UC Berkeley1

Fission Reactors –Options and Challenges

Ehud GreenspanDepartment of Nuclear Engineering

University of California, Berkeleygehud@nuc.berkeley.edu

GCEP-CNESFission Energy Workshop

MIT, November 29-30, 2007

UC Berkeley2

Introduction

Describe 4 options that are either not being explored bycurrent programs or explored at a low level of effort(GCEP statement of interest)

Consider “reactor systems”; including the fuel cycle

Not focus on the session title: “Closing the Fuel Cycle”

Describe reactor concepts very briefly

Express my personal opinion

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Presentation outline

Light-water cooled breeding reactors

Liquid-salt cooled high temperature thermalreactors

Nuclear battery type reactors

Deployment of fast reactors without separatingTRU from LWR spent fuel

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1. Light water breeding reactorsResource Renewable Boiling Water Reactor

RBWR

Design concept developed by Hitachi

Three design variants: Thermal spectrum (high power density alternative to ABWR; Uenriched )

Fast spectrum for sustainable energy (TRU; CR = 1): RBWR-AC

Fast spectrum for minimizing inventory of TRU (Burner): RBWR-TB

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General layout of fast spectrumRBWR core

Ref: "Status of advanced light water reactor designs." IAEA TECDOC-1391, 2004

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Selected operating characteristics ofRBWR cores

Ref: "Status of advanced light water reactor designs." IAEA TECDOC-1391, 2004

ABWR392613568723710Uniform5214.5380.171573.6 235U

4013.51.3-7x10-4

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Pros and cons of RBWR

Pros LWR technology can be used for

Increasing U resource utilization ( X 102)

Putting a cap on the total inventory of Pu and MA accumulated per GWe-yr

Transmuting (fissioning) most of TRU at the close of the fission energy era

Proven technology (almost)

Capital cost likely comparable with LWR

Requires limited resources and time for commercialization

Industry is familiar with basic technology and has the required infrastructure

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Pros and cons of RBWR (Cont.)

Cons Limited thermodynamic efficiency

Limited temperature of heat supply for thermo-chemical hydrogengeneration

Shorter cycle length (RBWR-TB)

Higher fuel-cycle cost (relative to once-through LWR)

Limited breeding gain

Using internal blankets of depleted uranium (but no blanket only fuel rods)

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Challenges for RBWRs

Feasibility of safe operation with a tight lattice and high voidfraction

TRU oxide fuel performance (fabrication) validation

Economic viability

Hitachi:

“The RBWR has been designed based on proven BWR technology, but itsBR of 1.0 in high BU fuel, negative void coefficient, (stability) andmechanical integrity of (fuel and) in-core structure need be verified atoperating conditions. Hitachi has a design for a 180 MWth demonstrationRBWR (75 hexagonal fuel bundles)”

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2. Advanced High Temperature Reactors*(AHTR)

Liquid-salt (Flibe) cooled, graphite moderated,TRISO fueled

Two design approaches:

Graphite blocks (ORNL-ANL; AREVA, FRAMATOM)

Pebble-bed (UCB: PB-AHTR; DELFT)

Preliminary conceptual design

* Supported by GNEP (low level)

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Liquid fluoride salts have highvolumetric heat capacity

Compared to graphite-moderated He-cooled reactors, highthermal inertia enables:

High power density

Passive safety at large unit capacity

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Power density of PB-AHTR is ~ 2 times that ofPBMR; passive safety is comparable

CurrentPB-AHTR

(2400 MWt)

PBMR(400 MWt)

ModularPB-AHTR(900 MWt)

PBMR(400 MWt)

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An ORNL top-down economics study in 2004showed potential for low capital cost for AHTR

Study compared capital cost of 2400 MWth AHTR withmulti-module GT-MHR and S-PRISM plants

Capital cost of AHTR was found ~50% that of S-PRISM plant of comparable capacity

If the S-PRISM is ~1.25 times higher cost than LWRs,then the AHTR capital cost is 1.25 x 55% = 70% of thecapital cost of a LWR

D. T. Ingersoll, et al., "Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)," ORNL/TM-2004/104, pg. 69, May 2004.

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PB-AHTR fuel costs could be lowerthan for LWRs

Relative to LWR (4.5% enrichment, 50 MWd/kg, 33% efficiency),fuel costs of a PB-AHTR (10% enrichment, 129 MWd/kg, 46% powerconversion efficiency): Natural uranium cost: 64.2% Enrichment cost: 86.2% Fuel fabrication cost: 150% Total fuel costs: 80.7%

Assumptions: Tails assay 0.3%; LWRfuel cost breakdown 60% uraniummining/conversion, 28% enrichment,12% fabrication

If fueled with TRU, AHTR can fission ~70%in one pass: “Deep-burn” capability similar tothat proposed by GA for GT-MHR

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Pros and cons of AHTRsPros High efficiency electricity generation

High efficiency hydrogen production

May be economically competitive with LWR

Suitable heat source for high-quality liquid fuel production (e.g. tar sand)

Effective transmutation of Pu (TRU) in one pass (GA’s “Deep-burn”idea)

Relative to He-cooled HTRs:

Passive safety at high unit capacity

High power density

Low system pressure

Cons Not sustainable

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Challenges for AHTRs Compatibility of structural materials with liquid salt above 700OC

(The baseline design has a conservatively low 700°C core outlettemperature to assure high corrosion resistance using metallicstructural materials. Extensive test data available for:

Hastelloy N has well understood corrosion resistance with fluoridesalts

Alloy 800H – to provide structural strength and is ASME Section IIIcode qualified for use up to 760°C; ORNL now extending code caseto 900°C)

Component design and testing

Detailed safety analysis

Identify markets and corresponding reactor designs for liquid fuelproduction Minimize dependence on imported oil

Minimize green-house gases emission

Assess contribution from “deep-burn” capability

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3. Nuclear battery type reactors

Featuring Once for life core; no fuel handling on site

Factory assembly-line fabrication (massproduction)

Transportable

Design variants Many combinations of coolants, fuels,

spectra, concepts

Will focus on lead-alloy cooled reactors(LFR)

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GEN-IV perceived role for LFR’sThe LFR battery is a smallfactory-built turnkey plantoperating on a closed fuelcycle with very long refuelinginterval (15 to 30 years)cassette core or replaceablereactor module. Its featuresare designed to meet marketopportunities for electricityproduction on small grids,and for developing countrieswho may not wish to deployan indigenous fuel cycleinfrastructure to support theirnuclear energy systems.The battery system isdesigned for distributedgeneration of electricity andother energy products,including hydrogen andpotable water.

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Typical features of LFRs Nearly zero “burnup reactivity swing”

(fissile inventory is preserved)

Long life core (~20 years) withoutrefueling; no fueling on-site

Power level: ~40 to 400 MWth

Natural circulation cooling (no pumps,no pipes, no valves)

Autonomous load-following capability

Factory assembly-line fabrication; QA

Low cost per unit

Short construction time

Properties of Pb and Bi: Very high boiling temperature

No violent reaction with air/water

SSTAR (ANL, LLNL, LANL, INL and UCB)

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Pros and cons of LFRsPros Provide sustainable energy along with energy security

Can minimize the nuclear waste Serves as “above ground” repository for TRU from LWRs

Maintains constant TRU inventory while fissioning 238U

Utmost proliferation resistance and safeguard-ability

“Maximum possible” safety

No need for emergency evacuation zone beyond plant boundary

Easy to operate

Easy to finance

Low financial risk

Russians have a proven technology for first generation; Tcoolant<~500oC

Shared bylarge FSSfast reactors

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Pros and cons of LFRs (cont.)

Are suitable for developing countries

distributed generation of electricity

high-quality liquid fuel production (App. 5;Forsberg’s presentation at GCEP)

collocation with industrial plants

possibly, multi-module central powerplants

High efficiency hydrogen production – ifhigh-temperature structural materialscould successfully be developed

Could possibly contribute to more of theGNEP objectives than any other singlereactor technology (Slide 49)

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Pros and cons of LFRs (cont.)

Cons Economic viability is not convincingly proven

Mass-production is a pre-requisite for economic viability

Integrity of components (clad) over 20 to 30 years of operation

Transportation of spent modules (core “cassettes”)

Institutional arrangements for handling spent modules (cores)

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Challenges for LFRs Master the material technology required for a long-life operation with

infrequent inspections and maintenance intervals

Identify markets and corresponding nuclear battery reactor designs for developing countries and remote sites

oil extraction (tar-sand/shale/heavy-oil) and making of low-carbontransportation fuel (App. 5; Forsberg’s presentation); for making fuel frombiomass

desalination

central power plants

How to initiate a factory assembly-line mass-production of nuclear batterytype reactors

Institutional arrangements to support worldwide deployment of nuclearbatteries, to provide energy security and to enable fuel recycling

Early construction of a demonstration plant

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Challenges for LFRs

Develop structural material to withstand Pb and Bicorrosion and maintain mechanical integrity above 600oC(desirable at 800oC)

Develop structural material to withstand ~ 4 times dpacurrently proven for HT-9*

Develop a technology to accommodate fission gas buildupfor up to ~40% burnup*

* Common requirements for all fast reactors

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4. Deployment of fast reactors withoutseparating TRU from LWR SNF

Decouple deployment of fast reactors (FR) from development of commercialreprocessing of LWR spent nuclear fuel (SNF) that is expensive

Example of approaches:

(a) Start with weapons Pu + depleted uranium (68 ton WG Pu 10 GWe)*

Design reactor to be fuel-self-sufficient (FSS; CR ~ 1+ ε ) ( ~ 68 t TRU)

Use AIROX-like process to recycle FSS fuel. May get ~40% dischargeburnup.

(b)The MIT breed & burn GFR design approach, MIT-GFR-035, 2005

(c) Start with medium-enriched uranium; build TRU concentration to equilibrium level. Recycle FR fuel (significantly less expensive thanrecycling LWR SNF)*

* Can use LWR SNF as makeup fuel for above FR after removing volatile FP using anAIROX-like process (no TRU separation)

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The AIROX process AIROX = Atomics International Reduction Oxidation

Adopted by Korea for the DUPIC program

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Transition from enriched U to TRU-Ucore (SVBR-75/100 LFR; Appendix 4)

(Toshinsky et al., GLOBAL’07)

% Pu 0 4 7 9 11 12 13 14

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Pros and cons of decoupling FRdeployment from LWR TRU separation

Pros Early development of the technology-base for FR

Including early availability of fast spectrum for irradiationexperiments for development of advanced fuel and structuralmaterials

Early deployment of fast reactor technology for sustainability

Improving uranium resource utilization

Reducing volume of nuclear waste

Cons The economic viability of FR is not proven

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Challenges for early deployment of FRs

Feasibility of AIROX (CARDIO) like processes for recycling FR discharged fuel LWR SNF as a makeup fuel into FR

Develop processing method for relatively inexpensive crude(but better than AIROX) separation Of fission products and of uranium from LWR SNF (no partitioning of

TRU) Of fission products only for recycling FSS reactor fuel

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Promising alternative reactor options -summary

It is suggested to explore the following alternativeoptions that have the potential to increase thebenefits from fission energy:

Light-water cooled breeding reactors (RBWR)

Liquid-salt cooled high temperature thermal reactors (AHTR)

Nuclear battery type reactors (LFR)

Deployment of fast reactors without separating TRU from LWR spent fuel

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Summary of challenges

1a Feasibility of safe operation with a tight lattice and high void fraction1b TRU oxide fuel performance (fabrication) validation1c Economic viability

2a Compatibility of structural materials with liquid salt2b Component design and testing2c Detailed safety analysis2d Identify markets and reactor designs for liquid fuel production

Minimize dependence on imported oil green-house gases emission

2e Assess contribution from “deep-burn” capability

3a Master the material technology required for a long-life operation withinfrequent inspections and maintenance intervals

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Summary of challenges (cont.)

3b Identify markets and corresponding nuclear battery reactor designs for developing countries and remote sites oil extraction (tar-sand/shale/heavy-oil) and making of low-carbon transportation fuel

(App. 5; Forsberg’s presentation); for making fuel from biomass desalination central power plants

3c How to initiate a factory assembly-line mass-production of nuclear batteryreactors

3d Institutional arrangements to support worldwide deployment of nuclear batteries,to provide energy security and to enable fuel recycling

3e Early construction of a demonstration plant3f Develop structural materials for operation > 600oC3g Develop structural material to withstand ~40% burnup3h Develop a technology to accommodate fission gas buildup for up to ~40%

burnup

4a Feasibility of AIROX (CARDIO) like processes for recycling FR discharged fuel LWR SNF as a makeup fuel into FR

4b Develop processing method for crude (but better than AIROX) separation of fission products and of uranium from LWR SNF (no partitioning of TRU) of fission products only for recycling FSS reactor fuel

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Appendices

References

AHTR

Nuclear battery type reactors

Use of LWR spent fuel without actinide partitioning

Application of nuclear power for making high qualitytransportation fuel

Transmutation capability of hydride fuel in LWR

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Appendix 1: ReferencesRBWR

R. Takeda, J. Miwa and K. Moriya, “BWRs For Long-term Energy Supply and forFissioning Almost All Transuraniums,” GLOBAL’07, Boise, ID, 1725, September 9-13, 2007

AHTRC.W. Forsberg, P. Pickard and P.F. Peterson, “Molten-Salt-Cooled AdvancedHigh-Temperature Reactor for Production of Hydrogen and Electricity,” NuclearTechnology, 144, pp. 289-302 (2003) See also 3 papers of Peterson et al., GLOBAL’07, Boise, ID, September 9, 2007

3. Nuclear battery type reactorsIAEA, “Status of Small Reactor Designs Without On-Site Refuelling,” IAEA-TECDOC-1536, January 2007

4. Use of LWR spent fuel without actinide partitioningG.I. Toshinsky et al., “Opportunities To Reduce Consumption of Natural UraniumIn Reactor SVBR-75/100 When Changing Over to the Closed Fuel Cycle,”GLOBAL’07 Boise, ID, 1226, September 9-13, 2007

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Appendix 1:References (2)

M. Driscoll et al., “Engineering and Physics Optimization of Breed and Burn FastReactor Systems,” MIT Report No MIT-GFR-035, December 9, 2005P. Yarksy, M.J. Driscoll, and P. Hejzlar, “Integrated Design of a Breed and BurnGas-cooled Fast Reactor Core,” MIT-ANP-TR-107(September 2005)

AIROX like processesS. Jahshan and T. McGeehan, “An Evaluation of the Development of AIROX-Recycled Fuel in Pressurize Water Reactors,” Nuclear Technology, 106, 350,June 1994H. Feinroth, J. Guon, D. Majumdar, “An Overview of the AIROX Process and ItsPotential for Nuclear Fuel Recycle,” Proc. Int. Conf. On Future Energy Systems,GLOBAL’93, Seattle, Washington, Sept. 1993J.S. Lee et al., “Research and Development Program of KAERI for DUPIC (DirectUse of Spent PWR Fuel in CANDU Reactors),” Proc. Int. Conf. On Future EnergySystems, GLOBAL’93, Seattle, Washington, Sept. 1993

For CARDIO process see above:MIT Report No MIT-GFR-035, December 9,2005

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Appendix 2: The AHTR combines two oldertechnologies

Liquid fluoride salt coolants (LiF-BeF2)Boiling point ~1400ºCReacts very slowly in airExcellent heat transferTransparent, clean fluoride salt

Coated particle fuel

1600°C

Fuel failure fraction vs. temperature

max.PB-AHTR

temp

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The AHTR can produce electricityand/or hydrogen

03-154

Water

H2

Oxygen

Thermochemical

Plant

Buffer Salt

Tank

Reactor Vessel

Fuel

PRACS Heat

Exchanger

DRACS Loop

Air Inlet

Control

Rods Pump

Heat

Transport

Passive Decay

Heat Removal

AHTR-MI

Reactor

Electricity or Hydrogen

Production

Turbo-

compressors

Reheaters

Coolers

Electricity

Metallic Internals

Tout = 700oC

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Comparison of building volume, concrete volumeand steel consumption - LWR values used as

reference (source: Peterson, UCB)

IntermediateTemperatureTout ~ 850oC

ORNL data Estimated from plant arrangement drawings

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Appendix 3Nuclear battery type LM cooled

reactorsExamples:

4S: Super-Safe, Small and Simple reactor (TOSHIBA; CRIEPI, Japan)

SVBR-75/100: Lead-Bismuth Cooled Fast Reactor(IPPE, Obninsk; Russian Federation)

SSTAR: Secure Transportable Autonomous Reactor (ANL/LLNL/LANL/INL/UCB)

ENHS: Encapsulated Nuclear Heat Source Reactor(UCB/LLNL/ANL/Westinghouse)

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Selected characteristics of LM coolednuclear battery reactors

# HLM Coolant Power loops type outlet T Circulation (MWt)

4S 1 Na 485oC Forced 135 (30) SVBR-75/100 1 Pb-Bi 480oC Forced 280 ENHS 2 Pb-Bi <600oC Natural 60 - 400 STAR-LM 1 Pb 590oC Natural 400 STAR-H 1 Pb 800oC Natural 400 SSTAR 1 Pb 570oC Natural 45

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STAR designs

SSTAR (ANL, LLNL, LANL, INL and UCB)

Possible configuration to enhanceprotection (STAR-LM)

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The ENHS30

m27

m

8m

2m

3m 2m

Number of Stacks = 4Cross Section of Stack

3m

3.64m (O.D; t=0.05)

17.6

25m

ENHS module

Reactor pool

Reactor Vessel Air Cooling System (RVACS)

Steam generators6.94m (I.D.)

Seismic isolators

Underground silo

Schematic vertical cut through the ENHS reactor

underground silo no LOCA

no pumps

no pipes

no valves

factory fueled

weld-sealed

20 years core

no fueling on site

Module is replaced

shipping cask

no DHRS butRVACS

the Module

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SVBR-75/100 Reactor Proposed by IPPE, EDO“Gidropress”, Russian

Railway transportable

Standard design for multi-purpose applications -renovation of PWRs -steam supply (oil shale mining) -developing countries -floating power plants

Power level = 75÷100 MWe

8.8 (15) EFPY core life

Lead-bismuth coolant

Conversion ratio = 1

Forced cooling (mechanical)

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SVBR-75/100 nuclear islandOne application: Renovation of the 2nd, 3rd, and 4th units of the

Novovoronezhskaya VVER NPP

Concrete vault

Reactormonoblock

Reactorcompartment

PHRS tank

Steam separator

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1600 MWe plantbased on 16 SVBR-100 modules

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Economics of SVBR 75/100

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SVBR 75/100 has many fuel options

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Fuel cycle flexibility“The flexibility of SVBR-75/100 in relation to fuel cycle

technologies is realized in compliance with the principle: “To

operate using the type of fuel and fuel cycle that are demonstrated

and most economical today”

[Russian developers of the SVBR-75/100 (slightly modified by EG]

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GNEP goals LFRs could contribute to

Technology to be Where LFR GNEP Goal developed could contribute?

1. Expand N-E (construct NPP) +2. Prol.-Res. recycling UREX+/Pyro + 3. Minimize waste Recycling/High BU +4. Develop ABR SFR/MA fuel +5. Reliable F-C services +6. Demonstrate SMR SMR +7. Develop safeguards +

Additional Goal8. High quality liquid fuels VHTR +

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On economic viability of nuclearbattery reactors

Their COE has a chance to be competitive with that of largecapacity commercial power plant due to the following:

Simpler design with fewer components No fuelling and no fuel hardware on site Factory mass production Short on-site assembly High availability Small staff Very long plant life Close match between demand and supply Can install several modules in one plant No need for long transmission lines Needs only depleted U or LWR SNF for makeup fuel

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Low financial risk

Relatively small investment per reactor unit

Highly modular standard design

Factory fabrication with good quality assurance

Cost over-runs during construction are unlikely

Superb passive safety assures against physical damage

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Trends in power plant capacity evolution

Reference: Ning Li, “Size matters: installed maximal unit size predicts market lifecycles of electricity generation technologies and systems” Presentation at UCB, Sept. 2007

“The saturation of themax size envelopesignals the end of thecorrespondingtechnologies andsystems in that market.So far no technologies orsystems have made asignificant comebackwith incrementalimprovements inperformance andeconomics”

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US nuclear power plant capacity evolution

Reference: Ning Li, “Size matters: installed maximal unit size predicts market lifecycles of electricity generation technologies and systems” Presentation at UCB, Sept. 2007

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Illustration of a new technology taking over

Reference: Ning Li, “Size matters: installed maximal unit size predicts market lifecycles of electricity generation technologies and systems” Presentation at UCB, Sept. 2007

“The emergence of peak unit sizedistributions and rapid marketpenetration of the recent systems,namely power plants with gasturbines and combined cycles,demonstrate the overwhelmingeconomic advantages ofstandardization, modularizationand factory-based massproduction. It also indicates thatthe 100-200 MWe units are wellreceived and perhaps optimal foreven a large and well-developedUS market”

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Trends in power plant capacity

Reference: Ning Li, “Size matters: installed maximal unit size predicts market lifecycles of electricity generation technologies and systems” Presentation at UCB, Sept. 2007

“100-200 MWe unitsare well received andperhaps optimal foreven a large and well-developed US market”

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Appendix 4Recycling SNF without separation of

TRU

Example (b): The MIT breed & burn approach

Example (c): The IPPE breed & burn approach

The Cardio process

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Example (b): Driscoll et al., MIT-GFR-035 (2005)

Load GFR core with enriched U and operate on a once-throughfuel cycle to high burnups; a fraction of the Pu produced isfissioned in-situ; overall natural uranium utilization is > of once-thru LWR

startup core has 8 w/o avg., 10 w/o max.

Makeup fuel is 5 w/o 235U

Discharge burnup ≥ 150 MWd/kg in 6 batches

Natural uranium utilization >3 that of current LWRs, with no recycling

Can double the U utilization by recycling fuel using the CARDIOprocess to remove only volatile fission products (Appendix 3)

May also recycle to LWR

Reduced fuel cycle cost may make it practical to commercialize fastreactors without first deploying reprocessing and fuel recyclingfacilities

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The MIT deep & burn GFR design MIT-GFR-035, 2005

Fuel cycle performance: Natural uranium utilization >3 that of current LWRs, withno recycling

GFR technology is not ready for near-term deployment; feasibility of safeeconomical GFR is yet to be proven. Similar fuel cycle can be implemented usingLM cooled reactors

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Example (c): G. I. Toshinsky et al.IPPE, Obninsk, Russia (Ref: GLOBAL’07): In the nearest future start LFR (SVBR-75/100; Appendix 4) using enriched

uranium dioxide fuel and operate in the open fuel cycle. TRU builds up to~4 % of HM (at 69 GWD/tU average discharge BU; no fuel shuffling)

When cost of U increases to the point that makes recycling of TRUeconomical, start recycling the FR SNF (do not partition TRU from LWRSNF); cost of TRU ~ 1/(TRU % in SNF)

After TRU concentration reaches equilibrium, use LWR SNF as themakeup fuel for the FR after undergoing an AIROX process that removesonly volatile FPs (Appendix 3)

In the meantime develop more advanced fuels (better performance isexpected if using metallic or nitride fuel) and compatible reprocessing and

refabrication capability

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Cumulative consumption of natural U(Toshinsky et al., GLOBAL’07)

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The CARDIO (CArbon DIoxideOxidation) dry process for UC fuel* UC spent fuel can be converted into UO2 via

UC + 3CO2 →UO2 + 4CO at T>670°C

Apply AIROX process to remove volatile fission products

Apply carbothermic reduction of oxide fuel in high purity inertatmosphere

UO2 + 3C →UC + 2CO

* Indian fast breeder work on carbide fueling: B. Raj, “The Core of Stage Two,”Nuclear Engineering International, Vol. 50, No. 614, Sept. 2005MIT recently confirmed feasibility (MIT-GFR-035, 2005)

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Appendix 5Application of nuclear power for making

high quality transportation fuelMore comprehensive coverage – in Charles Forsberg’s presentation

SMR without on-site refueling (such as LFR) and modularHTR/AHTR are particularly suitable for co-location with manyindustrial plants, including for oil production from tar sand (Alberta,Canada) or for ethanol production from corn ( Mid-west)

By providing process heat (steam) the nuclear reactors couldincrease the amount of oil extracted from tar sand (heavy oil) whilereducing CO2 emission

By providing hydrogen the low quality (low H/C) oil can be convertedto high quality transportation fuel thus additionally reducing theCO2 emission per fuel heat of combustion

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Tar sands = very large oil resource

Red Deer

Edmonton

Calgary

Peace River

Alberta

Lloydminster

Fort McMurray

Cold Lake

Athabasca

Wabasca

Production from oil sands in Alberta could be 2.8million BOPD in 2015, up from 1.2 million BOPD in2004

Current tar sands carbon intensity is 15 to 40% higherthan for conventional oil production

By using nuclear steam & H2, one can reduce thecarbon intensity to zero

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World’s largest oil accumulations

119CanadaTar SandWabasca

123RussiaTar SandMelekess160VenezuelaOil FieldBolivar Coast190KuwaitOil FieldBurgan190Saudi ArabiaOil FieldGhawar271CanadaTar SandCold Lake869CanadaTar SandAthabasca1,200VenezuelaX-Heavy OilOrinocoOOIP (109 Bbl)CountryTypeName

Source: Roadifer 1987

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Appendix 6: Transmutation capability ofhydride fuel in LWR

Recycling either only Pu, Pu+Np or all TRU

Fuel is inert matrix of ZrH1.6 with PuH2 or PuH2+NpH2 etc

Cycle duration is that of reference UO2 fueled PWR

3-batch core

Fuel rod diameter and lattice pitch as in reference PWR

All reactivity coefficients are required to be negativethroughout the cycle for each cycle

UC Berkeley66

What do we know on hydride fuel? U-ZrH1.6 is successfully operating in dozens of TRIGA research reactors for

> 40 years (water temperature; ~ 40oC)

The only U.S. designed nuclear reactor to fly in space, SNAP-10A, wasfueled with U-ZrH1.6. It operated with NaK coolant inlet/outlet temperatures of393/560oC. The SNAP-8 reactor was designed and tested with NaK coolantoutlet temperature of 702oC!

In late 90’s, Prof. M. Yamawaki et al. fabricated a number of U-Thz-Zry-Hx fuelsamples for actinides burning in fast reactors and performed in-coreirradiation tests. The maximum permissible linear heat rate is estimated at ~50 kW/m

PuH2 is even more stable than ZrH1.6

Relative to oxide fuel, hydride fuel has

~5 times higher thermal conductivity

Very low fraction of fission gas release (can operate to high burnups)

But higher swelling – can be accommodated using LM bonding

UC Berkeley67

Hydride fuel is safely operating at higher linearheat rates and higher burnup than oxide fuel

Characteristic High Power TRIGA BWR (Romania)

Fuel pin O.D. (cm) 1.294 1.056Cladding Material SS ZyThickness (mm) 0.40 0.86Fuel loading (kg U/m) 0.489 0.795Avg. (peak) linear-heat-rate (kW/m) 37 (74) > 21 (44)Specific power (W/g-HM) 75.7 > 25.9Power density (W/cm3) 138.7 > 56Discharge burnup (MWd/kgHM) 120 > 50Energy extracted from fuel (MWd/m) 59.2 > 39.8Peak fuel temperature (oC) 550(a) 2000Coolant exit temperature (oC) ~70 ~310

(a) Peak hydride fuel temperature at BWR or PWR operating conditions < 550oC

UC Berkeley68

Findings of UCB study on multi-recycling Most promising hydride fuel is PuH2-ZrH1.6 (PUZH); i.e., has

no U loading. It enables infinite number of recycling in PWRas significant H is left in core even when voiding 100% ofH2O

Most promising oxide fuel is PuO2-ZrO2 (MOX). It is limitedto only few (2?) recycles due to positive reactivity insertionby large voiding

At first recycle, PUZH has only slightly better TRUtransmutation performance than MOX

Recycle-dependent transmutation capability of PUZH fuel

UC Berkeley69

Findings of UCB study on multi-recycling

The transmutation fraction with PUZH fuel in PWR is ~64%in the first recycle and gradually decreases with recyclingfrom ~50% in the second recycle until it stabilizes at ~20% atthe equilibrium cycle (next slide)

It appears possible to multi-recycle with hydride fuel in PWR up to 8-12 times Pu and Np (corresponding to 110-170 years).

Only few (2-3 ?) all the TRU (work in progress)

Recommendation: study: Compatibility of hydride fuel with Zircaloy cladding and with high

temperature/pressure water

Recycling feasibility of ZrH1.6 matrix fuel

UC Berkeley70

UCB results for PuH2-ZrH1.6 recycling in PWR

Cycle

#

Pu

loaded

(g/cc)

Burnup

(GWD/

MT)

Pu

consumed

(g/cc)

Pu

recycled

(g/cc)

PU

destruction

fraction (%)

TRU

destruction

fraction (%)

Fiss Pu %

at

discharge

Specific

power

(W/gHM)

kgPu/

MWt-

yr

KgTRU

/MWt-

yr

0 0.734 628.4 0.539 73.4 64.4 21.6 456.7 0.4 1 0.36 1 0.931 508.2 0.585 0.195 62.9 50.7 23.7 359.1 0.45 0.36

2 1.083 442.1 0.590 0.346 54.5 45.4 23.4 308.1 0.45 0.37

3 1.209 391.4 0.605 0.492 50.1 40.2 24.0 275.3 0.46 0.37

4 1.327 357.9 0.613 0.604 46.2 36.8 24.4 250.3 0.47 0.37

5 1.430 330.6 0.626 0.714 43.7 34.1 24.8 231.3 0.48 0.37 6 1.520 311.0 0.634 0.805 41.7 32.1 25.0 217.5 0.48 0.37

7 1.602 293.2 0.640 0.886 40.0 30.3 25.3 205.3 0.49 0.37

8 1.675 281.0 0.647 0.962 38.6 29.0 25.4 196.5 0.49 0.37

9 1.745 268.5 0.651 1.028 37.3 27.8 25.7 187.8 0.50 0.37

10 1.803 259.9 0.657 1.093 36.4 26.9 25.7 181.8 0.50 0.37 11 1.861 250.7 0.660 1.147 35.4 25.9 26.0 175.3 0.50 0.37

12 1.910 244.5 0.664 1.202 34.8 25.3 26.0 171.0 0.51 0.37

13 1.951 236.5 0.666 1.246 34.1 24.5 26.1 166.7 0.51 0.36

14 1.996 233.2 0.665 1.285 33.3 24.1 26.1 163.1 0.51 0.37

15 2.034 228.5 0.672 1.331 33.1 23.6 26.2 160.1 0.51 0.37 16 2.073 223.6 0.672 1.362 32.4 23.1 26.3 156.4 0.51 0.37

17 2.104 220.4 0.676 1.400 32.1 22.8 26.3 154.2 0.52 0.37

18 2.132 217.6 0.678 1.428 31.8 22.5 26.3 152.2 0.52 0.37

19 2.158 215.1 0.680 1.455 31.5 22.2 26.3 150.5 0.52 0.37

20 2.187 211.3 0.680 1.478 31.1 21.9 26.4 147.8 0.52 0.36 21 2.209 209.3 0.682 1.507 30.9 21.7 26.4 146.4 0.52 0.36

22 2.234 208.2 0.683 1.528 30.6 21.5 26.4 144.8 0.52 0.37

23 2.249 205.7 0.688 1.551 30.6 21.3 26.4 143.9 0.52 0.36

24 2.272 205.5 0.686 1.561 30.2 21.3 26.4 142.5 0.52 0.37

25 2.282 202.9 0.692 1.585 30.3 21.0 26.4 141.9 0.53 0.37 26 2.300 200.4 0.686 1.590 29.9 20.7 26.5 140.2 0.52 0.36

27 2.314 198.4 0.687 1.613 29.7 20.5 26.5 139.3 0.52 0.36

28 2.326 197.5 0.687 1.626 29.5 20.4 26.5 138.6 0.52 0.36

29 2.338 197.3 0.687 1.639 29.4 20.4 26.5 138.0 0.52 0.36

30 2.348 196.6 0.690 1.651 29.4 20.3 26.5 137.4 0.53 0.36

31 2.356 195.8 0.691 1.658 29.3 20.2 26.5 136.9 0.53 0.36 32 2.364 195.1 0.691 1.665 29.2 20.2 26.5 136.5 0.53 0.36

33 2.372 194.6 1.673 20.1 26.4 136.1

UC Berkeley71

Equilibrium cycle transmutation ability of PWRloaded with Pu hydride fuel is significant

0.36*

20*

PWRHydride*

M – 0.27Ox – 0.27

M – 0.17Ox – 0.18

KgTRUtransmuted per MWt-yr

M - 27Ox - 30

M - 19Ox - 22

% TRUtransmutedper recycle

ABRCR=0.25

ABRCR=0.5

Transmutationcharacteristic

* Recycling only Pu using PuH2-ZrH1.6 fuel while ABR recycles TRU; not afair comparison; brought only to give preliminary indication