. Those results
were fairly consistent over a wide range of seepage relevant parameters
and conditions.

What you see here is, in blue, the seepage curve of time for an example
case where we have fairly high percolation fluxes, so seepage is likely.
And, you see in red the comparison of ambient seepage, if heat was not
included. Three different climate stages, and you see that you have a
delay in seepage, and then at some point, you get pretty close to your
steady state ambient results. And, at some point later, you would
probably approach them exactly.

Next slide. So, we use that to come up with a bounding case abstraction
of thermal seepage. Realizing that all the complexity of those time
dependent results cannot be implemented in TSPA, we essentially said
that seepage exceeds zero as long as temperatures are above boiling, or
clearly above boiling. And, as soon as that ends, we go up to the
respective ambient seepage results, and follow that curve. So, there\'s
some conservatism in that abstraction method.

Advantages. We can use the same look-up tables that we use for ambient
seepage, and we can use results of another model scale, thermal model,
that provides estimates for the drift wall temperature. And, that is
done location by location, so if you have a hot waste package and
you\'re in the center of the repository, that time scale would be larger
compared to being at the edge where you might not even have boiling
conditions. In that sense, then you wouldn\'t have that zero seepage
period.

Next slide. This is a brief summary of the abstraction we\'re proposing.
As I mentioned, ambient seepage is calculated from look-up tables. There
are different look-up tables for collapsed and intact drifts, different
parameter distributions for the units that the repository is residing
in. There is no seepage for wall temperatures above 100 degrees C.
Below, seepage is always like ambient predictions. There are no seepage
changes currently due to geomechanical or geochemical processes.

What I didn\'t talk about is that we do not assume a flow diversion for
the case of volcanic intrusion, simply because we are too uncertain
about the future properties of the lava that would fill a drift. Is it
going to be cooling joints? We just don\'t know. No seepage during
preclosure because of ventilation, and no seepage increased from rock
bolts. That stems from a study that I\'m not talking about today.

Next slide. I\'m giving you some results. Again, it\'s a probabilistic
calculation, similar to TSPA, not identical. Essentially, we take the
parameter distributions, move them into the look-up tables and get
results.

I should point out that a similar probabilistic evaluation was done to
understand the South Ramp seepage that was observed last year in
response to those strong rainfall events. And, there was pretty good
qualitative agreement.

Next slide. Okay, this is a result that shows seepage percentage over
time. Those are the different changes in climates, and a blue curve is
your mean climate scenario. It shows that we have about a 2 per cent
seepage of all samples that we looked at at present day up to 600 years,
about 5 per cent later, then 7.5 per cent, I think, for the glacial
transition climate. Meaning, that here, for example, 98 per cent of all
water would divert.

In terms of the number of waste packages affected by seepage for that
same present day climate, there would be 8 per cent that might see
seepage, 18 per cent monsoon, and then I think 21 or 22 for the latest
climate, that glacial transition.

Next slide. Now, if we look at the mean seepage rate of all non-zero
seepage samples, meaning, wherever we have seepage, we take a mean, it
would have about 40 kilogram per year per waste package, at present,
100, and then 150. Now, translating that, 100 kilogram per year per
waste package is about one drop of water every 50 seconds, if you assume
that there\'s only one drip location of a waste package. This is not a
lot of water.

What I didn\'t include in that graph is that there is a zero seepage
period because of the vaporization barrier. You see here, the predicted
timing of the boiling period for different locations in the repository
on average may be 800,000 years. So, you could actually cut that part
off and put it to zero on average for the first several hundred to
thousand years, in most cases. It depends on the location, though.
There\'s quite a variability of seepage rates, depending on the
variability of your input parameters.

And, we can go to the next slide. This shows you where that variability
comes from. These are your input parameters sampled over 10,000 samples.
These are all samples for permeability, capillary strength, percolation
flux, and the reddish ones here are those that have given us seepage.
So, you can see that the larger fluxes would tend to create seepage.
Smaller capillarities would tend to create seepage. Smaller
permeabilities would tend to create seepage. There\'s not a one on one
relation, though, even small fluxes sometimes will result in seepage if
you have low capillarity and low permeability.

Next slide. Just a few sensitivities. All results so far were presented
for intact drifts. These are the mean values again. That would be the
result for a collapsed drift. You see that seepage increases quite a bit
just because of the size of the drift and the less effect of diversion.
This is for different units. So far, I presented the lower lithophysal
unit. Up here is the middle non-lithophysal unit, which covers about 10
per cent, 12 per cent of the repository, versus 80 per cent here.
Seepage increases mostly because the permeabilities of those units are a
little lower. So, it\'s harder to divert flow sideways.

Next slide. This is just giving you a flavor of what happens if you do
not account for spatial variability of permeability and capillarity over
the repository. Seepage would decrease by a factor of 2, roughly. If you
wouldn\'t account for uncertainty in those properties, you might see a
decrease in 20 or 30 per cent, showing us that it is important to
include those in our assessment.

Next slide. This is my last slide here. Conclusion. I think, many of us
think that the seepage predictions in TSPA are soundly based on science,
and fairly realistic, rather than extreme in terms of extreme
conservatism. So, there are conservatisms, obviously. Ambient seepage in
fact is sampled without any simplification to the process models we
have. So, this propagates right into TSPA.

Thermal seepage is handled with a bounding case treatment. I think that
we have adequately addressed spatial variability and uncertainty.
Seepage varies in time and space, as you have seen. And, finally, the
flow diversion that we see is important in reducing seepage, or even
preventing it, in most cases.

Thank you.

GARRICK: Thank you. I\'m going to ask the hydrologist member of the
Board to lead this discussion. So, George Hornberger, why don\'t you do
that.

HORNBERGER: Thanks. I\'ll go first. Okay?

So, you have this fairly thorough discussion that you just gave us in
terms of how you have calculated the seepage flux in the past. Is
anything changing as you move this model forward?

BIRKHOLZER: I don\'t think the plans are set for changes that we would
do in the future. I guess you\'re probably asking in terms of the
changes in our compliance period? I don\'t think there\'s anything
definite yet. I guess from what I heard from Tim, the future climate
conditions wouldn\'t be drastically different from what we already
assume currently for the period after 2,000 years, what we call the
glacial transition period, which has an average of I think 20 to 25

per year. And, if you assume that that is stable, you wouldn\'t have to
change that. I don\'t think there is a great need of revisiting a lot of
the process models. I think you might want to try to work on some of the
conservatisms that are in that model, if necessary.

HORNBERGER: Okay. Just one other thing. I know the presentations are
broken up appropriately, seepage into the drift, seepage out of the
drift. And, one of the things, of course, we\'re interested in trying to
figure out is you have water coming down, deep percolation, Tim called
it, and let\'s say under current conditions 98 per cent of it is
diverted. In TSPA, that 98 per cent no longer has any impact on the
calculation, is that right, as it goes forward? Nobody cares about that
98 per cent because it\'s not going to carry any radionuclides?

BIRKHOLZER: Well, the 98 per cent doesn\'t pick up any radionuclides.
That\'s true. Only the amount of water that seeps picks up
radionuclides. But, the 98 per cent or, in fact, the 100 per cent, does
account for the radionuclide, or the flow in the unsaturated zone below
the repository, because that is obviously similarly analyzed with the
model that includes the entire flux and doesn\'t take any 98 per cent
out of it. But, locally, in terms of picking up radionuclides and then
moving those into the UZ flow and transport, that\'s only the 2 per
cent.

HORNBERGER: Okay. And, do we know that the 2 per cent that comes in is
the same amount that\'s going out?

BIRKHOLZER: That\'s a question for Ernie. But, I think it is.

GARRICK: David?

DUQUETTE: Duquette, Board.

Could you go to Slide 11, please? In the right-hand picture, the cartoon
that\'s on the right-hand side, you\'ve shown a drip shield being
intact. I believe anyway. Does anything change if the drip shield
collapses because of a collapse of the drift?

BIRKHOLZER: That again is a question that others could probably answer
better than I do. Obviously, if your drip shield is gone, it\'s easier
for\--I mean, water gets towards your waste packages, and it wouldn\'t
in the other case. But, how it\'s handled in TSPA, I might relay that
question to Bob, maybe. Oh, okay.

DUQUETTE: Okay, if it\'s going to be answered this afternoon, that\'s
fine.

BIRKHOLZER: Yes.

GARRICK: Bill?

MURPHY: Bill Murphy, consultant.

You stated that your model accurately captures or predicts seepage data
for all the test sites, and I was pleased to see that you looked at the
South Ramp occurrence of seepage and you see qualitative agreement
there. I\'m curious what you mean by qualitative. What\'s good and
what\'s not so good about your model? And, I\'m also curious about
whether or not you have considered looking at other analog systems
perhaps? For example, I don\'t know if the tunnels are open at Ranier
Mesa, but there are many places there where there is seepage which might
represent a different percolation.

BIRKHOLZER: Yes. Let\'s talk about the South Ramp study first. I\'m
saying qualitative (a) because our analysis is still ongoing, and (b)
because it\'s very hard from the\--what we\'re seeing in the South Ramp
to come up with quantitative numbers of how much water would have
seeped. What we kind of know is how many locations or what spots there
were, and you could make an argument that maybe 10 per cent of those 300
meters that were affected would have shown at least a bit of wetness.
So, that we can compare. We can\'t really compare quantity.

Now, also, we haven\'t really assumed to model the exact conditions to
do a process model for that. But, we took that abstraction model,
adjusted a little bit the look-up tables, because there\'s a larger
tunnel diameter, and we also had to adjust the amount of water getting
there. You have to know that the South Ramp is located above that
Paintbrush unit that I talked about, that sort of porous median,
non-fractured unit. So, there isn\'t really a dampening mechanism, and
that rainfall might have just gone down there pretty hard. And, putting
that into this probabilistic analysis, we came up, I think, with a
seepage fraction of 20 to 30 per cent, which is a little bit higher than
the 10 or 13 per cent. So, that\'s what I mean with qualitative. We\'re
still looking at maybe we can improve it.

In terms of analogs, we are looking at analogs. We have looked at
analogs. We have looked at analogs in caves and in Spain, for example,
there\'s caves that have a similar geology to Yucca Mountain, yet even
higher deep percolation, and they seem to show seepage of 1 or 2 per
cent, or so. So, it\'s really not an uncommon phenomenon. Also, maybe
the ECRB cross-drift might be a sort of analog if you want to, because
it doesn\'t have the impact of ventilation. So, if you were to see
seepage, you probably would see it. We haven\'t so far. There\'s water
in there, but that\'s mostly due to other reasons. So, yes, we have
looked at that, and I guess we continue looking at that.

MURPHY: Thank you.

PETROSKI: Petroski, Board.

Slide 20. The thermal seepage curve, the slope changes suddenly at 2,000
years. Maybe you said what happens then, but what prompts that?

BIRKHOLZER: I probably didn\'t point that out too clearly. What we do
have is a change, an abrupt, the assumption of an abrupt change in
climate at 600 years and at 2,000 years, with an increase in the
infiltration rate. And, that increase causes essentially more water to
arrive at the drift. At the same time, it kind of cools the system. So,
the increase in seepage is really related. It\'s a pretty fast response
to that change in infiltration pattern.

PETROSKI: So, then, what happens at 2,100 years? Why is there another
change in slope?

BIRKHOLZER: I guess that\'s a matter of plotting it. We plotted a few
time steps here. So, the drastic change in response to the change in
percolation is ended here, and now you have this sort of slow
approaching of the thermal field, with decreasing temperatures and less
thermal impact, up to the eventual sort of back to ambient temperature
seepage. And, that is then identical to the steady state seepage
calculation that doesn\'t even include these.

PETROSKI: So, you\'re saying your time step changes, or do you use the
time step throughout?

BIRKHOLZER: In that simulation, it changes in response to the numerical
necessities for that model.

PETROSKI: But, there is no corner there. That\'s the question.

BIRKHOLZER: Here?

PETROSKI: Yes.

BIRKHOLZER: Well, if you probably zoom into that, you have a curve
there. But, really, what it means is that if you change your climate and
you change your percolation, that pretty much fast changes your entire
UZ field in a matter of a few years, maybe 10 or 20 years, and once that
change is done, you\'re back to your sort of slow change in decreases in
temperatures, slow getting back to the ambient situation, and that\'s
shown here.

PETROSKI: So, you\'re changing your time step in response to numerical
needs? Is that what\'s happening?

BIRKHOLZER: Yes.

PETROSKI: Well, then, I would have a related question with regard to the
grids, such as for the 2D model on Page 19. It\'s shown up on the upper
right there, too. Why is the grid different to the left of about 20
meters, 19 meters?

BIRKHOLZER: Here?

PETROSKI: Yes, why are you changing the grid the way you do there, and
how is the grid decided upon?

BIRKHOLZER: Well, there are a few criteria. A, I should have probably
explained it better. This is a half drift model, and it\'s essentially
orthogonal to your tunnels, and you have tunnels at Yucca Mountain every
81 meters or so in parallel. So, what we\'ve done is we actually\--there
would be another half assumption, and then this is the center axis
between two adjacent tunnels. So, if you copy those, you would have a
sequence of tunnels.

PETROSKI: I understand that.

BIRKHOLZER: No, if we zoom in very close, we know that our processes are
mostly\--the processes that we are interested in mostly occur right next
to where our heat is. Our boiling zone, our vaporization or
condensation, our flux processes. So, we are very fine here, and then we
kind of get closer as we move out. Yet, on the other hand, we want to go
all the way up to the ground surface and all the way down to the water
table. So, we focus on what we feel is important, and we were less
refined where it\'s not important.

PETROSKI: So, in a way, you\'re anticipating the nature of the result.
That\'s what determines the nature of the grid?

BIRKHOLZER: Yes, to some extent. And, that\'s kind of what you do.

PETROSKI: Yes, but how do you know then in your results that there
aren\'t any artifacts on the grid? In other words, have you done
sensitivity studies for the grid?

BIRKHOLZER: Yes, you want to do that all the time.

GARRICK: Andy?

KADAK: Kadak, Board.

Now, at our last Board meeting, I think we talked quite a bit about the
81 meter basis. Now, is it your work that justifies that number and the
water flow around this drift? Or, whose work is it?

BIRKHOLZER: Well, I think it supports it. I don\'t think that our work
was\--if you want to really design your spacing, you want to do lots of
runs, and you probably don\'t want to look at all the complexities in
terms of seepage and heterogeneities. That\'s what we focused on. But,
it does support the results in terms of that water can easily drain in
between the waste packages.

KADAK: And, the 81 basis again was you needed that separation to get the
water to flow, or could be 79.5 meters?

BIRKHOLZER: Well, I couldn\'t tell that on the basis of this model
because we didn\'t really change the grid spacing. It was other studies
that would have done that. But, what we have seen here is this is the
centerline, we have only about 5 to 10 meters of boiling at peak. But,
if you march on with time, you are sort of 90 degrees, 80 degrees, 70
degrees, moves out, and actually I think your peak temperature here goes
to 70, 75 degrees currently. There\'s probably a margin that you could
work with, but I doubt it\'s significant. You might be able to go to 70,
others might have more insight in that.

KADAK: And, relative to your modeling versus what we just heard from the
NRC, what does your model show that\'s different than their model?

BIRKHOLZER: I am really not in a position to answer that, because I
don\'t know their model very well. Maybe Tim can comment on that.

MC CARTIN: I think this is probably a good example of one of the
questions asked to me earlier, the difference in sophistication. We have
a much, much simpler model than the Department in terms of the thermal
effects, and we have really almost a bucket type of approach in looking
at temperatures and for what happens during the thermal phase. So, we
have done some what we call off-line analyses with sophisticated 3D
thermal hydrologic models to get some sense, but we have a very
abstracted model, very simple almost close to a table look-up for the
thermal aspect.

GARRICK: Okay, Ali?

MOSLEH: Mosleh, Board.

Two questions. Your abstraction basically is translating the result of
your calculation into kind of a table look-up. Right? Is that the level
of abstraction? That\'s what you meant by abstraction?

BIRKHOLZER: Well, I think it\'s two things. It\'s using those, it\'s
developing those look-up tables. It\'s then most importantly populating
those tables in terms of defining what input parameter distributions do
you need to use to capture the variability that you would have in your
entire repository. It\'s also about making some choices and decision
what to do with thermal, or maybe how to categorize results, in example,
from drift degradation analysis where you sometimes have full collapse,
sometimes you have partial collapse, and yet you want to simplify it a
bit and make it handable by TSPA. That\'s what I mean with abstraction.

MOSLEH: And, in those choices, the decision points, did you do a
sensitivity on the impact on the results?

BIRKHOLZER: Well, this probabilistic analysis that I\'m presenting here
really was for me, in designing abstraction, was a tool for
understanding what these choices would mean in changes in seepage. In
other words, let\'s say if I were not clear about what parameter
distribution log-normal or normal to use for a certain parameter, I
would maybe just run that analysis and I would realize there\'s only a
few per cent change in seepage, and I wouldn\'t be worried about it. So,
I could do that in terms of seepage. What I couldn\'t do with that
analysis is use those results and then propagate them for the entire
TSPA to see what that does. That\'s up to TSPA itself.

MOSLEH: And, I\'m not so sure now about what you mean by sensitivities
of the type that you have on Slide 28. Does that mean that those
sensitivities, you determine that those are important variabilities and
uncertainties to capture, and, therefore, the results actually already
capture these things?

BIRKHOLZER: Yes. And, in some cases, I would decide it\'s not worth it,
you know, if you do something and you realize that, you know, there\'s
no change virtually in seepage, we shouldn\'t impose that on TSPA. So,
again, based on sensitivity choices.

GARRICK: Okay, quickly, David?

DIODATO: This is Diodato, Staff. I have just two quick questions.

And, thanks for your presentation. First, on Slide 20, I wanted to
follow up on Howard\'s observation here. You said you showed how rapidly
the unsaturated zone responds to climate changes, instantaneous climate
change. And, my question is, in this case then, what does that mean in
terms of the ideas we heard discussed earlier about long-term average,
steady state climate, versus the dynamics of the system? That\'s the one
question.

The second question is related to Slide 12. And, here, we have the
seepage look-up table. The Board sent DOE a letter a couple years ago
asking a question about this table. Just quickly, on the one axis
that\'s probably not visible to anybody who is not looking at the thing
in front of them, is the capillary strength. That\'s increasing as it
comes towards us out of the slide. And, then, the other axis of
permeability is increasing as it comes from me to you. So, when we look
at the rocks that have high permeability and high capillary strength,
have lower seepage, in a nutshell.

In lower lith, you show these things plot around 600 pascals on the
capillary strength, then about 10^-11.5^ on a permeability. The question
the Board sent DOE was are there any rocks that DOE had identified or
measured so far that plot in this red area of high permeability and high
capillary strength? Has DOE found any of these rocks so far, or not? I
mean, this is the area\--

BIRKHOLZER: Did you get an answer to that letter?

DIODATO: We never got an answer to the letter. That\'s why I ask you
now.

BIRKHOLZER: Are you asking me are there any rocks that would\--

DIODATO: That have that characteristic where they plot in that high
permeability, high capillary strength field in your look-up table here.
Can you think of any examples of rocks where that\'s the case?

BIRKHOLZER: Well, on that look-up table, if you want, in terms of those
two properties covers the range of values that we essentially cover with
all parameter distributions, if you want, going to the extreme cases.
So, the mean values of those that are most likely to occur, and mean in
this case means assuming that we have like average percolation of 5 or
10 millimeters, means that you don\'t\--mean phase doesn\'t really get
you seepage. You have to go to the extremes. So, most of the rocks are
typical for not really tending to have seepage. The reason that we go so
high in percolation is really because of future climates, because of
very focused, local fluxes that may occur close to faults, and issues
like that.

DIODATO: Yes, percolations are the question here. The question is the
realism inherent in this approach, whether there is realism, whether
there are really rocks that have this kind of characteristic that we can
identify, or not.

BIRKHOLZER: That range is based on the air permeability testing that was
done in many locations, done in the repository units, basically reflects
whatever came out of those slides I had on permeability. And, that range
is essentially based on the inverse modeling to the seepage tests.

Now, as I pointed out, those capillary values are fairly small, smaller
than what you would usually expect for just physical capillarities of
fractures of maybe .1 millimeter, or so, and they are small because they
do not just include the physical capillarity, but they also include some
effects like coarseness, sort of local roughness of walls. So, in that
sense, if you, you know, you would tend to think they are small, but
they are really calibrated, a range of calibrated results here reflected
in that.

DIODATO: Well, I guess I just wondered if you intended at some point to
give us a rock that you\'ve analyzed that has these properties, to show
us an analysis that has these properties.

BIRKHOLZER: Then, the other one was\--

DIODATO: The other one was the dynamics question.

BIRKHOLZER: Okay, the first thing I should point out, I mean, I chose
that example here because of these very high fluxes. Those are fluxes of
deep percolation fluxes like ten fold those that we have on average,
just because I wanted an example where seepage occurs. Now, if you have
ten fold the percolation flux that you would have, you know, on average,
obviously, you would see a much faster response to a drastic change in
climate.

The other thing is that a drastic change in climate, is an instantaneous
change, is an assumption that is done for some indications. It is not
anything that we would expect, I guess.

DIODATO: But, your analyses do show that with that set of assumptions,
you know, rapid climate changes, the system responds in a dynamic and
rapid fashion to that?

BIRKHOLZER: Yes. Again, if you have a ten times less average in deep
percolation, that would tend on slowing.

DIODATO: I understand. Thank you.

GARRICK: I think we\'d better go to our next speak, and I\'ll turn it
back over to George.

HORNBERGER: Our next speaker.

HARDIN: I\'m Ernie Hardin with BSC. I hope I don\'t keep you all from
your lunch.

With regard to Dr. Kadak\'s first question, yes, we do have a report
that describes the extent of dryout into the pillar. It runs cases for
different values of the flux and thermal conductivity, and our
conceptual design report gives the condition that the pillar centerline
remains at or below 96 degrees C.

KADAK: Send it to us, okay?

HARDIN: Okay. I\'m going to go back here. I want to recognize James
Schreiber, Stephen Webb and Alfred Reed of Sandia, who are major
contributors to the work I\'m going to describe.

Okay, next slide. Here\'s an outline. Very quickly, I want to talk about
the seepage water mass balance, that is, where the water goes in the
EBS, the flux diversion algorithm that we use to describe diversion by
the drip shield and the waste package. And, then, I want to spend a
little extra time, relatively speaking, on the in-drift condensation
model, which I believe you\'ve never been briefed on before. So, I\'ll
go through a couple details of the model development and present some
example results. And, finally, talk very briefly about how flow in the
EBS is partitioned back into the unsaturated zone.

Next, please. This slide addresses the question what key processes
affect these estimates. And, so, we\'re going to address here seepage,
thermal seepage and drift-wall condensation and evaporation, also, flux
diversion and flow partitioning back into the UZ model.

Next, please. This slide is really the interface between my talk and the
one you just heard. Jens, in fact, showed you this histogram here, and
these are for the present day monsoonal and glacial transition climate
states. These are estimates of the amount and frequency of seepage where
seepage occurs, and for the conditions specified here, seepage occurs at
24 per cent of waste package locations.

This is the inflow into the EBS from seepage, and in our EBS flow model,
we simply apply continuity. So, the outflow is equal to the inflow.

Temporal and spatial variation of this information then is propagated
through the EBS flow model. And, the flow in the EBS is correlated in
magnitude to percolation flux because of the seepages. So, that answers
those questions.

Next, please. Insofar as continuity is concerned, this is our conceptual
view of the flows or the fluxes of water in the EBS. Seepage from the
crown, part of it is diverted, if you have a breach in the drip shield,
which might be caused by fault displacement in the event of an extreme
event, or by general corrosion. Then, part of the flux goes through the
breach and can contact the waste package, and then ditto for the waste
package.

The fluxes are collected in the invert, and in addition to coming into
the invert, there is a calculated small flux, which we call imbibition,
which is a capillary flow from the host rock. And, then, going out of
the invert, we have flow into the matrix of the host rock, and into the
fractures.

Next, please. This slide addresses the technical basis for the flux
splitting algorithm. For the drip shield and for the waste package, we
use equations of this form. N is the number of breaches, l is the
characteristic length of a breach. The big L is the length of the drip
shield. And, this is the incident flux, and F~2~ would be the flux
through the breach. And, this little f here is a random variate that
when you sample over a range, to represent uncertainty. I\'ll talk more
about that in just a minute.

We take the minimum here because the expression is designed to represent
an arbitrarily large number of breaches, and, obviously, the flux
through the breach. All the breaches cannot exceed the total incident
flux, so we take the minimum.

Same type of relationship for the waste package. And, one of the key
uncertainties here is what happens to rivulet flow after contact with
the surface of the drip shield or the package. It flows off, and the
angle at which it flows off, and the likelihood with which it will
interfere with a breach is an uncertainty.

Next, please. I threw this in here to show the laboratory work that was
done to quantify the uncertain variate in that flux diversion
expression. And, of course, this is a full-scale mockup of a drip
shield. Breaches such as these would only occur due to general
corrosion. It could take a very long time.

Next, please. This slide addresses the key assumptions and uncertainties
associated with the EBS flow, seepage mass balance and flux diversion
aspects. One of the key assumptions here is that seepage from the drift
crown impinges directly on top of drip shields, and this is a
simplifying assumption. It tends to increase or maximize the
availability of that seepage for flow through breaches, if they exist.

Another key assumption here is that all leakage through drip shields,
through breaches in drip shields, impinges on the waste packages. This
is also an assumption that increases the availability of flow for
interaction with the waste form. There is an alternative mode of flow
here, which is that the flow that finds its way through a breach in the
drip shield could cling to the underside of the drip shield and be
diverted. That is not included in this model.

There is also the possibility of a mineral scale or debris accumulation
on the surface of one of these components, and the approach that we\'ve
taken here we think ensures that the average response when a model is
applied over many waste packages or drip shields would be honored.

And, I already talked about the key uncertainty in flux diversion, which
is just how does flux run off of a curved surface.

Next, please. Okay, now we switch gears and talk a little bit about
in-drift condensation. This model that I\'m about to describe was
developed a couple years ago in response to observations underground at
Yucca Mountain, particularly the closed interval of the ECRB cross-drift
that you may be familiar with.

I should point out from the onset that evaporation and condensation
processes are limited by the availability of moisture in the
environment, not by the availability of heat, that is, the waste heat
from spent fuel is sufficient to evaporate plenty of or all of the
incident percolation on a drift. That has been taken into account in the
development of this model.

The objective for developing this model is to represent drift-wall
condensation as a contribution to advective transport of radionuclides
in the EBS. I think we recognize that condensation is going to occur
perhaps everywhere eventually in the emplacement areas. As the system
cools down, there is water present, and the temperature differences from
point to point in the system may become very small, and those are all
conditions with which we are very familiar, and they give rise to
condensation in our everyday experience.

What the objective here is is to look at what kinds of fluxes in the
environment, similar to seepage, do we need to consider for inclusion in
the assessment of advective transport of radionuclides in the EBS.

So, how do we implement this? I haven\'t described the model yet, but we
will implement it as an additional source of water. It\'s treated as
seepage, so it finds its way through breaches in the drip shield, and so
on, and it\'s included in the downstream water mass balance, so it\'s
included in the water that is partitioned back into the UZ.

Next, please. Okay, this is kind of a conceptual run-up to the
condensation model. We recognize that condensation occurs in three
stages as the repository cools down. In the first stage, the entire
emplacement area is at or above the boiling temperature of water, and,
so, you get evaporation, but no condensation. You would get condensation
over here in the unheated region because of transport.

In the second stage, part of the repository, I think the edge of the
layout, has cooled to below the boiling temperature of water. So, now,
the drift wall is cool enough to sustain condensation.

And, in the third stage, the entire emplacement area, within that area,
the drift-wall temperature is at 96 degrees or less, and now you can
have evaporation or condensation anywhere in that area.

The model was developed to describe this stage. This stage here is
relatively brief, and we have other modeling results that suggest that
the results we did up here can be used here for the purpose of the
model.

Next, please. Condensation, or I should say this evaporation transport
condensation process we\'re talking about here is a drift-scale, even
repository-scale process. So, you have to start with a repository-scale
model for heat transport, and this is how we have approached that here.
The repository is modelled using a super-position solution of heat
sources, all with decaying characteristic, so its line average heating
in a number of drifts.

And, for discussion purposes, I will present results for Drift 7 as
examples. So, the super-position then gives us temperature profiles from
one end of the drift to the other. The dotted line here and here
indicate the extent of waste packages emplaced. This is the boiling
temperature up here. And, we\'re going to apply the model, we\'re going
to develop it and then plug in numbers to use it for 1,000, 3,000 and
10,000 years. These are time slices. It\'s a steady state model.

Next, please. Okay, describe the model. This is a network model, or you
may have heard such models referred to as lumped parameters. It\'s the
type of modeling approach that you would use if you are interested in
the bulk response. We want to know how much water could be deposited on
the wall of the drift. We\'re not interested in the fine details of how
that, or where that water would be applied, whether it would be, for
example, at the lower limb of the drift, or at the crown. We\'re not
interested in that detail. We just want to know the overall magnitude of
the effect.

So, what we have done here, this particular schematic gives some of the
information. Waste packages are represented as nodes. Drip shield also.
The under and upper surface of the drip shield, separate nodes. The
drift wall is another node. And, the top of the invert is another, and
the bottom is another. And, then, within the air space, we have nodes,
and we allow a dispersive mixing type transport process axially within
the air space.

There is one process implemented above the drip shield, and another one
below. So, you have about 1,000 nodes for a drift, and, so, you have
several thousand equations, solve them simultaneously. Again, we do it
at these three time steps. We have these nodes. We do include the
unheated end regions. That\'s important because a lot of condensation
can occur there.

Next, please. The source conditions are important here. What we have
done is we have used that super-position solution to calculate drift
wall temperature, and as the moisture source condition, we use the
saturation vapor pressure at that temperature. And, then, we use a mass
transfer correlation to allow that vapor to issue into the air stream.

So, that source condition is applied at the drift wall, or either at the
top of the invert or at the bottom. We\'ll talk more about that shortly.
The evaporation rate in the model is limited in a now linear way by the
percolation in the rock incident on the drift footprint. There is a
limit there.

Next slide, please. Now, this slide will attempt to briefly explain
where the dispersion coefficients come from that we use to describe
axial transport. Steve Webb at Sandia has done a number of 3-D CFD
simulations. They are steady state, single component simulations using
air. The waste packages and drip shields are represented explicitly. He
uses 14 waste packages in an array with no flux boundary conditions at
the end.

The CFD model also includes 5 meters of surrounding rock, and then he
uses a conduction only super-position type solution to fix the
temperature on the outside of that annulus of rock.

I showed you a temperature profile down the drift. Towards the limbs of
that profile have slope, and we call it tilt, and it makes a big
difference whether the boundary condition you apply to the outer surface
of the annulus of rock is uniform or whether it is a sloping or tilted
boundary condition. That affects the flow. He did it both ways.

To calculate a transport behavior, we put a non-buoyant tracer in the
gas phase. We introduced the tracer at one concentration, pulled it out
at another concentration, and then calculated the flux. The result is if
D-naught is the binary diffusion coefficient in the gas phase of 2 times
10^-5^ in SI units, then the range of dispersion coefficients is roughly
200 to 4700 times. So, this is the tilt case, and this is the uniform
boundary condition.

Next, please. So, those are our high and low d-values. Key assumptions
used in this model, I described the source condition for evaporation.
This certainly maximizes the availability of water to evaporate. The
invert is modelled as transmissive to vapor, but not to heat. That turns
out to be just a necessary model simplification. But, for the high
invert cases, and I will talk about this more in a minute, this
certainly, what we\'re doing is we are applying a saturated vapor source
condition at the top of the invert. Now, this is sort of analogous to
seepage entering the drift, being diverted by the drip shield and
getting into the invert and wetting up the top surface of the invert to
a high degree of saturation.

In addition, this model is a mass balance on water, but it is not a heat
balance, in terms of the latent heat of water. So, that\'s an assumption
and limitation on the model.

Some of the uncertainties, I\'ve listed three here. Each one is
addressed in the model results by using a range of parameters. For the
uncertain liquid saturation distribution in the invert, we either set
the source vapor pressure condition at the top or the bottom of the
invert.

Another condition I haven\'t talked about yet is whether the air volume
under the drip shield is ventilated to the air volume outside. We ran it
both ways. Clearly, ventilation will set the vapor mass fraction under
or equal to the vapor mass fraction outside the drip shield. It makes a
difference.

The dispersion coefficient we talked about. Now, there are mechanisms
such as barometric pumping and large-scale gas phase circulation, in
both the drift and the rock that could increase transport of vapor
axially down the drift. Those will tend to increase the transport of
vapor from the emplacement area outside where it condenses, and doesn\'t
wet up waste package locations. So, there are processes that are not
included in our development of the dispersion coefficients that might
decrease the amount of water available.

Next, please. Okay, here\'s example results. This is for the ventilated
drip shield. Condensation rate here is given in kilograms per year per
waste package location. This is condensation on the drift wall.
Condensation occurs if the vapor mass fraction in the air is greater
than the equilibrium vapor mass fraction at the temperature of the
surface where you\'re contemplating condensation, so, in this case, the
drift wall.

Said differently, if the red curve crosses the blue one, you get
condensation. But, in this range here, the curves coincide and we get
condensation at about 100 kilograms per year per waste package location.
Again, this is the emplacement region between the dotted red lines, and
we did not plot condensation in the unheated regions outside, but if we
had, it would be a strong signal.

So, the magnitude of condensation as calculated by the model then is
comparable to seepage magnitude that you saw in those histograms
previously.

Next, please. Now, here\'s the same plot I just showed you, vapor mass
fraction for the low dispersion, low invert case at 1,000 years, with
the mean percolation. And, here\'s a comparable plot, but the dispersion
parameter now has the high value, 4700. So, now we are looking at the
other end of the range for axial transport of vapor. In this case, there
is no condensation in the emplacement area. The curves do not cross.
They cross outside. So, what\'s happening here is the\--well, the
emplacement area stays relatively dry during this time interval.

Now, this result is corroborated by some other modeling that the Science
and Technology Program is in the process of completing. Basically, if
you have enough transport axially in the drift, you do not see
condensation in the emplacement area.

KADAK: Why is that?

HARDIN: Well, you can get the water vapor to go outside the emplacement
area, it\'s much cooler there, that\'s where it prefers to condense.

KADAK: So, you are arguing for circulation in the drift to avoid
condensation on the waste packages?

HARDIN: We have identified that this process is opprent (phonetic), and
these two cases are equally weighted in our abstraction that we
recommend for modeling.

KADAK: Okay, go on.

HARDIN; Okay. Next, please. Okay, this slide addresses some of the
assumptions and uncertainties attendant to this model. We have
propagated the three major uncertainties that we\'ve identified into
eight cases. We selected the ventilated cases for use in the system
model. That is consistent with the controlling assumption in the
multi-scale model, which is our thermohydrologic model that we used for
the system dose assessment. So, for consistency, we have adopted this
one.

If you use this one, if you use the non-ventilated drip shield
assumption, you get a different set of cases, and the system behaves a
little differently. Now, you have a tube within a tube, and, so, you can
have transport axially within the air space under the connected drip
shield segments. But, as I think I will show you on the next slide, the
condensation on drip shields and waste packages is not included in our
abstraction, and is dealt with in a different way in our screening
evaluation for features, events and processes.

Now, remember, I introduced the talk by saying that condensation can
occur eventually throughout the facility. It\'s quite likely that
condensation will occur. And, so, the consequences of that low-level
condensation occurring throughout the facility are excluded. The model
describes some of the thermally driven, stronger condensation effects
that meet the objective for developing the model.

So, finally, how do you implement results such as I described in a
system model? The way we have done this is we had developed statistical
correlations, combining for each drift, and for all percolation
conditions for each of the time slices. We developed a statistical,
it\'s a regression, and we have one for the probability of condensation
to occur on the drift wall at a waste package location, and we have
another for the magnitude.

Next slide, please. Now, we get to switch gears and talk very briefly
about how we partition flow back into the UZ. This is a schematic from a
more comprehensive description of the abstraction. This is how the
system model implements the partitioning.

If you start from, you have water effectively flowing from the waste
form, through corrosion products, into the invert, then the interface
we\'re talking about is right here. And, very simply, what we have done
is if there is seepage, that represents generally a stronger flow than
any other capillary flow in the system. So, that seepage, once it comes
in, it\'s coming from fractures, it goes back into the UZ in the
fractures. And, the imbibition flux I pointed out is a much weaker flux.
It comes from the matrix of the rock, it goes back into the matrix.
Drift wall condensation, if it occurs, is treated as seepage.

So, what we have done here, it addresses the drift- shadow effect. The
drift-shadow effect is the modification of the flow field by an opening
in this unsaturated rock. And, the key to the drift-shadow effect is
that below the drift opening, you don\'t really have much fracture flux.
And, so, we have honored that in principle in this abstraction by
putting radionuclides, if released from the EBS, into the matrix, unless
they are seepage, and then they go into the fractures.

Next, please. So, in summary, I have described the EBS flow model based
on continuity. The seepage water mass balance, it\'s a macro-scale mass
balance. Condensation is then added to seepage. The thing that\'s not
included in the mass balance is the source of the condensation. That\'s
evaporation somewhere where you have less water. So, that\'s an inherent
assumption to the representation of condensation. I have described the
experimental basis of flux diversion.

For the condensation model, I\'ve been very careful to point out the
modeling objective. It\'s a mass balance. We have identified three key
uncertainties, and we deal with them in the abstraction by specifying
ranges on key parameters. And, we have summarized those results in
statistical correlations, percolation flux being the dominant
independent variable.

Next, please. In regard to EBS and UZ flow partitioning, this is really
part of an EBS-UZ interface model that governs the transport of
radionuclides. But, I have just told you about the flow part of it, and
we believe, based on our own sensitivity studies, that we have captured
the essence of the drift-shadow effect by doing it this way, and that
it\'s realistic.

So, I\'ll leave you with one final point. Seepage, plus drift-wall
condensation gives you advective flow in the UZ, gives you advective
flow in the EBS. So, those are the\--I know that\'s a key distinction.
Without seepage or drift-wall condensation, the transport of
radionuclides in the EBS is limited to diffusive processes.

Thank you very much.

HORNBERGER: Thanks, Ernie. So, if I have this right, you actually can
have up to roughly twice as much water leaving the drift as coming in,
because you\'re condensing an amount about equal to what is seeping in;
is that right?

HARDIN: Right. Or stated differently, we\'re not accounting for the
dry-out effect, whence came the vapor that condensed. Somewhere,
there\'s a very dry place.

HORNBERGER: So, where is that? Where does the water come from? Where are
you getting the vapor from?

HARDIN: In general, it comes from wherever it\'s warmer in the
emplacement area.

HORNBERGER: I understand the physics, but you\'re having condensation
every place relative where you have thermal gradients, and, so, you
don\'t have a place that\'s really drawing out.

HARDIN: I understand the question. The type of\--my remarks about
condensation being a process that may occur everywhere eventually apply
in the very long-term.

HORNBERGER: I know. I understand that.

HARDIN: Where you have the strongly thermally driven process, then what
I just said applies, in that it is the hottest part of the emplacement
area that is the potential source for the operation that causes
condensation.

HORNBERGER: Okay. So, let\'s take a drift, the center of the drift is
the hottest spot, as you\'ve shown, and, so, the condensation at the
ends comes from water vapor at the center of the drift. Where does that
water vapor come from, unless it comes from seepage that came into the
drift?

HARDIN: Okay. It is possible to have\--I understand the question. You
know, the near-field host rock is permeable to gas, so it\'s evaporating
from the rock, or from seepage.

HORNBERGER: But, you\'re not taking into account evaporation from
seepage?

HARDIN: Seepage is not represented explicitly in this model. That would
be another order of model complexity.

HORNBERGER: I forget what seepage was, F~2~, or something. You\'re not
decreasing F~2~?

HARDIN: That\'s correct. Because of the conservatism that you\'re trying
to box in is just that, not decreasing F~1~ where appropriate, where
seepage might occur.

HORNBERGER: One last question. You\'re considering condensation on the
drift walls at the cooler ends because the drift walls are cooler than
the drip shield or the waste package. Could you envision getting
condensation on the drip shield or waste packages if in fact the latent
heat transferred to the drift walls at the ends actually warmed the
walls?

HARDIN: Yes. Could I see backup Slide 26, please? It is possible to
calculate condensation directly on the underside of the drip shield, or
on the surface of a cold waste package. But, that only happens in this
model under certain conditions. And, basically, you have to have that
high invert, you have to have high saturation in the very top layer of
the invert in order to get the vapor pressure up in the air space under
the drip shield. This figure kind of represents that. These are the
ventilated cases that we used.

I should also point out you also have to have a non-ventilated
condition. You have to restrict the communication of gas. So, here are
the non-ventilated simulations, so it shows that yes, at the drip
shield, we do get condensations at these time slices, and so forth. For
the low invert condition, we get the same result we get over here for
the ventilated case.

HORNBERGER: Thanks. Other questions? Andy.

KADAK: I\'m kind of confused. I thought I understood what you were
saying, but then this last exchange confused me again. Is the drift
non-ventilated, as a matter of fact?

HARDIN: This is post-closure. We do not take account of processes driven
by barometric pressure flow.

KADAK: Right. And, why wouldn\'t you assume that to be the case for your
TSPA?

HARDIN: I think it\'s another order of model complexity. We can drive
this model with a signal, which represents the barometric effect, and
the result will be that more water will get transported away from the
waste packages. We haven\'t done that.

KADAK: But, what I\'m asking I thought you said, and correct me if I\'m
wrong, but in your modeling of the water flow, you used the ventilated
case as your representative scenario. And, I\'m asking if, in fact, it
is likely to be less ventilated than ventilated, why not use that as the
model?

HARDIN: Okay. Well, the argument for using this case, besides
consistency with other models, is that there are plenty of paths by
which gas can get around the drip shield. It is not sealed. So, there\'s
a gap in the joint between drip shield segments, and it is not sealed at
the bottom either, so gas can get in and out.

KADAK: It\'s a ventilated drip shield, not ventilated drift?

HARDIN: Yes, sir.

KADAK: Okay. I\'m sorry, I got confused. The other question is relative
to where the water is. The water is condensing in the ends of the
drifts?

HARDIN: Yes.

KADAK: Okay. At some point, the temperature of the waste package will go
to below 100 degrees Centigrade.

HARDIN: Right.

KADAK: When will that, roughly, occur?

HARDIN: That occurs within 200, 300 years at the edge.

KADAK: Okay.

HARDIN: At the end of the drift.

KADAK: For the waste package, I\'m talking about.

HARDIN: Yes, you will have a cool package at the end of the drift, at a
location where the drift wall temperature cools below 96 degrees, within
less than 300 years.

KADAK: And, how about the rest of them in the center?

HARDIN: They take much longer. They can take up to 1,000 more years.

KADAK: 1,000 more. Okay. Now, what happens then to the condensation on
the waste packages? And, does that not also make another source of water
which, as I recall, you said was not included in the waste package
analysis?

HARDIN: Yes, but I\'m trying to place the question.

KADAK: You said you didn\'t consider condensation as a source of water
in the waste package degradation analysis.

HARDIN: Okay. What we have done for the purposes of this model, we have
only considered condensation on the drift wall, because that\'s where a
lot of condensation occurs.

KADAK: I\'m asking\--

HARDIN: It also occurs on the package itself, based on some of our
results here.

KADAK: Okay.

HARDIN: It\'s not all cases. It\'s just a few cases. It requires, first
of all, that you restrict gas phased communication under and above the
drip shield first. It also requires that you use a high invert source
condition, as we\'ve postulated in our model, and that was the condition
where you have high liquid saturation in the very top layer of the
invert. The invert is a porous medium, very permeable, and it\'s
unsaturated. Water will tend to flow right through it into the rock
below. On having a high saturation condition in the top of the invert,
is a limiting condition really.

KADAK: Well, I\'m told that the drift environment is rarely quite humid,
100 per cent, and I\'m asking the question at your 10,000 or several
thousand years, will you get condensation on the waste package that will
be the source of water\--

HARDIN: Yes, it is possible.

KADAK: \--which will be equivalent to the seepage rate?

HARDIN: No. As you get out to later times, the deltees (phonetic) in the
system are smaller. It becomes harder to predict what the condensation
and evaporation rates might be from point to point in the system. But,
yes, there would tend to be condensation in the cooler places, which
might include cold waste packages.

HORNBERGER: Thure?

CERLING: Cerling, Board.

It seems that combining the two talks, that condensation at the end of
the drifts then would somehow influence the percolation flux. And, so,
can you guys give at least a qualitative description of how much that
might be?

HARDIN: Yes. It could be all of the flux that might\--it could be 100
per cent, and I showed you that in that figure that showed the
distribution of the vapor mass fraction. Do you know which one I\'m
talking about? It\'s the two blue humps. 17.

In this case right here, there\'s no condensation in the emplacement
area, because, again, with enough axial transport, the vapor is driven
out to the cooler regions where there are no packages. That\'s a
plausible effect. We have other models that predict the same type of
response when we put parameters in there to represent this effective
axial transport mode.

HORNBERGER: In fact, isn\'t that more likely than the top one, in terms
of realistic boundary conditions?

HARDIN: That\'s hard to say. I kept in mind the possibility that you
could have seepage in the source region, and that it would wet up the
invert. It\'s possible. Whether it deserves equal weighting or not, is
open to discussion.

HORNBERGER: Go ahead, John.

PYE: Okay. Pye, Board Staff.

I\'d like to compare the diagram in the previous presentation, five,
which shows the vapor barrier, and it indicates that vapor transport is
away from the drift, compare that to Slide 10 in your presentation,
which shows that the vapor moves into the drift. So, conceptually, can
you explain why the difference?

HARDIN: Yes, the difference is in the application of boundary conditions
to the simulation. In a two dimensional thermal hydrologic simulation,
it\'s a\--you do not have transport in the third axis, the inherent
limitation, dimensionality of the simulation. And, if you do that, then
all the vapor flow has to be outlined.

PYE: Going back to a point that was made earlier about the\--the concept
there is in the two dimensional things, the heat basically pushes
moisture away from the drift to drain, but in this case, that wouldn\'t
be the case. The water would move into the drift and down to the end.
So, why do you need\--

HARDIN: I think we\'ve said there was margin there.

PYE: Okay.

BIRKHOLZER: I\'d like to say something in response. I think even if you
open up your drift in kind of a thermal analysis that was presented
earlier, you would have part of the vapor go back towards the drift as a
boundary condition, but you\'ll still have part of the vapor going out
into the rock. So, having that non-vapor transport along the drift is
somewhat conservative, but it\'s not exclusive.

HARDIN: It\'s not unrealistic.

BIRKHOLZER: There is a current S&T project that includes that as the
transport, and it\'s seepage and the moisture that should be available
in the fractured rock, and it probably could lead us to relax those
assumptions a bit.

PYE: Thank you. In Slide 15 and Slide 19, you use the term ventilated
and non-ventilated. I think you need to explain that a little more
carefully.

HARDIN: Okay. We\'re talking about this slide here?

PYE: Yes.

HARDIN: The uncertainty is the amount of, the extent of gas phased
communication between the air spaces below and above the drip shield.
Stated a different way, the amount of gas and water vapor included that
would migrate in and out through the joints in the drip shield and
through spaces, gaps in the footings.

GARRICK: All right, to avoid me getting into more trouble with the
people anxious to get to lunch, I think we\'d better truncate the
discussion. And, I guess I\'m going to suggest that we be back at 1:30.

(Whereupon, the lunch recess was taken.)

[AFTERNOON SESSION]{.underline}

GARRICK: I think we will resume and end our recess, and go onto the next
item on our agenda. Now, we\'re going to go onto the radioactive side as
well, and get some indication of what happens as radionuclides move
through the mountain. So, why don\'t you introduce yourself, and we\'ll
take off.

SASSANI: Okay. The title of this talk is mass and activity of key
radionuclides potentially released from waste forms, waste packages and
drifts over time. I\'m Dave Sassani. I work with Management and
Technical Support Services contractor, providing consulting to the
Department of Energy on the technical areas of geochemistry, coupled
processes, and waste form. I\'ll be doing the first portion of this
presentation, focusing in on processes occurring in the waste forms that
define the source term, and the second half will be done by Rob Howard
from Bechtel SAIC Company, covering the transport within the waste
package, through the drift, and to the UZ over time.

Next slide, please. So, just a quick overview. The source term
presentation, I\'ll go through concepts on the source-term model, I\'ll
discuss descriptions of these models and the bases for those, focusing
specifically on waste form degradation and solubility-limited
concentration models. I\'ll cover then some additional project studies
that are relevant to these areas, and then transition a little bit and
try to cover some of the Science and Technology source term targeted
thrust projects. This will be a little bit of a transition, and then
hand it off to Rob, or we can do questions at that point, whichever.

Before going on, I want to acknowledge that all of the work that I\'m
going to present here today, in both project models and the program, S&T
source-term targeted thrust projects, have been done by numerous
investigators and scientists, both within Bechtel SAIC Corporation, and
at the national laboratories, as well as a number of universities when
we consider the Science and Technology aspects.

So, I\'m going to focus in on concepts now. This is a conceptual diagram
of the drift, with both processes and materials within the drift shown.
Of course, I\'m going to focus in my presentation within the waste
package on the waste form areas. What else is in the package besides the
waste form? Well, there\'s things like the basket materials, carbon
steel, aluminum alloys, which of course form corrosion products after a
waste package is breached, which contribute to some of the transport
aspects. I\'ll focus in on the waste forms themselves, and then Rob
Howard will follow up with the transport through the system to the UZ.

Next slide, please. This is showing the two primary waste package
configurations that we evaluate, one being commercial spent nuclear fuel
waste packages, which can contain spent fuel from either boiling water
reactors or pressurized water reactors, and co-disposal waste package
shown here, which contains both high-level waste, glass canisters
surrounding DOE spent nuclear fuel canister, shown in the center.

Today, I\'m going to primarily concentrate on the commercial spent
nuclear fuel aspects. So, let\'s go to the next slide, please.

Well, what do we know about degradation of commercial spent nuclear
fuel? It\'s uranium, primarily as uranium dioxide pellets, in the fuel
rods, and we have a wealth of information from both natural systems and
from laboratory testing on UO~2~ and spent fuel itself in terms of what
happens to uranium oxides during corrosion and alteration processes that
create uranyl phases.

This figure on the left is a photomicrograph of a sample showing some
original uraninite here, this grain, and some zonation of the uranyl
alteration phases formed during oxidative alteration of that. This
yellow phase at the top, the yellow crystals, are schoepite, which is a
uranyl oxide hydrate. This sequence in zoning of the alteration is a
very common occurrence. It\'s been observed in numerous deposits going
into uranyl silicates with further zones, and the diagram on the right
shows, again, the development of these phases, uranyl oxide hydrates,
uranyl silicates, at later times or in the outer zones, this time in a
case over weeks, which are based on laboratory testing done at Argonne
National Laboratory.

So, this information from both natural analogs and from laboratory
testing provides a very nice conceptual framework within which to
develop models of the alteration of the spent fuel and potential release
of radionuclides through time, which I will now go into.

Next slide, please. So, we\'re now going to talk a little bit about the
description and the integration of models and the technical bases for
those for the project.

Next slide. This conceptual diagram is fairly complicated, but it shows
the connections within the source term model. My presentation is going
to focus on these boxes here, which are two of the primary models for
the source term, the waste form degradation models, which give the far
from equilibrium aspects of the system as it oxidatively alters to
produce dissolved concentrations in the waste package. And, then, the
closer approach to equilibrium processes represented by this box,
dissolved concentration limits, usually defined by solubility limits, of
minerals.

These two models are used in conjunction, or meant to be used in concert
with the system process determining which one is the controlling
limiting factor for various radionuclides through time.

What else I\'d like to point out on this slide is these models within
the package, and the waste forms have upstream process connections to
both EBS chemical environment in terms of gas compositions in the
package, and, of course, the in-package chemistry model, which is
produced both for CSNF packages and the co-disposal packages with
separate chemistries providing inputs.

Downstream, the connections are primarily to the EBS radionuclide
transport models, which plays a large role in terms of the flux of these
away from the waste form in determining which of these becomes a
controlling factor for which radionuclides. Rob Howard will talk more
about this end of it, and out through the invert. You can see, again, we
do check in the invert, dissolved concentration limitations if there\'s
changes in chemistry to be looked at there.

So, I\'m going to go into some detail about commercial spent nuclear
fuel degradation models first, and then solubility limits second. So,
next slide, please.

Okay, there\'s two primary portions to the spent fuel degradation model
which are shown by these two bullets here. One is a gap and grain
boundary instantaneous release fraction of a number of radionuclides
based on inventories determined from, for example, Grade 92 for the
radionuclides primarily that migrate to the boundaries of UO~2~ grains,
or migrate to the interface between the uranium dioxide pellet and the
cladding. These are represented as triangular probability distribution
functions based on the variability of uncertainty in the inventory data.
And, you can see for Cesium 137 and Iodine 129, there\'s about 10 per
cent, or about 25 per cent of those as a maximum that can be contained
there. Those are instantaneously released whenever that fuel becomes
available for alteration.

In addition to this, there\'s matrix dissolution rates which depend on
chemical conditions and temperature, based on the flow through testing
that\'s been done. And, I\'m going to talk to some more details on this
in the next slide.

So, what I\'ve got here is, again, a conceptual diagram for the CSNF
degradation rate model, which is a function of temperature and
chemistry. And, in addition, what we\'re going to look at a little bit
here is the cladding degradation model, which determines the mass of
exposed fuel based on the cladding failure for that package.

The defective clad is taken as split instantaneously after a waste
package breach testing in humid air environments about 175 at Argonne.
It took about two years to split the cladding, which on the time frames
that we\'re concerned with is instantaneous. All of the mass of fuel in
that split rod becomes available then to alter. And this diagram down
here shows more detail of a rod with the cladding split apart, and you
can see the unreacted UO~2~ here in gray, the alteration rind shown in
orange forming, which is generally schoepite, and the cladding itself.
This area splits apart as the fuel alters to schoepite, due to a volume
increase as this alteration rind forms.

The porosity of that rind is treated as uncertain based on determination
of porosity from schoepite, and the alteration rind itself is assumed
saturated, which plays a role both, it\'s used for the water volume into
which the radionuclides can dissolve from the altering fuel, and it
produces the diffusion transport path for radionuclide transport out.
Rob Howard will talk about this a little bit later. I want to
reiterate\--

KADAK: Just a clarification? This will occur regardless of the\--I mean,
suppose the waste package remains intact and the canister is sealed and
welded?

SASSANI: No, I\'m sorry, this instantaneous failure of the cladding and
splitting of any defective cladding occurs once there\'s a breach in a
package.

KADAK: Once there\'s a breach.

SASSANI: So, for unbreached packages, none of this is happening.

KADAK: Okay. And, when you say the package, you\'re talking about the
waste container itself has to be breached, not just the waste package?

SASSANI: The waste container. That\'s correct.

KADAK: So, you\'ve got two breaches theoretically before this happens?

SASSANI: No, you just need one breach through the package so that the
external environment can invade.

KADAK: Okay, I guess I\'m confused. What is that cylindrical gray thing?
Is that the waste package or is that a container containing spent fuel
that you will be putting into the waste package, which I would call the
over-package?

SASSANI: Okay, I\'ll clarify here. All the models that we\'re going to
be talking about today are all for our previous design, which is a waste
package, which is a dual layer waste package. Only the outer Alloy 22
layer which we take performance credit for. This is not for the
revisions that are designed which would include the TADs.

KADAK: But, even that waste package, wouldn\'t that be a seal welded
probably inerted container?

SASSANI: Yes. And, so, once you get a breach through the container wall,
such that the\--

KADAK: But, you need two breaches, is what I\'m trying to get to, the
waste package breach, and then that breach, to get access to the fuel?

SASSANI: I guess I\'m confused because that is one container that needs
to be breached. There\'s no inner-container. If this is what
you\'re\--you\'re referring to this up here?

HOWARD: I believe what your question is you believe we\'ve got the Alloy
22 outer waste package, and then we\'ve got the 316 stainless steel
inner shell.

KADAK: Correct. Is that 316 inner shell, which will contain these spent
fuel elements, is that a sealed and inerted can?

HOWARD: Yes, sir, it is.

KADAK: Okay. So, am I correct in saying two breaches have to occur?

HOWARD: We take no performance credit for that, once the Alloy 22 outer
shell is breached, we assume that the inner shell fails rapidly
thereafter.

SASSANI: You could add up here, not only does the effective clad split
instantly after the waste package breach, but you could clarify that
with once the Alloy 22 is breached, the inner stainless steel container
is assumed to instantly breach.

KADAK: Thank you.

SASSANI: Sure. So, going back to this alteration rind, it\'s used both
for the water volume and for radionuclide dissolution from the altering
fuel and the diffusive transport path. And, just to reiterate,
radionuclides are released from the fuel both by the instantaneous gap
and grain boundary right upon breach of the cladding, split of the
cladding, and also then by the matrix degradation alteration rate, some
more details of which are shown in the next slide.

So, the overall rate for spent fuel matrix degradation is constrained in
our system under two major conditions. One, under acidic conditions
where the rate is a function of both pH and the oxygen fugacity, or
oxygen partial pressure, or redox conditions, whichever you prefer, and
also under basic or outgoing conditions where the rate is then a
function of the oxygen fugacity and the total dissolved carbonate.

These rate equations are both based on the flow-through testing of
commercial spent nuclear fuel and UO~2~ dissolution under various
conditions. An example is shown here in this plot where log of the
corrosion rate per unit area is shown as a function of inverse
temperature, and the negative log of the oxygen fugacity in the system.
The data points are from both CSNF and UO~2~ dissolution rate,
experiments under various conditions, and the surface is the fit to
those data points.

An example calculation of the degradation lifetime based on these rate
models is shown here, years on this axis as a function of pH, and this
is the low temperature result, going to higher temperatures as the rate
goes up at higher temperatures. So, you have longer lifetimes at low
temperatures, and you can see no pH dependence in the outgoing region,
and a pH dependence here.

So, these provide that radionuclides are potentially available for
release, but those masses from the altered fuel are turned into
concentrations using the rind water volume for comparisons to
limitations based on solubility controls. For something like technetium,
there really are no solubility controls that are imposed, so technetium
would be limited by the degradation rate of the fuel, its release. But,
for something like neptunium, we do have solubility limiting phases, and
I\'ll talk about those now.

On the project, we\'ve considered a number of different models for
constraining. Neptunium concentrations based on mineral solubilities.
We\'ve looked at both pure phase models, which I\'m referring to
neptunium oxides here, both Np~2~O~5~, which is a metastable pentavalent
neptunium oxide, and neptunium dioxide, which is the stable tetravalent
neptunium oxide. And, these models are based on thermodynamic data
reviewed by the NEA, primarily taken from Lameer, et al., 2001.

The solution itself is dominated by neptuneal species. These are
pentavalent dissolved species, which indicates that we\'d be dealing
with a reduction reaction for precipitation of NpO~2~, or reduction
equilibria. That\'s not the case for the pentavalent neptunium, which,
in fact, precipitates very readily in over saturation studies in simple
systems, in which the more stable NpO~2~ phase is not observed at very
short times in these simple systems.

In secondary phase models that we have considered, which are
coprecipitation of neptunium within the uranyl alteration phases, the
uranyl silicates and oxide hydrates, it\'s expected that the neptuneal
line would substitute for the uranyl line in these phases based on the
theoretical studies from Burns, et al., 1997, which would involve, of
course, a coupled substitution to maintain charge balance in the crystal
structure.

But, the proximal alteration phases, for example, schoepite, in
particular, don\'t appear particularly effective at even taking up
neptunium in any manner. And, so, the project currently does not
implement this kind of a model. What the project does do currently, for
our expected system, is to use the neptunium dioxide constraint within
the waste package for a number of reasons. The kinetic barriers for
reduction are not expected to be an issue in this case, primarily based
on the numerous and massive reductants that exist within the package
itself, primarily the waste form, in which the neptunium resides, and
the steels, the carbon steel and the steel alloys, which comprise a very
large mass of reductive material inside the package.

In addition, neptunium is expected to be tetravalent within the spent
fuel itself, and, in fact, may even be controlled by the corrosion
process of the spent fuel at an even lower redox condition than the
atmospheric FO~2~ that we impose.

And, lastly, for our expected system, it\'s going to produce a gradual
increase of dissolved neptunium concentrations, starting from well under
saturation with all of these phases. The first phase that would saturate
would be this phase, if the system can even reach saturation with one of
those pure phases.

However, outside the package in the invert, which is a more simplified
system, does not contain quite the mass of reductants. As a hedge
against the uncertain precipitation kinetics for neptunium dioxide, we
apply the constrains from Np~2~O~5~. Now, there are additional studies
relative to what the extent of those kinetic rates may be for NpO~2~,
and they are shown on this next slide.

Finch did a study in 2002, where he formed NpO~2~ directly by humid air
alteration of neptunium-doped U~3~O~8~ as an oxidized form of UO~2~ at
elevated oxidation conditions, hydrogen peroxide was added to the
system.

In about three weeks at 150 degrees C, he formed very nicely crystalline
neptunium dioxide, and at 16 weeks at 90 degrees C, there was a
formation of neptunium dioxide and also Np~2~O~5~, although this was
determined by XRD because there is less overall reaction progress of the
starter materials, and, so, very little of the U~3~O~8~ neptunium had
reacted at this point.

In addition, for precipitation of NpO~2~ directly from solution, simple
solutions without any reductants added, Roberts, et al. in 2003 observed
at moderate pHs in less than three months at 200 degrees C, a
precipitation formation of nicely crystalline neptunium dioxide.

Well, there is additional information besides these tantalizing pieces
of information of relatively short times from a geologic standpoint, or
even from a standpoint of decades time frame, relatively short time
frame PO~2~ formation. In our expected system, I think we would even
expect to have a much longer than these periods of time, even at the
lower temperatures.

Next slide, please.

LATANISION: Excuse me.

SASSANI: Sure.

LATANISION: The precipitation would effectively mean that neptunium is
not mobile, not available for transport?

SASSANI: Well, it would be precipitated out as neptunium dioxide, and it
would set the remaining concentration and solution to the solubility and
equilibrium with that phase, which is a little bit lower, and I\'ll show
you some of those values here.

LATANISION: Okay.

SASSANI: So, this is a comparison of the spent fuel drip and batch
dissolution test data with the solubility models I\'ve just referred to.
Now, this has been shown a number of times, this plot is showing lot to
the neptunium concentration dissolved in solution as a function of pH.
And, there is a couple of things on here. I want to talk about the data
points first, which are from batch tests done by Wilson at PNL, where
spent fuel was put effectively in a beaker or container and let to
dissolve and monitored over time of months to years. And, at Argonne
National Lab, the drip testing, where spent fuel pellets had water
slowly dripped on them over time for up to a period of nine years,
almost a decade.

These data sets represent very nicely empirical, very nicely applicable
empirical data for what we should expect from the evolution of spent
fuel in our system. The data points are dissolved concentrations of
neptunium in those tests. They are not necessarily corresponding to
controls by a solubility limiting phase. But, the mechanisms in these
tests are all expected to be mechanisms that would apply in the system
that we will have, which is dissolution of spent fuel.

The lines on here, the dark line is the NpO~2~ model for solubility
limits, showing with the uncertainty band around it by the dashed
curves. And, the data sets themselves give us a window of insight into
the time frames that we have involved in expectation of what kind of
times do we have for NpO~2~ to precipitate and form.

In almost ten years of testing, these concentrations have not even risen
to the level of the solubility of this phase calculated at 25 degrees C.
Our calculation is consistently higher than any of these observations,
which suggests that we\'ll have reasonably long times to rise above here
and super-saturate that phase, and allow it to precipitate, possibly
years or possibly even decades.

In addition to the mechanisms that are occurring in here, a number of
recent studies indicate neptunium retention in uranyl solids. The
mechanism is not quite certain at this point, but this may offer some
explanation to where some of the neptunium might be going, or possibly
could go in our system.

The project has done additional studies on neptunium, both in the
commercial spent nuclear fuel pellets from these tests, to try to
characterize mechanisms, and also to look at specific formation kinetics
as a function of temperature for direct precipitation of NpO~2~ from
solution. I\'d like to talk about those now.

These additional studies, the data were supplied from Jim Kinane and Art
Gillis and Jeff Fortner from Argonne National Laboratory, and this shows
an image of one of the cores taken from a spent fuel pellet. The spent
fuel itself is shown in the pink color. The bluish color here is
actually the alteration phases, and the corrosion front is right here
where this arrow is pointing. I have two diagrams here of analyses,
essentially perpendicular to that corrosion front, not in exactly the
same location, but going across that.

The spent fuel is effectively to your left of this point right here, as
I\'ve lined these diagrams up, and the alteration phases are to your
right. And, what we have seen is the x-ray absorption spectroscopy shows
the neptunium is tetravalent in the fuel matrix. So, that, we\'re
starting with the reduced form of neptunium to begin with.

Also, looking at this plot, you can see that the neptunium to uranium
ratio is peaking just at that corrosion front before dropping off in the
mixed valence region, where the uranium transitions from tetravalent
from the UO~2~ to hexavalent in the uranyl phases shown in red here.

In addition, both neptunium, shown in blue, and plutonium in green, peak
in their concentrations right near that corroding surface. And, all of
these results indicate that there\'s evidence that the redox conditions
act as corrosion front and near that front are controlling the behavior
of neptunium. And, in fact, the neptunium itself may be remaining within
the fuel as a possible solid solution of NpO~2~, within the UO~2~. The
structures are very similar.

Additional studies on the next slide are for precipitation from
neptuneal pentavalent solutions, kind of after the Roberts, et al. study
in 2003, which was done at 200 degrees, and what I\'m showing here are
two diagrams of x-ray absorption spectra for solids that have been
collected from these experiments. These are all preliminary from Argonne
National Laboratory. From 200 up to 280 degrees C, the solids are shown
here, and for comparison, the tetravalent neptunium standard and the
pentavalent neptunium standard, and you can see they correspond very
well to the characteristics of the tetravalent species. So, we\'ve got
tetravalent solids forming in about three weeks at these temperatures.

However, at 150 degrees C, and I believe this is after about a couple of
months, the solids in that system shown here in red have a mixed
signature between both the tetravalent and the pentavalent species
themselves. So, the expected slower kinetics in this system at lower
temperature are being observed, and the work is ongoing to address both
homogenous and heterogeneous precipitation at the lower temperatures,
and to develop the temperature-dependent rates for these. If these turn
out to be favorable, that it might allow us to apply this constraint for
NpO~2~ across the board, even in the simple parts of the system.

So, I want to switch gears here now, and go onto the next slide, and
talk very briefly about the Science and Technology source-term targeted
thrust projects, a little bit of overview, a few examples that are
extremely relevant to what we are talking about here today.

Next slide, please. The prime directive that was focused on for this is
that the research program is focused on the changing conditions over
time, identifying the critical processes in each time interval, with
attention to the radionuclides that are the major contributors to dose.
So, they have the appropriate focus, and, in fact, both Professor Rodney
Ewing and Dr. Mark Peters, University of Michigan and Argonne National
Laboratory, have put together a very well integrated set of
collaborative projects across six national laboratories, seven
universities, with international collaboration with the European
community also.

Next slide, please. There\'s four primary areas of research, spent
nuclear fuel dissolution mechanisms and rates, formation and properties
of uranyl secondary phases, including neptunium incorporation, waste
form/waste package interactions, this is more in-package chemistry, and
transport of radionuclides through those materials, and then integration
of in-package chemical and physical processes. There\'s been a couple of
recently funded modeling studies. Carl Stefert at Lawrence Berkeley
National Laboratory, and Udo Becker (phonetic) at University of Michigan
just got funded this past year to work on some of this.

I\'m going to talk three examples from these three areas, starting out
with spent fuel dissolution results next. These data are from Pacific
Northwest National Laboratory under the direction of Brady Hanson and
co-workers. And, what we\'re looking at are low pH single pass
flow-through testing on commercial spent nuclear fuel dissolution. This
is very low pH, pH of 2 for the diagrams on the left, pH of 4 over here
on the right, where we have plotted the dissolution rate as a function
of time in the top plots, and the cumulative release fractions as a
function of time in days on the bottom plots, based on analysis of these
constituents in the effluent.

A couple of things to note. In the top plot, you can see that the
dissolution rate drops as a function of time in this far from
equilibrium dissolution test. That\'s quite important. In the bottom
plot, you can see that the dissolution test goes to about 90 per cent of
the entire sample dissolved out here at 120 days, based on looking at
the uranium coming out. It\'s bending over, based on this dropping of
the rate.

The really, really interesting thing for me from these data, if you look
at both bottom plots, the release of Technetium 99 is much, much lower
and it\'s below 20 per cent in all instances out here, all the way out
at 90 per cent of the fuel reactor. This may reflect technetium being
bound up in the epsilon phase, which is a separate metal particle phase
in commercial spent nuclear fuel containing rhodium, ruthenium,
technetium, I think molybdenum also, and some palladium. Those particles
tend to be extremely resistant to dissolution, and may represent a
mechanism for containing the Technetium 99.

LATANISION: Another question.

SASSANI: Sure.

LATANISION: Are these dissolutions oxygenated?

SASSANI: These are done in equilibrium with the atmosphere, I believe.

LATANISION: So, they\'re oxygenated.

SASSANI: Yes.

KADAK: Do you put a pellet in a beaker of water, or something?

SASSANI: These are nitric acid solutions. These are column tests where
they have liquid flowing through the material in a column, and then
coming out the other end and being sampled for analysis for these
constituents.

KADAK: How does that relate to the repository?

SASSANI: Well, these are representative of the far from equilibrium
dissolution. This is not meant to actually represent the repository
overall. It\'s really focused in on the matrix degradation rate. The far
from equilibrium oxidative dissolution of the UO~2~, the first step in
the process, as one of the major constraints on how does this alter as
you react through time. It\'s not looking at alteration products. This
is trying to get after what\'s the highest rate at which these things
can be dissolved as the fuel oxidatively dissolves. So, these are full
dissolution tests.

In fact, you can see on this plot that some of the plutonium looks like
it\'s also remaining behind. This may reflect some plating out of
plutonium dioxide on the fuel surface, which in these types of tests,
you don\'t want that to happen. You want to get the far from equilibrium
rates for the dissolution of the fuel, which is one end of the spectrum
that the system is going to analyze. Look at the far from equilibrium
dissolution, compare it with other constraints, which are based more on
approaches to equilibrium, precipitation of the uranyl phases,
precipitation of other phases, like plutonium dioxide or neptunium
dioxide.

KADAK: Here\'s my problem. I\'m trying to figure out how does this fit
into understanding how spent fuel in casks or waste packages in the
repository environment will dissolve or degrade. Tell me how this helps?

SASSANI: Well, this is a very specific piece of the puzzle which allows
us to define what the oxidative alteration rate of the UO~2~ can be in
an oxidizing environment. This is basically the conversion rate
to\--this is the highest level of conversion rate you could have to form
other uraneal phases, because UO~2~ itself is oxidatively unstable.

KADAK: Go ahead.

GARRICK: I guess one way of following up with Andy\'s question, given
that technetium is a dose contributor, especially in the earlier years,
a few thousand, how does this relate to the possibility of trapping
technetium in the degradation of the waste package?

SASSANI: Well, if in fact the technetium is bound within a separate
phase, this epsilon phase, in these metal particles within the fuel
itself, we currently represented as distributed throughout the UO~2~
matrix, which the matrix dissolution is represented by this right here,
if it\'s actually bound up in a separate phase, then in fact that phase,
based on some other observations, may undergo dissolution at a much,
much slower rate and, therefore, the technetium may be constrained to
come out at a much lower rate. This is all fairly preliminary
information.

GARRICK: Okay.

LATANISION: Just one additional question.

SASSANI: Sure.

LATANISION: These oxides are going\--their solubility is going to be a
function of both the pH and the oxidizing conditions and the potential,
just as a metal would be. Do you have equilibrium diagrams for these?

SASSANI: I believe we do for the neptunium system and the plutonium
system and the uranium system. These tests, in and of themselves, are
designed in such a manner that the concentrations in solution, because
of the flow-through application, are not high enough to actually
saturate with any of those phases. So, in these types of tests, you
don\'t want other phases to form. You want to see what\'s the forward
disequilibrium rate of dissolution.

LATANISION: I\'ll buy that. But, I\'m thinking back to Andy\'s question.
Depending on what conditions arise in a compromised\--these conditions
may or may not be appropriate.

SASSANI: No, that\'s correct. That\'s why at the beginning of the talk I
was saying the dissolution rate models are used in conjunction with the
solubility limiting models to see what\'s the controlling factor. In
those cases, that\'s where the solubility limits would kick in and say
if you degraded the fuel at these rates, and your fluid was not moving,
and you weren\'t taking it away fast enough, that coupling to the
transporter, if it\'s not being transported away fast enough, the
concentration rises to a point where it hits one of those solubility
limits and is constrained there to reflect precipitation of those
phases. So, those have to be taken as two sides of a coin almost.

The next slide is another set of results. You may have seen this before.
This is studies of pentavalent neptunium incorporation into uranyl
phases form. It\'s work by Burns, et al. at Notre Dame University, where
they have synthesized both uranyl phases that have interlayer cations
and those that do not, and have found that the uranyl phases with
interlayer cations actually do take up the neptunium from solution
reasonably in proportion to what it is in the initial solution. And,
those that don\'t have interlayer cations, which should facilitate the
chart balancing, do not seem to. And, this washes off very readily.

The major point here is not so much the quantitative aspect of it, but
it\'s one of the first demonstrations of the theoretically based
estimates from 1997 that were done by Burns, et al. where these minerals
actually take it up. What they synthesized here, though, were fine grain
masses of material, and, so, they were a little unsure of exactly what
the mechanism was, if it was actually in the crystal structure or not.

And, so, they also did some single crystal studies shown in the next
slide, where they have synthesized becquerelite in this system, which is
calcium/uranyl phase, and there\'s a photomicrograph of the becquerelite
single crystal here. And, then, they analyzed this using laser ablation
ICP mass spectrometry, which is shown, the track of the laser ablating,
essentially vaporizing the crystal is shown here, and the analysis is
shown here with counts versus time. Prior to the laser being turned on,
there are no background Neptunium 237 counts. Once the laser ablation
starts, you can see about 800 parts per million were measured to be
within this crystal, and it demonstrates nicely that that neptunium is
reasonably homogeneously distributed throughout the crystal here.

In addition to these studies on spent fuel and incorporation of
neptunium, there\'s also in package work shown on the next slide done at
Sandia National Laboratory by Pat Brady, Jim Jergen and Ken Korkup
(phonetic) at Pacific Northwest Laboratory, which is evaluating that
once we have a breached package and the basket materials begin to
corrode, what might these iron oxides do in terms of controlling the
redox conditions inside the package, and in particular, maybe occluding
the gas phase. You can see that the minerals equilibria for the iron
minerals are much more reducing conditions than even the oxidative
alteration of the UO~2~ itself, and well below atmospheric conditions
that are evaluated.

Within these types of systems, they\'re looking at both the generation
of the corrosion products, but they\'re also looking at interactions of
neptunium with those, and technetium via rhenium as an analog, the
rhenium ion as an analog for technetium, and they\'re finding both that
neptunium is reduced onto magnetite surfaces, and they are actually
finding that a lot of the rhenium is removed from solution when they
interact it with these types of corrosion products.

And, the next slide is just a picture, scanning electron micrographs of
some of those corrosion products from their coupon studies at low and
high magnifications in very dilute effectively tap water, J-13 type
solution. And, the only thing to take from here is that they see various
types of morphologies of these products, depending upon water chemistry,
and sometimes getting silica precipitants also.

These studies are more about the in package chemistry and transport
pathways for radionuclides, and the transport aspect is what Rob Howard
will cover in his presentation.

That\'s where I\'m done.

GARRICK: Thank you. Questions from the Board? Yes, go ahead, Bill.

MURPHY: That was a very interesting talk, and I\'ll try to stop asking
questions at some point. I am curious about your Figure 6-3 on Page 11,
first of all, where you show dissolution rates. And, there\'s quite a
substantial scatter, which doesn\'t surprise me. I understand these are
hard to determine, and the surface areas are difficult to characterize,
and I\'m just curious in your judgment, whether you think this rather
broad range is a consequence of uncertainties in measurements, or is it
a function perhaps of other independent variables controlling the rate
besides O~2~ and temperature, for instance, pH, or other aqueous
species?

SASSANI: Well, that\'s a long question. I\'ll try to stick to it a
little bit. You know, I think if you look at a lot of these studies, the
primary variables, things like the temperature aspects, the oxidation
state of the system, in particular, you dissolve carbonate under
ablating conditions by a very large controlling role. In particular, the
carbonate because of uranyl carbonate complexation at very high pH. The
uncertainty in these data of course are included in the functions that
are derived from these to try to capture that uncertainty. These are
shown per unit area.

Surface areas, in fact, for the overall rates are explicitly in the rate
relations, and surface areas can vary over a fairly wide range of
values. The surface areas that are used vary over about two orders of
magnitude for these dissolution rate studies. These reflect 10^-6\ 1/2^
meters squared per gram of material for the surface areas there. Those
are derived from backing them out of dissolution rate studies by knowing
the initial surface areas, and measuring the constituents as they are
released. They are also derived from characterization of the geometric
surface areas of the fuel. But, of course, the geometric surface area
may not be the full story if there are cracks and grain boundaries that
play a role in the dissolution rate process, to a large extent.

If you look at things like teller VT (phonetic) measurements of surface
areas, nitrogen gas measurements, those surface areas can be even a
couple orders of magnitude higher. The surface area effect itself, and
the uncertainty in that, the range of that that\'s used is, for the
model itself, one of the larger controls on where these\--it goes
directly up and down in time frame relative to those.

The actual measurements, some of it is possibly characterization of the
surface area, particularly for dissolution tests, you can characterize
that surface area when you begin, but it\'s very difficult to
characterize the change in that surface area through time in the test.
And, in fact, a number of folks have written some methodologies for
getting around that, in particular, some of the PNL co-workers, Brady
Hanson, looking at the actual cumulative release of material and trying
to normalize the rates in a different manner.

But, in terms of the chemical environment, there might be some effects
also from radiation interactions and radiolytic processes, which are
sort of rolled in here implicitly from the spent fuel tests, but
there\'s some uncertainty in how you deconvolve that out explicitly, and
whether or not we\'re seeing in a lot of our testing a large effect of
this radiolytic process that we\'re not explicitly representing.

MURPHY: That addresses some of my questions. I guess a rather basic
concern is that given this rather broad range of maybe an order of
magnitude, or more, is it possible that it reflects differing sets of
conditions, and how can you be comfortable with this particular
regression plane through the data as a conservative estimate of the
rate?

SASSANI: Well, again, these are far from equilibrium conditions. Now,
some of those other conditions which might be important in the specific
experiments, of course, would be the water chemistry, and in the
discussion of the SPFT tests in the S&T area, I talked about do you want
to run these experiments in dope form phases on the surface, which would
(a) occlude surface area and effectively remove a portion of the mass
from reaction in your test. There might be some of that. That\'s a
really hard question to answer.

The types, again, the types of lifetimes, we\'re getting to the fuel
range from hundreds of years to thousands, if not, an extremely long
lifetime. I think Tim McCartin referred earlier to we aren\'t taking an
enormous amount of credit for the alteration of the fuel, the front end
alteration of it.

MURPHY: I\'m curious if the relatively new Hanson, et al. data that you
showed in Figure 21, how they compare to these data for rates, and I
can\'t quite compare, the units aren\'t given in the first figure.

SASSANI: Well, I haven\'t seen the explicit comparison of these in terms
of the actual dissolution rates, and I\'d have to sit down and talk with
Brady about that to really comment directly. I don\'t know if Brady has
done that comparison or not. I suspect, from talking with him in
general, Brady has a very strong feeling from not just these studies,
but a lot of studies that have been done on spent fuel, that the
degradation rate of spent fuel that we use is probably a bit fast
relative to what reality might be.

MURPHY: One additional question concerning Figure 16. You say that
neptunium may remain within the fuel while the alteration occurs, and
I\'m wondering if you attempted a kind of mass balance. Because, it
seems to me that there\'s a substantial quantity of fuel pellet that\'s
been oxidized here, as represented by the blue, and if you would
accumulate all the neptunium in that at the background level of
neptunium, it certainly would be more than that little peak near the
surface, it seems to me.

SASSANI: Right. The results here, and also results in the S&T area, are
tantalizing in terms of seeing mechanisms. But, the quantification of
these I would say is at a very early stage. In fact, you look at these
peaks, and in fact, these are relative concentrations in arbitrary units
based on the absorption and spectroscopy that\'s been done. And, in
fact, if you look closely, the neptunium concentrations are a factor,
multiplied by .002, and the plutonium .006. So, these are just shown for
the qualitative behavior.

I\'ll say one thing in terms of the cores that have been taken, which
were not real easy to do, but these are actually more oriented than the
grab samples that were looked at before, Jeff Fortner indicated to me,
and I think they discuss it a bit in their report, when they do this
coring into these small fuel pellets, the coring device actually wraps
some of the bottom material up on top. You can see again here more spent
fuel up in this region, and it\'s unclear probably at this time how much
of the alteration phases have been rolled up onto here. So, I wouldn\'t
put too much quantification into that altered zone at this point, and it
has not, frankly, yet been done.

I think these types of studies, and these types of measurements, which
are not real easy measurements to make, is just the first step in that
process.

MURPHY: Thank you.

GARRICK: Okay. Yes, George?

HORNBERGER: So, some of these S&T results you showed, my question is are
there plans to move these forward to the TSPA? If so, how do you
anticipate that will be done, and over what time period? And, depending
upon that answer, how about future results of projects in the S&T
program?

SASSANI: The future of the S&T program, I\'m not really sure about. Russ
could probably answer that better. But, speaking with him, it\'s not
really clear what\'s going to go on with that at this point in time.

I\'ll speak to the S&T program as it was set up up to this point, and
its intent was actually to investigate and constrain more detailed
scientific questions, as opposed to be used for the regulatory process.
Now, in fact, that\'s a good thing, but that\'s a really basic question
to ask, of course, what do you do with this? The project\'s point of
view has been, you know, in fact, if there are results found which would
indicate that the project was doing something that was non-conservative,
then absolutely, the results would get rolled into a TSPA. I mean,
you\'d have to go update your models in terms of public health and
safety issues.

But, these types of data from the S&T program, from the regulatory
standpoint, may never get rolled directly into the TSPA model used for
regulatory purposes because they could be pointed to just as
demonstrating the point that we\'re reasonably conservative in the
approach, or reasonably cautious. For the regulatory standpoint, that\'s
okay.

Now, from the standpoint of answering detailed technical questions,
you\'d like to utilize those results also. So, that\'s where I believe,
I mean, which is a long way of saying I don\'t know.

GARRICK: I have some questions, but I think I\'ll wait until the next
presentation. Andy, you had a question?

KADAK: Yes. I\'m trying to get a perspective here of this work and the
relationship between the design or analysis of what the repository will
really do. What I got from the thrust of your presentation is you are
completely focused on trying to understand better the neptunium
dissolution rate from a waste package\--from a pellet, forget the waste
package. We earlier heard you\'re not taking any credit for the
container in which the spent fuel is placed, and yet we also heard this
morning that the NRC folks don\'t want to take credit for the cladding,
and you want to take credit for the cladding, and yet you don\'t want to
take credit for the can. Will you help me understand what you\'re doing
in this regard?

SASSANI: I\'ll clarify a little bit, and Rob may be able to clarify
more. But, in terms of the container, there is a lot of credit taken for
the container.

GARRICK: No, he\'s talking about stainless steel\--

SASSANI: I\'m sorry. Okay.

KADAK: And, you think the cladding is better, apparently, because
you\'re taking credit for it?

SASSANI: There is no representation of the stainless steel liner because
it\'s assumed that once you penetrate the Alloy 22, you have an
environment that is potentially corrosive and will corrode reasonably
fast relative to Alloy 22 corrosion.

KADAK: Okay.

SASSANI: That\'s the way I understand it, but I\'m not an expert on
corrosion.

KADAK: Then, talk to me about cladding then.

SASSANI: Cladding itself, we do have defective cladding included in the
modeling. That defective cladding has instantaneously failed upon waste
package breach. But, there is a lot of cladding that has not failed,
and, in fact, there are cladding models for the corrosion which under
the current operation for looking at the tens of thousands of year
period, did not produce any other cladding failures. The stainless steel
cladding is assumed to be failed. There is no credit taken for it. It\'s
only zircaloy cladding and commercial spent nuclear fuel, and there\'s a
distribution of defective cladding based on looking at the reports and
studies of defects in cladding that have been observed when cladding has
been moved or handled.

KADAK: So, zircaloy is better in this environment than stainless steel;
is that true?

SASSANI: Zircaloy is fairly corrosion resistant. From my understanding,
it\'s pretty much high fluoride environments that attack it, and that\'s
not expected in our case. But, again, I\'m not an expert on the
corrosion.

KADAK: What we as a Board are interested in, and maybe you could explain
when you get up, is how does this stuff actually get out of the fuel
pellet into the environment. SASSANI: That\'s what he\'s going to talk
about.

KADAK: And, he\'s going to talk about that? I\'m not sure we still got
the first part, and that\'s how we are corroding this fuel pellet to a
point where it can get down into the bottom of the can, given the amount
of water that we\'re talking about, and given the dissolution or some
kind of a corrosive mechanism. I understand you dissolved it in nitric
acid, but I\'m hoping that there isn\'t nitric acid in this bloody
repository.

SASSANI: That is just, and I really want to clarify, those experiments
are for characterizing properties, not for stainless steel.

KADAK: I understand that. But, my point was how do you take these
properties, which I know you\'re looking at the oxidation potential, but
how do you translate that into the repository environment?

SASSANI: Well, there are environment models, which I couldn\'t go into
here today because of the time frame, and those models provide all the
chemical parameters that are required to these waste form models, as the
repository evolves through time.

GARRICK: You\'ll get another shot at this after the next speaker. So,
thanks a lot. It\'s time for the next talk.

HOWARD: Good afternoon. I\'m Rob Howard with Bechtel SAIC. I want to
acknowledge that I did get quite a bit of help from Jim Schreiber on
this presentation. He was the lead author for the engineered barrier
system, radionuclide transport abstraction AMR, and much of the material
that I\'m presenting today was derived from that piece of work.

Dave went through this slide already, and I\'m not going to dwell on it.
We can come back to it if we have questions.

Next slide. Let\'s talk for a couple minutes about what the key
radionuclides are that we\'re concerned with in the repository. Start
off by noting that Strontium 90 and Cesium 137 are the highest
contributors to the initial activity in the repository, but they decay
off rather rapidly. They both have half lives on the order of about 30
years, and after several hundred years, they are not going to be
significant contributors.

Bill Arnold is going to show you a sensitivity study in the saturated
zone later on this afternoon that\'s going to show you some of the
effectiveness of the saturated zone with reducing the total repository
activity. So, you might want to keep this slide in mind when he goes
through that sensitivity analysis.

We also have plutonium and americium, and these two radioelements behave
similarly chemically. They\'re strongly sorbing, and that\'s important
when we talk about transport characteristics. They also are moderately
soluble. Technetium 99, we talked a little bit about already today.
It\'s in the GAP inventory. It\'s fairly soluble under oxidizing
conditions, and it persists for a relatively long time. Neptunium 237,
Dave went into great detail about its solubility, and it, again,
persists for a long amount of time in the repository.

Next slide, please. The uranium isotopes are interesting because they
are abundant. They\'re weakly sorbing. They also have relatively low
specific activities. In other words, there\'s U-238, for example,
there\'s a high\--a large amount of U-238 in the repository, but its
specific activity is rather low. Just to give you an idea, I think the
activity is about 3.4 times 10^-7^ curies per gram, whereas Carbon 14,
for example, is on the order of 4 1/2 curies per gram, and Cesium 137 is
on the order of 87 curies per gram. So, we\'ve got a high mass content,
but a low activity. We do track all of the uranium isotopes. Iodine 129,
we\'re interested in, has a relatively soluble element. It also has a
large fraction in that GAP inventory that Dave discussed earlier.

Next slide, please. The only thing I want to point out on this slide is
that we do have some thorium in there that we are concerned about, and
that comes from the high-level waste form.

Next slide. I guess this is where I\'m going to get some more questions
from Andy. Releases from the repository are scenario dependent. So,
let\'s spend a little bit of time going through some of the scenarios
and what the modeling implications are for those scenarios.

In the nominal scenario, we have two cases, general corrosion failure
case and early failure case. For the first 10,000 years of repository
evolution, we don\'t really see any waste package or drip shield
failures due to corrosion mechanisms.

In the early failure case, waste package failure is represented by a
poisson distribution, and in that case, the failed waste package, and
that\'s the Alloy 22, and the 316 stainless steel, performs no barrier
function. The drip shield is still in place, so you\'re going to have
just diffusive releases out of the waste package, but you\'re not going
to get any barrier performance from that waste package other than the
sorptive capacity of the internal materials.

Cladding damage in the nominal scenario is specified by the initial
conditions of the repository. There\'s about 1 per cent of the total
inventory that we receive of spent fuel rods is going to be stainless
steel clad. We don\'t take any performance credit for the stainless
steel clad. There\'s a distribution that we get on the initial
conditions for the zirclad stainless steel, and that ranges from about
zero to 1 per cent failed.

For the igneous scenario, the drip shield and the waste package, neither
one of them provide a barrier function. So, the seepage flux that we get
into a degraded drift is available to contact the waste form in that
case.

The cladding has no barrier function. The high temperatures involved in
the igneous case where we assume some magma will contact the waste form,
we don\'t get any credit for the zirclad. We apply basaltic chemistry
conditions to the engineered barrier system transport environment, and
I\'ll talk a little bit more about that later.

In the seismic scenario, we\'ve got two cases. We\'ve got a mechanical
damage case and a fault displacement case, and we use the seepage
abstraction for degraded drifts in both those cases. And, the mechanical
damage case, the drip shield remains intact. It still performs its flow
diversion function, so releases from the waste package are going to be
diffusive only.

For the seismic fault displacement case, the drip shield loses its
functionality when we have an initiating event with an exceedence
frequency smaller than 2 times 10^-7^.

Damage for the mechanical damage case for the waste package, it\'s a
network of tight cracks, they are very tortuous cracks that the waste
packages can bang into each other, or bang into the pallet, and get
these stress corrosion cracks. No advective flow through those cracks.
It\'s going to be diffusive only. And, the fault displacement case where
waste packages that are sitting on a fault can be disruptive, we can get
what we characterize as a crimping of the waste package. And, that
damage area is uncertain, and it ranges from zero damage to the total
surface area of the lid.

You can also get localized corrosion in the seismic fault displacement
case, but it\'s also going to, keep in mind, that the thing has already
been damages by the crimping conditions. We do have some cladding
performance for the seismic cases, but it\'s going to be a function of
the peak ground velocity. So, it varies with the ground velocity. And,
we adjust the temperature and relative humidity for collapsed drifts in
both seismic cases.

Next slide, please. Key transport processes within the EBS. We consider
transport of dissolved radionuclides, transport of radionuclides that
are reversibly sorbed to three types of colloids. We have groundwater
colloids. We have corrosion product colloids. And, we have waste form
colloids in the high-level waste glass.

We also consider transport of radionuclides irreversibly sorbed onto
iron oxyhydroxide colloids, transport of embedded colloids, that\'s
plutonium and americium that are irreversibly attached to those smectite
clay-like waste form colloids. And, irreversible sorption of
radionuclides, this is just plutonium and americium only, onto
stationary corrosion products.

Next slide, please. Okay, in order to solve the mass transport equations
within the engineered barrier system, we discretized the EBS into four
different domains. We have a waste form domain that\'s going to consist
of the commercial spent nuclear fuel, or the high-level waste glass. We
have corrosion product domain, which includes the damaged area to the
waste package. The invert, which is crushed tuff, and the EBS/UZ
interface domain that Ernie Hardin spoke about already today, that
establishes the boundary conditions for the UZ flow and transport model.

In order to solve the mass transport equations, we\'ve got to specify
water volume, saturation, porosity, diffusive area, the diffusive path
length, diffusion coefficients, what the advective flux is, if any, and
these can vary by the waste form type, whether it\'s commercial spent
fuel or high-level waste glass, or whether it\'s in a dripping
environment or non-dripping environment. Another way to think about that
is whether or not that drip shield is intact once the waste package is
damaged or not.

Next slide, please. For commercial spent nuclear fuel, the way we
conceptualize this, the CSNF waste form domain represents the breached
and axial splitting of the fuel rod that has degraded into the rind and
the alteration products from uranium oxide. And, Dave showed you a
couple slides on that.

The rind is assumed to saturate quickly and completely for both dripping
and non-dripping environments at temperatures below 100 degrees C. We
don\'t have any releases at temperatures above 100 degrees C.

There\'s a continuous thin film that we assume exists at all
temperatures below 100 degrees C. And, so, that always creates diffusive
path length out of the waste form.

The rind water volume is going to be a function of the time dependent
fraction of the degraded waste form, its porosity and the saturation.

The rind porosity is epistemic. I won\'t say any more than that.

The amount of each radionuclide mobilized is going to be a function of
the rind water volume, the waste form degradation rate, and the
dissolved concentration of that radioelement in water.

There are no colloids present in the commercial spent nuclear fuel waste
form domain.

Next slide. The diffusive area is going to be the product of the fuel
rod length, the split opening for those rods that have failed cladding,
and the number of failed rods that are in the waste package. That\'s
going to vary amongst the different scenarios.

The diffusive path length to the corrosion products is the time varying
thickness of the rind, and the diffusion coefficient, we start out with
free water diffusion coefficient, and we modify that by porosity and
saturation, or water content, using power law to form Archie\'s Law, and
based on experimental data from Concha and Wright in the early Nineties,
and some work we did I think out at Low See Road (phonetic) in the late
Nineties.

Advection is going to occur in seismic fault displacement and igneous
intrusion, so we specify what the advective flux is using the flux
splitting algorithm that Ernie went through this morning.

Next slide, please. Co-disposal waste form. The waste form domain here
represents the degraded high-level waste glass that forms the clay like
alteration product. It saturates quickly. The high-level waste glass is
hydroscopic, and that\'s for both dripping and non-dripping conditions.
Again, a thin film exists at temperatures below 100 degrees C.

The rind water volume is a function of the time dependent fraction of
the degraded waste form, but we set the porosity and saturation are
deterministic values, in this case, not uncertain values or variables.
The amount of each radionuclide mobilized is, again, going to be a
function of that rind water volume, the waste form degradation rate, and
the solubility of the radioelement in water, just like it is for the
commercial spent fuel. The alteration of the high-level waste glass does
include embedded plutonium and americium. That is different from
commercial spent fuel.

Next slide. The diffusive area is the sum of the surface area of the
five high-level waste glass logs. That\'s what we consider. So, Dave
showed you a slide of the co-disposal package that\'s got either four or
five logs in it. We use the surface area for the five log case. The path
length is the time varying alteration rind thickness.

Diffusion coefficient is handled essentially the same as we did for
commercial spent nuclear fuel, although we do reduce it by a factor of
100 for radionuclides that are bound to colloids, and that\'s based on
the Stokes-Einstein relationship.

Next slide, please. Corrosion product domain is the portion of the waste
package that has degraded. It\'s all the internal components, like the
basket materials that surround the commercial spent nuclear fuel, and
any other components that are internal to the waste package, other than
the spent fuel or high-level waste glass.

The degradation of the internals results in corrosion products. They can
be a large mass of stationary corrosion products, or they can be
colloidal in form. The mass of the corrosion products is going to be a
function of the mass of the stainless steel and carbon steel that\'s
available, along with their respective degradation rates. Just to give
you an idea, there\'s about 5,000 kilograms of carbon steel in these
commercial spent nuclear fuel waste packages, and on the same order for
stainless steel, and that includes that inner shell that was causing so
much confusion today.

We don\'t account for the consumption of water through chemical
reactions in this domain. The water volume is going to be the product of
the pore volume, and the saturation of the corrosion product mass. The
pore volume is a function of corrosion product mass, porosity, and the
corrosion product\'s density.

GARRICK: Is there a flow rate at which accounting for consumption of
water would be important? We are talking about very low seepage, and
very low flow rates.

HOWARD: I\'m trying to think of the best way to answer that. I see Ernie
on the edge of his chair.

HARDIN: Ernie Hardin, BSC. He\'s talking about a degraded waste package
where the moisture got in through the gas phase. At least that\'s the
conceptual model. In the case of advective transport of liquid through a
breach into the package, you would not be using this particular mode of
transport.

GARRICK: All right.

HOWARD: I think that not counting for the consumption of water, there\'s
been some sensitivity studies that I think Jim Schreiber and others have
done that looked at this, and you could get upwards of 4,000 or 5,000
years, or even more, of delay if you did account for it. But, we
haven\'t incorporated any of those alternative conceptual models.

GARRICK: Okay, thank you.

HOWARD: Water volume is a product of the pore volume, and the
saturation. Pore volume\--actually, we went through this already.

Next slide. The effective saturation in commercial spent nuclear fuel is
a little bit different. It\'s based on absorbed water and is a function
of relative humidity, and the specific surface area of the corrosion
products for non-dripping environments. For dripping environments, for
those cases where you have advective flow into a waste package, we just
set the saturation at 1.

The diffusion coefficient is handled essentially the same way as we do
for the waste form domain.

Next slide, please. Diffusive path length is epistemic and it ranges
from the waste package outer shell thickness to the radius of the waste
package, whether it\'s co-disposal or commercial spent fuel.

Sorption. Sorption is important. There\'s a large mass of iron in here,
the 316 stainless steel. Iron content is about 63 per cent. The A-516
carbon steel has an iron content upwards of 97 or 98 per cent. So,
there\'s a lot of iron available here, and we account for that in
sequestering radionuclides. So, we have sorption, irreversible sorption
to stationary corrosion products, and that irreversible sorption is
based on some field observations and experiments that we have done where
less than 1 per cent of these radionuclides actually desorbed in the
column experiments that we did over a period of several months. So, we
have irreversible sorption on the stationary corrosion products,
reversible and irreversible sorption onto the mobile corrosion products,
and reversible sorption onto groundwater and waste form colloids.

GARRICK: So, you don\'t take credit for the stainless steel for
corrosion resistant, but you do take credit for it as a corrosion
product sorption, absorption bed.

HOWARD: Yes, sir, we do. I should have been a little clearer on that.
That is correct.

Next slide. The invert domain. I\'ll touch briefly on this. The
concentration is going to depend on the solubility limits. We do check
for solubility, as Dave mentioned earlier, in the invert. Colloidal
stability, which is going to be a function of pH and ionic strength, the
transfer between the corrosion product domain and the invert, so, you
know, whatever that difference is in concentrations is going to control
the diffusive transport, and the boundary concentrations at the
invert/UZ interface.

The advective source fluxes are those ones that were defined earlier by
Ernie in the EBS flow abstraction. They generally dominate diffusive
releases in the invert domain. We don\'t account for the transport path
along the waste package pallet, so that is a simplifying assumption that
we made.

Next slide. Diffusion coefficient is handled essentially the same as
they are elsewhere within the EBS. The diffusive length is simply the
average thickness of the invert, and the diffusive area is just the
product of the invert top surface and the length of the waste package.

Next slide. EBS/UZ interface domain. Ernie touched upon it earlier. We
use a semi-infinite zero-concentration boundary condition and we apply
that approximately three drift diameters below the invert. And, we did
sensitivity studies to make sure that we were okay there.

The near-field UZ interface is modelled as a dual continuum consisting
of overlapping UZ matrix and UZ fracture continua, so that we can
account for releases into the fractures and releases into the matrix.
And, I think Bruce is going to touch upon that just a little bit.

Next slide. Sorption. Reversible sorption onto stationary corrosion
products is not considered. I said earlier that we had some laboratory
experiments that suggested less than 1 per cent actually desorbs.
Sorption onto corrosion products in the invert path is not considered.
So, we do have sorption in the invert on tuff when we use the Kds for
crushed tuff, the same Kds that are used in the unsaturated zone. But,
any of the steel plates and rails that are in the invert that corrode,
we don\'t account for that because of the high uncertainty of whether or
not radionuclides will actually cross that transport path.

Irreversible sorption rate constants for plutonium and americium on
corrosion products, whether they be stationary or mobile, is epistemic
and it depends on Goethite sorption density. The pictures that Dave
showed you earlier was a mixed assemblage of different forms of iron
oxide, of hydrous ferric oxide sorption site density, Goethite surface
area, the fraction of total iron oxide that is Goethite versus other
species, and the amount of high-affinity hydrous ferric oxide sites that
are available.

We use Kds for reversible sorption onto crushed tuff, and they\'re again
the same Kds that we use for the unsaturated zone.

Next slide. Colloidal transport. The sorption coefficients for
reversible radionuclide sorption for all three types of colloids, we use
the Kd approach. For the irreversible sorption, we use the rate
constant, just like we do for stationary corrosion products.

For the irreversible sorption of plutonium and americium onto the waste
form colloids is going to be a function of ionic strength and pH.

Some processes that we don\'t consider when we are evaluating colloid
transport in the engineered barrier system include physical filtration.
We don\'t take any credit for retardation at the air/water interface,
interaction with organics, or settling.

Next slide. To sum it all up and bring you back to the different
scenarios, for the early failure modeling case, releases are going to be
dependent obviously on the number of early failed waste packages. Again,
it\'s a Poisson distribution. In-package solubility limits for uranium
and neptunium. Irreversible sorption onto stationary corrosion products,
and the diffusion characteristics of the waste form and waste package.

For igneous intrusion, it\'s a little bit different. It\'s the number of
waste packages and drip shields that are disrupted by that event. The
solubility limits are still important. Irreversible sorption on the
stationary corrosion products, and in this case, we may see some more
releases because we have that advective flux. The diffusive
characteristics aren\'t that important to us, even that factor of 100
that we reduced the diffusion coefficient by, doesn\'t seem to matter.
And, so, colloidal concentration limits are going to be important for
releases out of the EBS in the igneous case.

For the seismic ground motion case, it\'s essentially the same
characteristics as it is for the early waste package failure cases.

All right, Andy?

GARRICK: Let\'s start with David.

DUQUETTE: Duquette, Board.

I thought I understood what was going on, but now I\'m a little
confused. Can I walk you through what I think you said, or at least what
between the two of you have said? And, that is, in order for the fuel to
swell and to cause splitting of the cladding, I have to get water inside
the cladding. So, I have to penetrate the cladding somehow, unless it\'s
a damaged cladding.

HOWARD: Right.

DUQUETTE: To get the water inside. And, as long as I don\'t get water
inside, I will not have splitting of the cladding; is that correct?

HOWARD: That\'s correct.

DUQUETTE: Okay. So, somehow I\'ve got to get the water through whatever
you use for the outer casing, whether it be stainless steel or C-22 or
some of the new alloys that are being looked at, the water then has to
get down onto the cladding surface, has to corrode the cladding surface,
and has to enter the cladding, and then it causes the fuel to swell, and
everything goes backwards, that is, it now allows the fuel to exit. Is
that correct?

HOWARD: That\'s essentially correct.

ARNOLD: Excuse my interrupting. I asked the question this morning how
would the cladding fail in the first place, and the answer came back
imbrittlement, seismic activity, various things that had nothing to do
with corrosion. There were physical effects.

HOWARD: Yes. And, as I said earlier, it\'s mainly mechanical damage,
either it\'s the cladding is damaged as it arrives at the repository
or\--

DUQUETTE: Then, the real question I have is does your model take into
account the fraction of bundles, if you will, where the cladding is
actually damaged?

HOWARD: When it\'s received?

DUQUETTE: Yes.

HOWARD: Yes, we don\'t take credit for the stainless steel, but there is
a distribution that we use, it\'s on the order of zero to 1 per cent
that we sample on for the failed cladding, failed fuel that is received
at the repository.

DUQUETTE: So, barring igneous or seismic events, it\'s only about 1 per
cent of the clad fuel packages are expected to have failed beforehand.
The others probably will not fail at all because these are zirconium
alloy is corrosion resistant.

HOWARD: For the 10,000 year cases that we\'ve evaluated. I\'d think you
wouldn\'t want to say that for\--

KADAK: There\'s another caveat though. Even if it does fail, the
temperature of the pellet has to be below 100 degrees to make any
problem relative to corrosion. That\'s what I also heard you say.

HOWARD: At temperatures above 100 degrees. If the cladding is split, it
degrades instantaneously.

KADAK: Instantaneously. But, what about this 100 degree thing, you had a
chart that said if it\'s above 100 degrees, you don\'t get something
going on. I forgot what it was.

HOWARD: Dave\'s degradation chart? It was water, right.

HARDIN: Would be no transport.

KADAK: So, it degrades, but doesn\'t transport if it\'s above 100
degrees?

HOWARD: That\'s correct.

KADAK: So, we\'ve got a lot of boundary conditions on this stuff
actually getting into what we would call the environmental portion of
this problem, which is the movement into the water?

HOWARD: Yes, sir.

DUQUETTE: Maybe we\'re going to hear it this afternoon. Duquette, Board.
But, so far, I haven\'t heard anything about how much of this stuff is
going to reach the surface? How much of the radionuclides will reach the
biosphere, given 1 per cent of the clad packages will have failed
beforehand, only those will split, assuming that water gets into them,
and then that most of it\'s going to be tied up with colloidal
absorption?

HOWARD: Again, that\'s going to be scenario dependent. It\'s also a TSPA
question. We don\'t have TSPA results.

DUQUETTE: Okay.

GARRICK: Ron?

LATANISION: Latanision, Board.

If we could go to your Slide 6? I just want to make sure I have the
perspective clear. In the nominal case in terms of the waste package, am
I reading this correctly? You have general corrosion, microbially
influenced corrosion and stress corrosion cracking. Are they occurring,
but they are not penetrating, so you don\'t have a failure, or are they
not occurring?

HOWARD: They are occurring, but they\'re not penetrating in 10,000
years.

LATANISION: Okay. And, in the case of a seismic event, you say stress
corrosion cracking damage\--

HOWARD: It\'s a function of the peak ground velocity.

LATANISION: Now, why do you think stress corrosion cracking occurs in
that instance?

HOWARD: Because the waste packages can bang up against each other end to
end.

LATANISION: I mean, how does that implicate stress corrosion cracking as
opposed to just mechanical failure of the packages if they\'re banging
into one another? Where does stress corrosion enter into a seismic
event?

HOWARD: Well, you have residual stresses in this case, you know, we
stress relief the waste packages in their initial conditions. So, in
this case, we have now induced residual stresses, which can allow for
stress corrosion cracking to occur.

LATANISION: So, you take the position that the fact that these packages
are banging into one another does not cause them to fail because of
mechanical reasons, but it may in the presence of a corrosive
environment?

HOWARD: That\'s correct. And, we assume that corrosive environment is
there and they do crack.

LATANISION: Okay. Yes, that\'s a good point. I mean, stress corrosion
cracking phenomena are called delayed failures because they don\'t occur
instantaneously. There\'s an initiation and a propagation stage. But,
this is obviously part of the conservatism of the process; is that it? I
mean, I would not have thought of stress corrosion cracking in a seismic
event as being a very realistic likely probability, given the time
element, the kinetics that are involved in the propagation, the
initiation of propagation of the stress corrosion crack.

HOWARD: Right. It will take some time to propagate, but on repository
time scales, you know, it\'s important to note that there is a lot of
aleatory uncertainty in that seismic event. We don\'t know the timing of
the event, when it\'s going to occur.

LATANISION: One other point. On the next item down, waste package damage
due to localized corrosion. The comment there is that it\'s very
unlikely to occur because the drip shields function, but I thought we
heard from Ernie earlier today that condensation may actually occur.
And, why is that not\--

HOWARD: That would be distilled water.

LATANISION: Right. Well, distilled water produced from, you know, the
constituents in the water phase, whatever they carry along with it would
be carried along with the water, and either the dust that\'s on the
surface of the canisters.

HOWARD: Okay. Ernie?

HARDIN: Ernie Hardin, BSC.

The dust has high nitrate, and there\'s no reason why the distilled
water contacting the dust would go to acidic pH. So, we will not see
localized corrosion under those conditions.

LATANISION: Okay.

GARRICK: Howard?

ARNOLD: I\'m frustrated by the answers on the cladding. Arnold, Board. I
got the impression this morning that you assumed that all the cladding
was damaged somehow.

HOWARD: We don\'t assume that. I think you\'re talking about the NRC\'s
presentation, Tim McCartin?

ARNOLD: Yes.

HOWARD: Yes. Tim I think could speak to that.

ARNOLD: He did mention that you all had different assumptions. That\'s
right. I do recall that.

The other issue with the stress corrosion cracking, a seismic event, as
has been said, is going to be a short time, unless it leads to stress
because things are leaning against each other, or something. I don\'t
see how it can be a factor in stress corrosion cracking.

HOWARD: We believe it will lead to residual stresses and it will dimple
these waste packages. They can bang up against each other, and they
could bang on the pallet. I think Mark Board went through a presentation
back in November where that was covered.

ARNOLD: Just one remark. The 1 per cent failure was used as a design
specification on our part to design clean-up systems for operating
reactors. It\'s very much higher than the actual experience rate.

HOWARD: Right. And, we use a distribution, so we sample on 1 per cent,
but it could also be lower than that. So, that\'s not a fixed parameter
at 1 per cent, up to about 1.2 if I recall correctly. And, that\'s added
to 1 per cent, which is stainless steel.

KADAK: Could you go to Slide 21? It was one slide that you said the
things you don\'t consider. Maybe it\'s the one after that.

HOWARD: Right before that, Slide 20.

KADAK: 20, yes. All the things that you don\'t consider I think are
pretty significant in terms of potentially holding back or retarding the
movement of these colloids, which I understand is one of the major
vehicles for movement. So, why did you not consider all those?

HOWARD: They were simplifications that we made to\--

KADAK: Are they hard to analyze? I mean, everybody knows that there is
some filtration going on.

HOWARD: Yes.

KADAK: All these mechanisms I think people know exist.

HOWARD: Yes. And, they are admittedly conservative.

KADAK: And, they could be significant relative to trying to meet even
your compliance standard.

HARDIN: Ernie Hardin, BSC.

Any sort of proof testing that you want to do in the laboratory with
colloids is, by its nature, time consuming. And, so, this is a
simplification that may have saved us a great deal of effort in the lab.

KADAK: Okay, I\'ll accept that. But, you spent a lot of time in the lab,
nine years dripping something on something, that may or may not be
important. So, I don\'t know where you\'re going with that.

GARRICK: This morning, the NRC gave us a dose profile for long-term
performance that included peak dose, and it was obvious that those of us
who also have seen the TSPA/LA results, at least of earlier models,
earlier versions, could see that the results were very different from
DOE\'s. And, we asked why, and, of course, the NRC model was an
illustrative model, and very well caveated, and we are not supposed to
read too much into it at this point. But, still, when asked about the
basis of the difference, one of the things that was mentioned was
release rate.

And, I guess I would like some elaboration on that, because I didn\'t
think release rate was all that important when you start talking about a
million year dose profile. But, if release rate is important for the
long-term performance, then it seems that the things that we have been
hearing about today relative to the source term need to stabilize some.

And, I guess my real question is when are you going to have a source
term that you really have confidence in in the kind of terms that we\'re
very interested in, namely realism? And, we know that you have somebody
working on the source term that has been asking that question for ten,
fifteen years, Rod Ewing, and now he\'s working on it, so there\'s no
excuse. So, when can we expect to see some stabilization of the source
term work?

And, also, comment on the sensitivity of the long-term peak dose, for
example, to release rate from the engineered barrier system.

HOWARD: Well, there were several comments and questions in there. When
are we going to have a source term that we have confidence in, the
answer to that from our perspective, we have a source term that we have
confidence in now, and that\'s for regulatory purpose.

GARRICK: And, that\'s where we have the chasm. It may be okay, because,
in a sense, you\'ve bounded the problem, and you end up with doses that
are in compliance. But, there\'s still this nagging concern that we have
about fundamental understanding of the source term. And, the comment was
also made by David, I think it was, that unless the S&T work
demonstrates that you\'re not conservative enough, probably won\'t
impact future TSPA\'s. And, I guess my reaction to that is supposing it
reduces the dose by a factor of 100, are you still\--obviously, are you
still going to ignore it as a part of the source term? The one you would
use in a performance assessment?

HOWARD: Yes, I\'m better at predicting the future at 10,000 years rather
than 10 years from now, which means I\'m not very good at all. Total
System Performance Assessments are iterative processes and they are
going to change over time. And, when the data becomes available for
regulatory purposes, and we\'ve got the validated models and it\'s
amenable to incorporation into the TSPA, you know, it\'s rational and
reasonable that we would incorporate it at that time. But, I\'m in no
position to tell you when that will be.

GARRICK: I guess the question partly is what\'s the convergence time
here? There must be a time\--well, you already answered it in one
respect by saying that you have a source term now that you\'re satisfied
demonstrates the compliance.

HOWARD: Okay. Mark?

PETERS: Mark Peters, Argonne.

Rod is not here, so I guess\--I\'ve only been working on it ten or
fifteen years, but I\'m the co-lead for the\--we\'ve got a program in
the S&T Program and source term looking at not only secondary phases,
but also a lot of the package type processes Rob just talked about, and
we thought about it in terms of a three to four year program, to give
you a time frame. We\'ve got the beginning of a task to start thinking
about how we would take all the interesting experiments and put into a
modeling framework. It might fit into a downstream TSPA as we iterate on
TSPA. But, that\'s the kind of time frame that Rod and I are talking
about in the program. Does that help?

GARRICK: That helps. Do you agree with the observation that was made
this morning, though, that the differences, these analysis results, one
of the major contributing factors is the release rate of radionuclides
from the engineered barrier system?

HOWARD: It makes sense to me.

GARRICK: It makes sense for 10,000 years, but I\'m not sure, I guess
what I\'m getting at is how much sense does it make for your peak dose,
which we\'re going to hear later is maybe on the order of 700,000 years.

HOWARD: And, that again may be scenario dependent as well. You know,
there\'s lots of aleatory uncertainty in these disruptive events and the
timing of them. And, how that plays out, until the calculations are set
up and run, I\'m loathe to speculate.

GARRICK: Okay. Okay, any other questions? Yes, Thure?

CERLING: Just as kind of a follow-on, but if we could go to Slide 1? I
really like the title of this talk, and it would make a very nice
caption to a figure, and I was wondering will we ever see a figure where
you could use that as the caption?

HOWARD: Noted.

CERLING: As long as we could remove potential.

GARRICK: Leon?

REITER: Reiter, consultant.

For the TSPA/SR, you convened an international review team to look at
it, and they looked at various things, and they had some strong
comments, some positive and some negative about what had been done. One
of the strongest negative comments they had was about the assumption of
a very thin continuous film of water that always exists that you have,
diffusive transport. I see it\'s still there. Have you looked at that at
all? And, what\'s been your conclusions?

HOWARD: Yes, we looked at that, and part of it is that we use the
absorption isotherm as a function of relative humidity. I think really
related to that is the idea that we don\'t consume any water in these
chemical reactions as these internals corrode. As I said, you know,
we\'ve done some sensitivity cases that would suggest that, you know,
you could get on the order of thousands of years of delay if you
accounted for that, or if there wasn\'t a continuous film.

Now, again, that assumption is probably more important in the nominal
early failure case than it is in the other cases where you can have
advective fluxes. So, we have looked at it, but we still haven\'t found
the basis to move very far beyond where we were with those site
recommendation models.

REITER: Just is the basis something that you believe it\'s there, or you
can\'t find proof that it\'s not there?

HOWARD: It\'s a recently cautious assumption.

GARRICK: Andy assures me it\'s a quick question.

KADAK: Relative to the failed fuel, most of the utilities know where
they have failures. Now, I\'m not sure exactly what they\'re doing
relative to packaging of these failed fuel assemblies. But, my hope
would be that they would put it in either some kind of zirconium clad
device that is corrosion resistant, so you\'d have the same performance,
or some other kind of system. Is that going to happen to reduce your 1
per cent number, because if you go all the way back and you work through
all the analysis, it says if you can eliminate that 1 per cent, and 1
per cent of whatever the millions of curies there is, that\'s a lot of
curies. So, what are you doing with canned or spent fuel, packaging them
any differently to make them more corrosion resistant?

HOWARD: Well, there\'s certainly no requirement for the utilities to do
anything with them before they package them and send them to us beyond
what\'s in their standard contract. And, we don\'t anticipate any
specific designs where, you know, we segregate them and we bundle them
up into a special package. That\'s why we developed the model the way we
did. In fact, you know, if the releases are going to be dominated by
those from seismic or igneous, I\'m not sure that would be a real
effective use of resources to attempt something like that.

KADAK: I think it might be worth checking just to see if it would make a
difference, because it\'s clearly a lot of curies.

GARRICK: Okay, thank you. Thank you very much.

Well, we have a little bit of a dilemma. One of the things I don\'t want
to do is push the public comment item on the agenda into the dinner
hour. So, what I guess I would like to do is what I say I would never
do, and that is skip the break and suggest that people take breaks
individually and as needed, but allow us to go forward. So, I think
we\'ll do that, and I guess that brings Bruce to the podium.

ROBINSON: Okay, I\'ll get started then. Thank you for inviting me to
make this presentation on radionuclide transport in the unsaturated
zone. Between myself and Bill Arnold of Sandia National Laboratories, we
will describe the radionuclide processes and models that are going to be
used in TSPA analyses, taking the radionuclides from the base of the
repository through the unsaturated zone\--that\'s my

piece\--and then Bill will pick it up and talk about saturated zone
radionuclide transport.

I would like to acknowledge co-workers on this effort from Los Alamos
National Laboratory, Lawrence Berkeley National Laboratory, who did a
great deal of work in flow modeling, which I believe the Board has heard
about in past meetings, as well as the radionuclide transport modeling,
and then finally, BSC folks who helped me also prepare the model that I
will be presenting today.

Next. This is an outline. I would like to start with conceptual models
for transport of radionuclides through the unsaturated zone, talk about
those models, what they are, what their technical basis is, and then
briefly describe how we go from those conceptual models to a numerical
implementation that\'s suitable for use in the TSPA analysis. Then, I
will talk about results, show some representative results, both in terms
of what I call a representative case, and then a series of sensitivity
analyses to look at parameters that are uncertain that we need to look
at in terms of what impact they might have on radionuclide transport.
And, those would include flow model parameters, diffusion processes and
parameters.

Then, I would like to show two other simulations, one related to whether
or not the radionuclides are released from the EBS system into the
fractures versus in the matrix. I\'ll show you the differences that our
models predict based on that. And, then, address the question of spatial
variability. Are there differences in transport through the unsaturated
zone, i.e. releases to the saturated zone, depending on where those
releases take place at the repository horizon.

Next. This is a conceptual model diagram. The repository, as we all
know, sits within the Topopah Spring welded tuff units. The diagram here
denotes that because of low matrix permeabilities and highly fractured
tuffs in the repository horizons, we believe the flow to be principally
within fractures of the Topopah Spring tuff that the repository sits in.

When you get down below the basal vitrophyre at the base of the Topopah
Spring tuff, you then get into the Calico Hills, which is either
non-welded and vitric, or altered to the zeolytic alteration. The main
difference there being mineralogic differences, but also hydrologic
differences in the properties of the matrix, which has an important
impact on the flow processes.

What all this boils down to is that depending on where you are beneath
the repository, the flow can be fracture dominated, matrix dominated,
with a little role played by the fractures, or something in between. So,
to capture that in a numerical model, we use a dual permeability model
formulation for both the flow modeling in the unsaturated zone, but also
the radionuclide transport modeling as well.

The processes included here in radionuclide transport are advection to
flow of water, also diffusion, sorption and colloid-facilitated
transport. I\'ll go into those in a little bit more detail, and also
show you sensitivity analyses describing what role they play.

Another aspect to consider here is that in the time scale over which
radionuclides are likely to be released from the EBS system, we assume
that the ambient flow field that would have taken place in some future
climate would not have any residual heat effects. There are still heat
effects in the rock itself, but the assumption is that the flow fields
within the unsaturated zone, for the purpose of radionuclide transport
calculations, can be developed based on the ambient flow conditions, and
how they change with climate. And, again, the releases in this model can
occur either in the fractures or the matrix.

Next. The scientific basis is basically summarized in this slide, and we
can get into more detail if you like. I split it into two categories
that I\'ll touch upon. One is fracture versus matrix flow. The other
being what are the flow and transport parameters that one uses? Both of
those require a description of the scientific basis.

So, in terms of fracture and matrix flow, we had experiments at the
field scale in vitric Calico Hills unit at the Busted Butte site several
years ago, which we believe is reasonable to conclude from those
experiments that matrix flow dominates in the vitric Calico Hills units.

In contrast, more recent studies in the ESF and Alcove 8, Niche 3,
confirm in the Topopah Spring tuff, confirm that the process of fracture
flow and matrix diffusion in those units is the right model to use as
the fundamental model for radionuclide transport in the units at the
repository horizon.

Furthermore, Chlorine 36 results have gone back and forth on this
program over a period of a decade or more. Basically, at this point,
these results suggest the possibility of fracture-dominated transport,
especially of conservative species, through the unsaturated zone. So,
given the fact that that can\'t be discounted at this point, and that
there is also other data in the course of those studies that suggests
the fracture-dominated transport is a reasonable mechanism, we assume
that mechanism for some of the units in the unsaturated zone beneath the
repository.

And, then, finally, I would just point out that this model that we\'ve
developed really is a combined fracture and matrix flow and transport
model. I think it\'s consistent with lots of studies that occurred in
various vadose zone sites beyond Yucca Mountain, and I think there\'s
nothing particularly unusual about any of the model assumptions that we
make in this model.

Related to flow and transport parameters, the process flow and transport
models are either informed in a formal way through direct calibration
using data sets such as the water content of the matrix or the matrix
potential, pneumatic studies, et cetera, or in a softer fashion by
consistency checks, where you do your model calibration and then you
check to see whether other data sets, typically chemical data sets, are
consistent with those flow model results, and those are the types of
data sets that would go into developing the technical basis for this
model.

Next slide. Okay, that\'s the technical basis. Now, how do we go from
that to a numerical model that\'s suitable for use in a Total System
Performance Assessment. I mentioned it\'s a dual permeability model for
flow. The particle tracking model that we use in this regard is also a
dual permeability model. It accounts for sorption and diffusion via a
probabilistic travel time delay type of approach, which I could go into
with more time, or if anybody is interested.

There is a full decay chain capability in the model whereby those decay
products are tracked through the simulation. Particle release locations
are related to upstream in the modeling sense, upstream analyses of the
engineered barrier system, which say that I have a given radionuclide
release rate at a certain location at a certain time. That is
implemented or put into the UZ system via particles that are released
across the repository horizon. So, that spatial variability piece is
included in the model, as well as whether the fracture or matrix
continuum receives the radionuclides.

Then, once particles travel through the system in a TSPA type
calculation, it\'s then turned back into radionuclide mass flux, which
is then passed to the saturated zone model.

A word on validation of this model. This is, in a sense, or in reality,
an abstraction model. We abstract from the UZ transport process model,
thereby inheriting the model validation steps that were undertaken to
validate or build confidence in that process model.

The model I\'m describing was compared directly to that one, and also to
1, 2 and 3 dimensional calculations for validation purposes.

HORNBERGER: Am I right that there is no drift shadow in any of this?

ROBINSON: There is no explicit modeling of the drift shadow, but as I
will show, there are drift shadow-like effects that come into the model
for radionuclides if they are released into the matrix.

This is a list of the radionuclides. The reason I put this up is not to
run through them all, but basically to say that I\'m going to be showing
conservative, typical conservative radionuclide result Technetium 99, as
well as a strongly sorbing radionuclide Plutonium 242.
Colloid-facilitated transport is also included. I will go into that in a
little bit of detail in a second.

Next slide. First, advective transport. We use directly the 3D steady
state, dual permeability flow fields that are developed from the UZ flow
model. The uncertainty in infiltration is captured in a discrete fashion
by a median low infiltration and a high infiltration case. The
climate-related variability is treated by kind of going along in this
direction, unfortunately somewhat out of order. You start with the
present day, go to monsoon, and then glacial transition.

So, the uncertainty piece is handled along here by discretely, sampling
from three discrete flow fields. The variability in time piece is
captured by using flow fields that are simulations of future climate
scenarios.

There is this assumption of instantaneous transition of the flow field
from one climate state to another. The water table also rises during
those future wetter climates.

I just mentioned the uncertainty from infiltration model. We do a lot
also of sensitivity analyses in which flow model parameters that really
aren\'t pinned down that well in flow model because of a lack of
relevant data to pin them down, we do explore those through sensitivity
analyses what impact that might have on our uncertainty in transport.

Next slide. Okay, the transport parameters beyond simply flow are
summarized in this slide. We basically are incorporating probabilistic
stochastically defined parameters to propagate the uncertainty through
the model.

Sorption, that would be the Kd. Bill Arnold will talk more about
sorption in his part of the presentation. Diffusion parameters are also
stochastically defined. The parameters themselves are based on diffusion
coefficient distributions derived from laboratory measurements. It\'s
not just the diffusion coefficient, it\'s also geometric and other
parameters related to the\--basically, what it boils down to is the
surface area available for diffusion between the flowing fracture and
the matrix. That relates to the aperture of the fracture spacing, and
other parameters which end up controlling that surface area.

GARRICK: You may have said this, Bruce, but are all the fractures
connected?

ROBINSON: No. They are all connected, but they don\'t all flow. And,
some of our parameters relate to basically boiling down to how many of
the fractures, of all the fractures that are in there, how many of them
are actually flowing. And, that\'s a model that I will describe in a
moment, at least I\'ll show sensitivity analysis for it that relates to
that.

There is also conceptual model uncertainty for how that diffusive
transport between the fractures and matrix occurs, as well as colloid
transport, which I will describe in the next slide.

This is a schematic that shows the various forms that radionuclides can
take. You have radionuclides in the aqueous phase, or sorbed to the
rock. But, then, the purpose of this slide is to describe how we capture
colloid-facilitated transport.

There are reversible sorption type colloids in which you have a
partitioning coefficient which says that a strongly sorbing, typically
radionuclide, will sorb the colloids as well as to the rock surface. So,
if that occurs reversibly, that can be an enhanced transport mechanism,
depending on how the parameters themselves play out. The parameter then
is the partitioning coefficient Kc that we talk about, which is the
product of the colloid concentration, and the Kd to the colloid, Kd for
the colloid and the radionuclide.

The reason this is colloid-facilitated transport is that if that\'s
occurring in a fracture, then during the time when it\'s sorbed to the
colloid, it can\'t diffuse, effectively cannot diffuse into the rock
matrix.

There\'s also irreversible sorption type colloids. And, what this is
meant to denote are radionuclides that are bound to colloids either so
strongly or intercollated into a colloid to the point where they
basically are irreversibly attached to that colloid. So, then, for those
types of colloids, you have advective transport without diffusion into
the rock matrix, size exclusion which prevents the transport of those
colloids into the matrix if the pore size distribution is such that they
wouldn\'t be able to go into the matrix.

We also have two flavors of this type of colloid, one in which we assume
that there is reversible filtration. The colloids move through the
fractures. They are slowed via straining processes, but it\'s assumed to
be a reversible process in this model. And, then, we also take a small
fraction of the colloid inventory and assume it to be transported
without retardation. This is the so-called fast fraction colloids.

Next. So, I\'m going to present results now. I want to describe what
you\'re going to be looking at, and how I\'m displaying the performance
of the unsaturated zone. Basically, I\'m showing normalized breakthrough
curves. So, what that means is we are looking at the model-predicted
distribution of arrival times at the water table.

So, for most of these analyses, we release at time zero, we call time
zero the moment at which we release these particles into the unsaturated
zone. We do that over the entire repository footprint, except when I
tell you that I did a point release. But, for most of the simulations,
it will be over the entire repository footprint. We introduce the
particles at time zero, and then the breakthrough curves that you will
be seeing are the cumulative number that arrive at the water table over
time.

So, you can look at this as the response to the system to a step change
in concentration spread over the entire repository, and that\'s the way
to think about these results. They are normalized to 1, all
radionuclides. So, we\'re not looking at specific concentrations of
individual radionuclides. We\'re looking at a normalized response for
each radionuclide that I\'ll show.

The other point that I would like to make is that the decay chains are
included in the next slide that I\'m going to present, but they are
introduced both as the radionuclide and also as a parent, which is
decaying. And, so, when you see in one of these simulations a
breakthrough curve going above 1, it\'s because it was put in as the
radionuclide itself, and also as the parent. So, therefore, that\'s
basically because of the choice of how to perform these simulations, not
a feature of the model per se, but just in terms of explaining what
you\'re about to see.

A few of the radionuclide breakthrough curves go above 1, and it\'s
because of essentially the combination of putting them in at the
repository, and then also gaining those radionuclides via decay of the
parent.

Next slide. This is a series of breakthrough curves normalized. There is
the 1 right there. So, it\'s normalized breakthrough versus time.
Glacial-transition climate, this is the climate state that we expect for
the majority of the 10,000 year simulation, the mean infiltration
scenario. That\'s mostly what I will show you today.

Walking from left to right in terms of earliest breakthrough to latest,
the colloidal special that we\'re showing here have the earliest
breakthrough. And, remember, they are introduced into the fractures,
they\'re travelling through fractures, they are unable to get into the
rock matrix, and, so, you get short travel times.

You would have to actually combine this with how much of the inventory
is in that state to really get a sense for whether or not that\'s an
important contributor to dose. This is the normalized breakthrough
curve.

Conservative and sorbing species have the ability to interact with the
rock matrix, and either just diffuse in there and be slowed down that
way, or actually sorb to the rock matrix. And, so, as you go from left
to right, Technetium 99 is in there, neptunium, and then you get into
the plutonium and some of the other species that actually have longer
travel times due to matrix diffusion.

That also results in a very much broader arrival time distribution,
which probably would filter through and have some impact on
concentrations downstream.

So, when these curves don\'t go to 1, it\'s because of radioactive
decay, which is included in these simulations. So, if the UZ is able to
hold up certain radionuclides long enough for decay to occur, then it
becomes a barrier for that radionuclide, a significant barrier.

HORNBERGER: The ones that go above 1 are due to, again, the decay chain
product?

ROBINSON: Yes. This is this business of putting in the radionuclide as
the parent and also the species itself. It\'s more or less an artifact
at how the calculation was performed rather than something that would be
in the TSPA analysis itself.

This is a slide that shows sensitivity to the infiltration scenario.
Technetium breakthrough versus time for three different scenarios, the
lower, mean and upper infiltration scenarios. I want to point out two
things here. The first is that a comparison of the red and the black
curve shows that the abstraction model being used for TSPA represents or
reproduces the model results from the process model, which is a vastly
different numerical scheme, thereby providing confidence that the
abstraction model is actually valid by inheritance of the fact that
the\--the abstraction model is valid because the process model is valid,
and it compares well to it.

With respect to infiltration itself, this basically shows what we
probably knew already, but that the infiltration scenario, uncertainty
in infiltration has a very large impact on the transport of all species,
including conservative species.

Next slide. This is an examination of a parameter uncertainty related to
the flow model, and that is the active fracture model parameter. I
mentioned that not all fractures flow, and that there\'s a surface area
available for diffusion between the flowing fracture and the matrix.
This parameter controls that. It\'s an uncertain parameter because we
don\'t have good data to really hone in on what that parameter is, and,
so, this type of analysis lets us determine how conservative or
optimistic our models are with respect to that parameter.

As a general statement, we do this with a lot of different flow model
parameters, and most of the flow model parameters really don\'t have
great\--the transport doesn\'t have great sensitivity to those
uncertainties. This is an example of one where we call it basically
moderate amount of impact on the breakthrough curve. As you go to
greater and greater diffusive areas for release into the fractures, you
get less early time breakthrough. The curves typically converge at later
times.

So, the TSPA models in the case of this parameter use flow parameters
that are on the conservative, but reasonable. End of that.

Next slide. This is a similar look at the active fracture model
parameter, but for a sorbing radionuclide, Plutonium 242. It also has
impact on plutonium. Other flow model parameter results is that
basically, a lot of parameters have kind of a low to moderate impact on
these breakthrough curves.

Next slide. I\'m going to move on to diffusion at this point. Now, I
show breakthrough curves for Technetium 99, the orange, and Plutonium
242 on the same plot. And, what we\'re varying in the series of three
simulations for each radionuclide is the diffusion coefficient. This
shows that the diffusion coefficient has a rather large impact on the
breakthrough curves. Basically, the higher the diffusion, the more
ability the radionuclide has to diffuse into the rock matrix, where flow
rates are much lower. They do eventually reach the water table, but with
a delayed breakthrough curve. And, those effects, in general, are more
dramatic for sorbing species than they are for conservative species.

Now, that uncertainty in the diffusion coefficient is captured in the
TSPA model as a direct parameter, stochastic parameter that\'s varied in
the TSPA.

Next slide. This is a slide which gets at the question do we have the
right conceptual model for this fracture/matrix interaction. So, I have
two sets of curves here. In this case, they are all Technetium 99, but
in the one set of curves, they are the curves I showed previously, but
with kind of the base case, dual permeability flow and transport model,
which basically has one computational grid cell in the matrix for every
one in the fracture. So, the gradient that you capture from fracture to
matrix is basically\--the gradient on which the diffusive flux from
fracture to matrix is based is very coarse. It\'s basically a quasi
steady model for diffusion versus the discrete fracture model, blue
curves, which basically do a better job at capturing the gradients of
concentration between the fracture and the matrix.

So, what\'s well known about this type of a comparison is shown here,
and that is that although the two sets of models converge at the later
times, early time behavior is predicted to be longer breakthrough curves
for this discrete fracture model than for the base case model basically
that we\'re calling the base case model, which is the dual-k model.
That\'s for technetium.

The next slide shows the same results for plutonium. When you also
combine that with sorption in the rock matrix, it has quite a dramatic
impact which of these conceptual models you choose. So, that plutonium
transport for the dual-k conceptual model are these curves. When you go
to the discrete fracture model, it gives you dramatically increased
delays through the unsaturated zone.

Next slide. This is a simulation which gives you a feel for what is the
variability in the breakthrough curve if you have the releases occurring
into the fractures at the repository horizon versus the matrix. And,
basically, it\'s a common sensical result, but it\'s quite a dramatic
one in which if you release the radionuclide into the fractures, you get
much shorter travel times than if it has to actually diffuse out of the
matrix and get to a flowing fracture before it can transport down.

There\'s other aspects to this simulation in the sense of the effect of
the diffusion coefficient itself has the opposite effect for the matrix
releases, i.e. if you have smaller diffusion coefficients, you will push
these curves out even further in time, which is the opposite of the
fracture release case.

Next slide. This is a simulation, my final one, and it gets to the
question of spatial variability. This is a depiction of the repository
footprint. The nodes in the numerical model are each of the points. And,
what we\'re doing here is we are releasing at a point, at each point
across the repository, we\'re releasing radionuclide, in this case,
Technetium 99, and we\'re saying what is the travel time to the water
table. But, we\'re doing that for the case of diffusion into the rock
matrix, and all those things.

So, this brings me to the point of what we\'re plotting here, and doing
in color coding here, is the median travel time. So, we have a
breakthrough curve, this is just a small reproduction of the previous
slide, we have a median travel time as the measure that we\'re using to
plot here, and it\'s basically the 50 per cent travel time on one of
these breakthrough curves.

This shows a rather dramatic variability across the repository in the
median travel time, depending on basically location where that release
takes place. And factors that enter into this are the percolation flux
variability, but more importantly, hydrogeologic variability. The extent
of fracture flow from repository to water table really controls these
results. So, they go from, in the case of the red, less than 10 year
median travel times, to locations in which the predictions yield travel
times of greater than 1,000 years. So, that gives you a sense for
spatial variability, which is one of the questions you had.

A point I\'d like to make on this is the way to think about how this
plays out in a TSPA calculation depends on the release, the release
term. If only a few waste packages in limit one, let\'s say, fails, then
this type of variability is really an uncertainty in the sense that the
travel times through the unsaturated zone are vastly different for that
one failing waste package here versus down here. However, if most of the
packages fail, the way this plays out is in a spread in the distribution
of arrival times, and all of the simulations that I showed you, with the
exception of one, previously include this spatial variability as part of
the breakthrough curves. So, I wanted to point that out just to give you
a better perspective on spatial variability.

Next. Conclusions. I have described the transport processes. We feel we
have a reasonable model that considers the relevant transport processes
for the UZ. The uncertainties that I showed to be most important are
infiltration rate, diffusion model parameters, and diffusion conceptual
model.

We also looked at certain flow model parameters, and they basically have
a low to moderate impact on the travel times. And, in those cases when
we don\'t directly propagate those uncertainties through TSPA, we take a
reasonably conservative approach for those uncertainties that are not
actually explicitly represented via uncertainty distributions in the
TSPA.

Then, for matrix releases, they quite clearly yield much longer travel
times. But, if you have fracture releases, a lower diffusion coefficient
enhances transport, gets it to the water table quicker, but for matrix
releases, that same lower diffusion coefficient will give you much
longer travel times. So, where the release occurs, fracture versus
matrix, has an important impact on how to interpret some of the
uncertainties, things like diffusion coefficient.

And, then, finally, I just showed you the spatial variability of travel
times, which results from the different percolation fluxes that we have
across the mountain, represented in the model, and also very
importantly, the hydrogeology across that part of the model.

Thank you.

GARRICK: Okay, Ali?

MOSLEH: Mosleh, Board.

Bruce, start with a couple of questions, starting with your conclusion,
Bullet Number 4. What sort of factors influence the decision on carrying
certain uncertainties parametrically or using a conservative approach?

ROBINSON: There\'s a variety of factors. Basically, parameters that are
kind of determined early on to have a low to moderate impact based on
logical technical arguments tend to be set off into the sensitivity
analysis world, if you will, rather than propagating uncertainties that
really aren\'t going to matter much through the TSPA. That\'s a
computational burden that we don\'t want to place on the TSPA analysis
because it might not allow you to look at uncertainties as well as you
could for things that really do matter.

Another category that gets to your question of this decision is in the
way the models are set up, some parameters are very much more simple to
propagate that uncertainty through directly in the TSPA than others.
Now, that\'s not a very great criterion, just because something is
difficult to propagate, you wouldn\'t want to leave it out of the TSPA,
and we don\'t do that. What we do is we take, for the ones that we take
that are kind of in that range where there\'s really not a very big
impact based on sensitivity analysis, we choose to leave those ones out
of the TSPA analysis, and handle them differently, handle them with
sensitivity analysis instead.

MOSLEH: And, then, maybe you can explain what you have on 18. I think it
was a modeling choice\--well, actually, I\'m sorry, on 17.

ROBINSON: I was going to say that would not be a good example. But, we
actually have that in the model for fracture and matrix releases. So,
that\'s in the model.

MOSLEH: How about 17?

ROBINSON: This is an example of when you\'re building these models, the
scientific, and also the quality assurance, criterion is that the models
be \"valid.\" In other words, you\'ve done your best to build confidence
that the models are consistent with all the available data, and the
model, which was developed first, that was made, calibrated and made
consistent with all the available data, is the dual-k model.

More recent analyses, and the ability to do this type of modeling better
has led us to a point where we believe that this model could also be
made \"valid\" in the sense of comparing it to data that\'s available
from the site, et cetera. However, that work hasn\'t been done yet, and,
so, just from a procedural, if nothing else, standpoint we go with this
type of model, the dual-k model, with the anticipation being that as
time progresses, and things develop further, we would tend to try to
also take other reasonable models that we believe are consistent with
the data, and use those as well.

GARRICK: Andy?

KADAK: This was an excellent presentation, by the way, it was very
impressive.

ROBINSON: Thank you.

KADAK: A couple of observations, and maybe you could help me understand
this as well. How well, I know you\'re doing a lot of sensitivities,
analyses, and the last chart showed a whole bunch of different zones in
the repository regime. How well do you know that to be able to do a
better job of predicting what the flow will actually be of the
radionuclides?

ROBINSON: Well, I think the thing to keep in mind kind of at the outset,
is there is no full up repository that we can compare to, and if we
could, we couldn\'t wait long enough to do a direct simulation. So,
everything is based on experiments that are possible to do, which tend
to be smaller in scale, to try to get what some of the fundamental
processes that occur are, and then a mechanism for coping with the scale
issues, the scaling in time and a spatial scaling. So, we can do
experiments that go over 20 meters transport distances, and time scales
that last the average time for graduate student. Five years.

And, so, basically, that gives you a picture for the nature of the
problem that we\'re dealing with in terms of how well we know mountain
scale radionuclide transport. To account for the fact that you basically
have that fundamental uncertainty, that leads you to needing to go to
parameter distributions and models that are reasonable, but not
necessarily overly optimistic. It tends to push you, unreasonably
perhaps.

KADAK: Towards the cautiously?

ROBINSON: Towards the cautiously reasonable; right. Can you go to the
sensible way to inform a\--

KADAK: We heard that in spades last time. Can you go to the next to the
last slide there where you showed the repository in multi-colors? There
you go.

Now, that\'s somebody\'s best estimate of the let\'s call it the
geological structure, whether it\'s fracture or matrix type of rock, is
that what we\'re seeing here displayed as results for travel time?

ROBINSON: That is a very strong factor in what you\'re seeing here.

KADAK: Okay.

ROBINSON: Let me talk to you about Busted Butte and the matrix flow in
Busted Butte. When you\'re talking about releases in this portion of the
repository, our geologic and hydrogeologic studies have shown that below
the Topopah Spring tuff, you have large sections of vitric Calico Hills
where it\'s probably matrix dominated flow, based on the field
experiments that we did. And, so, therefore, the way that plays out in
this type of a simulation is longer travel times because there\'s not as
much fracture flow. And, the opposite is true in the areas, in the more
northerly areas of the repository.

KADAK: So, are you saying you can capture the geology well enough to run
your models to be able to predict radionuclide flow in this unsaturated
zone? Is that what I\'m hearing you say?

ROBINSON: Yes, for the purposes of performing a TSPA analysis to get at
what the UZ system does for dose, or does or doesn\'t do for dose, yes,
that would be the data.

KADAK: Earlier, our NRC friend said that they assume a fifty-fifty split
between fractured and matrix type rock. I think that was like an
arbitrary selection. You\'re more sophisticated than that in terms of
being able to identify where the fractured rock is relative to the sort
of diffusion type rock?

ROBINSON: I can only speak to this model and what that\'s based on is a
developed over a large period of time, a representation of the
hydrogeology and the stratigraphy of these rocks, based on bore hole
measurements.

KADAK: Here\'s the bottom line question. The bottom line question is why
do you have such short travel times compared to the NRC which has
relatively long travel times, given all the sophistication?

ROBINSON: Again, I, I guess, wasn\'t smart enough from the presentation
I saw today to know what those travel times were in the unsaturated
zone. Keep in mind my travel times that I\'m presenting here are once a
release takes place. So, any precursor delay in that, I don\'t know
about in that other model.

KADAK: Well, if you look at these curves, it\'s fairly steep and over
100,000 years.

MC CARTIN: Tim McCartin, I could give a quick explanation. I mean, it\'s
all fractured rock in the unsaturated zone. In terms of the geology,
there are places where the Calico Hills vitric unit pinches out and
might be very thin, on the order of a meter or less. And, roughly
speaking, where the Calico Hills vitric unit, which has a very high
matrix permeability, so the water will be flowing in the matrix rather
than the fractures, it\'s about a fifty-fifty split in our model where
it\'s thick enough that we have included it in our model, versus other
areas where it\'s very thin and we have not included it for purposes of
the calculation.

And, so, where it\'s not present, you have a Calico Hills zeolytic unit,
a very low matrix permeability and the flow is primarily in the
fractures.

Now, the travel time distance differences I was talking about is
primarily obtained due to retardation in the saturated zone alluvium. In
terms of in our model, if I had to give you a best guess, where it\'s
fracture flow primarily, it\'s on the order of tens of years, and we
essentially typically take no credit for that, and just assume it\'s
instantly transported to the saturated zone.

Where the Calico Hills vitric unit is, and its matrix flow, travel times
are on the order of hundreds to possibly a couple thousand years. And,
if it\'s retarded, clearly, the nuclides would take much longer. But,
that\'s half of the repository. But, in the curves you saw, that was
primarily saturated zone, the delay time.

ROBINSON: Observation would be that\'s quite consistent with what I\'m
showing here.

GARRICK: Okay, Bill, and then Ron.

MURPHY: Bill Murphy, consultant.

You mentioned that the fraction of fractures that flow in your active
fracture models, one of your big uncertainties, they\'re a hard question
to answer, and I wonder if you could tell me if the Chlorine 36 bomb
pulse data could be used as a measure of which fraction the fractures
flow?

ROBINSON: There are a lot of reasons for the answer of no.

HORNBERGER: First of all, he\'s from Los Alamos.

ROBINSON: There\'s a lot of reasons no. Let\'s leave aside whether or
not the Chlorine 36 observations indicate that bomb pulse even got
to\--let\'s leave that aside, please. Even if you do that, you take for
granted that the measurements are valid, the original measurements by
Los Alamos. That\'s not really getting at what is going to control
radionuclide migration from the repository because (a) it\'s under a
different climate scenario than these simulations, and, in fact, the
simulations that TSPA will be at, that\'s one reason. Another reason is
that flow in these major features, if they really exist, may be the ones
we can see with Chlorine 36 and not the totality of what the flowing
fractures in the unsaturated zone might be.

I think generally, this model would say that the flowing fracture
spacing is much less, much smaller, in other words, there are more
flowing fractures than would be represented in the Chlorine 36 data of
Los Alamos, even if you take that data at face value. So, I just think
there\'s too many uncertainties, including the well publicized one
about, you know, reproducibility of that data, to use that data set in
that manner.

MURPHY: Well, you already answered my follow-up question concerning
different climates. But, I wonder what other sorts of field data you
would make use of to try to determine that?

ROBINSON: I think it\'s something that we\'re going to have inherent
uncertainty and basically will need to quantify that uncertainty. I
think it\'s a very tough question. You can do intentional inputs of
water, like we\'ve done for seepage and have done for other reasons, and
try to test that, look at just what is the, I don\'t know what the word
would be, topography of how that water moves through fractures, and some
of that has been done. But, it boils down to having to make a prediction
in a climate scenario that we don\'t have today, and that\'s a very
fundamental uncertainty that\'s difficult to get around.

MURPHY: Okay, I have one other very brief comment, and it has to do with
the figure you showed a moment ago, Number 17. And, here, you have a
case where you have a strongly sorbing radionuclide, and you\'ve
maximized your fracture matrix interaction in the curves to the right,
and you\'ve included your explicit fracture model to really capture it
as well as possible. And, the difference turns out to be 20,000 years
between one end member of this graph and the other. So, that seems like
a short time scale in general in the context of a million year
repository. So, can I draw a conclusion from this that the unsaturated
part of the transport pathway doesn\'t contribute much?

ROBINSON: Not really. And, the reason is that a million year transport
calculation is only for radionuclide that sticks around for a million
years. So, if a barrier like this one predicts travel times which are
large compared to the half life of the radionuclide, then, of course, it
is a barrier, regardless of\--you know, you can go to 10 million years,
and if the radionuclide is not around anymore, I mean, I\'m stating the
obvious, but I\'m placing a caveat on what you said, it\'s not always a
million year. It\'s radionuclide by radionuclide, a comparison of travel
times to half life that really will enter into the peak dose analysis.
Does that make sense?

MURPHY: Thank you. Sure.

GARRICK: Okay, Ron?

LATANISION: I have a generic question and I think if we look at Slide
11, that\'s the best. Many of these trancients appear to reach a steady
state, some do not, and I\'m looking in particular at the green squares,
which is neptunium, and the yellow triangles, which looks like it\'s
uranium. Is there something in the simulation that distinguishes those
two from the rest of the set?

ROBINSON: Here\'s the way we think of it. The fact that they\'re going
above 1, it means basically in the limit, they\'re going up to 2 because
they\'re being produced from a\--now, that\'s just an artifact of the
way this calculation was performed. But, basically, if they are still
going up\--well, here\'s one that\'s down here that\'s still going
up\--it\'s because some fraction of that radionuclide that\'s been
placed there at time zero hasn\'t gotten to the water table yet. So,
there is delay in the unsaturated zone, maybe not for the most rapidly
moving part of the radionuclide inventory, but for some of it.

Another reason for plateauing is that radioactive decay. Had this
radionuclide had a longer half life, it would have been going up like
the others, but it flattens out when the half life is essentially short
enough to where there\'s no additional arrivals at the water table at
longer times.

LATANISION: Okay. So, would you expect at some point, you\'re going to
see a plateau in the others, or not?

ROBINSON: Yes, they would plateau\--this is for a unit release. You
picture this as a constant release over the entire simulation. That\'s
the way to think about these simulations. And, furthermore, if you\'re
able to do numerical convolution in your head, these are transfer
functions, essentially, that for any input of radionuclides can be
convolved to get what the breakthrough at the water table would be.

LATANISION: Thank you.

GARRICK: Bruce, is the bottom line to your presentation that the process
models provide confidence in the abstractions used for the TSPA?

ROBINSON: That would be a conclusion, and that the TSPA representation
of UZ, i.e. the abstraction model, which is what I\'m presenting here,
is a realistic representation of how the UZ would behave, given all the
uncertainties that we also build into the model.

HORNBERGER: Just a quick follow-up. So, I remember from the last TSPA
reading AMRs and PMRs and a lot of this looks very similar to me, have
you presented anything that we haven\'t seen before, except for the
sensitivities?

ROBINSON: The sensitivities are new. The model is probably the same
model that you\'ve seen before, but there are, for example, the median
travel time is a function of space, location in the repository, same
model, new analysis to try to provide you what you asked for.

GARRICK: Speaking of models that we\'ve seen before, in an earlier
TSPA/LA, there was an appendix that delineated conservatisms, and I\'m
curious as to whether or not these are still conservatisms, or they have
been addressed. Let me just address a couple. One of the conservatisms
was the dual permeability grid is used rather than refined gridding
methods, such as the Mink method.

ROBINSON: Remember the plutonium simulations? That would essentially be
an examination of what the differences are with the dual permeability
model versus a more discrete fracture model. So, that gives you a sense
that, yes, the dual permeability model is a significant conservatism
that\'s currently in the TSPA model.

GARRICK: And, there\'s another one, the active fracture parameter is set
to a high value.

ROBINSON: Yes, I showed the active fracture parameter simulations, I
called it moderate impact in terms of the uncertainty. There was this
family of curves that kind of had a whip that kind of goes from this one
down to about that one, that kind of an uncertainty.

GARRICK: Right. And, then you had indicated a water table varying with
climate. And, in the past, it was they fixed it at a specific height,
and claimed that as a conservatism.

ROBINSON: That is still in the model. It\'s changeable, but at the
present time, it\'s changeable because you can vary quite easily what
the lower boundary of the model is, and basically tell the particles
when they\'re to leave and be in the saturated zone. But, the
simulations here and the ones that we anticipate to be in TSPA will have
the 120 meter water table rise, which is I believe what you\'re
referring to.

GARRICK: Yes. And, we already mentioned this one, but the dual
permeability representation of fractures assumes complete
interconnectivity, and you addressed that earlier. Is that potentially
something that could have a significant effect? I\'m trying to get a
sense of the conservatisms that still exist and whether or not they are
important.

ROBINSON: If we\'re talking about within a hydrogeologic unit like the
Topopah Spring tuff\--

GARRICK: Right.

ROBINSON: \--there\'s lots of fractures, there\'s a place for water
that\'s percolating down those fractures to go in an interconnected way,
so I wouldn\'t characterize that, if that\'s what you\'re referring to,
as a conservatism. I think it\'s realistic, that particular unit, which
is where a lot of this uncertainty kind of plays itself out. It\'s in
the Topopah Spring that a lot of this type of uncertainty really lies.

GARRICK: I guess I\'m still troubled by\--well, if I were a hydrologist,
I wouldn\'t be troubled, I\'m sure.

HORNBERGER: That\'s right, you wouldn\'t.

GARRICK: I really set you up for that. But, the physical aspects of this
whole unsaturated zone, the problem that disturbs me is that there\'s
very little water involved. Very little water involved. And, there\'s a
massive amount of material and surface area and high residence time, and
these fast pass-throughs, I just have a hard time conceiving it, as if
there was no absorption capability whatsoever in the fractures.

ROBINSON: I will say that our model mutes those rapid travel times
through the diffusion mechanism that I showed there.

GARRICK: Yes, but on the other hand, you also say that this model
doesn\'t impact much the way in which the TSPA\'s abstractions have been
made.

ROBINSON: Well, this is the abstraction model. I wouldn\'t say that this
model basically says that the UZ is no barrier. I think you have to go
radionuclide by radionuclide and do the type of analysis that we walked
through earlier, you know, talking about what is the travel time
compared to the decay half life to really decide on a radionuclide by
radionuclide basis whether it\'s a barrier.

GARRICK: Well, I thought that\'s what we wanted very much to do, yes.
Okay.

All right, any other questions? Yes, David?

DIODATO: Thanks for the presentation, Bruce. I just wanted to address
one comment you made regarding Slide 19. I was pleased, in Slide 19, to
see your recognition that hydrogeologic variability and percolation can
make the difference in terms of travel times. So, that\'s an interesting
result. And, you made the statement that you thought, well, if you only
have a few packages, or one package, then maybe the spatial variability
really represents some spatial uncertainty function that you imply on
that. But, that\'s assuming that the role of water in, say, waste
package degradation is, you know, it\'s not a factor.

However, if the zones where the reds occur are corresponding with zones
where the environment of the waste package might be more hostile to
waste package corrosion, then you couldn\'t separate out.

ROBINSON: That\'s an interesting point. I appreciate that comment. I
didn\'t mean to simply leave it at this, frankly. But, the bottom line
is it\'s in the TSPA model, that type of coupling. In other words, if
part of the reason for this rapid travel time is high infiltration rate,
there\'s probably an increased high percolation rate at the repository,
there\'s probably a coupled effect whereby packages may fail more
frequently in that location, and that\'s in the model.

GARRICK: Okay. Any other questions?

(No response.)

GARRICK: Excellent presentation. Thank you, Bruce.

All right, now we\'re going to hear from Bill Arnold.

ARNOLD: Okay, thank you. I\'ll be talking about transport through the
saturated zone. So, next slide, please.

This is the outline of the talk. I will give a quick summary of the flow
and transport modeling approach and abstraction approach used for the
saturated zone. I will go through some of the key processes that affect
releases, some examples of the technical bases for our understanding of
those processes, and some sensitivity studies to try to illustrate them
for the Board. We will talk about key assumptions and key uncertainties
associated with those processes, and then, finally, talk about the
affects of spatial distribution of releases from the unsaturated zone
analogous to what Bruce presented. This is to address the specific
question posed by the Board.

So, next slide, please. The general approach for saturated zone flow and
transport abstraction in the TSPA has not changed significantly within
the last few years. Consequently, I will summarize it at a fairly high
level. The primary basis for this is the three dimensional saturated
zone site-scale flow and transport models, and these are used to
simulate radionuclide transport to the accessible environment. We
defined four source regions beneath the repository, so we do take into
account that kind of variability in transport through the unsaturated
zone. And, those sources are modeled at point sources, however, that
point is varied from realization to realization.

We use the convolution integral method to couple radionuclide source
term from the unsaturated zone with saturated zone transport simulations
that are done ahead of time.

Downstream from the saturated zone, radionuclide concentration in the
groundwater is calculated by dividing the total radionuclide mass
crossing the boundary of the accessible environment by the
representative groundwater volume of 3,000 acre feet per year.

Climate change is incorporated into these analyses by scaling the
radionuclide mass breakthrough curves that are simulated, and they\'re
scaled in proportion to our estimated changes in the flux of the
saturated zone.

And, finally, there\'s a separate model, an abstracted one dimensional
transport model that\'s used for radioactive decay chains in the TSPA
calculations.

Next slide. This figure shows this three dimensional site-scale
saturated zone flow model domain. The upper part of the figure shows the
topographic relief with a satellite image draped on it. This is Highway
95 right here. The lower surface shows the simulated potentiometric
surface that\'s contoured, and the red lines are the simulated particle
paths through the system. This domain is 30 kilometers by 45 kilometers.
It goes to a depth of about 2,700 meters below the water table .

The numerical methods that are used here include a particle tracking
method that includes radionuclide transport processes of advection,
dispersion, matrix diffusion in fractured volcanic units, and sorption.

I should back up one step here and explain that the flow model that
forms the basis for the transport model is calibrated to site data,
including numerous water level measurements within the model domain.
It\'s also calibrated to estimates of ground water flux along the
lateral boundaries from the regional scale groundwater flow model.

To get back to the transport, these simulated flow paths in the
repository generally occur in the upper few hundred meters of the
saturated zone, and they cross the boundary to the accessible
environment approximately 5 kilometers west-northwest of the highway
intersection at Amargosa Valley. So, these particle paths roughly
simulate particle paths flow underneath the lower Forty Mile Wash across
Highway 95 right here.

Next slide, please. To go further in the abstraction process, the
uncertainty analysis is accomplished using the Monte Carlo method, and
we sample all of the uncertain parameters in both flow and transport
parameters. The result is multiple simulations using the saturated zone
site-scale transport model.

A total of 200 realizations have been performed for the TSPA. So, these
radionuclide transport simulations consist of radionuclide mass
breakthrough curves, assuming a unit mass input. And, this is similar to
what Bruce showed for his example sensitivity runs in the unsaturated
zone.

And, these breakthrough curves result in a library of breakthrough
curves that will be used by the TSPA model for the probabilistic risk
assessment analyses via the convolution integral method, which is really
just a numerical short-cut that is a computationally efficient method
that can be used in the TSPA.

This brings us to the first process that I wanted to discuss, and this
is advection of groundwater. And, I want to start with our conceptual
model of how this process affects radionuclide transport. Advective flow
of groundwater is conceptualized to occur through the relatively limited
volume of fracture networks in the volcanic rocks in the saturated zone.

These flow paths also go through the alluvium, and in the alluvium, our
conceptual model is that flow is more uniformly distributed in this
porous medium with some channelization occurring in the more permeable
strata.

Another feature of the system is that simulated groundwater specific
discharge generally increases along the flow path from beneath the
repository to the boundary of the accessible environment, and this is
due to the convergent nature of the groundwater flow system at the
site-scale.

And, the specific discharge varies from point to point in space, and
it\'s a function of local hydraulic gradient, permeability, the
anisotropy in permeability in the fractured volcanic units, and, to a
lesser extent, temperature. This really only applies deeper in the
model, where the viscosity of water is a function of temperature.

Next slide, please. This is some information that forms the technical
basis for our conceptual model of advective flow in the saturated zone.
On the left, is an example of this information for the volcanic units.
What\'s shown here is a flowmeter survey in one of the C-wells. On the
left is depth in meters. These are the geologic units in the bore hole.
And, these curves here are the flowmeter survey, and there are two
different survey methods used here, a spinner survey and an oxygen
activation survey.

What they show is a high degree of channelization of flow within the
well when the well is pumped. So, these are cumulative flows. So, 21 per
cent of the cumulative flow occurs in this zone in the Tram tuff, 6 per
cent here, 62 per cent in this zone and the Bullfrog, another 4 per cent
here, I\'m not sure I can read that, 8 or 9 per cent up here. And, these
zones in which significant flow occurs are separated by significant
distances, on the order of tens to hundreds of meters.

KADAK: I\'m sorry. Could you explain that again? What is it that you\'re
measuring? Are you pumping water, or what?

ARNOLD: Yes. The well is being pumped, and a spinner log, or a log is
being taken of the flow rate through the bore hole.

KADAK: To some other place, or are you just pumping water at a certain
depth?

ARNOLD: Well, just the pump the well, or down the well bore in this
case.

KADAK: You\'re sucking it?

ARNOLD: Yes. And, this is the cumulative amount of flow as you go up the
well.

KADAK: And, those regions are specific regions where there\'s apparently
more water than others; is that right?

ARNOLD: That\'s right. These are the regions where water is flowing into
the well.

KADAK: I see.

ARNOLD: Here is something like 62 per cent of the entire water being
produced by the well is flowing in this region right here.

KADAK: Okay, thank you.

ARNOLD: On the right, is just a picture of the alluvium. It has the
visual appearance of a classic porous medium. However, examination of
the outcrops like this, and testing at some of the Nye County wells,
reveals that there is significant variability in the hydraulic
conductivity of this material, and that is accounted for in our
uncertainty in the parameters in the model.

Next slide, please. Okay, now I\'m going to show some breakthrough
curves to illustrate the effect of advection of groundwater and our
uncertainty in advection of groundwater. And, Bruce Robinson went
through an explanation of what these breakthrough curves are, but just
to remind you, this is relative mass as a function of time that\'s
produced at the downstream end of the saturated zone at the boundary of
the accessible environment.

This is assuming a unit input at the upstream end from the unsaturated
zone. For these breakthrough curves, this is for glacial transition
climatic conditions, and there\'s no decay and no sorption for these
breakthrough curves.

The black curves are the breakthrough curves for all 200 realizations of
the system. So, each is a separate realization. These embody uncertainty
in all of the uncertain parameters that go into the analysis.

Now, I\'ve also conducted a couple of other runs just for sensitivity
here. If we set the advective groundwater flow rate at its 90th
percentile value, we get the solid red curve here. If it\'s set at its
10th percentile value, we get the dashed red curve here.

So, the conclusion is that our uncertainty in advection encompasses a
significant portion of the overall uncertainty in radionuclide transport
rates for non-sorbing radionuclide, anyway. And, this conclusion is
borne out by other sensitivity studies that indicate that our
uncertainty in groundwater specific discharge is one of the most
significant uncertainties we have in the saturated zone.

Next slide, please. The next key process I want to discuss is
radioactive decay. Now, these are similar breakthrough curves. This is
for Technetium 99, so this is non-sorbing radionuclide, and most of the
breakthrough curves that I\'m going to show here do not include decay
primarily for the purpose of better illustrating or isolating the
processes I\'m trying to illustrate in the sensitivity analyses. And,
such is shown on the left here. So, this is the entire suite of 200
realizations with no decay.

For Technetium 99, it has a half life of 213,000 years, and you can see
the effects of decay only occur at very long transport times. And, you
saw this on the breakthrough curve shown for the unsaturated zone. The
breakthrough curves flatten out here at very long times.

Next slide, please. And, it\'s good to remind ourselves of how dramatic
this effect can be, as Bruce pointed out, for a radionuclide such as
Cesium 137, which is highly retarded in the saturated zone. On the left,
are the simulations with no decay, and on the right, shown with decay,
which of course is essentially zero breakthrough because Cesium 137 has
a half life of only 30 years, which is very short relative to the
simulated transport time.

GARRICK: I don\'t know why you show that curve. It\'s so obvious, I
mean, nobody is worried about Cesium 137.

ARNOLD: Yes, this is kind of an end number.

GARRICK: Let\'s get to the ones that are important.

ARNOLD: Okay, next slide, please.

Okay, another key process is climate change. This is done in a fairly
straightforward manner. We have scaling factors of groundwater specific
discharge in the saturated zone for monsoonal and glacial transition
climate states, and these have values of 2.7 and 3.9 respectively.

These scaling factors are based on the ratios of the average
infiltration in the unsaturated zone site-scale flow model for these
climate states.

In the case of the glacial transition climate, this is also corroborated
by steady-state flow simulations using the Death Valley regional
groundwater flow model for glacial climatic conditions.

Next slide, please. The next key process is matrix diffusion, and Bruce
spent quite a bit of time discussing this process in the unsaturated
zone. I thought I\'d summarize here what we do know in the saturated
zone. We have an extensive database of laboratory-scale studies and
measurements of effective diffusion coefficients in the volcanic rock
matrix.

In the saturated zone, we also have a field scale demonstration of the
matrix diffusion process in cross-hole tracer tests at the C-wells.

However, there are significant uncertainties in this process. We have
uncertainties in the degree of channelization of groundwater flow in the
fractured volcanic units, as illustrated by that flowmeter survey that I
showed earlier. We have uncertainty in the flow porosity in the
fractured volcanic rocks, and we have uncertainty in the effective
diffusion coefficient due to uncertainties in chemistry and
uncertainties in the specific rocks encountered along the flow path in
the saturated zone.

Next slide, please. This is an illustration of the kind of experimental
data that we have to support these conclusions. What\'s shown on the
left here are some breakthrough curves from cross-hole tracer tests at
the C-well complex in the Prow Pass tuff. This is a normalized
concentration of the tracers used in the test, pumped over a distance of
approximately 30 meters between the wells, versus the time here in
hours.

On the right is a laboratory diffusion cell experiment for the Prow Pass
tuff, the same rocks through which the tracer test was conducted. This
is a concentration versus time within this diffusion cell experiment.
And, let\'s start on the figure on the right. What this shows is two
different tracers or solutes of bromide and PFBA. These have different
diffusion coefficients in the tuff, and they differ by about a factor of
three. The diffusion coefficient for bromide is higher than the
diffusion coefficient for the PFBA because it\'s a larger molecule.

Now, if we go back to the results from the C-well tracer test, this is
the breakthrough curve for the PFBA. This is the breakthrough curve for
the bromide, and this is the breakthrough curve for lithium, which also
sorbs on the tuff matrix.

So, two points here. First of all, there\'s a significant offset in the
peak between the TSPA and the bromide concentrations, the breakthrough
curves. Really, the only way to account for this offset is by the
process of matrix diffusion from the groundwater flowing in the
fractures between the two wells, and the relatively immobile groundwater
in the tuff matrix.

Now, lithium is even lower, and this is a demonstration that the
sorption process is also operating here, where diffusion into the matrix
is providing access to the matrix for sorbing radionuclide and then it
is also then sorbing, and the peak is lowered even more for the sorbing
radionuclide.

A couple of other indications are that there were flow interruptions in
the tracer test, and the concentrations of all of these tracers
rebounded during the flow interruption. This is a diagnostic
characteristic of matrix diffusion. It also occurred at this point here.

Next slide, please. Now, some sensitivity runs to show you the impacts
of matrix diffusion. Again, a breakthrough curve, a relative mass versus
time. The solid breakthrough curve here is our expected case for a
non-sorbing species. By expected case, I mean that we\'re using the
expected value for most or all of the uncertain parameters used in the
analysis in the transport model.

The short dashed line here is our minimum diffusion case. This is with
the diffusion coefficient set very low, so there\'s essentially no
matrix diffusion occurring. So, all of the transport time out to this
short dashed curve is accounted for by transport through the alluvium,
or delay through the alluvium.

The long dashed curve is the maximum diffusion case. This is with the
diffusion coefficient set high enough that there\'s essentially full
diffusion into the rock matrix. So, this is the envelope of behavior
that can occur with regard to matrix diffusion for a non-sorbing species
in the model. And, what you should note here is that the expected value
case includes significant matrix diffusion, but a majority of the mass
arrives at times closer to the minimum diffusion case.

And, looking at the entire range of uncertain parameter values that are
used in the transport simulations, we actually span this behavior for
minimum diffusion to maximum diffusion among those 200 realizations.

Next slide, please. This is a similar sensitivity run conducted to
illustrate the impact of sorption in conjunction with matrix diffusion.
The solid black line again is our expected case for a non-sorbing
species. The solid blue line is using the same parameters but for the
sorption coefficient of neptunium in the volcanics. So, this is the
amount of additional retardation due to sorption in the volcanic matrix
only.

And, then, the dashed blue line is neptunium breakthrough curve for
sorption in both the volcanics and the alluvium. So, for these expected
parameter values, the sorption for neptunium in the alluvium provides
the majority of the retardation for neptunium.

Next slide, please. To look at an even more highly retarded radioelement
like plutonium, this is a similar sensitivity study. Again, the expected
case breakthrough curve\--now, this is for plutonium that is reversibly
sorbed onto colloids, so this is a colloid-facilitated model. Here, the
solid blue curve is for sorption in the volcanic matrix only, not in the
alluvium, and the dashed blue curve is sorption both in the volcanics
and in the alluvium.

So, for expected parameter values for neptunium, sorption in the
alluvium provides additional retardation of plutonium, and that\'s
particularly true for the mass that arrives before the 50 per cent of
the breakthrough value.

Next slide, please. Now, I wanted to describe next, the next key process
is sorption, and our conceptual model for sorption. Our conceptual model
is that we have local equilibrium between radionuclides in the aqueous
phase and the aquifer material. In other words, the sorption reactions
are rapid and reversible.

We have a linear relationship between radionuclide mass on the solid
phase and the aqueous phase. This is the Kd approach, linear Kd
approach. And, we recognize that sorption reactions are influenced by
the local chemical conditions. And, this includes water chemistry, such
a pH, Eh, concentration of carbonates, and so on. It\'s also influenced
by the rock types, the mineralogy of those rock types, devitrified
versus zeolitic tuffs in the alluvium, and potentially influenced by
radionuclide concentrations.

So, these factors and scaling considerations are taken into account when
we assess the uncertainty in sorption coefficients in setting our
uncertainty distributions for sorption coefficients in the transport
modeling.

Another point is that oxidizing conditions are assumed in the saturated
zone. This is a topic that we have discussed with some members of the
Board recently, and I will have more on that later.

Next slide, please. This is just an example of some of the sorption
data. There is a large amount of laboratory sorption data that are site
specific that are available for Yucca Mountain. This example is for
neptunium onto devitrified tuffs, where we have the neptunium Kd, these
are batch sorption experiments, as a function of experiment duration.
This gives some idea of how long you have to conduct these experiments
to achieve equilibrium. And, you can see that we\'re talking about
sorption coefficients in the range of 1 to 10 milliliters per gram for
neptunium. This is under oxidizing conditions, open to the atmosphere.

The other example is for plutonium onto devitrified tuff. Again,
plutonium Kd is a function of experimental duration. Now, we\'re talking
about sorption coefficients approximately in the range of 100 to 1,000
milliliters per gram for plutonium.

Next slide, please. This is an example of some of the data for sorption
in the alluvium from samples obtained from the Nye County Drilling
Program. These are for neptunium and for uranium. You can see that there
is some variability within the samples that have been taken from the
alluvium. But, we do have a good technical basis for our uncertainty
distributions for neptunium and uranium in the alluvium.

Next slide, please. Now, to look at a similar sensitivity analysis to
get an idea of the relative importance of neptunium sorption. You can
compare this to the sensitivity study that I showed for specific
discharge. So, these are our breakthrough curves. This is the entire
suite of 200 breakthrough curves for neptunium in the saturated zone,
shown with the black curves. And, then, the red curve is using the
expected value for all other parameters, and the 10th percentile value
of neptunium sorption coefficient.

The red dashed curve is using the 90th percentile of the neptunium
sorption coefficient. This is in both the alluvium and the volcanic
units.

So, our uncertainty in neptunium sorption encompasses really only a
moderate portion of the overall uncertainty in the radionuclide
transport rates. And, this is because we have a pretty good data set on
neptunium sorption, and we have a fairly high degree of confidence in
what those neptunium sorption coefficients are in the system.

Next slide, please. Now, to go through some of the key assumptions. Our
current information forms the basis for the following key assumptions
with regard to radionuclide flow and transport, groundwater flow and
radionuclide transport in the saturated zone.

Steady-state groundwater flow in the saturated zone. We assume an
instantaneous change in saturated zone groundwater flux with climate
change, and no change in the flow paths for the saturated zone.

Matrix diffusion occurs from uniformly spaced, parallel fractures in the
fractured volcanic units, as implemented with the analytical solution of
Sudicky and Frind. This is an obvious simplification of the system.

Equilibrium, linear sorption occurs in the tuff matrix and the alluvium.
There is no sorption of solutes on fracture surfaces or coatings.
Radionuclide mass from the fractures and matrix flow in the unsaturated
zone is input to the saturated zone in the fractures. So, the sum of the
radionuclide mass arriving at the water table in these two continua in
the UZ model is put into the fractures in the saturated zone for
transport.

Next slide, please. We assume oxidizing conditions in the saturated zone
with regard to sorption coefficients and solubility limits of
redox-sensitive radionuclides, such as Technetium 99 and Neptunium 237.
And, we do have some indications that local reducing conditions may
exist in the saturated zone. These reducing conditions can have a
dramatic impact on both the solubility and the sorption coefficients for
Technetium and Neptunium.

We have a high degree of uncertainty regarding the distribution of those
redox conditions in the saturated zone, but as I said, this is something
we have discussed with some of the Board members in more detail.

For transport of radionuclides reversibly attached to colloids, we
assume that there\'s local equilibrium among the colloids, the aqueous
phase, and the aquifer material. For radionuclides irreversibly attached
to colloids, it is assumed there will be no desorption of radionuclides
from the colloids. So, we assume that this attachment is permanent, at
least for the duration of the residence time in the saturated zone.

And, colloids are subject to attachment and detachment from the mineral
grains in the aquifer, and are, thus, retarded in their transport
through the saturated zone. But, there is no permanent filtration of
colloids in the saturated zone.

Next slide, please. This is just a list of the parameter uncertainties
in the saturated zone abstraction modeling. Here, I categorized these
into uncertainties in groundwater flow and geological uncertainty, and
transport uncertainty. We have uncertainty in groundwater specific
discharge, the degree of horizontal anisotropy in permeability in the
fractured tuff. We have geologic uncertainty with regard to the
alluvium/tuff contact in the subsurface, although this uncertainty has
been significantly reduced with the Nye County Drilling Program.

With regard to transport, parameters that influence matrix diffusion, we
have uncertainty in the flowing interval spacing, or the spacing between
fractures that conduct significant amount of groundwater in the
saturated zone, effective diffusion coefficients in the matrix, flow
porosity in the tuff. Then, of course, sorption coefficients,
dispersivity, effective porosity in the alluvium, the source location
beneath the repository, colloid retardation factors, sorption
coefficients onto colloids, and groundwater colloid concentration.

Next slide, please. This is an example of one of the uncertainty
distributions for this key uncertain parameter, which is specific
discharge, or this is really the multiplier in specific discharge as
it\'s applied in the abstraction model.

What this is is cumulative probability. So, this is CDF, and this is the
log of the specific discharge. So, our median value has a log value of
zero, so it\'s a multiplier of 1. Values greater than zero represent a
multiplier greater than 1, and less than zero, represent a multiplier of
less than 1. A devisor.

So, uncertainty in specific discharge is based on the results of the
saturated zone expert elicitation and on more recent well testing in the
alluvial tracer complex. And, the discrete cumulative distribution
function has 80 per cent of its probability between a factor of
one-third and 3 for the specific discharge multiplier.

And, this is really the range in uncertainty that was derived from the
tracer testing in the alluvium. However, we have residual uncertainty
with regard to specific discharge along the entire flow path, and that
uncertainty is accounted for in the tails of the distribution and these
tails are taken from the saturated zone expert elicitation.

Next slide, please. And, finally, this is that sensitivity study to
address the impacts of uncertainty in the source location beneath the
repository. So, what I looked at here, here\'s the repository outline. I
conducted transport simulations for non-sorbing species for four
locations near or beneath the repository. These correspond to the center
of those four source regions that are used in the saturated zone
abstraction model. The colors here, which I know don\'t show up on your
black and white copies, correspond to the four breakthrough curves shown
on the right here.

So, these blue flow paths correspond to the blue breakthrough curve for
source sort of in the northeast corner of the repository. Then, the
green is the northwest corner, and this other blue and purple, magenta,
correspond to sources in the southwest and southeast parts of the
repository. And, some things to note here. Generally, the transport
times are a little bit longer for source release to the further north in
the repository. That\'s because the transport distance is longer from
these locations to the boundary of the accessible environment. And,
also, releases on the western side of the repository seem to transport
somewhat faster than on the eastern side of the repository, particularly
the northeastern side of the repository. This is due to the flow path
which is further to the west, and it encounters less alluvium along the
flow path and, thus, has shorter transport times.

It\'s interesting to note that there\'s kind of a reverse correlation
with the pattern of transport times that Bruce Robinson showed for the
unsaturated zone. In the unsaturated zone, the longest transport times
occurred in the southern part of the repository down here. But, overall,
there is not a high degree of sensitivity to the source location
underneath the repository with regard to transport time, simulated
transport times in the saturated zone.

Next slide, please. So, in conclusion, I just described how the
calibrated three dimensional site-scale flow and transport models form
the basis for the abstracted radionuclide transport simulations for use
in the TSPA. We examined several key processes and their impacts on the
releases of radionuclide mass from the saturated zone. Of these, our
uncertainty in advection of groundwater is probably the most important
with regard to our uncertainty of transport and release rates from the
saturated zone.

And, also, spatial variations of releases to the saturated zone did not
have large impacts on the simulated releases from the saturated zone
relative to other uncertainties.

So, thank you.

GARRICK: Thank you. Dave?

DUQUETTE: Duquette, Board.

I\'ve been seeing these curves for about as long as I\'ve been on the
Board, about three years, and maybe I\'m finally starting to understand
them somewhat. Am I correct in assuming that basically you pick up where
Bruce left off, that is, they\'re simply additive. Your time zero is
when he has breakthrough from the unsaturated zone into the saturated
zone?

ARNOLD: That\'s correct.

DUQUETTE: Okay. And, if I take a look at your Figure 9, or Slide Number
9, at least some of the breakthrough occurs in less than ten years. And,
he also showed some breakthrough in less than ten years. Can I make the
assumption that you expect some nuclides to be transported to the
biosphere in about a 20 year period?

ARNOLD: You have to be careful how you use the word expect here, because
you actually have to look at this entire suite of breakthrough curves,
and what we\'re seeing here is basically our probabilistic assessment,
our uncertainty in these. So, one way to look at this would be that just
for the saturated zone for these simulations, it looks like there are
about four breakthrough curves that have a median breakthrough of less
than ten years. That\'s four out of 200. So, you might say that we have
a 2 per cent probability that transport through the saturated zone for
glacial transition climatic conditions could occur in less than ten
years.

DUQUETTE: Yes, that\'s what I thought I understood with the way the
curves are to be used. Doesn\'t that make it even more important, even
with a 2 per cent probability, that the containment of the radionuclides
in the engineered barrier becomes increasingly more important?

ARNOLD: I don\'t think you can make that direct comparison between these
results and the relative importance of other components of the overall
system.

GARRICK: Yes, Thure.

CERLING: Cerling, Board.

If we could go to Slide 23? So, now, where would you put your effort if
you could try to improve on the spread that Dave just alluded to in the
previous slide 9, or whatever it was, but where do you think your most
important area is in this long list of things that are the parameter
uncertainties that you could narrow down that very wide range of horse
tails.

ARNOLD: Well, to do that, it\'s a pretty subjective exercise, and I can
give a couple of ideas here, but I can\'t necessarily back them up, you
know, quantitatively. It\'s kind of a balancing act between how much
difference could additional understanding for that particular parameter
mean with regard to our transport simulations, and how amenable would
this parameter be to investigation through further work.

I would say that this is an important parameter of flowing interval
spacing. The degree of channelization of groundwater flow in the
saturated zone has a high degree of uncertainty based on the field
observations that we have at this point, and there are ways in which
that uncertainty could be reduced.

I think another one is specific discharge. Our uncertainty in the flux
of groundwater through the system is an area in which we could reduce
our uncertainty also. Those would probably be the highest priority
parameters that I would choose out of this list.

GARRICK: Bill, Andy and Ali.

MURPHY: Bill Murphy, consultant.

I was fascinated by the data from the C Number 3 well, the flow, the
channelized flow. I hadn\'t seen that before. And, I added up the
percentages, and it looks as if 100, actually 102 per cent of all the
flow is in the channels. Does this imply that there really is no flow in
the large intervening spaces?

ARNOLD: Well, there are limitations to the way in which this assessment
was made. These spinner logs, I think in particular the oxygen
activation survey, have limitations with regard to resolution. And,
again, there are other techniques that have a higher resolution that
could detect flow from fractures at a much smaller percentage than this.
So, this is not definitive, but these are the data that we have to work
with now.

Another thing to keep in mind that I sort of glossed over with regard
to, and this is on Slide Number 7, if you could put that up, is that
these features that are producing groundwater into the well bore are in
actuality probably dipping at a fairly high angle. You know, I think the
average dip angle for fractures and fracture zones is over 60 degrees.
So, the actual perpendicular distance between these is not as great as
the apparent distance shown by the separation and the well bore. So,
that\'s another consideration here. We did account for that in our
analysis.

MURPHY: That\'s a very important point, and I\'m glad you pointed it
out. It starts to address a related question I have. You said in your
particle tracking models, the transport is in the upper two or few
hundred meters. And, I\'m wondering to what extent the particle tracking
model accommodated the details of this channelized flow that\'s
recognized in the bore hole?

ARNOLD: Yes, the particle tracking itself uses a continuum
representation of the flow field. So, those flow paths do not include
the complexities of this channelization that we\'re inferring from this
information here. The transport simulations do account for it in the
dual porosity transport simulations. So, there is kind of a difference
here. The flow paths don\'t have the kind of detail that\'s shown here,
but the transport simulations do account for this in an implicit
fashion.

MURPHY: Would you expect, given the channelization that\'s evident here,
that flow is confined to a narrower top part of the saturated zone, or
would the dip on these channels mix at hundreds of meters?

ARNOLD: It would not necessarily be shallower or necessarily deeper. It
would probably be not as smooth as it is represented in the particle
tracking.

MURPHY: One final question. Are there comparable flow data for the
alluvium showing potential preferential flow paths, or channels?

ARNOLD: Again, in the alluvium\--in the particle tracking, the alluvium
is represented as a homogeneous medium. So, the particle paths are
fairly smooth.

HORNBERGER: He\'s asking about data.

ARNOLD: Okay, with regard to data, there are data that indicate that
there would be channelization, and we do have some preliminary data from
cross-hole tracer tests that suggest that the effective porosity of the
alluvium is lower than the total porosity of the alluvium. This would be
an effect of channelization. And, that hydraulic conductivity can vary
significantly between strata within the alluvium.

MURPHY: Thank you.

GARRICK: Thanks. There\'s a number of us that have questions, but we
have just simply run out of time, and I think we need to move on. And,
as a matter of fact, as a courtesy to the public members, I want to
sandwich in at this time the public comments, rather than having the
last speaker, if the last speaker will just be patient, we will come to
that after we have heard from those who have signed up to make the
statements.

I have two names here. They are familiar names. Is Sally Devlin in the
audience?

DEVLIN: Good evening, everybody. And, thank you all for coming to
Nevada. I hope next time you\'ll come to Pahrump and have your meeting,
in Nye County, the host county where I live, and I\'m from Pahrump, and
my name is Sally Devlin and I\'ve been doing this for 13 1/2 years. And
at my very first meeting, John Cantlon, who was the chairman then, he
said you\'re an idiot, go back to school. So, we did. And, thanks to
school, and I say that you have really made my ego go to the moon,
because all you did today was talk about my colloids, which I introduced
in \'95, and the microbes.

And, of course science has really blossomed these last five years with
microbiology and all the rest of the stuff, and I really didn\'t know
how far it would go, but I see that almost every report has one or the
other, or both, in them. And, this is wonderful, and I really do
compliment you on it because the science of Yucca Mountain has improved
so in the last five years, it\'s been just simply wonderful. And, John,
of course, has been a wonderful leader. He\'s my pet. But, anyway, it\'s
so nice to see so many familiar faces, as well as so many new faces on
the Board. And, I knew there would be quite a bunch new ones.

But, anyway, what I have to say is basically this. I want to thank one
of the presenters for saying something nice about Nye County wells. When
Nick Stellasoto (phonetic) was alive, and he did a hydrology report back
in \'99 in Amargosa on Yucca Mountain, it\'s one of the best meetings
I\'ve ever attended, and he did actually go in the mine and put his
equipment in there, and, of course, it disintegrated. And, I always
remember that for those of us who were there.

The other thing, you keep talking about the TSPA and that will come, and
what have you. We all know that the licensing is off for another two
years, and I know you will be preparing another million papers. You were
kind enough to send me a thousand pages on twelve topics, and you got my
report on it, and I said basically the same thing I\'ve been saying for
13 1/2 years. I love your reports. They\'re very informative. I love the
science that you are producing. I love the sophistication with which you
present your programs, as the toastmaster, which I told everybody to go
and join, I give you wonderful evaluations, very few ahs, very few
pauses, and it\'s been a delight. But, again, you have done what you
always do, and that is you are modeling.

And, I don\'t know how long, Mark said you\'re going to do his thing for
another five, six years, and so on. Now, I\'m 76 and I\'m on borrowed
time and this is my 13th and a half year here, and I\'m saying to myself
I might not be around for those five or six years, and then who\'s going
to come and yell at you, as I always do. Because I really resent the
modeling. I want some reality. I want to see a canister. You haven\'t a
sign of a canister. That\'s a lynch pin component. With the Alloy 22, my
bugs ate it immediately. I have all those reports. So, there\'s the
bugs. You have nothing for a canister, and that\'s number one.

Number two, you have no design for the mine. You have no transportation,
and so on and so forth. But, mainly the canister. If you have nothing to
put the waste in that will last less than 100 years, and I\'ll even give
you 50 years, I\'m very generous, I\'ll even give you another three,
Russ. Hear that? And, I do mean that, and it bothers me as the public
because as the public, and there are very few of us who have really
taken the interest, who are deeply affected, when we hear nothing but
modeling and not reality, reality is real, and the 16 billion, which is
the last figure that I saw for the monies spent on Yucca Mountain, I
think that is thoroughly unacceptable for continuing modeling.

I understand that you\'re doing the geography and the water flow, and
all that wonderful stuff. But, that doesn\'t make a canister. It
doesn\'t make the metallurgy that is really needed to be implemented and
put in the mountain and tested. And, that is number one with me. So,
that\'s what I\'m going to leave you with. You\'d better do a canister,
and you\'d better have a better alloy than 22 that my bugs will eat.
And, I don\'t know where it\'s going to come from, because the steel
won\'t work, I don\'t care whose steel it is, zirconium or whatever.
Remember my diamond? I thought zircaloids were diamonds. But anyway, I
did, they were zircons; right? Same family.

But, anyway, I\'m just saying that this science has got to stop that
you\'re doing now and do some real stuff. Do at least some kind of test
on some kind of thing. All these labs have been getting rich over the
years. You all, I hope, have had a very comfortable marvelous living,
but as I say, I want to be here to see something really in fruition, and
I don\'t get that feeling.

So, that\'s about my statement. I hope you heard it, because those of us
who have been on this project, as I say, for a very, very long time,
we\'re not going to be here. Our children are not going to be here to
fill up a Yucca Mountain with 77,000 metric tons of waste. It will take
100 years, and there will be a need for the second Yucca Mountain, and
I\'m looking at the other reports that I get from my wonderful friends
at NRC about on-site storage, 34 plants are putting it on site, and so
on and so forth. Will there be a need for Yucca Mountain? That\'s
another question. And, where will it go?

I heard President Bush last night talk about Yucca Mountain, and whether
it was positive or negative, I don\'t know, but he had an open mind, and
I think the open mind was hurry up, guys, you\'ve spent too much money,
we don\'t have the money. We\'re in serious financial straits, and we\'d
better get something done.

So, my suggestion is work on the canister immediately, start testing in
the mine, or something, or if you really don\'t have a lynch pin
component to do Yucca Mountain, then I think it should stop. And, so,
anyway, we are all sitting on banana peels and we\'ve got to do
something positive, and I think it\'s wonderful that you model, but I
want reality.

Thank you.

GARRICK: Thank you, Sally. We always enjoy your comments.

Our next speaker will be Grant Hudlow. Grant?

HUDLOW: Hi. I\'m Grant Hudlow. For those of you that don\'t know me,
I\'m a chemical engineer with nuclear engineering training and
experience. And, I have some good news and I have some bad news. The
only reason I have bad news is because we have new members on the Board
that I think need to hear it. The others have all heard it. And, I have
some very good news.

You have a valuable resource in John Garrick. He breathes the rarified
air that only industrial turnaround experts can access.

The bad news, we have 20 years of junk science, and Congress jumped all
over DOE in the Nineties, and it just goes on and on and on.

Investigation into the murder of Paul Brown shows that gangsters got a
large amount of the Yucca Mountain money. What they did was they bid on
projects, did a phony report, got paid, and then went and did another
one. And, the whistle blowers are now confirming that sort of nonsense.

Number three, there is no known material on the planet that can
withstand neutron imbrittlement for over 100 years. And, number four,
the California rate payers want a refund because of fraud.

Well, the good news is that you can run Yucca Mountain as a warehouse
and replace the waste packages every few years. The other good news is
that industry has now solved the waste problem by building high-rise dry
storage facilities. And, as a result, they are building new nuclear
facilities to generate electricity.

But, just a brief summary, I know we\'re out of time, and thank you for
bringing your brains to the backwoods of Nevada. I always appreciate
seeing you again.

GARRICK: Thank you. Thank you, Grant.

Are there any other comments? Yes.

TREICHEL: I have one question. Judy Treichel, Nevada Nuclear Waste Task
Force. Can I get either the backup or the additional slides that were on
Ernie Hardin\'s presentation. He had additional slides from what were in
the package.

GARRICK: Sure. Any other comments?

(No response.)

GARRICK: All right, I guess we\'ll go into our final presentation on the
Management and Technical Support Peak Dose Sensitivity Analysis by Mark
Nutt.

NUTT: It always seems that I get the opportunity to talk to the Board
when I get the pleasure to develop simplified models. All these other
people are talking about very complex processes, but my pleasure is one
about five years ago that I got to talk, and then today about simplified
models we put together. So, maybe my thought would be let\'s not do
anymore simplified. We\'ve got to keep making more complexity into it.

In my overview, I\'m going to talk about the objectives and limitations
of the analysis we put together. I\'ll give a quick overview of the
model approach, talk about the features, events and process evaluation
we went through to come up with this sensitivity analysis, the key
assumptions, the modeling approaches, and finally, results.

Next slide. Before I get into it, it\'s the big caveat. The objective
was to develop a scoping-level simplified model to identify those
factors that influence the repository performance over the period of
peak dose. This is not a compliance model. The results are informative
in nature only, and should not be used to compare to any proposed or
final regulations.

This effort was initiated well in advance of the issue of the proposed
rules 40 CFR 197 and 10 CFR 63. In fact, it was started well over a year
ago, we started looking into this effort. No attempt has been made to
reconcile the approaches, methods and the sensitive analysis with those
proposed revisions. And, an assessment of post-closure repository
performance for demonstration and compliance with those rules will be
done once they are finalized.

I\'ve been told early that the Board is somewhat disappointed at the
lack of result curves that are in here. To generate the results, or to
satisfy the objectives we were trying to meet with this effort, we did
have to generate result curves. They\'re in the report that\'s publicly
available. I invite anybody to go look at them. But, the reason I
didn\'t include them is I wanted the emphasis not to be on the curves,
but the differences between them in the sensitivities we found. And,
I\'ll be happy to talk to, and I will at the end of the talk, to what we
found out regarding the sensitivities, what are the key factors related
to repository performance over this period.

Next slide, please. In the overview, part of the reason we started doing
this and we looked into this is we, as the Board recalls, or has seen,
we\'ve done estimates for the FEIS over the period of the peak. We
hadn\'t done anything since then, as you\'ve seen all the models we
talked about today refer to 10,000 year period. So, the question came up
given the new changes we have had, what could be the potential impacts
over the period of the peak? So, we started doing this.

In doing it, we considered the FEPs, the features, events, the
processes, that were evaluated over the 10,000 year period. We developed
a simplified model that includes representative FEPs that could
potentially affect repository performance over the period of the peak
dose.

FEPs that either have a minor or no effect on the peak dose were either
not included in the model or were included in a bounding representation.
And, even though some of the FEPs may have an influence on the
repository performance over the 10,000 years, they may not have or
don\'t have an influence over the period of the peak, and can be
excluded or be included in the model in what we call the bounding
representation. And, that\'s that we took no performance credit for
those FEPs or series of FEPs, and I\'ll talk to some of those later on.

Next slide, please. We used documentation current as of early 2005 in
this effort. We also used historical information to essentially bridge
the gap between the 10,000 year analyses that were put together that you
have seen today, and the period of the peak. As Russ Dyer mentioned this
morning, we are looking at revising some of those models, and some of
the inputs that were used in this analysis may be revised.

This piece of work won\'t be revised. It\'s essentially served its
purpose and it\'s done.

This sensitivity analysis used the simplified model, was similar to a
performance assessment in 10 CFR 63. It\'s a fully integrated
system-level model, but what it is not, it\'s not a full TSPA. We
didn\'t include the level of detail in the various processes that you
would see in a full TSPA. And, you will see some of that later on when I
start talking through the model approaches we took.

Next slide, please. In order to look at the factors that were important
over the period of peak, it was necessary to go back and look at the
features, events and processes, the FEPs. So, we began evaluating the
exclusion arguments that were conducted over a 10,000 year period
against a longer time frame.

We did, when you find the report, or look at it, you won\'t see an
exhaustive evaluation blow by blow, FEP by FEP. We just didn\'t do that.
We looked through and made some judgments and I\'ll talk to those later.

But, prior to talking about that, we want to look at what we call slow
and infrequent processes over the period of repository performance.
Those are essentially of secondary importance to the primary degradation
modes that are the most significant to repository performance over the
period of peak. These less significant role of slow and infrequent
processes, if they were to occur, would actually tend to spread the
release rate over time. They\'d be mechanisms that may give smaller
releases from the EBS that would\--we\'d have a total amount of
inventory. You\'d be pulling it back further in time, rather than
leaving it out or calculating it out at the time of the peak.

These types of FEPs are also not likely to occur over longer time
periods due to cooling of the repository environment. And, we feel that
such processes can reasonably be excluded from the post-10,000 year
assessment, based on low consequence.

Next slide. I\'ll give a few examples. One is the slow degradation
processes of the engineered features, the drip shield, the waste
packages and the pallet, and these types of processes include, but are
not limited to creep, thermal sensitization, hydride cracking,
consolidation, and non-seismic induced mechanical degradation. They are
very slow degradation processes that are accelerated at higher
temperatures or require higher temperatures to initiate.

If they are either not initiated, they\'re in the 10,000 year period, or
they are sufficiently slow over that 10,000 year period that they won\'t
have an effect over the 10,000 years. But, as you go out further in
time, the temperatures will tend to slow down, or tend to go down below
the 10,000 year\--the values that are over 10,000 years, so these
processes, the rates will tend to decrease or they won\'t initiate, and
therefore, you could argue that since they don\'t play a role in 10,000
years, they\'re not going to play a significant role beyond 10,000
years.

As I mentioned earlier, even if they did affect the degradation rate,
the modes generally result in small holes and cracks in engineered
features, which would result in slower diffusion or less of a release
than would occur when the engineered features are essentially grossly
breached by the corrosion processes, the general corrosion that occurs
at the time of the peak dose.

Next slide, please. Slow stress-induced degradation processes of the
emplacement drifts that could affect engineered barrier system
performance. These include, but are not limited to, drift collapse,
consolidation induced by drift collapse, and included in these processes
is the creep and static fatigue of the rock. And, this mainly is dealing
with the non-lithophysal units of the repository horizon.

And, again, the rates of these processes decrease with lower
temperatures and, therefore, they become more stable with the repository
environment. They are more significant at earlier times when the
repository is thermally perturbed and you end up with a thermal
perturbation and mechanical stress modifications. And, if they can be
reasonably excluded from a 10,000 year period, the decreased rate of
deformation over the long period of time, we feel we can reasonably
exclude them for the longer period.

Next slide, please. There\'s also infrequent stress-induced degradation
processes of emplacement drifts affected by seismic events. And, the
drift over the period of the peak dose is expected to be affected by
seismic events, and they can induce stresses that lead to deformation
and degradation of the emplacement drifts.

Because this is the most significant effect of seismic, it would be such
degradation as drift collapse. This degradation mode should be
considered included in a seismic scenario evaluation.

There\'s degradation processes initiated by seismic and volcanic event
sequences. We feel that evaluating the risks associated with such events
can be reasonably approximated by continuing the 10,000 year assessments
to a longer time period. And, we tend to believe in how we approach it
that this analysis is evaluated in an uncoupled fashion would tend to
maximize the likelihood\--or minimize the likelihood of diluting the
risk. Essentially, again, if you look at them uncoupled, you are not
getting early failure to pull inventory release out in time. You\'re
waiting and letting it all come at a later period of time when the event
occurs. You are not distributing the EBS type releases over a long
period of time.

Next slide. So, using that kind of background, we went through the FEP
evaluation, again, looking at all the FEPs that were in the various
reports that are out there that covered the 10,000 year period, and feel
that the vast majority of the screening justifications applicable to the
10,000 year period are appropriate over a time period that covers the
peak dose.

This tends to result from screening justifications being made on either
time invariant or on an annual probability basis, or low consequence
basis that is not affected by time. Essentially, we\'ve made the FEP
justification that doesn\'t have an aspect of time in it, and those tend
to stay out.

What we found is those that can be excluded from a 10,000 year
post-closure performance assessment can be and were excluded from this
sensitivity analysis. Those that need to be included in a 10,000 year
post-closure performance assessment were included in the sensitivity,
simplified sensitivity model either explicitly or implicitly through a
bounding approach.

Next slide. The remaining FEPs essentially fall into three categories,
and the first one was that the screening justifications for several FEPs
that could be excluded from a 10,000 year post-closure performance
assessment can continue to be excluded with some augmentation. But, in
this sensitivity study, we continue to exclude them, and in the report
that\'s publicly available, you can see the augmentation to the
arguments that were made.

A few of the FEPs related to seismic effects that can be excluded from
the 10,000 year post-closure performance assessment were considered
appropriate for inclusion into the sensitivity analysis, and they were,
and we\'ll talk to those in a little bit.

Some of the FEPs, as I mentioned before, that need to be included in a
10,000 year post-closure performance assessment can be excluded from
analysis conducted over the peak dose based on low consequences, and if
they have a negligible effect on the measure of the peak dose.

Next slide. The key assumptions we made in the sensitivity analysis, and
I\'ll go into more detail of these later, so I\'m going to walk through
them pretty quick. We assumed that integrated long-term average climate
state, which resulted in a slightly larger infiltration rate than the
glacial transition climate.

We assumed the repository percolation flux was equal to the average
infiltration rate. We assumed collapsed drift conditions for seepage all
throughout the repository. We assumed the seismic activity will result
in drift collapse throughout the entire repository, and I\'ll show the
impact of that in a bit.

We did not consider diffusive radionuclide transport within this model.
We looked at previous TSPA analyses, the FEIS, the TSPA for the site
recommendation, the supplemental science and performance assessment
analyses, and tended to demonstrate that the engineered barrier system
radionuclide releases via advection were several orders of magnitude
larger than diffusive releases at the period of the peak dose.

And, since that large of a difference, and we found that the peak dose
tended to be dominated by these advective releases, that we didn\'t
consider diffusive releases. And, if you make that assumption, the
degradation mechanism results in you not having to consider stress
corrosion cracking, which we believe right now is a diffusive release
mechanism, given the tightness of the cracks and the tortuosity.

This also tends to maximize the inventory, the radionuclide inventory
available when gross breaching occurs, and you get these advective
transport pathways. It leaves everything in the analysis and in the
model until you get the gross breaching, and then you can let it all
come out under much more I\'ll call them aggressive conditions.

Next, please. So, general corrosion is the only corrosion that we
considered in the sensitivity analysis. The key aspect of controlling
the peak dose is the formation of large openings in both the drip shield
and the waste package, leading to these advective transport pathways.
And, again, not considering these other smaller breaches, such as stress
corrosion cracking, will maximize the inventory available when the gross
breaching occurs.

This is the one that goes back to Tim McCartin\'s talk this morning. We
assumed instantaneous degradation of the waste forms. The period over
which the waste forms from our modeling efforts and data are small as
compared to the time frame of the peak dose. We also, based on the
results of our seismic consequence work, I felt over the period of the
peak, the cladding is likely to be completely degraded by seismic
activity. So, even if the waste package may not fail, or it may be able
to withstand the seismic activity, you\'re going to shake it enough to
assume all the cladding was breached. So, in this model, you will see
there is no instantaneous waste form degradation with no cladding.

For simplicity, we assumed immediate transport through both the
unsaturated zone and through the fractured volcanic aquifer region in
the saturated zone. I\'ll talk to this a little bit more later.

Next, please. It\'s a little more detail in the approaches. Again, we
used the long-term average infiltration rates for the various climate
states to produce the time-integrated average rate for the long-term
average climate. Essentially, as Tim McCartin mentioned this morning, he
took the climate model we had that had these immediate transitions, and
did a time-integrated average, and for the medium infiltration case, you
will see the value he mentioned this morning of 26.6. So, I had an
independent verification. It was nice.

You also see the range that we\--it\'s a slightly lower than what the
NRC proposed. We also assumed the average percolation flux equals the
average infiltration rate, and recognized that the spatial variability
in the percolation flux at the repository horizon has a pretty
significant effect on seepage properties and advective transport rates.

So, you see right here is a curve out of one of our model reports, the
UZ flow models and submodels. That curve gives a normalized cumulative
percentage of the repository as a function of normalized flux, you can
take this curve, use these values, and come up with those curves. So,
these are actually the distributions of percolation flux over the
repository horizon that were used in the model. The inset shows the
glacial transition climate. So, you can see we\'re actually somewhat
larger in terms of percolation flux than that one.

Next, please. Why is that variability important? It drives the drift
seepage model, and as I\'ve mentioned, we used the collapse drift
seepage model. If you go back to Jens\' presentation, he has a
comparison of both a slide of the non-collapsed and the collapsed drift,
and they are significantly different. You get significantly more seepage
for a given percolation flux for the collapsed drift seepage.

So, we applied the distribution of percolation flux to these curves,
essentially had lookup tables to calculate both a fraction of waste
packages that would see seeping, and the average value of\--the average
seepage rate over those waste package that experienced seepage. It\'s a
much simpler approach that\'s used in the seepage abstraction model. We
essentially did a spreadsheet approximations of that, did comparisons
back to that and realized we had to do some adjustments for spatial
variability and flow focusing, which we did. It\'s documented in the
report.

And, this figure shows you kind of the ranges, and you can see between
the collapsed drift and non-collapsed drift a fairly significant
increase in seepage fraction, and the amount of water seeping into the
drifts and contacting those waste packages. You can see you get up a
mean of 88, 74 to 88 on these number of waste packages contacted. So,
with that collapsed drift seepage flux, we get an awful lot of waste
packages seeing water under this end. There\'s quite a significant
amount of water, which plays a role later on.

Next, please. Another important piece of this is the in-drift
environment, and I\'m going to focus on the thermal hydrology piece. We
did use some long-term chemistry from the engineered barrier system
physical and chemical environment model, and used very long-term
in-drift chemistry to calculate both partial pressure of CO~2~ and the
invert, the pH in the invert.

More importantly is the thermal response to the repository and how
that\'s used. Essentially, this is out of the thermal hydrology model,
multi-scale thermal hydrology model, and this shows the range of
temperatures waste packages will go through for about 20,000 years. And,
you see they will go through a cool range and they will go through a hot
range, and there\'s actually a wide variation within here.

We took this data and used the historical information that showed at
about 100,000 years, everything went to about 21 degrees C, and just did
a simple extrapolation. And, what you\'re seeing here is kind of a
tabulated value of the low and the high, and that will play a role in
the waste package degradation approach we took that I\'ll talk to I
believe on the next slide.

Before waste package degradation, we\'ll talk how we treated the drip
shield. We used the probability distributions of general corrosion rates
from the long-term test facility measurements, and those were determined
in the analysis of them to represent uncertainty. There is both weight
loss and crevice specimens, and consistent with the modeling approach
that\'s being done for the 10,000 year case, we applied the weight loss
to the underside and we applied both the weight loss and crevice
specimen to the outer surface to calculate kind of an average corrosion
rate of this drip shield. And, just calculated how long it would take
given a realized value of the corrosion rate to penetrate, and what
you\'re seeing on this curve is a distribution of time that the drip
shield fails. And, since this all represents uncertainty, this is a time
that they all fail completely, and that\'s how we handled it in this
model. So, roughly 50 per cent at 20,000, 30,000 years. I can\'t quite
see that.

So, how the model would work is the drip shield, we would calculate the
corrosion rate at a given point in time, it would be completely gone
with respect to a barrier for advective transport.

For the waste package degradation due to general corrosion, we did a
simplified approach to represent the results that are described in the
WAPDEG Analysis of Waste Package and Drip Shield Degradation analysis
and model report. Again, we applied the probability distribution of
general corrosion rates from the long-term test facility. They represent
variability, and essentially are applied in this approach as variability
in corrosion rates across the waste package surface. They have roughly
1,000 patches that are sampled, and they sample over a probability
distribution.

There\'s also a temperature dependence that\'s applied, and we applied
it down to a limit of 45 degrees C. So, as the temperature decreased,
the corrosion rate decreased until we hit 45 degrees C, and then we held
it. So, even though the temperature of the repository is going down to
21, we held it at 45 degrees. We did an analysis of what would happen if
we let the thermal, the temperature dependence go all the way down to 21
degrees, and I\'ll talk to that in a little bit.

So, we determined the time for the initial waste package general
corrosion penetration, the very first failure for the coolest and
hottest waste package. Essentially took that curve, and used the
temperature dependence, used the sampled corrosion rates across the
repository surface, found out the first one that would fail, and then
using the temperature dependence, came up with a range of time for
initial waste package for the coolest and hottest.

We then assumed a uniform distribution of failures between the coolest
and hottest waste package. And, that gives these curves here. What
you\'re seeing is the percentiles of the curves, kind of like what Tim
McCartin was showing earlier, they are very steep, you know, several
tens of thousands of years, but there\'s considerable uncertainty and
you can imagine you\'re moving along, and then you\'re going to follow
up one of these curves, you\'re going to start the first waste package
failure due to general corrosion, and then the last one would be right
there. But, that period of time moves around.

ARNOLD: Is it correct or a slip of the tongue that the upper curve shows
a 50 per cent at 2 or 300,000 years, not\--

NUTT: Sorry, thank you. I\'m trying to read at a slant the fuzzy chart.
Thank you.

We also adjusted the average number of general corrosion breaches on a
penetrated waste package. This model has a method for calculating how
many, of the waste packages that are failed, what is the average number
of general corrosion breaches on it. And, it\'s calculated in this
report. They have an example calculation that goes down to the 21
degrees C. We adjusted it upwards due to limiting this temperature
variability, and that has an effect on the EBS transport model.

Real quick on the chemistry. We essentially, in order to drive the
solubility models later, we needed to come up with estimates of
in-package chemistry. Essentially took two approaches. We realized that
looking at the results of some of the historical documents, that it
looks like it\'s about a fifty-fifty chance with our modeling results of
will the waste package initially breach by stress corrosion cracks or
will it initially breach by a general corrosion penetration.

What we assumed is when it cracks, there will be enough time for the
internals to fully degrade when you\'ll get the first general corrosion
breach. So, we wanted to apply those chemistry conditions at that point
in time. When we assumed that the other 50 per cent of the time, we said
well, it\'s going to be a general corrosion breach that initially
degrades the waste package, so we wanted to apply waste package
internals chemistry that was more along the lines of what would be
controlled by internal waste package degradation.

And, you will see that in here. There is not a difference in the low
value of the pH. There is some difference as you head up into these
lower CO~2~ fugacities where you will get a higher pH value. And, in the
end, this really wasn\'t an important parameter.

Next, please. We used the results of past TSPAs. This is not a full
inventory case. You won\'t see technetium. You won\'t see iodine. We
basically used the results of past cases to kind of come up with
somewhat of a limit. It didn\'t turn out to be too limited. We used the
TSPA-FEIS to point out these radionuclides we considered. EPRI pointed
out in their report probably Thorium 229 and Uranium 233, and we got the
significant daughter products in there, and additional radionuclides in
the decay series. And, this shows the example of the inventories and the
radionuclides we used.

Next, please. Radionuclide release from the engineered barrier system.
Again, we considered only radionuclide transport out of the engineered
barrier system via advection. So, as I pointed out, there\'s a fraction
of the waste packages that don\'t see water. In a sense, those were not
considered. They did not have, since they were diffusive transport
pathway, we did not consider them. We only looked at the ones that were
flowing water and would calculate the dose based on their advective
transport.

We used a 1-D transport using mixing cells, very similar to the approach
that\'s documented in the Project\'s documentation, essentially a waste
form, a waste package, internals and invert, both what Dave Sassani
mentioned and what Ernie and Rob had talked about earlier. The only real
difference is we\'re only using one column. We\'re not discretizing
across the repository surface or area.

We applied solubility limits in each mixing cell, and full dose out of
the dissolved concentration limits AMR that Dave had mentioned. Again,
we used NpO~2~ as the controlling phase, unlike that AMR and as part of
the difference is things have evolved since early spring 2005, as we
used it in both the invert and in the waste package, unlike currently
where this is in the waste package, and Np~2~O~5~ in the invert.

There\'s others we looked at in sensitivity analysis that I\'ll talk a
little bit. As I mentioned early, it was a function of the in-package
environment. We looked at reversible sorption on both the waste package
internals and the invert. This is another difference. As Rob pointed
out, they aren\'t looking at reversible sorption on the internals
anymore. But, they are looking at reversible and that\'s not considered
in ours.

Next, please. Natural barrier system beneath the repository. Mainly for
simplicity, we assumed immediate radionuclide transport through the
unsaturated zone and through the fractured volcanic aquifer in the
saturated zone, essentially took it right up to the alluvium. And, as
Bruce had mentioned, this does depend heavily on radionuclide specific.

For moderately sorbing radionuclides the UZ breakthrough is comparable
to the waste package failure time. So, the assumption may not be too
bad. You essentially end up with a steady state where the release of the
moderately sorbing radionuclides, where the mass flux out, essentially
becomes equal to mass flux in. Also saw, and this is a conservative
assumption that we made, and probably may not be appropriate for things
like plutonium, if you look into our report, you will see that Plutonium
242 is one of the dominating radionuclides. That may not be the case if
we had done a full UZ or when a full UZ is done.

We saw the saturated zone breakthrough on the order of several tens to
hundreds of thousands of years for strongly sorbing radionuclides. Most
of the delay was in the alluvium. If you refer back to some of Bill\'s
talks, you can kind of see where the alluvium got you. So, it results in
a significant retardation of protactinium, plutonium and thorium. And,
we also expect significant retardation of them in both the UZ and the
fractured volcanic portion.

That\'s part of the reason that we are looking at these results, is to
look at the comparisons. Don\'t look at the magnitude, don\'t do any
comparisons to the absolute magnitude of the numbers, is things like
that are not in it.

We used the biosphere dose conversion factors for the glacial transition
conditions with the new guidance. We used what is in the current rule
for the dose conversion factor. So, it\'s updated. I know you were
having a discussion tomorrow about the\--this morning about the
differences between the two.

Next, please. Treatment of seismic disruptive events. This is where the
fun kind of began for me. As you heard earlier, the waste package damage
from vibratory ground motion is expected to result in an increased
susceptibility to stress corrosion cracking. And, over the 10,000 years,
the waste package is protected by the drip shield from rock fall. The
consequence of rock fall on the waste packages haven\'t been analyzed
because of the protection of the drip shield over the 10,000 year
period.

Well, we have analyzed the consequences of rock fall on the drip shield
and, again, it is an increased susceptibility to stress corrosion
cracking. So, you can make the extrapolation, I\'ll say, of the likely
consequence of rock fall on a waste package would be increased
susceptibility to stress corrosion cracking. Again, this is mainly in
the non-lithophysal unit where you can get the real big rocks to come
falling down.

So, although we feel that stress corrosion cracking is likely to be the
dominant failure mode, and it will result in those diffusive transport
mechanisms which are of lower importance than the establishment of the
advective transport mechanisms, we did a sensitivity analysis where we
assumed that both vibratory ground motion and rock fall resulted in
gross breaching of the waste packages.

We essentially considered multiple seismic events over 1,000 years. We
let them occur to a poisson process. We randomly sampled the magnitude
of the event, the peak ground velocity, and then calculated or estimated
the amount of gross breaching as a function of magnitude. And, how we
estimated that, the amount, was we took the area that these models
calculated as being damaged on a drip shield, and said that area,
instead of being a stress corrosion crack, is gone. So, we essentially
just added up and assumed very gross breaching, and then accumulated
consequences as we went along. So, the waste package could see multiple
events and get an accumulation of damage with more and more holes.

Next, please. The treatment of the igneous disruptive events, we did
consider two cases of that, and there was simplified representation
based on the approach described in the igneous consequence AMR.

For the igneous intrusion, we assumed one event every realization, with
the timing and magnitude of that event uncertain. So, it occurred
anywhere within the time frame of the million year period we were
considering. We calculated the number of dikes that crossed or
intersected the waste packages, failed those waste packages completely.
Let the rest of them kind of follow the general corrosion mechanism. So,
in a sense, I want to call it the nominal scenario, but we used our base
scenario of general corrosion going on for those waste packages that
weren\'t intersected by the event.

We sampled the uncertain parameters of how many waste packages would be
intersected, and calculated what the dose would be. And, for the
volcanic eruption, this was more of a confirmation. As Tim McCartin
mentioned this morning, the radionuclides are going away. We wanted to
do a simplified reproduction of volcanic eruption to see if any other
radionuclides might be building in out in the period of time.

So, we assumed, generated very similar to the approach being taken,
recommended for the performance assessment, essentially generated 1,000
realizations with each realization producing a dose history, with a
series of eruptive events, essentially erupting one each time step, and
then back weighted the probability out to come up with a risk curve
that\'s similar to what you\'ve seen before.

The results. Again, I really want to focus the discussion on the
sensitivities and not the magnitude and timing of the peak annual dose.
And, like I said before, those curves were calculated to get at what
we\'re really after, is what is important, what\'s driving things.

So, what we found is the peak annual dose really depends on what I\'ll
call the nominal degradation processes. The gradual degradation of the
engineered barriers and subsequent release of radionuclides contained
within them. In particular, the dose depends on the timing and rate of
waste package failure due to general corrosion processes, and the rate
the water transports radionuclides out of the EBS. That\'s the source
term. To me, it\'s not surprising that our waste package performance out
at that period of time has a role. If they don\'t fail significantly and
grossly, you won\'t have a significant release.

We feel seismic events will occur, and although the seismic induced
mechanical damage may influence the annual dose prior to the onset of
significant waste package failure due to general corrosion, it is not
expected to have a significant effect on the peak annual dose, either
the mean or the median.

And, what we mean by that is it will tend to control\--it can affect the
magnitude in the transient phases, but it will be when you get up to the
period where you get the large scale general corrosion breach of the
waste packages, where we really see the magnitude of the peak dose go
up. And, that\'s even with that very what I feel conservative model
where we did the\--every seismic event caused some sort of gross
breaching, gross damage to the waste package.

Igneous intrusions are the same way, although, again, it\'s a low
probability event over the period of the peak dose. Unlike a seismic
event, which we will have over that million year period, the likelihood
of getting an igneous is still a low probability disruptive event.

Again, what we saw is it may control the curve, the risk/dose curve over
the period prior to the onset of significant gross breach in the waste
packages due to corrosion. And, we don\'t believe it will have a
significant effect on the measure of the peak annual dose, mean or
median.

We also, again, I said we looked at the volcanic eruption and found that
no, there are no radionuclides that are building in. The decrease that
you\'ve seen in the past in terms of the volcanic eruption will continue
onward, and essentially the risk of this event will be most significant
during the 10,000 year period.

Next, please. The sensitivity analysis. This is more what we\'re looking
for. We found that the infiltration rates and percolation rates through
the repository have what I will call a minor effect on the magnitude of
the peak dose. Some of these minor or significant, I\'m going to claim
are my judgment. The report has values of the changes that we saw. I\'ll
invite you to judge for yourself whether it\'s significant or not.

Over the range of repository average infiltration rates that were\--this
is true over the range of repository average infiltration rates that
were representative of this long-term climate scenario when we
considered collapsed drift seepage. We did the infiltration, kind of
looked at the high value and the low value and see what the differences
were, and we did that for this collapsed drift seepage.

We found that emplacement drift seepage has a significant effect on the
magnitude of the peak dose, and how we accomplished that was just to
switch to non-collapsed drift seepage, and there was a fairly
significant, about a 70 per cent drop in the mean, about an 86 per cent
drop in the median, and it tended to move things further out in time.

Again, no surprise, we saw the waste package performance and in
particular, general corrosion rates had a significant effect on the peak
dose, both the magnitude and timing. Essentially, we ran a series of
sensitivity studies. We look at increasing the general corrosion rates
of both the drip shield and the waste package by a factor of 5. That
caused both the magnitude to increase, and the peak to move closer in
time. It, again, affects how some radionuclides decay, but it also
affects the rate that the waste packages are failing, and releasing
material.

We also did the full temperature dependence. We let the temperature
dependence go down to 21 degrees C, and saw about a 30 per cent
reduction in the dose. But, more importantly, it moved the measure of
the peak way out in time, well past a million years.

The results of the sensitivity model indicated that the choice of a
controlling solubility phase for neptunium dissolved concentration
limits didn\'t really have a significant effect. And, again, that was
for the range of parameters included in the model. The high drifts, in
particular, the collapsed drift seepage. Even if you would invoke
solubility limits, you are flowing so much through that you can deplete
the inventory fairly quickly in a lot of the realizations.

We did see that if we lowered the seepage rates, we did the same
scenario where we went back to the non-collapsed drift seepage rates,
and rechecked the two different solubility approaches. This one, the
secondary phase solubility control when we did this one.

The lower seepage rates did have an effect. You saw more of the
solubility was controlling and having a role in the release rate.

We found that the drip shield performance had a minor effect on the peak
dose. It is important early on, it does provide, in the earlier time
periods, a barrier for water to get to the waste package, also provides
a barrier, a very protective barrier for the waste package against rock
fall.

But, at the period of the peak, it tends to be for the most part
completely degraded. Yes, there are realizations where it will be there
if you look back at that curve, but for the most of the realizations, it
tends to be gone by the time we get the general corrosion breach of the
waste package, and it doesn\'t play a significant role.

We feel that the natural barrier system below the repository, mainly the
alluvial portion of the saturated zone, is an effective barrier for
several key radionuclides. And, again, it kind of goes back to what
Bruce and Bill were saying that strongly sorbing radionuclides could be
significantly retarded, which the sensitivity analysis in there where we
said, well, we\'re just going to assume the I\'ll call it ludicrous
assumption of no alluvium, and just to test the model and see what would
happen, and there\'s a significant increase in the very highly sorbing
radionuclides, the protactinium, these long-lived, highly sorbing.

Next, please. On that, I\'m not going to repeat this slide to say more
than you saw at the beginning. But, I really want to caution you on what
you and anybody else who uses these, what these results mean. They are
not to be compared to any proposed standard.

GARRICK: Thank you very much, Mark.

Because of the late hour, and the fact that the Board has other
commitments this evening, I think I\'m going to not ask questions at
this point. So, I think we will end it.

But, on the other hand, I do want to make a couple of comments. The
first comment is, of course, to thank the presenters and the briefers.
Every time I sit through one of these, I envy much of the work that\'s
done, and that I can\'t be an intimate part of it. And, I was very
impressed with the quality of the individuals that made the
presentations. And, that\'s the good news.

The bad news is that I\'m afraid if I had to have another meeting like
this, you would force me to retirement, because in my opinion, what we
have been through is unacceptable for two reasons. One, we didn\'t get
what we wanted in terms of really getting an understanding, a
fundamental understanding of the mass and curie balance of the
repository.

Now, I understand full well why we didn\'t, and I know they did the very
best they could do. But, the point is the Board didn\'t get what they
wanted. And, until we do, we\'re not going to be able to evaluate, as
we\'re supposed to do, and convince ourselves that there is a
fundamental understanding of the issue having to do with the
radionuclide transport through the repository. And, as I say, we know
why there are extenuating circumstances. There\'s budget issues.
There\'s legal issues, and a number of factors that have contributed to
the constraints that the DOE is having to work for. And, my comments
apply to the DOE presenters.

The second thing that I want to comment on is that we failed miserably
to comply with what this Board has come to invoke, namely the
fifty-fifty rule, that for the allotted time for the presentations, half
of it is supposed to be the presentation, and half of it is supposed to
be for the Board to be able to ask the questions it asks. And, we have
had to unduly truncate questions from the Board all day because of that,
and that is completely unacceptable.

I don\'t know what happened to that. The last meeting, it worked very
well. It worked very well today for the NRC presenters. They followed
the fifty-fifty rule, but nobody else did. This last presentation, which
was an outstanding presentation, and the work is really important and
good, but if you think in terms of the time that was allotted for the
presentation, which was 25 minutes, and you invoke the 50 per cent rule,
it should have been 13 minutes, and the presentation was three times as
long as it should have been. And, we\'ve got to do something about that
if we\'re going to be effective in our ability to implement our mandate.
And, it\'s just not happening.

So, this is why I\'m unhappy, even though as far as the quality of the
presentations and the quality of the material and the circumstances
under which it was presented was exemplary, and I just hope that in our
next meeting, we are able to do a better job of putting ourselves in a
position to do our job, because I certainly don\'t think we were in that
position this time.

And, I guess I leave a note of frustration about that, but on the other
hand, I think that it\'s important for us to communicate directly to DOE
how we feel, and when we don\'t feel we\'re getting what we want, we so
say it, and we didn\'t get what we wanted in this meeting.

And, number two, you didn\'t, the whole approach didn\'t allow us to
have the kind of exchanges that I think we absolutely have to have in
future meetings.

Now, are there any other follow-ups or comments from other members of
the Board?

And, I know we\'re all very tired, and we will I\'m sure rectify the
situation and we\'ll see you again soon, and we\'ll look forward to
that, and we will now adjourn.

(Whereupon, at 6:11 p.m., the meeting was adjourned.)