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UCBNE-4225

Impacts of Waste Transmutationon Repository Performance

J. Ahn, P. L. Chambr�, E. Greenspan, W. E. Kastenberg, M. D. Lowenthal,B. Park, and J. Vujic,

Department of Nuclear Engineering,University of California, BerkeleyBerkeley, California, 94720-1730

Prepared for

Los Alamos National LaboratoryLos Alamos, New Mexico

Under Contract #G1772-0018-23

June 1999

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The authors invite comments and would appreciatebeing notified of any errors in the report.

Ehud GreenspanDepartment of Nuclear Engineering

University of CaliforniaBerkeley, CA 94720

USA

gehud@nuc.berkeley.edu

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Table of Contents1 INTRODUCTION....................................................................................................................................... 12 METHODS DEVELOPMENT..................................................................................................................... 3

2.1 Mass Flow in ATW System.............................................................................................................. 32.1.1 Introduction ............................................................................................................................... 32.1.2 Mass Flow Model for ATW ..................................................................................................... 3

2.2 Overview of Yucca Mountain Repository........................................................................................ 72.2.1 Geohydrology............................................................................................................................ 82.2.2 The Repository.......................................................................................................................... 82.2.3 Regulatory Context ................................................................................................................... 9

2.3 Models for Repository Performance Assessment .......................................................................... 102.3.1 Mass of Radionuclides in the Repository............................................................................... 102.3.2 Radiological Hazard at Boundary of Accessible Environment............................................. 10

3 Results of Repository Performance Assessment ............................................................................. 143.1 Input Data ........................................................................................................................................ 14

3.1.1 LWR Spent Fuels .................................................................................................................... 143.1.2 ATW System Parameters........................................................................................................ 143.1.3 Repository Performance Parameters ...................................................................................... 16

3.2 Results of Mass Flow Analysis....................................................................................................... 183.2.1 Mass of Radionuclide in ATW Waste.................................................................................... 183.2.2 Decay Chains .......................................................................................................................... 19

3.3 Change of Radionuclide Mass in Repository by Radioactive Decay............................................ 193.3.1 Radionuclides in LWR Spent Fuel and ATW Waste ............................................................ 193.3.2 Effect of Recovered Uranium................................................................................................. 21

3.4 Hazard from One Canister of LWR Spent Fuel ............................................................................. 223.5 Hazard from One Canister of ATW Waste with Realistic Waste Fraction................................... 233.6 Hazard from One Canister of ATW Waste with Minimum Waste Fraction................................. 243.7 Hazard from One Canister of Recovered Uranium........................................................................ 243.8 Comparisons of Total Hazard ......................................................................................................... 25

3.8.1 LWR and ATW without Recovered Uranium ....................................................................... 253.8.2 LWR and ATW with Recovered Uranium ............................................................................ 26

3.9 Discussions ...................................................................................................................................... 273.9.1 Radiological Hazard................................................................................................................ 273.9.2 Mass of Radionuclide in the Repository ................................................................................ 28

3.10 Summary.......................................................................................................................................... 284 MEASURES OF REPOSITORY IMPACTS ............................................................................................... 29

4.1 Licensibility ..................................................................................................................................... 294.1.1 Existing Criteria ...................................................................................................................... 294.1.2 Qualitative Evaluation of Licensibility .................................................................................. 314.1.3 Proliferation............................................................................................................................. 324.1.4 Criticality Safety ..................................................................................................................... 32

4.2 Desirability ...................................................................................................................................... 324.2.1 Radiological Hazards.............................................................................................................. 324.2.2 Proliferation Hazards .............................................................................................................. 324.2.3 Underground Criticality.......................................................................................................... 334.2.4 Summary of Hazard Reduction Factors ................................................................................. 33

4.3 Other Measures................................................................................................................................ 345 CONCLUSIONS AND DIRECTIONS ........................................................................................................ 35

5.1 Factors Increasing or Limiting ATW Desirability ......................................................................... 355.2 Future Research ............................................................................................................................... 36

REFERENCES........................................................................................................................................................ 38

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List of FiguresFigure 2.1 ATW treatment system [1] .............................................................................................................. 4Figure 2.2 Fractions of radionuclide inventories coming out of the ATW process as waste. (α's are

waste fractions. δ is the fraction transmuted. They are defined for each radionuclide.Subscript i for radionuclide is omitted.) ......................................................................................... 5

Figure 2.3 Waste fractions from the reprocessing part. Subscript i is omitted. ............................................. 6Figure 2.4 Waste fractions from the partitioning part. Subscript i is omitted................................................. 6Figure 2.5 In-drift waste-package design ......................................................................................................... 9Figure 2.6 In-drift emplacement configuration ................................................................................................ 9Figure 3.1 Uranium inventory in LWR spent fuel and in ATW waste. ........................................................ 20Figure 3.2 Plutonium inventory in LWR spent fuel and in ATW waste....................................................... 20Figure 3.3 Inventories of Am, Cm, and Np inventory in LWR spent fuel and in ATW waste. ................... 20Figure 3.4 Thorium inventory in LWR spent fuel and in ATW waste.......................................................... 20Figure 3.5 Radium inventory in LWR spent fuel and in ATW waste. .......................................................... 20Figure 3.6 Inventories of Pa, I, TC, and Ac in LWR spent fuel and in ATW waste. ................................... 20Figure 3.7 Masses of thermally fissile actinides in LWR spent fuel. ............................................................ 21Figure 3.8 Masses of thermally fissile actinides in ATW waste with realistic waste loss. .......................... 21Figure 3.9 Masses of thermally fissile actinides in ATW waste with minimum waste loss. ....................... 21Figure 3.10 Thermally fissile fractions in actinides in LWR spent fuel and ATW waste.............................. 21Figure 3.11 Uranium inventory in LWR spent fuel and in ATW waste plus recovered uranium.................. 22Figure 3.12 Thorium inventory in LWR spent fuel and in ATW waste plus recovered uranium.................. 22Figure 3.13 Masses of thermally fissile actinides in ATW waste with realistic waste loss plus

recovered uranium. ........................................................................................................................ 22Figure 3.14 Masses of thermally fissile actinides in ATW waste with minimum waste loss plus

recovered uranium. ........................................................................................................................ 22Figure 3.15 Hazard at the 5,000 m location from one canister of LWR spent fuel for the mobile

medium........................................................................................................................................... 23Figure 3.16 Hazard at the 5,000 m location from one canister of LWR spent fuel for the immobile

medium........................................................................................................................................... 23Figure 3.17 Hazard at the 5,000 m location from one canister of ATW waste with realistic waste

fraction for the mobile medium..................................................................................................... 23Figure 3.18 Hazard at the 5,000 m location from one canister of ATW waste with realistic waste

fraction for the immobile medium. ............................................................................................... 23Figure 3.19 Hazard at the 5,000 m location from one canister of ATW waste with minimum waste

fraction for the mobile medium..................................................................................................... 24Figure 3.20 Hazard at the 5,000 m location from one canister of ATW waste with minimum waste

fraction for the immobile medium. ............................................................................................... 24Figure 3.21 Hazard at the 5,000 m location from one canister containing only uranium recovered by

reprocessing for the mobile medium.............................................................................................25Figure 3.22 Comparison of hazard from 2230 canisters of ATW waste with that from 7640

canisters of LWR spent fuel, disposed of in YMR surrounded by a mobile medium. ............... 26Figure 3.23 Comparison of hazard from 2230 canisters of ATW waste with that from 7640

canisters of LWR spent fuel, disposed of in YMR surrounded by an immobile medium.......... 26Figure 3.24 Comparison of hazard from 2230 canisters of ATW waste plus 7285 canisters of

recovered uranium with that from 7640 canisters of LWR spent fuel, disposed of inYMR surrounded by a mobile medium. ....................................................................................... 27

Figure 3.25 Comparison of hazard from 2230 canisters of ATW waste plus 7285 canisters ofrecovered uranium with that from 7640 canisters of LWR spent fuel, disposed of inYMR surrounded by an immobile medium. ................................................................................. 27

Figure 4.1 The sum of the fractions of the actinide waste stream from the ATW (concentrationrelative to the concentration limits at three DOE LLW disposal facilities). The threewaste streams correspond to low process losses (clean), higher losses (current average),and spent fuel. ................................................................................................................................ 30

Figure 4.2 40 CFR 191 Criteria, summary of repository results. ATW cases comprise the ATWwaste from utilizing the entire stockpile of LWR spent fuel. ...................................................... 31

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List of TablesTable 3.1 Assumed Spent Fuel Canister ....................................................................................................... 15Table 3.2 Radionuclide Inventories in LWR Spent Fuel.............................................................................. 15Table 3.3 Fractions Transmuted per Passage Through the Burner [1]......................................................... 15Table 3.4 Fractions Lost as Waste per Passage Through the Material Recycle Process............................. 16Table 3.5 Input Data for Radionuclide Transport Analysis.......................................................................... 17Table 3.6 Data for Radionuclide Transport in Geologic Formations (Upper Row:ÒImmobileÓ

Case, Lower Row:ÓMobileÓ Case)................................................................................................ 17Table 3.7 Radionuclides Data and Results of ATW Mass Flow Analysis. ................................................. 19Table 3.8 Reduction of peak hazard at the 5-km location by the ATW system. ......................................... 27Table 3.9 Comparison of uncertainty ranges. ............................................................................................... 28Table 4.1 Summary of Thinking Regarding Safeguards of Fissile Materials During

Transportation and Geologic Disposal with Respect to Different Threats. MC&ARefers to Material Control and Accounting. (Reference [26] refers to intrinsic barriersrather than physical barriers.)........................................................................................................ 33

Table 4.2 Summary of the Relative Migration Hazards, Proliferation Hazards, and CriticalityHazards from ATW Fuel-Cycle Waste in a HLW Repository (No recovered Uranium)........... 34

Table 4.3 Summary of the Relative Migration Hazards, Proliferation Hazards, and CriticalityHazards from ATW Fuel-Cycle Waste in a HLW Repository (With recovered uranium) ........ 34

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1 INTRODUCTION

In March of 1998, a research group at the University of California at Berkeley (UCB) under con-tract from Los Alamos National Laboratory (LANL)1 began a study that examines the whole acceleratortransmutation of waste (ATW) concept from both evaluative and design-improvement perspectives. Thereare two major thrusts to our study. One thrust focuses on assessing the impacts the ATW may have. Thegoal of the other thrust is to identify improved designs for the ATW blanket and mode of operation. Thereis a close interaction between the two efforts. The first phase of our work is to evaluate the impacts of thewaste from the LANL ATW design concept [1]. Our analysis has identified the Òworst playersÓ Ð the iso-topes that make the dominant contributions to the radiological, proliferation or criticality hazard of theYucca Mountain repository. Guided by this information we are investigating ATW blanket design conceptsand modes of operation that can significantly reduce the inventory and, hence, the hazards from the Òworstplayers.Ó The impacts from our modified ATW design will then be systematically assessed, leading to thenext design iteration. This report summarizes the first phase of our work. It describes the measures adoptedfor evaluating impacts of the ATW, the methods developed for impact analysis, the obtained results andfuture directions.

Our activities broke down into 5 tasks: establishing appropriate measures of ATW performance,modeling mass flows in the partitioning and transmutation system to characterize waste streams, modelingmobilization and migration of radionuclides from repository emplacement to the accessible environment,evaluation and scoping of the impacts using the measures and models developed in prior tasks, and devel-opment of preliminary conclusions and insights.

The measures selected for evaluating ATW performance build on previous evaluations ([2],[3])and represent two major concerns: licensibility and desirability. Licensibility is measured both by com-paring the waste streams to several sets of existing criteria and by a more qualitative evaluation of the clar-ity or transparency of the license application claim. We understand desirability in terms of reductions bothin the magnitude and the duration of hazards and of costs including radiological hazards, proliferation haz-ards, hazards of underground criticality, and costs measured in a relative sense in terms of the disposal ca-pacity of a HLW repository. Current limitations of the project have not permitted evaluation with respectto other measures that we present in the text.

A detailed mass-flow analysis provides the basis for our analysis of the impacts on repository per-formance. A box model is employed in modeling mass flows in the transmutation system. In its simplestform, the model has a waste-fraction parameter for each component of the transmutation system: initialreprocessing of spent fuel, fuel fabrication, transmutation, and partitioning. (A more complex version ofthe model breaks each component into sub-components or steps with associated mass flows.) The trans-muter or burner also has a destruction coefficient representing the reactorÕs effectiveness of the burner.The destruction coefficient can be established in separate calculations using reactor physics codes, which isnot shown in this report. Given an inventory of spent fuel to be transmuted, the model yields mass frac-tions in the waste stream.

For the analysis of radionuclide migration, we assume no particular waste form for ATW wastes.We assume, however, that the hypothetical waste form has the same leach time, 100,000 yr, as that for theuranium dioxide spent nuclear fuel. We further assume that there is a path between each canister and theboundary of the accessible environment (which is considered to be 5 km distant from the repository). Us-ing methods described in Section 2.3.2 of this report (a modified version of the methods described in Ref.[4]), it is assumed that water flows through fractures steadily in a liquid column with a constant velocity,and that water in the pores of the rock matrix is stationary. Mobilization of a radionuclide depends on itssolubility in the water, the water flow rate, and the overall quantity of waste present. Migration is calcu-lated assuming fracture flow with retardation and matrix diffusion at the boundaries along the path.

Evaluation of the hazards in a HLW repository was generally carried out for both the ATW fuel-cycle wastes and the initial spent fuel stock.2 We evaluated the radiological performance of the repositoryunder the anticipated extremes of the range of conditions at Yucca Mountain. Other hazards, such as those

1 Contract number G1772-0018-23.2 Because this evaluation is confined to the repository, we have not pretended to consider factors either beneficial(electricity produced) or detrimental (risks of accidents) associated with operation of the reactors and the reprocessingfacilities.

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associated with proliferation and criticality, are based on the fissile-material content of the waste. Reposi-tory capacity is based on the relative number of waste packages.

Our preliminary conclusions and insights highlight the factors that tend to increase or limit the de-sirability of the ATW with respect to the repository.

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2 METHODS DEVELOPMENT

2.1 Mass Flow in ATW System

2.1.1 Introduction

Spent fuel from light-water reactors (LWR) contains uranium, plutonium, other actinides, andlong-lived fission products that require the long-term isolation. The ATW system [1] has been proposed forseparation and transmutation of long-lived actinides and fission products in the spent fuel from commercialLWR. It aims at avoiding the safety assessment required for the proposed Yucca Mountain Repository(YMR) [5] for an unprecedented long time period (ten thousand years or longer) by converting long-livedradionuclides in spent fuel into short-lived species. Because the ATW system still generates wastes thatrequire a geologic repository for the disposal, the extent to which the ATW system can reduce the difficultyof geologic disposal of spent nuclear fuel is determined by comparing the performance of YMR for the caseof direct disposal of spent fuel with that for the case of ATW waste disposal.

For the repository safety, one should consider the following aspects: (1) the potential radiologicalimpact on the public by radionuclides released from failed waste containers and taken up by human beingsvia transport through geologic formations surrounding the repository and through the subsequent foodchain, (2) the possibility of criticality events resulting from buried fissile materials, and (3) the possiblediversion and use of fissile actinides contained in the spent fuel for weapon construction.

In Section 2.1, mathematical models are established for investigating the aforementioned three as-pects of repository safety. More specifically, for each radionuclide, the radiological hazard at the boundaryof the accessible environment (5 km distant from the repository) and the mass (or inventory) existing in therepository are obtained. For both models, the mass of each radionuclide to be placed initially in the re-pository is the key information. Therefore, a model is established to quantify the mass of each radionuclidethat comes out of the ATW system as waste. We call this the mass flow model of the ATW system. Thus,the repository performance assessment model consists of (1) the mass flow model, (2) the inventory model,and (3) the radiological hazard model.

As a measure for the radiological impact of the repository, the radiological hazard of a radionu-clide is defined in this study as the ratio of the radioactivity per year [Bq/yr] of a particular radionuclidearriving at the accessible environment boundary to the annual limit on intake for oral ingestion [Bq/yr] ofthat radionuclide given in 10CFR20 [6].

The mass of fissile materials existing in the repository, which is obtained by the inventory model,is used as a measure for the criticality safety and the proliferation resistance.

The numerical results obtained by these models are shown in Chapter 3. The differences betweenthe repository performance by LWR spent fuel disposal and that by the disposal of waste from ATW arepresented in terms of the radiological hazard at the boundary of the accessible environment and the inven-tory in the repository. The observations obtained by such comparison give insights for optimization andimprovement of the ATW concepts.

2.1.2 Mass Flow Model for ATW

The ATW system is divided into four parts [1]: the reprocessing of LWR spent fuel, the fabrica-tion of ATW transmutation target assemblies (TA), the burner, and the partitioning of spent TA (see Figure2.1). Radioactive wastes are generated in each of these parts. The mass of each radionuclide coming out ofATW as waste is the basis for the repository performance assessment for the ATW system, and can be ob-tained by the following mass flow analysis.

Approximately 63,000 tons of LWR spent fuel and 7,000 tons of defense High-Level Wastes(HLW) are considered to be disposed of in YMR. In the ATW concept, the LWR spent fuel is first reproc-essed. In the reprocessing part, uranium and fission products except for iodine and technetium are removedfrom the spent fuel. The spent fuel pins are chopped, and the oxide fuel is separated from zircaloy clad-ding. The oxide fuel is sent to the reduction process, where the oxide fuel is converted to metal. The metalincluding virtually all uranium, transuranic elements, and fission products, will be the anode of the electro-refining process. The separated cladding material is used as the source of zirconium, which is the base-matrix material for the ATW target assembly. The off-gas released by the decladding process is collected,stored in a metal container, and eventually sent to the repository.

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Spent Fuel Decladding

Spent fuel

Electrorefining

Electrowinning TA Fabrication Electrowinning

Reductive Extraction

Electrorefining

Spent TA ChoppingATW Burner

Anode Polishing

U Storage

YMR

“Reprocessing” “Partitioning”

Direct Oxide Reduction

Figure 2.1 ATW treatment system [1]

In the electro-refining process of the reprocessing, the metal from the direct reduction process isplaced in the anode. Virtually pure uranium is collected on the cathode. In the molten salt, most of theactinides and the rare earth elements as well as a small fraction of uranium are included. This molten salt issent to the electro-winning process. The anode shrinks as the electro-refining process proceeds. A batchelectro-refining process is stopped after a certain time period. The anode still containing small fractions ofuranium, actinides and rare earth elements is polished to reduce the discharge of the actinides and the rareearth as waste. The actinides and the rare earth recovered by the anode polishing are returned to the elec-tro-refining process. The uranium recovered in the cathode of the electro-refining and the anode polishingstages is stored. In the current ATW concept, the recovered uranium is sent to a low-level waste repository(This would not be permitted under current regulations. See Chapter 4). The anode after polishing is re-garded as radioactive waste from the ATW system by LANL [1].

At the electro-winning process, the molten salt from the electro-refining process is further sepa-rated into three parts. On the cathode, uranium is concentrated. The recovered uranium is also sent to theuranium storage. On the anode, the actinide elements are extracted with small fractions of uranium and therare earth elements. This is the feed material for the ATW target assembly fabrication. In the molten salt,most of the rare earth fission products is included. The molten salt is sent to the reductive extraction proc-ess, where the fission products (FPs) are removed from the molten salt. The cleaned-up molten salt is re-utilized in the electro-refining process. The removed fission products are solidified, stored, and eventuallydisposed of in YMR.

The material from the electro-winning process is fabricated into an alloy fuel. Small fuel pelletsare fabricated, which are assembled into a fuel rod. Mechanical processes as grinding or cutting, whichgenerate wastes, are likely to be required to keep accuracy in dimensions of pellets and rods.

In the burner, the target assembly is exposed to neutrons. Some actinides undergo fission, gener-ating FPs, while others become heavier species by neutron absorption. After some residence time in theburner, the target assembly (TA) is discharged from the burner, and stored for cooling. In the dischargedTA, there are still actinides that have not fissioned. Therefore, the spent TA is sent to the partitioning part,where remaining actinides are separated from fission product elements.

The partitioning starts with chopping of the irradiated TA into small sections. The used claddingis sent to storage. The off-gas released by chopping is collected and stored.

The alloy fuel is processed at the electro-refining process to remove FPs. On the cathode, most ofthe actinides are collected with small fractions of uranium and FPs. This is sent to the target assembly fab-rication. In the molten salt, still considerable amounts of uranium, actinides, and the rare earths are in-cluded. Therefore, the molten salt is sent to the electro-winning process for further recovery of the acti-nides, the rare earths and uranium. The anode contains most of the uranium and very small fractions of theactinides and the rare earths. This is regarded as waste from the ATW process [1].

At the electro-winning process in the partitioning, the rare earths remaining in molten salt are sentto the reductive extraction process. Cleaned-up molten salt is re-utilized in the electro-refining process.The rare earths are solidified, stored, and eventually sent to the repository. The recovered actinides withsmall fractions of uranium and the rare earths in the cathode are sent to the target assembly fabrication.

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The total mass of the materials returned to the TA fabrication from the partitioning is smaller thanthe mass of the target before burning it in the burner because fission products are removed by partitioning.Also, some fractions of actinides are lost as waste from the processes in the partitioning. The mass deficitis made up by the feed materials coming from the electro-winning of the reprocessing.

Based on the aforementioned overview for the material flow in the ATW system, we determine themass of each radionuclide that comes out of the ATW system as waste, which is to be disposed of in therepository.

The mass flow depicted in Figure 2.1 is simplified as Figure 2.2. Figure 2.2 is applied for each ra-dionuclide. A unit mass of a radionuclide is considered as input to the reprocessing part, consisting of de-cladding, oxide reduction, electro-refining, electro-winning, and anode polishing, as shown in Figure 2.1and Figure 2.3. It is assumed that all the spent fuel is reprocessed once at the same time. It is assumed thatthe mass fraction, α1,i , of radionuclide i comes out from the reprocessing part. The fraction, 1 1− α ,i , is

sent to the TA fabrication. This is the starting material for recycling.

ATW Burner

TA Fabrication

Partitioning

Reprocessing

1α 2α

3α

δ

1

Figure 2.2 Fractions of radionuclide inventories coming out of the ATW process as waste. (α'sare waste fractions. δ is the fraction transmuted. They are defined for eachradionuclide. Subscript i for radionuclide is omitted.)

The fraction, α1,i , is determined by the detailed mass flow consideration in the reprocessing part.

Figure 2.3 shows the break-down of the fraction, α1,i . The fraction, αd i, , is assumed to be lost as waste by

the spent fuel decladding. This fraction represents the loss of radionuclide i as such forms as fine particlesgenerated by chopping or inclusions in the cladding material. At the direct oxide reduction process, thefraction, αo i, , is assumed to be lost as waste. The fraction, (1Ð αd i, )(1Ð αo i, ) is input to the electro-refining

process. The output from the electro-refining is assumed to be divided into four fractions: the fraction,γ r i, , included in the anode and sent to the anode polishing, the fraction, βr i, , included in the cathode and

sent to the uranium storage, the fraction, εr i, , included in the molten salt and sent to the electro-winning

process, and the loss, α r i, , as waste. Therefore, these four fractions satisfy

α β γ εr i r i r i r i, , , ,+ + + = 1. (2.1)

At the anode polishing, the fraction, γ a i, , is recovered and sent to the electro-winning process.

The fraction, βa i, , is included in the recovered uranium as impurity. The fraction, αa i, , is lost as waste.

Therefore, the relation,

α β γa i a i a i, , ,+ + = 1, (2.2)

is satisfied.The electro-winning process receives a fraction, (1Ð αd i, )(1Ð αo i, ) εr i, , from the electro-refining,

and a fraction, (1Ð αd i, )(1Ð αo i, ) γ r i, γ a i, , from the anode polishing. At the electro-winning process, a

fraction, αw i, , is lost as waste. A fraction, βw i, , is included as impurity in the uranium recovered in the

cathode.Thus, the total waste fraction, α1,i , from the entire reprocessing process is obtained by summing

the aforementioned waste losses as

α α α α α α α β γ β ε β γ α ε α1 1 1 1, , , , , , , , , , , , , , , ,i d i d i o i d i o i r i r i r i a i r i w i i a i r i w i= + −( ) + −( ) −( ) + + + + +( ) . (2.3)

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The mass fraction, 1 Ð α1,i , is sent from the reprocessing to the TA fabrication (see Figure 2.2),

and the first cycle starts. In the TA fabrication part, the fraction, α2,i , is assumed to be lost as waste. Thus,

for the first cycle, the fraction, (1 Ð α1,i )(1 Ð α2,i), is fabricated into a TA, and installed in the burner.

Spent Fuel Decladding

Electrorefining

Electrowinning

Anode Polishing

U Storage

Direct Oxide Reduction

1 – α1

α1

αd

αo

αrαa

αU αw

εr

γr

βrβa

γa

Figure 2.3 Waste fractions from the reprocessing part. Subscript i is omitted.

At the burner, it is assumed that the fraction, δ i , of radionuclide i becomes some other species (or,is transmuted). The spent TA after the first-time irradiation in the burner contains radionuclide i with themass fraction of (1 Ð α1,i )(1 Ð α2,i)(1 Ð δ i ).

The spent TA is sent to the partitioning. It is assumed that the fraction, α3,i , is lost as waste from

the entire partitioning process. From the partitioning part, the fraction, (1 Ð α1,i )(1 Ð α2,i)(1 Ð δ i )(1 Ð α3,i ) is

returned to the target fabrication for the second cycle.The detail of the partitioning process is depicted in Figure 2.4. At the spent TA chopping, the

fraction, αc i, , is lost as waste. Then, the fraction, 1 Ð αc i, , is sent to the electro-refining process. The

fraction, (1 Ð αc i, ) α r i, , is lost as waste from the electro-refining process. The fraction, βr i, , is included in

the molten salt, and sent to the electro-winning process. The fraction, γ r i, , is included in the cathode of the

electro-refining process, and sent to the target fabrication. The mass fraction lost as waste from the entirepartitioning process is obtained by summing the losses from the TA chopping, the electro-refining, and theelectro-winning, as

α α α α β α3 1, , , , , ,i c i c i r i r i w i= + −( ) +( ). (2.4)

Electrowinning

Electrorefining

Spent TA Chopping

1 – α3

α3

αc

α r

αw

βrγr

Figure 2.4 Waste fractions from the partitioning part. Subscript i is omitted.

The first cycle starts with the mass fraction, 1 Ð α1,i , of radionuclide i coming from the reproc-

essing part to the TA fabrication part, and ends in the fraction (1 Ðα1,i )(1 Ðα2,i)(1 Ð δ i )(1 Ðα3,i ) from the

partitioning part. The waste fraction, f i1, , occurring in the first cycle is obtained by summing the waste

fraction, (1 Ð α1,i ) α2,i , from the target fabrication, and the fraction, (1 Ðα1,i )(1 Ð α2,i)(1 Ð δ i ) α3,i , from the

partitioning as

f i i i i i i i1 1 2 1 2 31 1 1 1, , , , , ,= −( ) + −( ) −( ) −( )α α α α δ α . (2.5)

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The second cycle starts with the mass fraction, (1 Ðα1,i )(1 Ð α2,i)(1 Ð δ i )(1 Ð α3,i ), inherited from

the first cycle. The fraction of waste at the TA fabrication is obtained as (1 Ð α1,i )(1 Ð α2,i)(1 Ð δ i )(1

Ð α3,i ) α2,i . After the transmutation in the burner, the fraction, (1 Ðα1,i )(1 Ð α2,i)2(1 Ð δ i )

2(1 Ðα3,i ) remains

in the target. At the partitioning, the fraction, (1 Ðα1,i )(1 Ð α2,i)2(1 Ðδ i )

2(1 Ð α3,i ) α3,i is lost as waste.

Thus, for the second cycle, the waste fraction, f i2, , is obtained as

f i i i i i i i i i i i2 1 2 3 2 1 22 2

3 31 1 1 1 1 1 1 1, , , , , , , , ,= −( ) −( ) −( ) −( ) + −( ) −( ) −( ) −( )α α δ α α α α δ α α . (2.6)

In general, for the k-th cycle, the waste fraction, fk i, , is written as the sum of the waste fraction

from the target fabrication and the fraction from the partitioning as

fk i i ik

ik

ik

i i ik

ik

ik

i, , , , , , , , ,= −( ) −( ) −( ) −( ) + −( ) −( ) −( ) −( )− − − −1 1 1 1 1 1 1 11 2

1 13

12 1 2 3

13α α δ α α α α δ α α

for k ³ 1. (2.7)The total fraction, Fn i, , lost as waste at the end of the n-th cycle is obtained by summing fk i, from

k = 1 to n and the waste fraction, α1,i , from the reprocessing, as

F fn i i k ik

n

i i i i i ii

ni

ni

n

i i i

, , ,

, , , , ,, ,

, ,

= +

= + −( ) + −( ) −( ){ } − −( ) −( ) −( )− −( ) −( ) −( )

=∑α

α α α α δ αα δ α

α δ α

11

1 1 2 2 32 3

2 3

1 1 11 1 1 1

1 1 1 1

(2.8)

The total waste fraction, Fn i, , increases with the number, n, of recyclings, approaching the upper

bound, F i∞, , by taking n to infinity, as

F i i ii i i i

i i i∞ = + −

+ − −− − − −, , ,

, , ,

, ,

( )( )( )

( )( )( )α α

α α δ αα δ α1 1

2 2 3

2 3

11 1

1 1 1 1. (2.9a)

The total fraction, F iT

∞, , transmuted by infinite recyclings is obtained in a similar way as

F iT

ii i

i i i∞ = −

−− − − −, ,

,

, ,

( )( )

( )( )( )1

1

1 1 1 112

2 3

αα δ

α δ α. (2.9b)

The sum of F i∞, and F iT

∞, can actually be confirmed to be unity.

If the mass of radionuclide i in the spent fuel initially available is MiSF( ) [mol], then the mass of

radionuclide i in the waste from the ATW system destined to the repository is written as M FiSF

i( )

,∞ . This

mass is considered the mass of the radionuclide initially placed in the repository.Note that radioactive decay during the ATW process operation is neglected in the aforementioned

formulation. Considering that the operation period is approximately half a century, which is significantlyshorter than the half-lives of most actinides considered in this analysis, this assumption is reasonable.However, for 238Pu and 241Pu, whose half-lives are 87 and 14 years respectively, radioactive decay shouldbe taken into account in future analyses. Also to determine the waste fractions from ATW for short-livedfission products, especially for 90Sr and 137Cs (half-lives about 30 years), two major heat generators, moredetailed analyses are required, where such quantities as the residence time of the TA in the burner core, thecooling time before the partitioning after the discharge from the burner, and the age of the spent fuel areexplicitly treated as system parameters.

The mass of a radionuclide initially existing in waste canisters in the repository is utilized in thefollowing analyses. In Section 2.3.1, the initial mass in the repository is utilized as the initial condition forthe batch-decay equation (Bateman equation) to estimate the mass of the radionuclide existing in the re-pository. In Section 2.3.2, the initial mass is utilized as the initial condition for the radionuclide transportmodel.

2.2 Overview of Yucca Mountain RepositoryYucca Mountain is located in Nye County, southern Nevada. It is on the southeastern boundary of

8

the Nevada Test Site (NTS), approximately 120 km northwest of Las Vegas. The region is part of thesouthwestern Nevada volcanic field, and is characterized by linear mountain ranges and intervening valleyswhose orientation is primarily controlled by north-trending normal faults. Elevations range from approxi-mately 350 m to 1,600 m above sea level [5].

2.2.1 Geohydrology

The unsaturated zone beneath Yucca Mountain consists of stratified units of welded and unweldedtuffs of contrasting hydrologic properties. Some of the units are more highly fractured, which may influ-ence the flow of groundwater. In addition, both the faults bounding the Yucca Mountain block, and theGhost Dance fault within the block, also may be either pathways for flow or impediments to flow. Thedownward percolation may be offset by upward flow of water vapor.

A small fraction of the precipitation at Yucca Mountain that falls on the surface flows generallydownward through the unsaturated units, past the repository horizon to the water table. The static ground-water table underlies Yucca Mountain at depths of 600 to 800 meters, depending upon topography, pro-ducing a very thick unsaturated zone.

In the Yucca Mountain region, the dominant saturated flow appears to be generally southerly,from higher elevations in the northern NTS to discharge areas in the Amargosa Desert to the south. In thetuffs, fractures may provide paths for water flow. Although the tuffs are fractured, their tip, offsets, andvariation in physical properties between layers cause fractures in adjacent units to not necessarily align.Thus, fracture flow may not be continuous.

2.2.2 The Repository

Containment and isolation are to be achieved by the use of multiple barriers to mobilization andtransport of the contaminantsÑwaste containers that resist degradation, and the location of the repository ina geologic setting that reduces the probability that both expected and unforeseen processes and events canmake the contaminants accessible to humans.

Ventilation equipment, waste-receiving, and safety and access control facilities are located at theground surface. Access to the underground workings is by gentle-grade ramps for moving the waste pack-ages. Current design specifies access ramps at the north and south ends of the repository. The repositoryitself is bisected by two or three main tunnels that provide ventilation and access for waste-emplacementoperations. A perimeter drift runs around the periphery of the emplacement area. Waste is located in em-placement drifts that run between the main access drift and the perimeter drift.

The repository is expected to remain open for over 25 years while waste is placed. During the op-erational period, the drifts are ventilated and accessible to humans. At a later time, approximately 75 yearsfrom start of operations, the repository will be closed.

The potential repository is designed to hold the highly radioactive spent nuclear fuel from nuclearreactors and high-level waste resulting from activities at DOE defense facilities. The designed capacity ofthe repository is 70,000 MT of high-level waste. Of this amount, 63,000 MT are spent fuel and the balanceis defense HLW.

The Yucca Mountain site was chosen for characterization because the sequence of unsaturatedgeologic units would make exposure of the waste to pervasive groundwater unlikely. Distances of severalhundred meters from the surface to the potential repository horizon, and from the repository horizon to thewater table, would present barriers to rapid groundwater travel and contaminant transport. Furthermore, thesite is thought to be relatively geologically stable, and relatively free of economically desirable resources.

The waste package is a part of an engineered barrier system (EBS) that may include the wasteform, internal stabilizers, the container, and backfill or standoffs between the container and the adjacentrock. Several waste-package designs have been considered by the DOE. All consist of a cylindrical metalcontainer into which the waste is placed. The container is sealed with a gas-tight closure.

The spent fuel consists of assemblies from both boiling water reactors and pressurized water re-actors. The assemblies include not only the uranium oxide fuel, but also the fuel cladding and supporthardware, all of which are radioactive due to activation or intrinsic radioactivity. The HLW consists ofproducts resulting from physical and chemical processes associated with the separation of fissile materialsfor defense needs. These waste products are immobilized in a glass or calcine matrix.

We consider a waste-package design that contains 7.5 MTU of BWR spent fuel or 9.2 MTU ofPWR spent fuel. The package could weigh in excess of 50 MT. The containers can be used for all aspectsof the cycle including offloading from the reactor, transportation, and emplacement in the repository. The

9

packages are placed horizontally on the floors of the emplacement drifts. This configuration is called theÒin-driftÓ emplacement.

The waste packages have a heat output at time of emplacement that depends on the burnup anddecay of the spent fuel. Their spacing in the repository at emplacement therefore determines the local arealpower density (given in killowatts/acre). At the time of emplacement, thermal areal power densities areachieved by controlling the spacing between nearest-neighbor waste packages.

Figure 2.5 In-drift waste-package design Figure 2.6 In-drift emplacementconfiguration

2.2.3 Regulatory Context

The basic policy in the United States regarding disposal of radioactive waste was set forth in theNuclear Waste Policy Act of 1982. The Act outlined the procedure to be followed for choosing, character-izing, and approving a site for storage of radioactive waste. The Act has been implemented in regulationsof the US DOE, the US NRC, and the US EPA.

General siting guidelines are defined by the DOE in 10 CFR Part 960. Sites are to be evaluatedagainst a number of criteria having to do with favorable or unfavorable site characteristics. If the YuccaMountain site is determined to be a suitable location for a radioactive waste repository according to the 10CFR 960 guidelines and other assessments, the DOE would apply to the NRC for authorization to constructthe repository and for a license to operate it.

The NRC regulations governing repositories for HLW are given in 10 CFR Part 60. The regula-tion covers many aspects of repository siting, design, operation, and post-closure performance. Section60.112 states that the repository system must conform to applicable EPA standards, and Section 60.113introduces three additional performance objectives relating to particular barriers:• Containment of HLW within the waste packages must be Òsubstantially completeÓ for a period of 300

to 1000 years.• The release from the EBS following the containment period must not exceed a given rate (one part in

100,000 per year of the inventory at 1000 years following repository closure, with some important ex-ceptions).

• The pre-waste emplacement groundwater travel time along the Òfastest path of likely radionuclidetravelÓ from the disturbed zone to the accessible environment must be at least 1000 years.

Regarding criticality, 10 CFR Part 60, Section 131(b)(7) states:The Engineered Barrier Segment shall be designed to ensure that a nuclear criticality accident isnot possible unless at least two unlikely, independent, and concurrent or sequential changes haveoccured in the conditions essential to nuclear criticality safety. Each system shall be designed forcriticality safety under normal and accident conditions. The calculated effective multiplicationfactor must be sufficiently below unity to show at least a five percent margin, after allowance forthe bias in the method of calculation and the uncertainty in the experiments used to validate themethods of calculation.

The regulations governing such performance have been given by 40 CFR Part 191 [7]. However,the Energy Policy Act of 1992 requires new standards to be promulgated for the Yucca Mountain site. TheAct specifies that a study be conducted by the National Academy of Sciences (NAS) and that the EPA thenset standards for Yucca Mountain consistent with the recommendations of the NAS. At the point of writingthis report, the new standards have not yet been announced.

10

40 CFR Part 191 states that disposal systems shall be designed so that for 10,000 years the releaselimit set for each radionuclide per 1,000 MTHM of waste in 40 CFR191 will not be exceeded at the acces-sible environment with a likelihood of less than 1 chance in 10 of exceeding the limit, or less than 1 chancein 1000 of exceeding 10 times the limit. The boundary of accessible environment is understood located at 5kilometer distance from the boundary of the underground facility of the repository. EPA release limits areset in such a way that less than 1000 health effects (cancer fatality) occur in 10,000 years.

2.3 Models for Repository Performance Assessment

2.3.1 Mass of Radionuclides in the Repository

The mass of a radionuclide existing in the repository changes with time due to radioactive decayand release to the surrounding geologic formation after failure of waste canisters. In this section, we con-servatively overestimate the mass by neglecting the release to the surrounding geologic medium. Withoverestimated mass of actinides in the repository, the repository criticality safety and the attractiveness ofthe repository as the source of fissile materials are estimated conservatively. The preliminary calculationsbased on a model that includes the loss by radionuclide release from the repository show, however, that thecumulative radionuclide mass release from the repository to the surrounding geologic formation is smallcompared to the mass that decays within the repository, for most actinide radionuclides [8].

If the radionuclides initially placed in the repository are assumed to be kept in the repository, thechange of the mass with time is governed by the Bateman equations [9]:

dM t

dtM t M t ti

i i i i( )

( ) ( ),= − + >− −λ λ 1 1 0 , i = 1, 2, É, n, (2.10)

which are valid for i = =1 2 00, , ..., λ , (2.11)

where λi is the radioactive decay constant [yrÐ1] of radionuclide i, Mi [mol] the mass of radionuclide i ex-isting in the repository. Governing equations (2.10) are solved subject to the initial conditions,

M M ii io( ) , , , ...0 1 2= = . (2.12)

The mass of radionuclide i existing initially in the repository, Mio [mol], is expressed as

M M Fio

iSF

i= ∞( )

, . (2.13)

The analytical solution to this mathematical problem is readily obtained [9] as

M t M t Mt

i io

io

ij

k jkk j

ij

ii

( ) expexp

= −( ) +−( )

−( )

+ + −

=≠

==

−

∏∑∑λ λ λ λ λ

λ

λ λl l l l

l

ll

L1 2 11

1

, (2.14)

where any two radionuclides in the chain are assumed not to have an identical radioactive decay constant.Numerical evaluations have been performed with (2.14) for the mass of radionuclide in the re-

pository. Results are shown in Chapter 3.

2.3.2 Radiological Hazard at Boundary of Accessible Environment

The mass, M FiSF

i( )

,∞ , of radionuclide i coming out of the ATW system as waste is to be placed in

the repository. It is assumed that the mass, M FiSF

i( )

,∞ , of radionuclide i is placed in the repository immedi-

ately after it is generated, and that all the ATW waste is placed in the repository at the same time.The total mass, M Fi

SFi

( ),∞ , of radionuclide i is distributed over many canisters unevenly. In this

study, however, it is assumed that all the canisters have an identical mass of each radionuclide. Namely, ifthe number of waste canisters is N, then the mass of radionuclide i to be contained in one canister is as-sumed M F Ni

SFi

( ),∞ .

Buried waste canisters generate heat. Water will be expelled away from the waste canisters. Cor-rosion of waste canisters may slow down. As heat generation decreases, water comes back to the vicinity

11

of the canisters, and wets the waste canisters. When holes due to corrosion penetrate canisters, the wasteforms inside the canisters contact with groundwater. The waste form starts alteration or dissolution by re-acting with groundwater, and radionuclides contained in the waste form are released into the groundwater.We assume that all the canisters fail, i.e., start to release radionuclides into groundwater, at the same time.

Waste canisters are assumed to be placed in a two-dimensional array fashion in the repositoryarea. If d is the distance between neighboring canisters, the repository footprint per canister is d2. In YMR,waste canisters are initially placed in open excavated tunnels. We assume that the tunnels are filled by rub-bles of the host rock (i.e., tuff) by the time the canisters fail.

We assume that the rubble filling the tunnels is regarded as a homogeneous porous medium,where radionuclides released from failed canisters are transported by molecular diffusion. The averageporosity of the rock rubble is assumed εp.

The YMR is located above the water table [5]. The water flow through the repository is vertical,whereas it is horizontal below the water table. The tuffaceous rock in Yucca Mountain contains manyfractures of various sizes. Fractures are considered to be main conduits for groundwater flow. We assumethat there is a path between each canister and the boundary of the accessible environment, which is consid-ered 5 km distant from the repository in the EPA regulation (40CFR191 [7]). It is assumed that water flowsthrough fractures steadily in a liquid column with a constant velocity, v, and that water in the pores of therock matrix is stationary.

To simplify the analysis, a parallel planar fracture of aperture 2b and width d times n is assumed tobe the path connecting one canister to the accessible environment boundary [4, 10]. With the distance, d,between neighboring canisters, and the spacing, 2a, of fractures, the number, n, of fractures intersecting therepository footprint per canister is calculated as

nd

a=

2. (2.15)

Actual paths between individual canisters and the boundary of the accessible environment aretortuous. Actual travel distance between individual canisters and the accessible environment boundary isgreater than the straight-line distance between two points. The planar-fracture assumption is, therefore,conservative for the evaluation of the mass flux at some distance away from the source point because italways gives shorter travel times resulting in less radioactive decay loss during the transport.

Materials released from the waste form are assumed to be transported through fractures by advec-tion as solutes, and to diffuse into the rock matrix by molecular diffusion. As the pores in the rock matrixare partially filled with water, diffusion in the rock matrix is assumed to occur only through the intercon-nected water phase in the pores. The fraction of the pore spaces filled with water, i.e., the water saturationis assumed S. The solutes in the pores of the rock matrix are in sorption equilibrium with the solid phase ofthe rock, resulting in retardation of radionuclide transport.

For some radionuclides such as plutonium, transport in a colloidal form as well as in a solute formcan be considered. In Yucca Mountain conditions, colloids with positive surface charge would be expectedto attach to tuff with negative surface charge. According to the numerical results obtained based on a sim-plified colloid-facilitated transport model for plutonium [10], the transport distance of most colloids is rela-tively short (less than 50 m). Therefore, in this study, we assume that materials released from the wasteform are all transported as solutes.

The radionuclide transport equations for the solute concentrations C z tk ( , ) [mol/m3] of radionu-

clide k in the water in the fracture and C y t zkP ( , ; ) in the pore water of the rock matrix are, respectively,

RC

tv

C

zD

C

z

D

b

C

yR C R C t ze k

k k L k e kI

kP

y

e k k k e k k k( )( )

( ) ( ) , , ,∂∂

∂∂

∂∂

∂∂

λ λ+ − − ⋅ + − = > >=

− − −

2

20

1 1 1 0 0 0 (2.16)

and α ∂∂

∂∂

α λ α λe kkP

e kI k

P

e k k kP

e k k kPC

tD

C

yC C t y a z( ) ( ) ( ) ( ) , , , .= − + > < < >− − −

2

2 1 1 1 0 0 0 (2.17)

Here k = 1, 2, É, i, with λ0 = 0, and λk [yrÐ1] is the decay constant of radionuclide k. The coordinate z istaken along the fracture in the water flow direction. The coordinate y is taken perpendicular to the waterflow direction into the rock matrix surrounding the fracture.

The initial and boundary conditions are

12

C z zk ( , ) , ,0 0 0= > (2.18)

C y z y a zkP ( , ; ) , , ,0 0 0 0= < < > (2.19)

S vC t DC

zS Q t tf f k

L k

zkε ∂

∂( , ) ( ), ,0 0

02−

= >

= (2.20)

C t tk ( , ) , ,∞ = >0 0 (2.21)

C t z C z t t zkP

k( , ; ) ( , ), , ,0 0 0= > > (2.22)

∂∂C

yt zk

P

y a=

= > >0 0 0, , . (2.23)

The velocity, v [m/yr], and the longitudinal hydrodynamic dispersion coefficient, DL [m2/yr], areassumed to be constant with time and uniform in the fracture and common for all radionuclides, since thehydrodynamic dispersion is determined mainly by the geometry of the contaminant transport paths.

Molecular diffusion of radionuclides from water flowing in the fractures into the pores of the rockmatrix is an important retardation mechanism. De k

I( ) [m2/yr] is an intrinsic diffusion coefficient for radi-

onuclide k of element e in the rock matrix. Subscript e(k) indicates that the k-th member nuclide in a de-cay chain is an isotope of element e.

For sorption equilibrium in the porous rock, we define the capacity factor αe k( ) for radionuclide k

of element e as

α ε ρ εe k p p p dpeS S K( ) = + −( )1 , (2.24)

where ρp [kg/m3] is the density of the porous rock matrix, and Kdpe [m3/kg] the sorption distribution coef-

ficient of element e for the rock matrix. The saturation of the rock matrix, S, is included to account for thepartial saturation of the geologic medium in the Yucca Mountain. The retardation coefficient, Re k( ) , for

advection and longitudinal dispersion in the fracture is defined as

R Ke kf f

fdfe

( ) = +−( )

11ρ εε

, (2.25)

where ρ f and ε f are the density and the porosity of the material filling the fractures. Kdfe [m3/kg] is the

sorption distribution coefficient of element e for the material filling the fractures.The time dependent function, Q tk ( ) [mol/m2áyr], in (2.20) is the mass flux of radionuclide k that is

released from the region around one canister, entering the fracture. In [4], a mathematical expression forQ tk ( ) is obtained by considering the molecular diffusion flux at the interface between the rubble rock re-gion and the host rock. The rubble region between the waste solid and the surrounding rock formation issimplified as a single uniform porous medium. S2 is the surface area of the rubble rock region containingone waste canister. Sf [m

2] is the total cross sectional area of the fractures intersecting the repository area,d2, per waste canister, and is expressed as

S n b df = ⋅ ⋅2 . (2.26)

We assume that radionuclides are transported only by molecular diffusion through the stationarywater in the pores of the region around a waste canister filled with rubbles of the host rock. De k( ) [m

2/yr] is

the diffusion coefficient in the rubble rock region for element e. The retardation coefficient, Ke(k), for dif-fusion in the rubble rock region is defined as:

K Ke k de

( ) = + −( )1

1ρ εε , (2.27)

where ρ [kg/m3] and ε are the density and the porosity of the rubble rock, respectively, and Kde [m3/kg], a

constant sorption distribution coefficient. In [4], the concentration of radionuclide k in the water phase in the rubble-rock region was first

13

obtained by analytically solving diffusion equations that take into account retardation of diffusion by sorp-tion and a multiple-member decay chain. Then, the diffusive mass flux, Q tk ( ) , of the radionuclide at theinterface between the rubble-rock region and the surrounding host rock was obtained by the concentrationgradient obtained analytically multiplied by the diffusion coefficient, De k( ) , and the porosity of the rubble-

rock region.On the waste-form surface, if the solubility Ne

* [mol/m3] of element e is so low that all the radi-onuclides of element e released by the alteration of the waste form cannot dissolve into the water phase,then precipitate of element e will occur. The concentration of element e at the interface between wasteform and the rubble region is limited by its elemental solubility until the moment when the precipitate dis-appears. If there are multiple isotopes, then a precipitate of element e is assumed to consist of its isotopes,and the concentration of each isotope at the interface between waste form and the rubble rock region willbe a fraction of the elemental solubility. The fraction is assumed to be equal to the ratio of total mass of theisotope released to that of all the isotopes of element e, from a waste form by alteration.

If it is found that no precipitate occurs, then for the boundary condition at the interface a mass fluxof radionuclide k is assumed to be equal to its congruent release rate from the waste form. The congruentrelease is assumed to continue for the leach time of the waste solid, TL [yr] , which is the time period be-tween the beginning and the end of the waste form alteration.

The analytical solution for the concentrations C z tk ( , ) [mol/m3] of radionuclide k in the fractureare available for the mathematical problem described above [4]. The radiological hazard, H L tk ( , ) , forradionuclide k at the observation point L [m] distant from the waste canister is expressed as

H L t

vC L t DC

zS

ALIk

kL k

z Lf f

k

( , )

( , )

=−

=

∂∂

ε [Bq/yr]

[Bq/yr](2.28)

where ALIk is the annual limit on intake for ingestion of radionuclide k, given in 10CFR20 [6].A computer code was readily developed based on the analytical solutions [4]. The numerical re-

sults for the total radiological hazard (the sum of hazards of all radionuclides at the point L m away fromthe repository) based on (2.28) will be presented in Chapter 3 for the following cases: (1) the direct disposalof the LWR spent fuel, and (2) the disposal of wastes from the operation of the ATW system starting withthe same amount of LWR spent fuel.

14

3 Results of Repository Performance Assessment

3.1 Input Data3.1.1 LWR Spent Fuels

The total mass of LWR spent fuel to be disposed of in the proposed YMR is approximately 63,000metric tons (MT). If the in-drift emplacement scheme as described in Section 2.2.3 is used, waste packageshave the characteristics given in Table 3.1. The total number, 7640, of waste canisters is given in [11] isused. The mass (8.2 MTU) of one canister for disposal is obtained by dividing the total mass (63,000MTU) assumed in [11] by the total number of canisters (7,640). The leach time of 100,000 yr is assumedfor the spent fuel.

Masses and half-lives of radionuclides in the spent fuel are summarized in Table 3.2. These val-ues have been obtained and utilized in the previous performance assessment for the YMR [11].

The waste acceptance is assumed to start in 2010. The rate at which waste is placed in the reposi-tory and its age and amount of burnup are assumed based on the DOE Mission Plan Amendment [12]. Forthe present analysis, a Òdouble-blendedÓ waste stream is assumed as was done for TSPA-93 [11]. It con-sists of the equivalent of approximately 63,000 MT of spent fuel expressed as MT of uranium in the fuel asfabricated, or MTU, from both BWRs and PWRs. The use of the double-blended waste stream implies thatas the repository is filled, the waste being placed has fairly constant thermal-power output and radionuclideinventory. We therefore assume that the waste has uniform nuclear and thermal properties at the time ofemplacement, which makes the calculations applicable over the entire repository.

With the assumed double-blended waste stream, the weighted average and burn-up have been es-timated in Ref. [11]. For the purpose of determining the radiological properties of the spent fuel, both theBWR and the PWR fuels are considered to have an age of 25 years and burn-ups of 30,000 MWd/MTU and40,000 MWd/MTU, respectively. The inventory given in the third column of Table 3.2 is listed in Table 5-2 of Ref. [11], which was taken from the Characteristics Data Base [13]. The values in the third columnare the weighted average spent fuel inventory. The fourth column in the table is obtained by multiplyingthe value in the third column by 63,000 MTU. The values in the fourth column are the total masses of ra-dionuclides included in 63,000 MTU of the LWR spent fuel. The fifth column is obtained by dividing thevalue in the fourth column by the total number, 7640, of the canisters. The values for 246Cm, Am-242m,Pu-238, and 244Cm are not shown. These masses are lumped with those of their daughters due to their shorthalf-lives (less than 100 yr) or negligibly small inventories.

3.1.2 ATW System Parameters

The fractions, δ, transmuted by the burner per passage are given in Table 3.3. These numbers areobtained from [1]. We assume that the values in Table 3.3 are applied for each pass through the burner.Because actual values can be strongly dependent on the designs and neutronics of the burner core, determi-nation of these values will be a major task for future design studies for ATW.

In Table 3.4, fractions lost as waste from the recycling system are shown. Here two differentcases are assumed: minimum and realistic waste losses. The Òminimum waste lossÓ case is the one as-sumed in the LANL Report [1]. In this case, mass losses are assumed to occur only at the anodes of elec-tro-refining in the reprocessing and in the partitioning. No mass is assumed to be lost as waste from otherprocesses in the reprocessing and the partitioning. No waste generation is assumed at the TA fabrication.

The anode from the electro-refining of the reprocessing contains 0.37% of actinide processed.99.7% of actinide in the anode is recovered by the anode polishing. Thus, the polished anode contains0.3% of 0.37% of the mass processed at the electro-refining. The fraction, α1, is calculated as 1.1E-5 forthe minimum waste loss case.

At the partitioning, 9.47 ppm of the mass processed by the electro-refining is assumed to be in-cluded in the anode. 7.07% of actinide is included in the molten salt phase, and is transferred to the electro-winning process. It is assumed in Ref. [1] that all actinides are recovered by the electro-winning process,and are transferred to the TA fabrication.

In the Òrealistic wasteÓ loss case, in addition to the waste loss at the anode polishing process andthe electrorefining in the partitioning assumed for the minimum waste loss case, 0.3% of mass is arbitrarilyassumed to be lost as waste at other processes, based on previous experiences for EBR-II spent fuel treat-ment [14].

15

Table 3.1 Assumed Spent Fuel Canister

Parameters ValuesDiameter 1.52mLength 4.91mMass of spent fuel 8.2 MTUTotal number of canisters 7640 [11]

Table 3.2 Radionuclide Inventories in LWR Spent Fuel

Isotope Half life(years)

Inventory(Ci/MTU)

[11]

Inventory,

MiSF( )

(mol)

Inventory,

MiSF( ) / 7640

(mol/canister)

ALI(Bq/yr) [6]

Cm246 4.731E+03 3.22E-02 2.68E+01 1.75E+08Pu242 3.869E+05 1.74E+00 1.19E+05 1.55E+01 2.96E+04

Am242m 1.520E+02 9.62E+00 2.58E+02 5.62E+06Pu238 8.774E+01 2.43E+03 3.76E+04 3.25E+06U238 4.468E+09 3.15E-01 2.48E+08 3.25E+04 3.70E+05U234 2.445E+05 1.51E+00 6.52E+04 1.35E+01 3.70E+05Th230 7.700E+04 3.64E-04 4.96E+00 3.86E-02 1.48E+05Ra226 1.600E+03 2.22E-06 6.27E-04 1.53E-04 7.40E+04Cm243 2.850E+01 1.30E+01 6.55E+01 1.06E+06Am243 7.380E+03 1.71E+01 2.23E+04 2.66E+00 2.96E+04Pu239 2.406E+04 3.33E+02 1.41E+06 1.79E+02 2.96E+04U235 7.038E+08 2.28E-02 2.83E+06 3.75E+02 3.70E+05Pa231 3.277E+04 3.72E-05 2.15E-01 3.91E-04 1.48E+08Ac227 2.177E+01 1.80E-05 6.92E-05 2.52E-07 7.40E+03Cm245 8.499E+03 1.64E-01 2.45E+02 2.96E-02 2.59E+04Pu241 1.440E+01 4.07E+04 1.03E+05 5.02E-05 1.48E+06Am241 4.322E+02 3.26E+03 2.48E+05 9.35E+00 2.96E+04Np237 2.140E+06 3.78E-01 1.43E+05 5.53E+01 1.85E+04U233 1.585E+05 5.77E-05 1.61E+00 1.36E-02 3.70E+05Th229 7.339E+03 3.10E-07 4.01E-04 2.54E-05 2.22E+04Cm244 1.811E+01 7.62E+02 2.43E+03 6.70E+05Pu240 6.357E+03 5.09E+02 5.86E+05 6.93E+01 2.96E+04U236 2.341E+07 3.12E-01 1.29E+06 1.76E+02 3.70E+05Cs135 2.300E+06 4.79E-01 1.95E+05 2.55E+01 2.59E+07I129 1.570E+07 3.39E-02 9.37E+04 1.23E+01 1.85E+05

Sn126 1.000E+05 8.04E-01 1.42E+04 1.84E+01 1.11E+07Pd107 6.496E+06 1.14E-01 1.31E+05 1.71E+01 1.11E+09Tc99 2.130E+05 1.43E+01 5.36E+05 7.00E+01 1.48E+08Mo93 3.498E+03 1.80E-02 1.11E+01 1.19E-03 1.48E+08Nb94 2.030E+04 9.22E-01 3.30E+03 4.18E-01 3.33E+07Zr93 1.530E+06 2.23E+00 6.00E+05 7.86E+01 3.70E+07Se79 6.496E+04 4.53E-01 5.19E+03 6.72E-01 2.22E+07Ni59 8.000E+04 3.08E+00 4.34E+04 5.63E+00 7.40E+08Cl36 3.010E+05 1.10E-02 5.82E+02 7.61E-02 7.40E+07C14 5.729E+03 1.45E+00 1.46E+03 1.70E-01 7.40E+07

Table 3.3 Fractions Transmuted per Passage Through the Burner [1]

Element Values Element Values Element ValuesPu238 0.01 Cm244 0.0 Am242 0.0Pu239 0.47 Cm245 0.0 Am243 0.077Pu240 0.135 Cm247 0.0 Am241 0.4Pu241 0.19 Cm248 0.0 Tc99 0.168Pu242 0.0174 Cm242 0.0 U238 0.06

Cm243 0.0 Np237 0.385

16

Table 3.4 Fractions Lost as Waste per Passage Through the Material Recycle Process

Waste loss rateATW waste treatment Symbol

Minimum RealisticSpent fuel decladding αd 0 0.003Direct oxide reduction αo 0 0.003

Waste α r 0 0.003Impurity in U storage βr 0 0.003To anode polishing γ r 3.7E-3 3.7E-3

Electrorefining

To electrowinning εr 0.996 0.990Waste αa 0.003 0.003

Impurity in U storage βa 0 0.003Anode polishingTo electrowinning γ a 0.997 0.994

Waste αw 0 0.003Electrowinning

Impurity in U storage βw 0 0.003

Spent fuel re-processing

Total α1 1.1E-5 0.018Transmutationassembly fab-

ricationTotal α2 0 0.003

Chopping αc 0 0.003Electrorefining Waste α r 9.47E-6 9.47E-6

To electrowinning βr 0.0707 0.0707Electrowinning αw 0 0.003

Partitioning

Total α3 9.47E-6 0.00322

3.1.3 Repository Performance Parameters

In Table 3.5, the input data for the radionuclide transport calculations are summarized. The cylin-drical waste canister and buffer region are transformed to spheres with the same surface areas. The surfacearea of the cylindrical waste canister is obtained from Ref. [11]. Based on the radius of the excavated dis-posal drift and the spacing between two adjacent canisters [11], the surface area of the rubble region is de-termined.

The total mass of ATW waste is estimated to be approximately 21,000 MT [1], most of which iszircaloy LWR fuel cladding. The number of canisters for ATW wastes has been determined to be 2230 byassuming that 9.4 MT of waste is contained in one canister of the same dimensions as those for the LWRSF canister.

No specific materials have been decided for ATW waste solidification. It is likely that ATWwastes are categorized into several groups depending on the forms, radioactivity concentrations, and so on.Each group may be solidified with a different matrix, resulting in a different leach time, and in a differentconfiguration of the engineered barrier system around the waste canister. In this study, a single leach time,which is identical to that of the LWR spent fuel (SF) (100,000 yr), is assumed for the entire ATW waste.The repository configuration identical to the current YMR for LWR SF is assumed for the ATW waste dis-posal. By the aforementioned assumptions, it is hoped that comparison between the LWR SF disposal andthe ATW waste disposal can be made from the viewpoint of the ATW system performance. However, forfurther analyses, detailed waste characterization and optimization of repository design will be required.

The porosity of the host rock is assumed to be 0.1 [10]. The saturation of the host rock matrix isassumed to be 65% [10]. Thus, the effective porosity for matrix diffusion from water flowing fractures is0.065. The porosity of the rock rubbles filling the gap between the canister and the tunnel surface is as-sumed arbitrarily to be 0.3. The density of the rock matrix is 2.2 g/cm3 [10].

In Ref. [10], the water flow velocity through the vertical parallel fractures is assumed 40 m/yr.This value was used for the evaluation of the radiological hazard at the location 200 m below the repositoryhorizon, where the water table is considered to be located. The water flow is considered to be horizontalbelow the water table, and is assumed slower than the vertical flow in the unsaturated region. Within thetransport over 5,000 m, both the vertical transport through the unsaturated region and the horizontal trans-port through the saturated region are included. In this preliminary calculation, a single value of 10 m/yr is

17

used for the evaluation of the hazard at the location 5,000 m away from the repository.The same value is assumed for the diffusion coefficient in the buffer for all the radionuclides. The

value for the diffusion coefficient in the rock matrix is assumed smaller than that for the rubble in thebuffer region.

The retardation due to sorption on to the fracture surfaces is conservatively neglected. The spac-ing and the aperture of the fractures are taken from Ref. [10].

The hazard has been evaluated numerically for two different geologic conditions: ÒmobileÓ andÒimmobileÓ. A medium for which the upper-bound values of the ranges given in Ref. [11] are assigned forradionuclide solubilities, and the lower-bound values of the ranges are assigned for sorption retardationcoefficients, is referred to as a ÒmobileÓ medium. A medium with the lower-bound values for solubilitiesand the upper-bound values for sorption retardation coefficients is referred to as an ÒimmobileÓ medium. InTable 3.6, values are shown, which are determined from the ranges reported in previous performance as-sessments [4][11][15].

For the rubble-rock region, the same value for the sorption distribution coefficient is used for theretardation coefficient in the buffer (2.27) and for the capacity factor in the rock matrix (2.24).

Table 3.5 Input Data for Radionuclide Transport Analysis

Parameters ValuesSurface are of waste form 28.3m2

Surface area of buffer 726 m2

Distance between waste forms 30mEquivalent radius of spherical waste form, r1 1.5mEquivalent radius of buffer (crushed tuff), r2 7.6m

Leach time of waste form 100,000 yrMatrix rock (tuff) porosity 0.1

Saturation of the host rock matrix pores, S 0.65Buffer (crushed tuff) porosity 0.3Matrix rock and buffer density 2.2 g/cm3

Dispersion coefficient 1.0 m2/yrWater velocity 10 m/yr

Diffusion coefficient in matrix rock 2.05E-4 m2/yrDiffusion coefficient in buffer 3.15E-3 m2/yr

Retardation in the fracture 1Fracture spacing, 2a 1mFracture aperture, 2b 1.8E-4m

Number of fractures intersecting repository 2,000Repository dimension, LR 2000m

Table 3.6 Data for Radionuclide Transport in Geologic Formations (Upper Row:ÒImmobileÓCase, Lower Row:ÓMobileÓ Case).

ElementSolubility*(mol/m3)

Sorption dis-tribution co-

efficient*(cm3/g)

Retardationcoefficientin buffer

(2.27)

Capacityfactor in

rock(2.24)

1.E-03 100 5.14E+02 2.06E+02Cm,Am, Ac 1.E-07 2000 1.03E+04 4.11E+03

1.E-03 30 1.55E+02 6.18E+01Pu

1.E-07 200 1.03E+03 4.12E+021.E+01 0 1 6.50E-02

Np1.E-05 100 5.14E+02 2.06E+021.E+01 0 1 6.50E-02

U1.E-06 20 1.04E+02 4.12E+011.E-02 0 1 6.50E-02

Pa1.E-07 100 5.14E+02 2.06E+02

18

1.E-04 100 5.14E+02 2.06E+02Th

1.E-07 2000 1.03E+04 4.11E+031.E-02 100 5.14E+02 2.06E+02

Ra1.E-06 5000 2.57E+04 1.03E+04

100 5.14E+02 2.06E+02Cs ÐÐÐ

3000 1.54E+04 6.17E+030 1 6.50E-02

I, Tc, Cl ÐÐÐ0 1 6.50E-02

1.E-04 20 1.04E+02 4.12E+01Sn

1.E-08 300 1.54E+03 6.17E+0220 1.04E+02 4.12E+01

Pd, Mo ÐÐÐ300 1.54E+03 6.17E+02

1.E-04 100 5.14E+02 2.06E+02Nb

1.E-06 2000 1.03E+04 4.11E+031.E-04 100 5.14E+02 2.06E+02

Zr1.E-09 2000 1.03E+04 4.11E+03

0 1 6.50E-02Se ÐÐÐ

20 1.04E+02 4.12E+011.E+02 0 1 6.50E-02

Ni1.E-03 500 2.57E+03 1.03E+03

100 5.14E+02 2.06E+02C ÐÐÐ

300 1.54E+03 6.17E+02* SAN93-2675, [11] and SAND94-2563, [15].

3.2 Results of Mass Flow Analysis

3.2.1 Mass of Radionuclide in ATW Waste

In the present analysis, only the mass flow of actinides and 99Tc is considered. Other materialssuch as fission products except for 129I and 99Tc, and structural materials in LWR spent fuel, are not consid-ered because they are removed from the LWR SF by the reprocessing and from the spent TA by the parti-tioning. Because contribution of those neglected radionuclides is negligibly small, comparison between thehazard from the whole LWR SF and the hazard from those radionuclides shown in this table is still mean-ingful. For 99Tc, the mass flow analysis has been performed in the same way as for actinides, based on themodel shown in Chapter 2. The mass of 129I in the LWR spent fuel is assumed to be reduced by a factor of15 by the ATW system [1] for all the cases, without performing the mass flow analysis.

Uranium isotopes are also omitted from this consideration because the LANL report [1] assumesthat uranium is to be removed from the LWR spent fuel and sent to a low-level waste repository. However,as we observe later, the recovered uranium has considerable radiological impact. The effects of recovereduranium are discussed later in this chapter.

In Table 3.7, the total waste fraction, F i∞, , defined by (2.9), is shown for two cases: ÒminimumÓ

and ÒrealisticÓ (see Section 3.1.2 and Table 3.4) in the second and third columns. For those radionuclideswhose fraction transmuted, δ, is zero, such as 246Cm, 245Cm, and 243Cm (see Table 3.3), the total wastefraction is unity for both minimum and realistic cases. This means that all the masses originally included inthe LWR spent fuel will be transferred to the waste stream. For those radionuclides with δ greater thanzero, the total fraction, F i∞, , becomes less than unity. F i∞, is smaller for the minimum waste case than that

for the realistic case.In the fourth and fifth columns, the mass of radionuclide i in the waste from the ATW system des-

tined to the repository, M FiSF

i( )

,∞ ., is tabulated. The mass, MiSF( ) , initially included in the LWR spent fuel

is given in the fourth column of Table 3.2.In the sixth and seventh columns, the mass of a radionuclide per canister is shown, which is ob-

tained by dividing the values in the fourth and fifth columns by the number of canisters. The number ofcanisters for wastes generated by ATW is estimated by assuming that 9 MT of waste is contained in onecanister. With the total mass of waste (21,000 MT), 2,230 canisters are assumed for ATW waste disposal.The values for 246Cm, 244Cm and 243Cm are not given in the table. These masses are lumped with those of

19

their daughters to simplify the decay chains. For example, the mass of 246Cm is lumped with that of 242Pu.Thus, the mass of 242Pu shown in the table includes this contribution already.

Table 3.7 Radionuclides Data and Results of ATW Mass Flow Analysis.

Total waste fraction, F i∞,Inventory (mol)

M FiSF

i( )

,∞

Waste per canister

M FiSF

i( )

, /∞ 2230(mol/canister)

Isotope

Minimum Realistic Minimum Realistic Minimum RealisticCm246 1.00E+00 1.00E+00 2.68E+01 2.68E+01Pu242 1.34E-04 9.40E-02 1.59E+01 1.12E+04 1.87E-02 4.88E+00Cm243 1.00E+00 1.00E+00 6.55E+01 6.55E+01Am243 1.25E-04 8.87E-02 2.78E+00 1.97E+03 2.98E-02 8.90E-01Pu239 2.20E-05 2.77E-02 3.09E+01 3.89E+04 1.35E-02 1.70E+01Cm245 1.00E+00 1.00E+00 2.45E+02 2.45E+02 1.07E-01 1.07E-01Pu241 5.17E-05 4.61E-02 5.32E+00 4.75E+03 2.32E-03 2.07E+00Am241 2.55E-05 2.99E-02 6.33E+00 7.42E+03 2.76E-03 3.24E+00Np237 2.64E-05 3.05E-02 3.77E+00 4.34E+03 1.64E-03 1.90E+00Cm244 1.00E+00 1.00E+00 2.43E+03 2.43E+03Pu240 7.20E-05 5.83E-02 4.22E+01 3.42E+04 1.08E+00 1.60E+01I129 6.67E-02 6.67E-02 6.24E+03 6.24E+03 2.73E+00 2.73E+00Tc99 5.82E-05 5.01E-02 3.12E+01 2.68E+04 1.36E-02 1.17E+01

3.2.2 Decay Chains

For actinides, decay chains effect must be taken into account. The following chains are consid-ered:240Pu → 236U → 232Th → 228Ra → 228Th228 → 224Ra224 →245Cm → 241Pu → 241Am → 237Np → 233U → 229Th → 225Ra →242Pu → 238U → 234U → 230Th → 226Ra →243Am → 239Pu → 235U → 231Pa → 227Ac → 227Th →

3.3 Change of Radionuclide Mass in Repository by Radioactive Decay

3.3.1 Radionuclides in LWR Spent Fuel and ATW Waste

Masses of radionuclides in LWR spent fuel and in the waste from the ATW system with time areshown as functions of time in Figure 3.1 to Figure 3.6.

The masses of uranium isotopes in the repository as LWR spent fuel and as the ATW waste areshown in Figure 3.1. For the ATW waste, the results for both minimum and realistic cases are shown. Al-though uranium is assumed not to exist in the ATW waste initially, due to the decay of higher actinides,masses of uranium isotopes in ATW waste increase. From the ATW system, if the waste generation is assmall as assumed in the minimum waste loss case, the mass of uranium in the repository is reduced to lessthan 100 moles, or 24 kg. For the realistic case, the total mass of uranium is at most 40,000 moles, or 10MT.

The inventory change of the plutonium is shown in Figure 3.2. Due to transmutation of 239Pu inthe ATW burner, the plutonium inventory in ATW waste for the realistic waste loss case is 0.8% of theinventory in LWR spent fuel. If the waste loss is as small as assumed for the minimum waste case, lessthan 100 moles of plutonium isotopes exist in the repository.

Due to the transmutation of Am and Np, inventories in the ATW waste with realistic waste lossfractions are significantly smaller than those of LWR spent fuel as shown in Figure 3.3. The inventory ofCm is unchanged, however, because of no transmutation assumed by the ATW burner. The inventorychanges of other radionuclides are shown in Figure 3.4, Figure 3.5, and Figure 3.6. Like for Pu, with theminimum waste losses, radionuclide masses in the repository can be as small as less than 100 moles,whereas with the realistic waste losses, radionuclide masses in the repository range between 1 and 10 MT.

From the results shown above, masses of thermally fissile actinides are depicted in Figure 3.7,Figure 3.8, and Figure 3.9. The masses of thermally fissile nuclides can be used as a measure for the criti-cality safety of the repository [16]. The repository with LWR spent fuel contains about 1000 MT of ther-mally fissile materials (Figure 3.7). This can be reduced to 10 MT (Figure 3.8) by ATW with the realistic

20

waste loss, and to 10 kg (Figure 3.9) with the minimum waste loss.For the realistic waste loss case, the judgement is more complex. As shown in Figure 3.10, the

fraction of thermally fissile actinides in the total actinide mass is as high as 40% at early times. The fractionof fissile actinides in LWR spent fuel is 1.7% initially and decreases continuously.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

109

Comparison of Uranium Inventory

U234

U235U238

U234

U235U238

U235

U234

U238

U236U233

LWR Spent Fuel ATW Realistic ATW Minimum

Inve

ntor

y (m

ol)

Time (year)

Figure 3.1 Uranium inventory in LWRspent fuel and in ATW waste.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

Comparison of Plutonium Inventory

Pu242

Pu241

Pu239

Pu240

LWR Spent Fuel ATW Realistic ATW Minimum

Inve

ntor

y (m

ol)

Time (year)

Figure 3.2 Plutonium inventory in LWRspent fuel and in ATW waste.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

Am243

Comparison of Am, Cm, Np Inventory

Am243

Cm245

Np237

Am241

LWR Spent Fuel ATW Realistic ATW Minimum

Inve

ntor

y (m

ol)

Time (year)

Figure 3.3 Inventories of Am, Cm, and Npinventory in LWR spent fuel andin ATW waste.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

109

Comparison of Thorium Inventory

Th230

Th232

Th227

Th232

Th230

Th232

Th228

Th230

Th229

LWR Spent Fuel ATW Realistic ATW Minimum

Inve

ntor

y (m

ol)

Time (year)

Figure 3.4 Thorium inventory in LWRspent fuel and in ATW waste.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

Ra228Ra225

Comparison of Radium Inventory

Ra226

Ra226

Ra228Ra225

LWR Spent Fuel ATW Realistic ATW Minimum

Inve

ntor

y (m

ol)

Time (year)

Figure 3.5 Radium inventory in LWR spentfuel and in ATW waste.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

Pa231

Ac227

Pa231

Comparison of Pa, Ac, I, Tc Inventory

I129

Pa231

Tc99

Ac227

LWR Spent Fuel ATW Realistic ATW Minimum

Inve

ntor

y (m

ol)

Time (year)

Figure 3.6 Inventories of Pa, I, TC, and Acin LWR spent fuel and in ATWwaste.

21

100 101 102 103 104 105 106 107 108 109 101010-3

10-2

10-1

100

101

102

103

104

105

106

107p

Total

U235

U233

Pu239

Pu241

Mas

s (k

g)

Time (year)

Figure 3.7 Masses of thermally fissileactinides in LWR spent fuel.

100 101 102 103 104 105 106 107 108 109 101010-3

10-2

10-1

100

101

102

103

104

105

106

107

Total

U235

U233Pu239

Pu241

Ma

ss (

kg)

Time (year)

Figure 3.8 Masses of thermally fissileactinides in ATW waste withrealistic waste loss.

100 101 102 103 104 105 106 107 108 109 101010-3

10-2

10-1

100

101

102

103

104

105

106

107

Total

U235U233

Pu239

Pu241

Mas

s (k

g)

Time (year)

Figure 3.9 Masses of thermally fissileactinides in ATW waste withminimum waste loss.

100 101 102 103 104 105 106 107 108 109 10100

10

20

30

40

50

60

70

80

ATW - Minimum

ATW - Realistic

LWR Spent Fuel

Fra

ctio

n (%

)

Time (year)

Figure 3.10 Thermally fissile fractions inactinides in LWR spent fuel andATW waste.

3.3.2 Effect of Recovered Uranium

In the ATW scheme proposed by LANL [1], uranium is removed from the LWR spent fuel at thereprocessing stage. The separated uranium is assumed to be disposed of in some LLW repository, which isnot identified. Here, we observe the effect of the uranium recovered from the LWR spent fuel in case thatit is disposed of in the YMR together with the wastes from the ATW operation.

By combining the ATW waste with the recovered uranium, the total mass of uranium to be dis-posed of in YMR becomes comparable to that included in LWR spent fuel. Thus, as shown in Figure 3.11,the masses of 238U, 235U, 236U, and 234U are almost the same among the three cases: LWR spent fuel directdisposal, ATW waste with realistic waste losses plus recovered uranium, and ATW waste with minimumwaste losses plus recovered uranium. The major difference is observed in the mass of 233U, which is thedecay daughter of 237Np. Since 237Np is transmuted by the ATW burner, the mass of 233U in ATW wastebecomes smaller than that in LWR spent fuel.

The mass of thorium also shows a similar trend with the uranium mass due to the influence of ura-nium, as shown in Figure 3.12. Th-229 is the daughter of 233U, resulting in effects of 237Np transmutationsimilar to 233U.

With the ATW system, the total mass of thermally fissile actinides other than 235U is reduced sig-nificantly (see Figure 3.7, Figure 3.13 and Figure 3.14). Because the mass of 235U contained in the recov-ered uranium is dominant, there are negligible differences among three cases. By dilution with the recov-ered uranium and by the transmutation of thermally fissile actinides and their precursors, the mass fractionof thermally fissile actinides in the repository is reduced significantly.

From the viewpoint of the attractiveness to the proliferator, the ATW waste seems to have an ad-

22

vantage over the LWR spent fuel in case the waste generation from the ATW system is minimized. On theother hand, the attractiveness would not be linearly proportional to the mass of fissile materials, butstrongly dependent on ease and cost with which weapons materials are recovered from ATW wastes. Thus,the judgement for the realistic waste loss case is more complex, and requires more detailed analyses.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

109

Comparison of Uranium Inventorywith Recovered Uranium

U235

U234

U238

U236U233

LWR Spent Fuel ATW Realistic + U ATW Minimum + U

Inve

nto

ry (

mo

l)

Time (year)

Figure 3.11 Uranium inventory in LWRspent fuel and in ATW wasteplus recovered uranium.

100 101 102 103 104 105 106 107 108 109 101010-5

10-4

10-3

10-2

10-1

100

101

102

103

104

105

106

107

108

109

Comparison of Thorium Inventorywith Recovered Uranium

Th227

Th232

Th228

Th230

Th229

LWR Spent Fuel ATW Realistic + U ATW Minimum + U

Inve

nto

ry (

mol)

Time (year)

Figure 3.12 Thorium inventory in LWRspent fuel and in ATW wasteplus recovered uranium.

100 101 102 103 104 105 106 107 108 109 101010-3

10-2

10-1

100

101

102

103

104

105

106

107Mass of Fissile Actinides in ATW+U - Realistic

Total

U235

U233

Pu239

Pu241

Ma

ss (

kg)

Time (year)

Figure 3.13 Masses of thermally fissileactinides in ATW waste withrealistic waste loss plusrecovered uranium.

100 101 102 103 104 105 106 107 108 109 101010-3

10-2

10-1

100

101

102

103

104

105

106

107Mass of Fissile Actinides in ATW+U - Minimum

Total

U235

U233

Pu239

Pu241

Ma

ss (

kg)

Time (year)

Figure 3.14 Masses of thermally fissileactinides in ATW waste withminimum waste loss plusrecovered uranium.

3.4 Hazard from One Canister of LWR Spent FuelDue to the decay and the different behaviors of the radionuclides in the geologic medium, the ra-

diological impact of the repository should be compared in terms of the impact of radionuclides reaching theaccessible environment boundary set at the 5-km location from the repository boundary. The transportmechanism of the radionuclides through the geologic medium is described in Section 2.3.2 in detail.

Aiming at bounding the uncertainties of the geologic data, we have used two sets of input data forthe hazard calculation. One is referred to as the mobile medium, i.e., low retardation and high solubility.The other is as the immobile medium, i.e., large retardation and low solubility (see Table 3.6). The defini-tion of the hazard is given by (2.28).

23

101 102 103 104 105 106 10710-10

10-9

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10-6

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10-4

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10-2

10-1

100

101

102

103

Ra224

Th228

Ra228Th232

Th227

Ni59

Pa231

Ra226Ac227

U235

U236U238

U234

Ra225

Se79

Th229

U233

Np237

Cl36

Tc99

I129

From one waste canister containing 8.24 MTULowest Kd's and highest solubilitiesYucca MountainLWR spent fuel

Ha

zard

at

50

00

m f

rom

EB

S

Time, year

Figure 3.15 Hazard at the 5,000 m locationfrom one canister of LWR spentfuel for the mobile medium.

101 102 103 104 105 106 10710-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

100

101

102

103

Cl36

Tc99

I129

From one waste canister containing 8.24 MTUHighest Kd's and lowest solubilitiesYucca MountainLWR spent fuel

Ha

zard

at

50

00

m f

rom

EB

S

Time year

Figure 3.16 Hazard at the 5,000 m locationfrom one canister of LWR spentfuel for the immobile medium.

101 102 103 104 105 106 10710-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

100

101

102

103

Th230

Ra224Th228

Ra226

Th227

Ra231

From one canister of ATW waste containing 9 MTHMLowest Kd's and highest solubilitiesYucca MountainRealistic waste fraction lost as waste

Ra228Th232

U234U238

Ac227 U235Ra225

U236

Th229

U233

Np237

Tc99

I129

Haza

rd a

t 5000m

fro

m E

BS

Time, year

Figure 3.17 Hazard at the 5,000 m locationfrom one canister of ATW wastewith realistic waste fraction forthe mobile medium.

101 102 103 104 105 106 10710-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

100

101

102

103

From one canister of ATW waste containing 9 MTHMHighest Kd's and lowest solubilitiesYucca MountainRealistic waste fraction lost as waste

Tc99

I129

Ha

zard

at

50

00

m f

rom

EB

S

Time, year

Figure 3.18 Hazard at the 5,000 m locationfrom one canister of ATW wastewith realistic waste fraction forthe immobile medium.

Numerical results for the hazards at the 5-km location, resulting from one canister of the LWRspent fuel disposed of in YMR, are shown in Figure 3.15 and Figure 3.16. The former is for the case withthe mobile medium assumption, while the latter is for the case with the immobile medium assumption.

In Figure 3.15, it is observed that the hazard is dominated by 237Np. Other important contributorsare 234U, 236U, 233U and 238U. The contribution of all uranium isotopes is as small as 11.9% of the total haz-ard arising from LWR spent fuel while about 98% of the mass of the LWR spent fuel is due to them. Bothuranium and neptunium are assumed to have a small retardation and a large solubility, resulting in thesimilar behavior in the geologic medium. However, the toxicity of 237Np is greater than that of uraniumbecause the ALI for 237Np is smaller by a factor of about 30 than those of uranium isotopes (see Table 3.2).Thus, the hazard of the neptunium appears at the top in the figure.

3.5 Hazard from One Canister of ATW Waste with Realistic Waste FractionThe hazard from 237Np is the main contributor in the ATW waste for the mobile medium. The

237Np inventory in the entire ATW waste is 4340 moles (see Table 3.7), which is smaller by a factor of 33than that in LWR spent fuel (143,000 moles, see Table 3.2). The hazard from one ATW canister observedat the 5 km location is smaller than that from one canister of LWR spent fuel by the factor of 9 (compareFigure 3.15 and Figure 3.17).

24

In the case of 99Tc, the per-canister inventory is decreased by a factor of 6 after the ATW opera-tion. Because 99Tc is assumed to be soluble, it is released congruently with the waste matrix; the magnitudeof the hazard is proportional to the mass. Thus, the hazard is decreased by a factor of 6 from that of LWRspent fuel (compare Figure 3.15 and Figure 3.17). In the case of 129I, the per-canister inventory is decreasedby a factor of 5 after the ATW operation. Thus, the hazard is decreased by a factor of 5 from that of LWRspent fuel (compare Figure 3.15 and Figure 3.17), due to the congruency with the waste matrix.

In the immobile medium, the transport of the actinide is so slow that we do not observe any othercontributors than those of 129I and 99Tc. For 99Tc and 129I, the curves shown in Figure 3.18 are the same asthose shown in Figure 3.17, due to the assumption that these two radionuclides are soluble and no sorbing.

3.6 Hazard from One Canister of ATW Waste with Minimum Waste Frac-tion

The hazard from 237Np is the main contributor in the ATW waste. The 237Np inventory in the en-tire ATW waste becomes 3.77 moles (see Table 3.7), which is smaller by a factor of 38,000 than that in theentire LWR spent fuel (143,000 moles, see Table 3.2), and by a factor of 1150 than that in ATW waste withthe realistic waste losses (4340 moles, Table 3.7). The hazard from one ATW canister observed at the 5 kmlocation is smaller by the factor of 566 than that from one canister of LWR spent fuel (compare Figure 3.15and Figure 3.19), and by a factor of 80 than that from one canister of ATW waste with the realistic wastelosses (compare Figure 3.17 and Figure 3.19).

101 102 103 104 105 106 10710-10

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10-8

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10-1

100

101

102

103

Pa231

From one canister of ATW waste containing 9 MTHMLowest Kd's and Highest solubilitiesYucca MountainMinimum waste fraction lost as waste

Th232U234

U238

Ac227

U235Ra225

U236

Th229

U233

Np237

Tc99

I129

Haz

ard

at 5

000m

from

EB

S

Time, year

Figure 3.19 Hazard at the 5,000 m locationfrom one canister of ATW wastewith minimum waste fraction forthe mobile medium.

101 102 103 104 105 106 10710-10

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10-8

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101

102

103

From one canister of ATW waste containing 9 MTHMHighest Kd's and lowest solubilitiesYucca MountainMinimum waste fraction lost as waste

Tc99

I129

Haz

ard

at 5

000m

from

EB

S

Time, year

Figure 3.20 Hazard at the 5,000 m locationfrom one canister of ATW wastewith minimum waste fraction forthe immobile medium.

In the case of 99Tc, the per-canister inventory is decreased by a factor of 5150 after the ATW op-eration. Because 99Tc is assumed to be soluble, it is released congruently with the waste matrix; the mag-nitude of the hazard is proportional to the mass. Thus, the hazard is decreased by a factor of 5150 from thatof LWR spent fuel. In the case of 129I, the per-canister inventory is decreased by a factor of 5 after theATW operation. Thus, the hazard is decreased by a factor of 5 from that of LWR spent fuel, due to thecongruency with the waste matrix.

In the immobile medium, the transport of the actinide is so slow that we do not observe any othercontributors than those of 129I and 99Tc. For 99Tc and 129I, the curves shown in Figure 3.20 are the same asthose shown in Figure 3.19, due to the assumption that these two radionuclides are soluble and no sorbing.The curve for 129I in Figure 3.20 is identical to that shown in Figure 3.18.

3.7 Hazard from One Canister of Recovered UraniumIn this section, the hazard from one canister containing only recovered uranium is considered. The

recovered uranium is assumed to be placed together with the ATW waste at the same repository. In Section3.8, we compare the hazard from the entire LWR spent fuel with that from the ATW waste plus recovereduranium.

If we assume that one canister contains 8.2MT of recovered uranium, the total number of can-

isters for recovered uranium is 7,285 for 60,000 MTof uranium.

25

The total initial inventory of uranium iso-topes and the decay chains are as follows:236U(1.29E6mol) → 232Th → 228Ra → 228Th → 224Ra→233U(1.61E0 mol) → 229Th → 224Ra →238U(2.48E8mol) → 234U(6.52E4) → 230Th → 226Ra→235U(2.83E6 mol) → 231Pa → 227Ac → 227Th →

Figure 3.21 shows that the main contribu-tors to the hazard are 234U, 238U and 236U. The haz-ard level of the recovered uranium is smaller by afactor of 16 than that of LWR spent fuel, for whichthe main hazard contributor is 237Np. No figure isshown for the immobile medium, since there is nohazard from recovered uranium at the 5 km location.

101 102 103 104 105 106 10710-10

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10-1

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101

102

103

Ra224Th228

Ra228Th232

Th227

Pa231

Ra226Th230

U235

Ac227

U238U236

U234

Ra225

Th229

U233

From one canister containing 8.24 MTLowest Kd's and highest solibilitiesRecovered Uranium onlyYucca Mountain

Ha

zard

at

50

00

m f

rom

EB

S

Time, year

Figure 3.21 Hazard at the 5,000 m locationfrom one canister containingonly uranium recovered byreprocessing for the mobilemedium.

3.8 Comparisons of Total Hazard

3.8.1 LWR and ATW without Recovered Uranium

The hazard from 2230 canisters of ATW waste is compared for the mobile medium case, with thatfrom 7640 canisters of LWR spent fuel, disposed of in YMR in Figure 3.22. The same comparison is madefor the immobile medium case in Figure 3.23. For both figures, the results for the 5 km location are shown.

For the total hazards from all LWR spent fuel canisters in the repository observed at the 5 km lo-cation, the main contributors to the total hazard in the mobile medium are 237Np, whereas main contributorsin the immobile medium are 129I and 99Tc. The difference between the mobile and immobile cases is about350-fold.

At the 5-km point, practically no hazard is observed until 50,000 years regardless of the mediumassumption. This is because, even for 129I and 99Tc for which no retardation is assumed, radionuclide mi-gration through fractured media can be delayed significantly due to the diffusion of radionuclides fromfractures into surrounding rock matrix [17]. If the 10,000-year time frame is the major concern for the re-pository safety as is specified by the EPA regulation [7], the repository already complies with the regula-tion.

It is hoped that by pursuing the ATW (1) the hazard of the waste will be reduced significantly, (2)the length of time for the repository safety assessment can be reduced significantly, and (3) the uncertain-ties due to the heterogeneous geologic formations can be reduced. Hereafter, we observed how or whetherthese goals are achieved by the current ATW concept.

For the total hazard of ATW waste with realistic waste loss assumptions, the main contributor inthe mobile medium is 237Np, whereas in the immobile medium, the main hazard comes from 129I and 99Tc.At the 5 km location, the difference between the mobile and immobile cases is about 150-fold. The totalhazard is smaller for the ATW realistic waste loss case than that for the LWR spent fuel, by a factor of 35for the mobile medium, and by a factor of 15 for the immobile medium.

Thus, from the first viewpoint (reduction of hazard magnitude), it is observed that with the realis-tic assumptions for waste loss from the ATW system about a factor of 30 reduction is expected for the ra-diological hazard from the repository.

From the second viewpoint (length of time frame for repository safety assessment), it is observedthat nothing has changed due to ATW. The hazard rises at about 50,000 year at the 5km location.

From the third viewpoint (reduction of the uncertainties), it is observed that the difference betweenthe mobile and immobile for the ATW realistic case is about 150 fold, whereas the difference for the LWRspent fuel is about 350-fold. Thus, the uncertainties are reduced by a factor of two.

For the minimum waste fraction case, the main contributors at the 5000 m location in the mobilemedium are 237Np and 129I, whereas in the immobile medium, the hazard comes from 129I only.

26

From the first viewpoint (reduction of hazard magnitude), with the minimum waste loss from theATW system, about a factor of 1300 reduction is expected for the radiological hazard from the repository,compared with that of the total hazard of LWR spent fuel disposal for the mobile medium case. For themobile case, also observed is that the difference between the realistic and minimum is about 35-fold. Forthe immobile case, due to the fixed reduction for 129I (the factor of 15 in any case), the difference betweenthe realistic and minimum waste loss cases are negligibly small.

It is also pointed out that the reduction is greater for the mobile case, for which the radiologicalsafety of the repository is most jeopardized. Thus, the introduction of the ATW system is more justifiablewhen the host geologic formations turn out to be ÒmobileÓ, from the viewpoint of radiological safety im-provement.

From the second viewpoint (length of time frame for repository safety assessment), it is observedthat nothing has changed due to ATW. The hazard rises at about 50,000 year at the 5km location. Thus, bythe ATW introduction, the time frame for the repository safety assessment will not be reduced regardless ofthe ATW system performance.

From the third viewpoint (reduction of the uncertainties), it is observed that the difference betweenthe mobile and immobile for the ATW minimum case is about 4 fold, whereas the difference for the realis-tic waste loss case is about 150 fold, and about 350-fold for the LWR spent fuel. Thus, the uncertaintiescould be reduced significantly with the minimized waste loss from ATW.

101 102 103 104 105 106 10710-4

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101

102

103

104

105

106

107

Mobile Medium

ATW, Minimum

ATW, Realistic

LWR

Haz

ard

Time, year

Figure 3.22 Comparison of hazard from2230 canisters of ATW wastewith that from 7640 canisters ofLWR spent fuel, disposed of inYMR surrounded by a mobilemedium.

101 102 103 104 105 106 10710-4

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101

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104

105

106

107

Immobile Medium

ATW, Minimum

ATW, RealisticLWR

Haz

ard

Time, year

Figure 3.23 Comparison of hazard from2230 canisters of ATW wastewith that from 7640 canisters ofLWR spent fuel, disposed of inYMR surrounded by animmobile medium.

3.8.2 LWR and ATW with Recovered Uranium

In this section, the hazard of the recovered uranium is estimated, under the assumption that it isdisposed of at the same repository together with the ATW waste. For the hazard comparison, the hazardfrom ATW waste is combined with that from the recovered uranium, and is compared with that from LWRspent fuel.

The results for the mobile medium are shown in Figure 3.24. The results for the 5000 m locationare shown for the cases of (1) LWR spent fuel direct disposal, (2) ATW waste with the realistic wastelosses plus recovered uranium, and (3) ATW waste with the minimum waste losses plus recovered ura-nium. The results for these cases are shown in Figure 3.25 for the immobile medium, which turned out tobe identical to those shown in Figure 3.23 due to the assumed immobility for uranium isotopes.

For the mobile medium, the hazard reduction from LWR to ATW with realistic waste fractionwithout recovered uranium is 35 fold. The hazard reduction from LWR to ATW with minimum waste frac-tion without recovered uranium is 1,300 fold (see Figure 3.22). However, by combining the hazard fromthe recovered uranium, the hazard reduction becomes only 14 fold (from LWR to ATW minimum plusrecovered uranium) or 12 fold (from LWR to ATW realistic plus recovered uranium).

27

3.9 Discussions

3.9.1 Radiological Hazard

(1) Reduction of radiological hazard arising from waste disposal:The total hazard from ATW waste is always smaller than that from LWR spent fuel. The hazard

from ATW waste combined with the recovered uranium is also smaller than that from LWR spent fuel.Thus, ATW always gives the reduction in the radiological hazard by transmuting the actinide and the long-lived fission products, which are the significant contributors to the hazard.

Table 3.8 summarizes the effect of peak hazard reduction by the ATW system for different me-dium conditions and different operation conditions. The hazard is compared with the peak hazard arisingfrom the entire LWR spent fuel for the mobile medium.

For the mobile medium, recovered uranium could contribute significantly to the total hazard, sothat the difference between the realistic waste loss case and the minimum waste loss case becomes negligi-ble, and that the difference between the LWR spent fuel and the ATW waste could also be as small as 14fold.

101 102 103 104 105 106 10710-4

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101

102

103

104

105

106

107

Mobile Medium

ATW+U, Realistic

ATW+U, Minimum

LWR

Haz

ard

Time, year

Figure 3.24 Comparison of hazard from2230 canisters of ATW wasteplus 7285 canisters of recovereduranium with that from 7640canisters of LWR spent fuel,disposed of in YMR surroundedby a mobile medium.

101 102 103 104 105 106 10710-4

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101

102

103

104

105

106

107

Immobile Medium

ATW+U, Realistic

LWR

ATW+U, MinimumHaz

ard

Time, year

Figure 3.25 Comparison of hazard from2230 canisters of ATW wasteplus 7285 canisters of recovereduranium with that from 7640canisters of LWR spent fuel,disposed of in YMR surroundedby an immobile medium.

Table 3.8 Reduction of peak hazard at the 5-km location by the ATW system.

Medium Option LWR ATW realistic ATW minimum Major contributor

Without U 1/35 1/1,300Mobile

With U1

1/12 1/14Np-237

Without U 1/5,250 1/5,250Immobile

With U1/350

1/5,250 1/5,250I-129

In the immobile (large retardation and small solubility) medium, the hazard from ATW is reducedby a factor of 15, compared to that from LWR spent fuel for the realistic or minimum waste loss assump-tions. This is due to the dominant hazard contribution of 129I. In the immobile medium, the hazard fromactinide becomes negligible, due to its slow transport through geologic formations. We have used the samewaste fraction of 129I in both ATW process assumptions.(2) Reduction of the length of time for the repository safety assessment:

For any case investigated, it is observed that the hazard rises at about 50,000 year at the 5km loca-tion. Thus, the introduction of the ATW does not reduce the time frame for the repository safety assess-ment regardless of the ATW system performance. If the 10,000-year time frame is the major concern for

28

repository safety as is specified by the EPA regulation [7], the repository can comprise with the regulationwithout ATW application. (3) Reduction of the uncertainties:

Table 3.9 shows the comparison of uncertainty ranges. The values show the ratio of the peak haz-ard for the mobile medium to that for the immobile medium for each case, which can be obtained from thevalues in Table 3.8. This table shows that with recovered uranium in the same repository, there is negligi-ble effect of uncertainty reduction. If recovered uranium is not included in the repository, then some de-gree of uncertainty reduction is observed, especially the uncertainties for the minimum waste case withoutrecovered uranium shows significant reduction. Thus, significant reduction of uncertainties associated withthe radiological impact evaluation has not been observed.

We need to compare uncertainties based on statistical analyses, where the probability distributionfunction is obtained for each case by assuming probability density distribution functions for parameters.The comparison made here is merely a crude estimate for the uncertainties.

Table 3.9 Comparison of uncertainty ranges.

ATWRealistic MinimumLWRWithout U With U Without U With U

Uncertainty range (ratio of themobile to immobile)

350 150 440 4 380

3.9.2 Mass of Radionuclide in the Repository

For the minimum waste loss case, the total mass of uranium existing in the repository is reduced toless than 100 moles, or 24 kg. For the realistic waste loss case, the total mass of uranium is at most 40,000moles, or 10 MT.

The repository with LWR spent fuel contains about 1000 MT of thermally fissile materials. Thiscan be reduced to 10 MT by ATW with the realistic waste loss, and to 10 kg with the minimum waste loss.With 10 kg, the possibility of criticality accident can be completely eliminated.

By dilution with the recovered uranium and by the transmutation of precursors to thermally fissileactinides, the total mass of thermally fissile actinides in the repository is reduced significantly by the ATWsystem, which implies that the possibility of criticality accident in and around the repository becomes neg-ligibly small.

From the viewpoint of the attractiveness to the proliferator, the ATW waste seems to have a clearadvantage over the LWR SF because of the reduced mass of actinides. But, for more discussions, wastecharacterization should be done in a more detailed fashion, including waste forms, concentrations of radi-onuclides in waste forms, volume of wastes, the engineered barriers around the waste packages, and so on.

3.10 SummaryImpacts of the Accelerator-driven Transmutation of Waste (ATW) system on performance of the

proposed Yucca Mountain Repository (YMR) have been evaluated by quantifying the radiological hazardat the 5-km location from the repository and the mass of fissile materials existing in the repository, basedon the mass of each radionuclide that comes out of the ATW system as waste. The waste is assumed to beplaced in YMR.

The radiological hazard from ATW waste has been found to always be smaller than that from LWRspent fuel, while the time length for the repository safety assessment is not shortened by introduction of theATW. From the viewpoint of the attractiveness to the proliferator and criticality safety, the ATW wasteseems to have an advantage over the LWR spent fuel in case the waste generation from the system is mini-mized. The judgement for the realistic waste loss case is more complex, and requires more detailed analy-ses. Detailed discussion of implications can be found in Chapter 4.

For more detailed and realistic analyses, the destination of recovered uranium and characteristicsof the ATW wastes should be identified. Based on detailed waste categorization, proper engineered barriersystem for each waste category should be determined. The ATW design should be optimized, based on therepository performance as well as on such factors as energy utilization and cost. Although the presentanalysis has shown results of two bounding cases regarding the geochemical parameters, statistical analysesshould be performed to clarify whether or not the differences in performance of the repository found in thisanalysis are statistically meaningful.

29

4 MEASURES OF REPOSITORY IMPACTSWe focus in this section on the measures to be used to evaluate the impacts an ATW system could

have on the performance of a high-level waste repository. Our examination is not complete, so while wepresent results for some measures in this document, for others we describe the measures without applyingthem.

Measures for evaluating ATW impacts reflect two independent but related features of the ATWsystem with respect to the repository system: licensibility and desirability. Licensibility bears on the rela-tive ease or difficulty of persuading a regulator that the repository meets societyÕs standards for protectionagainst hazards. Desirability here is a broader category of features, aside from legal and regulatory con-straints, that factor into oneÕs judgment whether society is better off pursuing or not pursuing the ATW.An alternative fuel cycle could make the repository more licensible and still be undesirable; and it could bedesirable, but make a repository no more licensible. Overall desirability depends on features of the entireATW scenario including the lifecycles of the ATW reactors, the storage and transportation involved, andthe repository. We confine ourselves for this discussion to considering desirability with respect to the re-pository.

Current regulations pertain primarily to radiological hazards, with less thorough coverage of criti-cality and proliferation hazards, whereas at least five prominent areas of concern pertain to relative desir-ability: radiological hazard, proliferation hazard, criticality hazard, disposal capacity, and economics.

In our evaluation of the impacts that alternative fuel cycles may have on the performance of aHLW repository, we examine the relative radiological hazards, the relative proliferation hazards, and therelative criticality hazards. Our methods draw on previous studies ([2],[3],[18]) and measure performancein terms of reductions both in the magnitude and the duration of the most prominent hazards. Currentlimitations of the project have not permitted evaluation with respect to other measures, including economicfactors.

4.1 LicensibilityLicensibility is measured in this study in two ways: by comparing the waste streams to several sets

of existing criteria and by a more qualitative evaluation of the clarity or transparency of the license appli-cation claim.3

4.1.1 Existing Criteria

Waste streams can be compared to regulations governing disposal of radioactive waste in generaland to waste-acceptance criteria (WAC) for particular disposal sites. General regulations governing radio-active waste management include 10 CFR 60, 10 CFR 61, DOE Order 5820.2A and generic standards re-garding environmental protection, such as those supporting clean air and clean water.

The concentration limits for disposal of low-level waste in near-surface facilities are based on cal-culations of radiological hazard, reflecting an intruder dose. The ratio by which the waste stream exceedsthe limits is then an indicator of a hazard associated with the waste [19], although this is of limited utilitydue to the difference between near-surface and deep-geologic disposal. Each DOE disposal site has a set ofWAC that includes the concentration limits. Anticipated fuel-related waste streams from the ATW far ex-ceed concentration limits for all of these sites. Figure 4.1 displays the sum of the fractions4 for the actinidewaste stream at three LLW facilities [20][21][22]. We emphasize that this is not, by itself, a measure of theacceptability of the waste for near-surface burial; while the sum of the fractions for the clean projection foractinides at Hanford is below 1, the actual waste would not qualify for near-surface burial due to the ac-companying fission products. The sum-of-the-fractions for only three fission productsÑ90Sr, 99Tc, and137CsÑexceeds the limit at Hanford by nearly a factor of 30. Thus, we would not expect disposal as LLWto be a possibility. But the scenario with low process losses reduces the actinide hazard by as much as afactor of nearly 9000 (at the Savannah River E-Area Vaults).

3 If one can base a license application on simple reasoning and strong conservatism, without resorting to arcane andcomplex models, then the system can be said to be more licensible.4 The sum of the fractions is the sum of the ratios of the concentration of a radionuclide in the waste to the respectiveconcentration limit. A value greater than 1 indicates that the waste is greater-than-class C (exceeds the highest limitsfor near-surface burial) and is not generally suitable for disposal near the surface.

30

0.1

1

10

1 00

100 0

1000 0

1 00000

Hanford Savannah River NTS

Act

inid

es-o

nly

Sum

of

the

Fra

ctio

ns

LWR spent fuel

ATW (curren t average)

ATW (clean projec tion)

Figure 4.1 The sum of the fractions of the actinide waste stream from the ATW (concentrationrelative to the concentration limits at three DOE LLW disposal facilities). The threewaste streams correspond to low process losses (clean), higher losses (currentaverage), and spent fuel.

Another waste stream, waste from the spallation source and coolant, is more complicated thanother waste streams from a licensing standpoint. Based on parallels to calculations done for the acceleratorproduction of tritium (APT) project, we do not expect the lead waste stream to qualify for near-surface dis-posalÑit fits the definition of low-level waste (LLW) but is likely to be greater-than-class-C LLW (GTCC)due to contamination with activation and spallation products. Calculations are needed to substantiate con-clusions on this point (see Section 5.2, Future work). GTCC waste is currently stored in anticipation ofdisposal with HLW in a deep-geologic repository. Because lead displays the characteristic of toxicity, it isconsidered a hazardous waste regulated by the U.S. EPA under the Resource Conservation and RecoveryAct (RCRA). ÒActivated leadÓ is then both a listed waste regulated by the EPA under RCRA, and a radio-active waste regulated by the DOE or NRC under the Atomic Energy Act (AEA). Due to the mixed juris-dictions and the mixed hazardous and radioactive features, this category of waste is termed Òmixed low-level wasteÓ (MLLW). The dual jurisdiction substantially complicates management of the waste becausethe competing regulators have some conflicting requirements for management and because a disposal facil-ity for MLLW must satisfy both regulations. RCRA typically requires that hazardous waste be treated be-fore it is disposed of even in a facility licensed for disposal of hazardous waste. It is not clear what theappropriate treatment for the hazardous component of this waste would be. If the treatment results in dilu-tion, then near-surface burial may be possible.

The United States does not now have a repository for disposal of HLW and spent nuclear fuel.The DOE is characterizing a site at Yucca Mountain in Nevada to support a proposal and a license applica-tion to construct a repository there. As of the date of publication for this report, the regulations governing arepository at Yucca Mountain, and therefore the criteria against which a license application would bejudged, had not been set. U.S. EPAÕs generic regulation containing standards for HLW-disposal sites (40CFR 191) has been rejected by Congress in favor of a site-specific standard that has not yet been promul-gated.5 We nonetheless compare results from our model to the limits in the regulation, still noting thatwhile this regulation is unlikely to fit the form of future standards for Yucca Mountain, but it does reflect aregulatorÕs approach to setting standards for a HLW repository.

We have evaluated the radiological performance of the repository containing ATW fuel-cycle

5 Congress asked the National Academy of Sciences (NAS) to recommend a site-specific, dose-based standard forYucca Mountain to the standard-setting body, the Environmental Protection Agency (EPA). In 1995, the NationalResearch Council (the research arm of the NAS) published a report recommending a standard based on the risks posedto critical groups [23]. We have not yet performed the more complex modeling required to compare the results fromATW waste streams to these suggested criteria.

31

wastes under the anticipated extremes of the range of conditions at Yucca Mountain using a simplifiedmodel of waste dissolution and transport. The results of this evaluation and a description of the model usedto carry it out can be found in Section 3 of this report. We have used the same model to compare the re-pository performance to the generic standards in 40 CFR 191. Both the assumptions of the model and thefeatures of the regulation are described below.

For this discussion of the ATWÕs impact on the radiological performance of a deep-geologic re-pository, we use the expected stock of spent fuel from existing U.S. LWRs as the reference source of haz-ard, and the conditions and physical configuration expected at Yucca Mountain as the reference repository.Because there is uncertainty regarding the conditions at the repository, we consider two sets of hydrologicaland geochemical conditions: a mobile-conditions case with low Kds (solid phase-aqueous phase distributioncoefficients) and high solubilities, and an immobile-conditions case with high Kds and low solubilities.These should bound the expected performance of the repository. We also consider two operating scenariosfor the ATW facility: a scenario with low process losses (the clean projection, with values provided byLANL), and a scenario with higher process losses that are consistent with current experience in commercialfacilities (current average).

Note that the model does not account for colloidal transport, which is anticipated to be the domi-nant transport mode for plutonium at Yucca Mountain (making 242Pu second only to 237Np in the ViabilityAssessment dose calculations [24]).

The regulation limits the integrated mass flux of particular radionuclides into the accessible envi-ronment during the first ten thousand years. The regulation has two tiers of limits that correspond to differ-ent confidence levels. The long transport times anticipated from the repository to the accessible environ-ment at Yucca Mountain make it unlikely that this measure will be meaningful if the accessible environ-ment is defined to be 5 km from the repository: The plume arrival time for the most troublesome radionu-clides is projected to be after the ten thousand year cutoff. Our intent here is to compare waste scenariosthat include an ATW system to those that do not. Thus we first note that there is little difference betweenthe two scenarios for a 10,000 yr standard at 5 km. If we now apply the same standard at the plumeÕs en-trance to the water table (~250 m from the waste packages), we can see differences between the two sce-narios. Figure 4.2 illustrates the results from 10,000 yr to 1 million yr for the ATW and the unprocessedLWR spent fuel, each under mobile and immobile conditions. The value Ò1Ó on the vertical axis of Figure4.2 corresponds to the 90% confidence limit in 40 CFR 191. While these results could neither support norundermine a license application, we observe that in contrast to the findings from previous figures, it is un-der the mobile conditions that differences between the two waste streams are insignificant. Under immobileconditions, the difference is a factor of ~60. This is due to the difference in weighting of the individualradionuclides in the two metrics: the regulation weights uranium more heavily than it weights neptunium.

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+ 04 1.E+05 1.E+ 06Time (years)

Reg

ulat

ory

Inde

x (E

PA

crit

eria

) Mobile LWR

Mobile ATW (best ) + U

Immobile LWR

Immobile ATW (best) + U

Figure 4.2 40 CFR 191 Criteria, summary of repository results. ATW cases comprise the ATWwaste from utilizing the entire stockpile of LWR spent fuel.

4.1.2 Qualitative Evaluation of Licensibility

We postulate that if one can base a license application on simple reasoning and strong conserva-tism, without resorting to arcane and complex models, then the system can be said to be more licensible.

32

Initial investigations indicate that the ATW system does not eliminate the need for complex calculationsand therefore still relies on expertise and complex representations to make a case for improved perform-ance. The destruction efficiency for relevant radionuclides is not high enough in the current design toeliminate concerns. Transparency arguments are particularly difficult to apply to disposal facilities due tothe complexity of the system and to the long timeframes involved. The ATW may contribute more toregulatory transparency if it realizes significant increases in destruction efficiency or if waste forms areoptimized for ATW wastes. A strength of the ATW with respect to transparency may emerge more clearlyin the context of reactor safety.

4.1.3 Proliferation

Long-term safeguards may be required for spent fuel in a repository in a non-weapons states, ac-cording to the IAEA [25]. Current plans for the U.S. repository call for designs that will allow for volun-tary IAEA inspection. IAEA does not currently regard reprocessing wastes from an actinide-burning fuelcycle as proliferation hazards.

4.1.4 Criticality Safety

As described in Section 2.2.3 of this report, limits are on the allowable keff in the repository.

4.2 DesirabilityIn the context of the ATW systems impacts on a repository, we understand desirability in terms of

reductions both in the magnitude and the duration of hazards and of costs. The hazards include radiologicalhazards measured in migration; proliferation hazards measured according to several key parameters (con-centration, total inventory, chemical form, isotopic composition, radiation barrier, and precision, accuracyand verifiability of accounting); criticality hazards measured in concentration and isotopic composition offissile materials. The internal costs are measure in a relative sense in terms of the disposal capacity of aHLW repository. These are discussed in more detail below.

4.2.1 Radiological Hazards

The radiological migration hazard, as described in Section 3, is assessed using a model of radionu-clide transport from the waste form to the accessible environment, 5km away. These hazards are exten-sively discussed in other sections of this report.

4.2.2 Proliferation Hazards

Proliferation resistance is measured in both crude and subtle terms. Broadly, three or four featuresrepresent proliferation resistance for a system that includes fissile material (see, e.g., Ref. [22]): a high de-gree of difficulty of theft or diversion, sometimes called accessibility; a high probability of timely detec-tion, sometimes called observability; a low rate of possible diversion; and high degree of difficulty of utili-zation, sometimes called utility. Difficulty can be economic, technical, or institutional and is generallyexpressed in terms of costs and time delays. Notice that few of these can be measured simply by featuresof the material itself; the context and institutional factors are generally considered to be crucial elements.Table 4.1 summarizes common thinking regarding safeguards of fissile materials with respect to differentthreats.

The proliferation concerns for a repository in a weapons state, such as the United States, aresomewhat different from those for a non-weapons state if only because the weapons state presumably hasthe primary and ancillary facilities for weapons production already in place. A subset of non-weaponsstates is commonly distinguished as virtual proliferants: states such as Japan and Germany, that have virtu-ally all of the technical expertise and infrastructure that would be required to become a weapons state, butthat have decided for political reasons not to develop a nuclear arsenal. The remainder of the non-weaponsstates do not, at least publicly, have the expertise and infrastructure required for production of nuclearweapons. These designations are only meaningful, however, in the short term. The expertise and infra-structure are bound to change in the 300-600 years it would take for the fissile material in spent fuel to be-come appreciably attractive. Access to material then becomes the crucial point. But the question remainswhether a state (weapons or non) would mine a repository or choose alternative methods of production.We have not yet addressed this question and focus for now on theft by a subnational group.

The emphasis in preventing diversion or theft of fissile materials proceeds in order from (using theterminology from [26]) institutional modes to intrinsic features: prevention of accessibility, enhancement ofobservability, then reduction of utility. If we assume that long timeframes undermine our confidence in

33

institutional barriers, we are left to focus on the intrinsic physical barriers. Most of these barriers to theft bya subnational group can be measured quantitatively, e.g., a radiation barrier, or an environmental barrier(the depth of the repository).

Table 4.1 Summary of Thinking Regarding Safeguards of Fissile Materials DuringTransportation and Geologic Disposal with Respect to Different Threats. MC&ARefers to Material Control and Accounting. (Reference [26] refers to intrinsicbarriers rather than physical barriers.)

Proliferant è Host Nation SubnationalWeapons State Nonweapons State Group

Objective of prolifera-tion

Increased arsenal forstrategic purposes

Capability as a nu-clear power

Threat of nuclearterrorism

Mode of proliferation Diversion(overt & covert)

Diversion(overt & covert)

Theft

Physical Barriers: time, cost time, cost haz, diffclty, cost radiological minor barrier minor barrier major barrier chemical mod barrier mod barrier mod barrier isotopic U-maj; Pu-min U-maj; Pu-min U-maj; Pu-mod physical form minor barrier minor barrier mod barrier environment minor barrier minor barrier mod-maj barrierInstitutionalmodes of prevention

International pressure MC&A, Surveilance,IntÕl pressure

Physical security & Detection

Looking first at the environmental barrier, Peterson [27] has examined the relative ease or diffi-culty of covertly mining a repository for weapons material and uses concentration of fissile material as anindex of the attractiveness of the repository as a source. We have calculated the overall reduction of fissilematerial achieved by the ATW system assuming different processing efficiencies (a factor related to theTRU ratio and depletion ratio devised by Pigford and Choi [18][3]). Peterson utilizes a quantity he termsequivalent kg of Pu, or Ekg, based roughly on the relative bare-sphere critical mass of each fissionableisotope: 237Np has a bare-sphere critical mass 5.6 times that of 239Pu, 5.6 kg of 237Np = 1 Ekg Pu. We cannow compare the rate of recovery of fissile material from a repository for different waste streams. Assum-ing 1.9 MT of waste is extracted each day, intruders could recover three thousand Ekg per year from a re-pository containing LWR spent fuel. Compare this to just over 400 per year for the ATW under currentaverage reprocessing, and 7 per year for ATW waste with the cleanest projected reprocessing. Elevatedlevels of 242Pu in ATW waste would make the material less attractive due to a high rate of spontaneous fis-sions. In Table 4.2 and Table 4.3 we have adjusted the fissile material content to reflect quantity of Puequivalent.

Future work will characterize other barriers and features of the waste stream. Barriers will becharacterized by the gamma-dose rate per gram of material (from the waste form and also from the acti-nides alone). Chemical barriers depend strongly on the waste form, which has yet to be selected. The at-tractiveness of the fissile material will be characterized by the bare-sphere critical mass of the actinides inthe waste stream, the heat-generation rate, and the spontaneous-neutron emission rate.

4.2.3 Underground Criticality

It is difficult to measure the underground-criticality hazards associated with waste containing fis-sile material without rather sophisticated calculations [16][28][29]. But we can note that ATW waste isreprocessing waste, that fissile material from several canisters would be required for one critical mass, andthat Pu (85% of the fissile material) is expected to be immobile in the Yucca Mountain environment (ex-cept with respect to colloidal transport). In this context, the recovered uranium from the LWR spent fuel(~1.1% enrichment) is of no consequence except that it can isotopically dilute the 235U produced by decayof 239Pu, if these wastes are disposed together. The mass of fissile materials existing in the repository, ob-tained by the inventory model, is used as a measure for the criticality safety and the proliferation resistance.

4.2.4 Summary of Hazard Reduction Factors

In Table 4.2 and Table 4.3, we present a summary of the relative migration hazards, proliferationhazards, and criticality hazards from ATW fuel-cycle waste in a HLW repository with respect to the haz-

34

ards from the LWR spent fuel inventory in the same repository.For the mobile base case for LWR spent fuel, actinides (primarily 237Np and isotopes of U) domi-

nate the hazard downstream from the repository: Of the fission products, 129I generates the greatest hazard.But the peak hazard is due to 237Np, which contributes more by a factor of ~300.

4.3 Other MeasuresWe have identified a set of other measures of the impact of ATW systems on performance of a re-

pository. Developing the tools to evaluate the ATW with respect to these measures lay beyond the scope ofthe work in the present study. They are described below as possible guidance for future work.• Evaluation with respect to the recommended NAS Yucca Mtn standards (critical group with scenario)

[23].• Evaluation with respect to protection of ground water• Cost analyses. One approach is to carry out an ALARA evaluation following the US NRC decontami-

nation and decommissioning cleanup methodology for $/person-rem averted.• Evaluate the differential risk burden through time. This follows an idea by Holdren [30] regarding the

risk we are willing to incur now to avert future risks, i.e., the ratio of increased present burden to de-creased future burden. (HoldrenÕs opinion suggests a ratio 10:1.) The present burden would includeroutine emissions and accident risks while the future burden includes waste hazards.

• A comparison of HLW hazards and LLW hazards from ATW waste streams.• Refinement of the proliferation risk reduction considering radiation barrier, chemical form, isotopic

composition (bare-sphere critical mass), and accounting (precision, accuracy, and verifiability). Allevaluated relative to once-through fuel cycle.

Table 4.2 Summary of the Relative Migration Hazards, Proliferation Hazards, and CriticalityHazards from ATW Fuel-Cycle Waste in a HLW Repository (No recoveredUranium)

Radiological Health 1 Proliferation 2

Management Option Mobile Reposi-tory

Immobile Re-pository

&Criticality

ATW process waste(clean projection)

1/1300 1/156E-5

7E-5 - 2E-4ATW process waste(current average)

1/35 1/150.030.09

1. Measured in dimensionless units of peak hazard (mass flux ratio to Annual Limit on Intake at 5km) relative to hazardfrom US stock of LWR Spent Fuel.2. Measured in total Pu equivalent relative to that in US stock of LWR spent fuel, and Pu equivalent kg per canisterrelative to LWR spent fuel.

Table 4.3 Summary of the Relative Migration Hazards, Proliferation Hazards, and CriticalityHazards from ATW Fuel-Cycle Waste in a HLW Repository (With recovereduranium)

Radiological Health 1 Proliferation 2

Management Option Mobile Immobile & Criticality

ATW process waste + U(clean projection)

1/14 1/15U 1.1%

6E-57E-5 - 2E-4

ATW process waste + U(current average)

1/12 1/15U 1.1%

0.030.09

1. Measured in dimensionless units of peak hazard (mass flux ratio to Annual Limit on Intake at 5km) relative to hazardfrom US stock of LWR Spent Fuel.2. Measured in U enrichment (for the 67,000MT of U), total Pu equivalent relative to that in US stock of LWR spentfuel, and Pu equivalent kg per canister relative to LWR spent fuel.

35

5 CONCLUSIONS AND DIRECTIONS

5.1 Factors Increasing or Limiting ATW DesirabilityWithout judging whether the values in Table 4.2 represent sufficient reductions in hazard to make

an ATW system favored, we can observe the factors that tend to increase desirability and those that tend tolimit the desirability of the system.

First, we note a factor that does not appear in the tables: the ATW reactors would generate an ad-ditional ~500GWe-years and the waste could be disposed of in ~45% of the repository space (depending onthe waste form). No risk or cost is worth enduring without benefits, but the ATW system performing asdesigners hope would provide benefits.

Simple inspection of the tables reveals that decreases in process losses in reprocessing and fuelfabrication (represented here by transitions from the current average cases to the clean projection cases) donot reduce the radiological hazards under immobile conditions in the repository. Under mobile conditions,looking first at the case that ignores uranium, the reduction is more than a factor of 35 and would be evengreater if the losses were lower. Thus, under mobile repository conditions, if we ignore the impacts of theseparated uranium, the hazards from ATW waste are strongly linked to the process losses. This materialprocessing might then be an area to focus efforts to improve the ATW with respect to waste; but only if themobile conditions are seen as more likely than the immobile conditions, or as a kind of insurance policyagainst the worst case.

One notes, however, that the radiological picture is quite different if we consider the impacts ofthe separated uranium. According to legal definitions for waste classes in the United States, the separateduranium stream must be managed as HLW or spent fuel if it is considered to be waste. And this is notmerely a legal distinction: The oxide form of the uranium waste stream exceeds concentration limits fordisposal at LLW facilities by factors ranging from ~10 at Hanford [18] to over 5000 at Savannah RiverÕs E-Area Vaults [19]. Thus we assume that the separated uranium is co-disposed with the ATW fuel-cyclewaste. Under mobile conditions in the repository, ATW HLW with current-average waste fractions exhibita peak radiological hazard that is lower than that from LWR spent fuel by a factor of 12. Lower processlosses only reduce the radiological hazard by another 15%. Further, this is the limit to improvement be-cause the uranium hazard dominates at that point. These statements represent the expected upper and lowerlimits of the mobility. Results would change if neptunium and uranium were not found to behave similarlyin the repository. Because it is not legitimate to simply ignore the separated uranium stream, the resultsassuming co-disposal are our reference scenario whereas the others are not. If the ATW is to reduce haz-ards beyond the 1/14 ratio to LWR spent fuel hazards, then alternative strategies for dealing with the ura-nium must be explored.

Reprocessing and burning substantially reduces the attractiveness of a repository for recovery ofweapons-usable material over spent fuel. A repository loaded with ATW HLW yields a factor of 7-400lower rate of recovery of weapons-usable material 1000 years after disposal, for a clandestine mining ef-fort. The concentration of actinides is sufficiently low and the concentration of fission products sufficientlyhigh to make this a relatively expensive and dangerous source of special nuclear material for hundreds ofyears. While the proliferation resistance of the repository is improved by the ATW, the safeguards burdenis shifted to the ATW facilities, themselves.

The waste from the ATW should resemble reprocessing HLW with respect to criticality concerns,as well. The potential for underground criticality in vitrified HLW should be dramatically lower than fordirectly-disposed LWR spent fuel by virtue of the low concentration of fissile material in the waste. Meas-ured simply by the fissile material content of the repository, the ATW improves the situation by a factor of10 to 15,000, depending on the process losses.

We must also examine implicit assumptions regarding the scale and thoroughness of the spent-fuel-burning campaign. In all of our calculations, it is assumed that all of the US LWR spent fuel is proc-essed through the ATW. It is possible that only part of the campaign would be completed leaving bothATW waste and LWR spent fuel for disposal in a repository. Further investigation is needed to determinethe benefits as functions of the fraction of the spent-fuel stockpile that has been processed.

Finally, if we are concerned with the benefits that the ATW can provide for a US HLW repository,we must consider all of the waste to be disposed in that repository. The overall radiological hazard of therepository can be reduced by reducing the contribution to that hazard by the commercial LWR spent fuel.But the reduction is limited by the impact of collaterally disposed waste such as defense high-level waste

36

(DHLW) and naval reactor spent fuel. It might be possible to lower the contribution from commercialHLW to a level such that the collateral waste dominates the hazard, but further reductions reap little re-ward. Further reductions require transmutation of naval spent fuel and DHLW.

5.2 Future ResearchFuture research will focus on improving the performance of the ATW system capitalizing on the

systems capabilities in tailoring the wasteÕs form, concentration, heat load, and other features, and will alsofocus on refining some of the analyses done to date. More specifically, we describe below future directionsexamining the effects of waste form, repository design, and type of repository on ATW impacts, designflexibilities in the ATW system.A. Radiological Impacts of ATW system:

• Lead activation. We will calculate the radioactive inventory in the PbBi coolant-spallationsource. We need to acquire a tool and library for calculating the spallation product inventory andcouple this to tools for neutron activation calculations that are already available.

• Routine emissions. We will asssess the routine emissions from the ATW system. This requiresknowledge of the inventories of the materials streams (PbBi and fuel-cycle materials) and sepa-ration efficiencies of the processing equipment for each stream.

• Accident hazards. We will assess the accident hazards based on radioactive inventories. Similarto the item above, this requires knowledge of the time-dependent inventories in different compo-nents of the ATW system. In addition, this will include criticality safety (margins under crediblereconfigurations) in the reactor and in reprocessing.

B. Radiological impacts of the waste:• Maximizing ATW transmutation efficiency. We will continue to search for ATW blanket designs

and modes of operation that will maximize the fractional destruction of the more problematic iso-topes.

• Alternate waste forms. Current calculations assume leach times of 100 ky. (Leach time is the timeto complete dissolution of the waste form at a constant rate.) As with any reprocessing system, arange of waste forms are available and the performance of each waste form is uncertain. Lesssoluble waste forms might result in better performance (reduced radiological hazards) and moresoluble waste forms might result in worse performance. An effect will be apparent only if the ra-dionuclides are not solubility limited. The actinides could be immobilized in a stable ceramic ormineral waste form with superior performance characteristics (i.e., better than spent fuel or bor-osilicate glass). Criticality and proliferation considerations might make the separate waste streamsless attractive. We plan to explore the relationship between waste form and radiological impacts.

• Alternative repository loading. The flexibility in waste form and loading provides flexibility tooin the loading of the repository and could afford increased storage capacity and different configu-rations that could improve repository performance. Larger packages and closer spacing for recov-ered uranium are simple examples of such possible changes.

• Alternative repository loading and disposal environments. The flexibility in waste form andloading provides flexibility too in the loading of the repository and could afford increased storagecapacity and different configurations that could improve repository performance. Larger packagesand closer spacing for recovered uranium are simple examples of such possible changes. Differentrepository environments might also exhibit different behavior with respect to ATW fuel cycles.For example, one could send actinide wastes (uranium and any other separated actinide wastes)along with 99Tc to a repository in a reducing environment and send the remainder to a YuccaMountain-like setting. One might argue that comparable benefits could be achieved by separationwithout the burning; but these issues bear further examination.

• Probabilistic calculations. The current calculations of groundwater-migration hazards for the re-pository are deterministic, represent what seem to be upper and lower bounds. But the parametersrelevant to repository performance might be better represented by statistical distributions. MonteCarlo treatment of the parameters in groundwater-transport calculations will reveal the conserva-tism of the calculations and verify whether the calculations are actually bounding calculations.

• Dose and Risk-based measures of performance.• PbBi waste-stream impacts. .

C. Proliferation Impacts of ATW system:• Inherent barriers to proliferation. We will calculate the bare-sphere critical masses and radiation

37

barriers of the waste streams as functions of time.• Institutional barriers to proliferation. We will assess the materials accounting capabilities and

demands of ATW systems, and the safeguards demands based on inherent barriers of the wastestreams.

D. Evaluation of the system with respect to the measures in Section 4.3

E. Scenarios for managing uranium. (disposal or recycling)

38

REFERENCES1. Los Alamos National Laboratory, Presentations Prepared for the MIT Technical Review, available at

http://www-adtt.lanl.gov:80/ATW_papers.html2. National Research Council. Nuclear Wastes: Technologies for Separations and Transmutation. Com-

mittee on Separations Technology and Transmutation Systems, Board on Radioactive Waste Manage-ment, Commission on Geosciences, Environment, and Resources. National Academy Press, Washing-ton, D.C., 1996.

3. Ramspott, L.D., J.S. Choi, W. Halsey, A. Pasternak, T. Cotton, J. Burns, A. McCabe, W. Colglazier,and W.L. Lee. Impacts of New Developments in Partitioning and Transmutation on the Disposal ofHigh-Level Nuclear Waste in a Mined Geologic Repository. Lawrence Livermore National Labora-tory, UCRL ID-109203, 1992.

4 . Ahn, J. Integrated Radionuclide Transport Model for a High-Level Waste Repository in Water-Saturated Geologic Formations. Nuclear Technology, 121, 24-39, 1998.

5. Office of Civilian Radioactive Waste Management, Department of Energy, Viability Assessment of aRepository at Yucca Mountain, DOE/RW-0508/V3. December 1998.

6. 10CFR20 Standards for Protection against Radiation, Appendix B - Annual Limits on Intake (ALIs),September 29, 1995.

7. 40CFR191 Environmental Radiation Protection Standards for Management and Disposal of Spent Nu-clear Fuel, High-Level and Transuranic radioactive Waste, U.S. Environmental Protection Agency

8. Hatanaka, K., and J. Ahn, Repository-wide Performance Assessment Model by Object-Oriented Ap-proach, Proc., 8th International Conf., High-Level Radioactive Waste Management, Las Vegas, Ne-vada, May 11-14, 1998, American Nuclear Society, 1998.

9 . SAND93-2675, Total System Performance Assessment for Yucca Mountain (TSPA-1993), Sandianational Laboratory, April 1994.

10. Benedict, M., T. H. Pigford, H. Levi, Nuclear Chemical Engineering, 2nd ed., McGraw-Hill,11. Ahn, J., Transport of Weapons-Grade Plutonium and Boron Through Fractured Geologic Media, Nu-

clear Technology, 117(3), 316-328, 1997.12. US DOE, Draft Mission Plan Amendment, DOE/RW-0187, Office of Civilian Radioactive Waste

Management, Washington, DC, 1988.13. US DOE, Characteristics of Potential repository Wastes, DOE/RW-0184-R1, Office of Civilian Ra-

dioactive Waste Management, Washington, DC, 1992.14. Argonne National laboratory, EBR-II Spent Fuel Treatment Program Monthly Report, ANL-NT-58,

October 1997.15. Sandia National Laboratory, Performance Assessment of the Direct Disposal in Unsaturated Tuff of

Spent Nuclear Fuel and High-Level Waste, SAND94-2563, 1995.16. Kastenberg, K. E., P. F. Peterson, J. Ahn, J. Burch, G. Casher, P. L. Chambr�, E. Greenspan, D. R.

Olander, J. L. Vujic, B. Bessinger, N. G. W. Cook, F. M. Doyle, and L. Brun Hilbert, Considerationsof Autocatalytic Criticality of Fissile Materials in Geologic Repositories, Nuclear Technology, 115(3),298Ð310, 1996.

17. Ahn, J., Mass Transfer and Transport of Radionuclides in Fractured Porous Rock, PhD Dissertation,University of California, Berkeley, 1988.

18. Pigford, T. H., and J.S. Choi, ÒInventory Reduction Factors for Actinide-Burning Liquid Metal Reactors,ÓTransactions of the American Nuclear Society, 64, 1991.

19. Holdren, J.P., D.H. Berwald, R.J. Budnitz, J.G. Crocker, J.G. Delene, R.D. Endicott, M.S. Kazimi,R.A. Krakowski, B.G. Logan, and K.R. Schultz, ÒReport of the Senior Committee on Environmental,Safety, and Economic Aspects of Magnetic Fusion EnergyÓ Lawrence Livermore National Laboratory,UCRL-53766, 1989.

20. Westinghouse Hanford Company, Low-Level Waste Acceptance Criteria, Section 3, WHC-EP-0063-4.21. Westinghouse Savannah River Company, WSRC 1S SRS Waste Acceptance Criteria: Procedure 3.10 E-Area

Vaults Low-Level Radioactive Solid Waste Acceptance Criteria, Rev. 2, July 31, 1996.22. Nevada Test Site Waste Acceptance Criteria, NTSWAC (Rev. 0), September 1996.23. National Research Council, Technical Bases for Yucca Mountain Standards, National Academy Press,

Washington, D.C., 1995.

39

24. Office of Civilian Radioactive Waste Management, Department of Energy, Viability Assessment of aRepository at Yucca Mountain, Volume 3: Total System Performance Assessment. DOE/RW-0508/V3.December 1998.

25. IAEA Department of Safeguards, Report of the consultants' group meeting on safeguards for the directfinal disposal of spent fuel in geological repositories, STR-305 VIC 27 Nov. to 1 Dec. 1995.

26. Hinton, J. P., R.W. Barnard, D. E. Bennett, R. W. Crocker, M. J. Davis, G. A. Harms, L. W. Kruse, J.A. Milloy, W. A. Swansiger, K. J. Ystesund, H. J. Groh, E. A. Hakkila, W. L. Hawkins, E. E. Hill.Proliferation Vulnerability Red Team Report. SAND97-8203, UC-700, October 1996.

27. Peterson, P. F. Relative Attractiveness of Reprocessing Waste and Spent Fuel: Implications for Long-Term Safeguards Technical Requirements. Energy Future in the Asia/Pacific Region: Research &Education for Nuclear Energy. Proc. of the Internation Symposium Sponsored by Tokai UniversityEducational System, Japan, and the Department of Nuclear Engineering, University of California,Berkeley, March 27-28, 1998, Honolulu, Hawaii, 1998.

28. Greenspan, E., J. Vujic, and J. Burch, Neutronic analysis of critical configurations in geologic reposi-tories. I. Weapons-grade plutonium, Nucl. Sci. and Engin., 127, 262, 1997.

29. Vujic, J., and E. Greenspan, Neutronic analysis of critical configurations in geologic repositories. II. Highlyenriched uranium, Nucl. Sci. and Engin., 129, 1, 1998.

30. Holdren, J. P., Radioactive-Waste Management In The United-States - Evolving Policy Prospects AndDilemmas. Annual Review of Energy and the Environment, 17, 235-259, 1992.