Fast Reactor Development: Motivation, Challenges and Recent Advances

Presented at

NE50 - Symposium on the Future of Nuclear Energy Georgia Institute of Technology November 1, 2012 By

Hussein S. Khalil Director, Nuclear Engineering Division Argonne National Laboratory

Overview

Fast reactor characteristics and role in a closed fuel cycle

International development status of sodium cooled fast reactors (SFR)

SFR development in the U.S.

– Transition to metallic fuel

– Current R&D focus

– Breed & burn concepts

2

PWR

SFR

0.00

0.20

0.40

0.60

0.80

1.00

Fast Reactors Enable a (Fully) Closed Fuel Cycle

0.00

0.10

0.20

0.30

0.40

0.50

1.00E-03 1.00E-01 1.00E+01 1.00E+03 1.00E+05 1.00E+07

Energy, eV

Norm

aliz

ed

Flu

x/L

eth

arg

y

PWR

VHTR

SCWR

SFR

GFR

LFR

Neutron energy distribution

3

Neutron excess (η > 2) allows fissile regeneration and breeding

Via conversion of 238U to 239Pu

Alternatively, can “waste” excess neutrons to consume Pu and MA

... η = (νΣf/Σa)fuel

Heavy-mass (or low-density) coolant limits neutron moderation

Fast spectrum enhances actinide fission probability

Limits buildup of higher Pu isotopes and MA

Key for full recycle

Energy, eV

Norm

aliz

ed F

lux /

Leth

arg

y

Probability of fission per absorbed neutron

Motivation for Closing the Fuel Cycle

4

High-Level Waste Toxicity Normalized to Natural Uranium Ore

1.00E-01

1.00E+00

1.00E+01

1.00E+02

1.00E+03

1.00E+04

10 100 1000 10000Time (years)

1X

1XT

1Z

1G

2X

2Z

2G

3M

3T

ALWR spent fuel

No

rmal

ize

d H

azar

d

LWR

LWR + FR Strategies

Time, Years

Radiotoxicity of waste relative to mined U

Increase uranium utilization to >95%

– Related benefits from reduced need for uranium mining & enrichment

Improve waste management, through sharply reduced

– TRU content of waste

– Long-term (>100 y) and integrated heat emission by waste

Typical

Architecture

for Recycle

System

Requirements/challenges

– Favorable impacts: economic, safety, environmental, and proliferation risk

Key technical challenges

– Cost reduction, especially for FR

– Robustness of safety & reliability case

– Fuel fabrication and performance

– Minimal (<1%) TRU losses in recycle

– Durable, leach resistant waste forms

– Cost-effective fuel cycle safeguards

Key coolant characteristics – Low pumping power – Significant heat capacity – High thermal conductivity – Non-corrosive to structural steels – Large margin to coolant boiling (880°C) with low-pressure RCS

Can be exploited in SFR design to achieve passive (intrinsic) safety – Pool configuration for RCS enhances heat capacity – Negative reactivity from coolant temperature increase (+low stored heat in fuel) – Natural circulation flows in heat transport circuits

No coolant voiding or fuel damage in LOF or LOHS without scram accident

Nevertheless, coolant voiding and core compaction hypothesized (non-mechanistically) as reactivity addition mechanisms

High chemical reactivity with air and water must be controlled in design and operation

Sodium Cooled

Fast Reactor (SFR)

5

Nearly 400 reactor-years experience w/SFR in several countries – Experimental, prototype and demonstration units

Currently operational units: – Experimental reactors – Joyo (Japan), BOR-60 (Russia), FBTR (India), CEFR (China)

– Demonstration reactors – Monju (Japan), BN-600 (Russia)

A larger 880 MWe SFR (BN-800) is under construction in Russia at Beloyarsk – Two more planned in China

– Pool type layout with 3 IXH

– Primary sodium outlet temperature ~550⁰C

– Capital/operating cost estimated to be 20%/10% greater than VVER

A 500 MWe prototype fast reactor (PFBR) is under construction in India at Kalpakkam

– MOX fueled

– Th blanket to breed fissile U-233

Sodium Cooled Fast Reactor (SFR)

6

U.S. Fast

Reactors

7

Facility

Location

Capacity (MWt)

Mission

Dates

EBR-I Idaho 1.4 R&D – built to demonstrate fuel breeding; first usable nuclear electricity

1951-1963

EBR-II Idaho 62.5 R&D – built to demonstrate integral recycling of metallic fuel; first pool-type fast reactor

1963-1994

Fermi-1 Michigan 200 Power generation – employed U-Mo metallic fuel

1963-1972

SEFOR Arkansas 20 Safety testing – built to measure Doppler reactivity feedback for oxide fuel

1969-1972

FFTF Washington 400 Fuel & material testing – built in support of the U.S. LMFBR program

1980-1992

CRBRP Tennessee 975 LMFBR demonstration plant – construction was initiated before project was terminated in 1983

Designed but not Built

Fast Reactor Fuel Characteristics

Metal

U-20Pu-10Zr

Oxide

UO2-20PuO2

Nitride

UN-20PuN

Carbide

UC-20PuC

Heavy Metal Density, g/cm3 14.1 9.3 13.1 12.4

Melting Temperature, ºK 1350 3000 3035* 2575

Thermal Conductivity, W/cm-ºK 0.16 0.023 0.26 0.20

Operating CL Temperature a, ºK

(T/Tmelt)

1060

(0.8)

2360

(0.8)

1000

(0.3)

1030

(0.4)

Fuel-Cladding Solidus, ºK 1000 1675 1400 1390

Thermal Expansion Coeff., 1/ºK 1.7E-5 1.2E-5 1.0E-5 1.2E-5

Heat Capacity, J/gºK 0.17 0.34 0.26 0.26

Reactor Experience US, UK RF, France, Japan

US, UK

India

Research & Testing Experience US, Japan, RoK,

China

RF, France, Japan,

US, China

US, RF, Japan India, France

a at 40 kW/m

8

Intensive 10-y effort through 1994 in the IFR and ALMR programs

– IFR R&D and technology demonstrations at ANL (in IL and ID)

– ALMR design by General Electric (PRISM)

Key IFR features adopted by GE in the PRISM design

– Sodium coolant

– Metallic Fuel

– Pool layout

– Electrometallurgical fuel recycle (aka pyroprocessing)

Fast reactor R&D resumed in the US starting in 2003

– At much reduced scale, as part of the DOE Generation IV and Advanced Fuel Cycle programs

– Currently part of the DOE-NE Advanced Reactor Concepts (ARC) program

U.S. Fast Reactor Program following CRBR

9

Passive shutdown and decay heat removal

Schematic of Metal Fuel Pin (not to scale)

Injection Casting of Metal Fuel Slugs

10

Fuel Slug (U-Pu-10%Zr)

End Plug

Cladding (SS316, D-9 or HT-9) Sodium Bond

Gas Plenum

Pore morphology of irradiated U-10Zr

Fuel Periphery (dominantly a-phase)

10 μm

Fuel Center (dominantly g-phase)

100 μm

Metal Alloy Fuel

(U-Zr or U-Pu-Zr)

Over 40,000 U-Fs pins irradiated in EBR-II through early 1980’s

Approx. 17,000 U-Zr and 700 U-Pu-Zr fuel pins irradiated in 1984-1994

– U-Pu-Zr fuel reached peak burnup of ~20%.

7 metal fuel assemblies irradiated in FFTF

– Lead assembly achieved peak burnup of 16%.

– One assembly contained U-Pu-Zr, which achieved peak burnup of 10%.

Three MA bearing pins fabricated and irradiated to 6% burnup

– Initial composition: 68.2%U, 20.2%Pu, 9.1%Zr, 1.2%Am, and 1.3%Np

– Approximately 40% of the Am was lost during casting due primarily to volatile impurities in the Pu-Am feedstock

– Judicious selection of the cover gas pressure during the melt preparation and the mold vacuum level during casting is expected reduce the Am loss by ~200 times

– Extensive PIE revealed

• Similar U, Pu behavior as in non-MA bearing fuel

• Am follows Zr; precipitates in pores

• Np is sessile

11

Irradiation Experience with Metallic Fuel

Sample history of a typical driver fuel in EBR-II:

– 40 start-ups and shutdowns

– 5 15%-overpower transients

– 3 60%-overpower transients

– 45 loss-of-flow and loss-of-heat-sink tests

Excellent off-normal performance observed

– Transient accommodation

– Run-beyond-cladding-breach performance

– Transient overpower failure margins

– Pre-failure axial extrusion behavior

RBCB Test of Metal Fuel with 12% Burnup

RBCB Test of Oxide Fuel with 9% Burnup

12

Transient and Off-Normal Performance

Intrinsic Safety Demonstration

13

Time, s

-100 0 100 200 300 400

800

700

600

500

400

Outlet Temperature

Tem

per

atu

re, °

C

Time, s

-100 0 100 200 300 400

100

50

0

Primary Flow Rate Pe

rcen

t Time, s

-100 0 100 200 300 400

0.0

- 0.1

- 0.2

- 0.3

- 0.4

Reactivity

Do

llars

Time, s

-100 0 100 200 300 400

Power

Perc

ent

100

50

0

EBR-II Loss-of-Flow w/o Scram Behavior

Two major accidents were simulated in EBR-II tests conducted in Apr. 1986 LOF without scram from

full power LOHS without scram

from full power Tests demonstrated the passive safety potential of SFRs … Pool design provides

thermal inertia Low stored Doppler

reactivity due to high thermal conductivity of metal fuel

Negative feedback from core expansion

Fuel Temperature at Full Power

(Oxide Core)

Initial Coolant Temperature

Fuel Temperature at Full Power

(Metal Core)

Asymptotic Temperature

After LOF

Positive Doppler

Reactivity

Negative Expansion Reactivity

With metallic fuel, lower operating fuel temperature and hence smaller stored Doppler reactivity leads to a much lower asymptotic temperature

14

Benefit of Metallic Fuel in LOF w/o Scram

Current SFR R&D in the U.S.

15

Target significant cost reduction and enhanced assurance of safety & reliability – Commercial deployment by ~mid century

R&D areas – Advanced technologies (power conversion, ISI, refueling, …) – Improved materials – Safety research exploiting HPC capabilities

Research focused and integrated through conceptual

development a sodium cooled SMR concept (AFR-100)

– Reactor power of 250MWt/100MWe

– 30-year refueling interval; no on-site fuel storage

– Transportable from pre‐licensed factory

– Compact layout, with fission gas vented fuel (option) and advanced shielding materials

– High operating temperature

LMRs

16

4S 10 MWe SFR

Toshiba, Japan

PRISM 311 MWe SFR

General Electric, USA

• High power density hence compact core facilitates design of transportable reactor modules

• Efficient U conversion facilitates design for long refueling interval

• Outlet temperature enhances power conversion efficiency and expands output beyond electricity

• Intrinsic safety behavior may expand site options

ARC-100 100 MWe SFR

Adv. Reactor Concepts, USA

SMFR 50 MWe SFR

ANL-CEA-JAEA

Small Modular SFRs Benefits of SFR technology for SMR missions

Breed & Burn

Fast Reactor Concept

17

Fast-spectrum concept fueled with DU DU is converted to Pu w/irradiation Fissile material required only for the

first core, to initiate the conversion

Fission wave propagates from fissile “starter” through adjacent DU zone

Kinf vs. burnup for DU in a fast spectrum

Traveling wave reactor is a particular variant

Candle Concept

(Tokyo Inst. of Tech.)

17

Breed & Burn Concepts

Contemporary interest – CANDLE Reactor (Tokyo Institute of

Technology) – Traveling Wave Reactor (TerraPower) – Energy Multiplier Module (General Atomics) – National Lab and university studies

Drawbacks/development challenges – Low power density in DU assemblies – Large power swings – Reactivity management and control – Stability of power distribution – Demands on fuels and materials

Fuel recycle or “reconditioning” after discharge an option for all concepts

18

Annular layout with central DU zone

– Fission power shifts inward over time

Configured for extended (54 y) operation on the initially loaded fissile and DU assemblies

Enriched U (6.2% avg.) fuels initial core

Only DU assemblies subsequently loaded

– 1.5 y refueling interval

– 12 DU assemblies loaded in each refueling

– Each assembly “shuffled” to new location

Slowly transitions to equilibrium cycle

Stationary Fertile

Assembly (SFA) Concept

Moving Fertile Assembly

(MFA) Concept

Breed & Burn Concepts Studied at Argonne

19

Characteristics of B&B Concepts (3000 MWt Basis)

Parameter PWR

Typical SFR

B&B: SFA

B&B: MFA

HM inventory, t 89 70 320 178

Specific power, MW/t 33.7 42 9.4 16.9

Power density, W/cc 80 250 58.2 96.2

Refueling interval, yr 1.5 1 54 1.5

Fissile enrichment of starter, % 12 12.2 6.2

Excess reactivity (max / min), %Δk 1.5 / 0.5 3.9 / 0.5 3.1 / 0.6

Avg. power density (BOC/EOC), W/cc ‒ Fissile (starter) region ‒ Fertile (DU) region

350

50 - 100

171 / 48 2.8 / 63.3

177 / ----- 5.5 / 96.1

Avg. discharge burnup (GWd/t) ‒ Fissile fuel ‒ Fertile fuel

100

30

316 198

----

277

Peak fast fluence, x1023 neutrons/cm2

‒ Fissile fuel ‒ Fertile fuel

3.5 2.0

22.1 21.6

23.4 21.7

Discharge rate (t/GWe-yr) ‒ UNF total ‒ Plutonium ‒ MA

19.7 0.24 0.02

-- -- --

4.93 0.37 0.01

2.90 0.28 0.01

20

= uranium utilization

Summary

Fast reactor characteristics enable full actinide recycle and provide flexibility in actinide management

Considerable experience gained with experimental/prototype/demonstration SFRs in several countries

New prototype SFR plants are under construction in Russia and India, and planned in China, France and Korea

The U.S. IFR/ALMR program (1984-1994) demonstrated key advances achievable through the use of metallic fuel, including passive safety

Current fast reactor R&D in the U.S. is targeting technology innovations to support commercialization in the 2050 time frame

Several commercial organizations are pursuing the development of fast reactors, including concepts that avoid, limit or defer fuel reprocessing

21

Questions & Discussion

Uranium Utilization

24

PWR-50GWd/t PWR-100GWd/t HTGR SFR

Burnup, % 5 10 10.5 12

Enrichment, % 4.2 8.5 14.0 12.5

Utilization, % 0.6 0.6 0.4 0.5

LWR LWR-SFR SFR

UOX MOX LWR-UOX Fast Burner Fast Converter

Power sharing, % 90 10 57 43 100

Burnup, % 5 10 5 9 -

Enrichment, % 4.2 - 4.2 12.5 -

Utilization, % 0.7 1.4 ~99

Once-through systems (LEU)

Recycling Systems

Is it possible to improve U utilization significantly … • without recycle? • with limited recycle?

What are the implications for waste management?

Discharged Fuel Composition

PWR50 CANDLE CBZ MB3

Discharge rate (t/GWe-yr)

UNF total 19.7 3.42 4.93 2.90

Uranium 18.4 2.33 3.68 1.77

Plutonium 0.24 0.24 0.37 0.28

Minor Actinides (MA) 0.02 0.01 0.01 0.01

Fission products 1.03 0.84 0.87 0.85

Plutonium isotopic vector (%)

Pu-238 2.5 0.6 1.0 0.6

Pu-239 51.6 76.8 82.6 74.3

Pu-240 23.7 21.1 15.0 22.3

Pu-241 14.9 0.7 1.1 2.1

Pu-242 7.2 0.8 0.3 0.6

Pu discharge rates for B&B systems comparable to or higher than PWR MA discharge rates are lower

25

Comparison of Fuel Cycle Options

26 26

Open Cycle

Breed & Burn

Limited Separations

Full Recycle

Uranium Utilization

<1% 1% to <30%

Consumes DU “waste”

30% to ? >95%

SNF/Waste Management

Interim storage and long term (geologic) isolation

Slows discharge of UNF

Comparable UNF characteristics

TBD Only recycle losses and FP are disposed; maximum benefit (targeted)

Proliferation Risk

Reference Potentially reduced:

Reduce enrichment

Higher Pu fraction in discharged fuel

TBD Potentially increased (separations & breeding capacity vs. enrichment)

Safety Reference TBD (concerns include stability and fuel loading errors)

TBD Partially established (significant distinctions between FR and LWR)

Economics (Cost)

Reference Penalized by low power density and local power swings

TBD Strongly impacted by FR cost; expected to increase

Modified Open