Liquid Metal Cooled Fast Reactors: Worldwide Overview Worldwide ...

Liquid Metal Cooled 

Fast Reactors: 

Worldwide Overview 

www.inl.gov 

Alexander Stanculescu 

17-20 April 2012

Fast! 

Source: Astérix le Gaulois (René Goscinny, Albert Uderzo) 

International Workshop on Innovative Nuclear Reactors Cooled by Heavy Liquid 

Metals: Status And Perspectives, Pisa, Italy 17-20 April 2012 

2

Thermal vs. Fast 

fission in fissile 

capture in fertile 

LWRs use low enr U, 

HWRs even nat U 

• Fuel temperature increase ⇒ 

Doppler broadening 

⇒ Loss of reactivity 

• Loss of liquid coolant 

⇒ loss of moderator 

⇒ loss of reactivity 

fission in fissile 

fissile to fertile concentration 

fission/capture and neutron multiplicity 

BREEDING 

• Locally positive void effect 

• Doppler effect smaller 

• Fuel expansion ⇒ loss of reactivity 

• Power increase ⇒ loss of reactivity 

• Not in most reactive geometry 

• Smaller β eff ⇒ less margin to prompt 

criticality 

International Workshop on Innovative Nuclear Reactors Cooled by Heavy Liquid Metals: 

Status And Perspectives, Pisa, Italy 17-20 April 2012 

3

Another Way to Show Fast Reactor 

Breeding Potential 

Isotope 

235 

U 

239 

Pu 

233 

U 

Spectrum Thermal Fast Thermal Fast Thermal Fast 

σ f (barn) 582 1.81 743 1.76 531 2.79 

σ c (barn) 101 0.52 270 0.46 46 0.33 

α=σ c /σ f 0.17 0.29 0.36 0.26 0.09 0.12 

ν 2.42 2.43 2.87 2.94 2.49 2.53 

η=νσ f /σ a 

Neutron Multiplicity 

2.07 1.88 2.11 2.33 2.29 2.27 

β eff (pcm) 650 210 280 

Breeding Requires η = ν 

1 + α > 2 


Status And Perspectives, Pisa, Italy 17-20 April 2012 4

… and Minor Actinides Utilization Potential 

Isotope 

Thermal 

α = σ c / σ f 

Fast 

235 

U 0.22 0.29 

238 U 8.3 7.5 

239 

Pu 0.58 0.3 

240 

Pu 396.6 1.6 

241 

Pu 0.40 0.19 

242 

Pu 65.5 1.8 

237 

Np 63 5.3 

241 

Am 100 7.4 

243 Am 111 8.6 

244 

Cm 16 1.4 

245 

Cm 0.15 0.18 

α = σ c 

σ f 

is more favorable to minor actinides 

fission in the fast spectrum. 



5

Fast Reactors (FRs) Offer Great Potential 

Increased uranium utilization 

– High fission neutron yield per neutron absorbed in 239 Pu in a fast 

spectrum 

– At 2–3% fuel cycle losses and 100–200 GWd/t burnup it is possible to 

fission more than 80% of the natural uranium resources 

– FR thermal efficiency ~ 0.4 (~ 0.3 for a thermal reactor) 

amount of energy that can be recovered per unit uranium mass is 

greater 

Higher nuclear power growth rates with available resources 

thanks to fissile material breeding 

– Advanced denser fuels (carbide, nitride and metal) have higher 

breeding potential compared to oxide fuels, allowing even higher 

nuclear power growth rates 



6

FRs Offer Great Potential (cont’d) 

Utilization of used fuel and reduction of waste 

– Long-lived minor actinides ( 237 Np, 241 Am, 242 Am, 243 Am and 244 Cm), posing 

radiological hazard in the repository, are produced in thermal reactors since 

Σ cap ˃ Σ fiss 

– MA are produced at a lower rate in FRs, due to smaller Σ cap /Σ fiss , and even 

utilized through fission 

In FRs, even with multiple recycle, the Pu isotopic composition can be 

maintained at a nearly constant value 

Potential for thorium fuel cycle 233 U breeding with low 232 U content 

– 232 U contamination (strong source of γ radiation) in a thorium-fuelled thermal 

reactor: 2000 - 4000 ppm, vs. 500 -1000 ppm in FR 

Utilization of fission products and other radioactive isotopes 

– Production of fission products having commercial value in the FR cycle, e.g. 

137 

Cs (~31% of total Cs yield); 90 Sr (~60% of total Sr yield; Pd (produced nearly 

free of radioactive contamination); Rh and Ru (both radioactive but decaying to 

acceptable levels within a few years); 212 Pb; 212 Bi; 223 Ra; 227 Ac 


Metals: Status And Perspectives, Pisa, Italy 17-20 April 2012 7

Fast Reactor Coolants 



Conclusion 

Sustainability 

and breeding 

recycle used fuel utilization 

fast reactor indispensable 

For long-term sustainable nuclear power closed fuel cycle 

with fast reactors will constitute the main nuclear energy 

system 



WORLDWIDE CLOSE TO 400 FR-YEARS CUMULATED OPERATION 

China 

– CEFR (23 MWe) July 2010 

India 

– FBTR (13 MWe) 1985 

– PFBR (500 MWe) 2011/12 

Japan 

– Joyo (140 MWth) 1977 

– Monju (280 MWe) 1994, 

May 2010 

Russia (USSR) 

– BR10 (8 MWth) 1958 – 2003 

– BOR60 (12 MWe) 1968 

– BN350 (130 MWe) 1972 – 99 

– BN600 (600 MWe) 1980 

– BN800 (870 MWe) ~2013 

EU (D, F, UK) 

– Rapsodie (40 MWth) 1967 – 83 

USA 

– DFR (15 MWe) 1959 – 77 

– KNK-II (20 MWe) 1972 – 91 

– Phénix (250 MWe) 1973 – 2010 

– PFR (250 MWe) 1974 – 94 

– SNR300 (300 MWe) not started 

– Superphénix (1200 MWe) 1986 – 98 

– EFR Proj. (1580 MWe), cancelled 98 

– EBR-I (a few 100s We) 1951 – 64 

– EBR-II (20 MWe) 1961 – 1998 

– FFTF (400 MWth) 1980 – 1996 

– CRBR Proj.(380 MWe), cancelled 83 



Phénix 

2010 shut down > 35 yrs 

Technology demonstrator 

and R&D support 

1999 – 2002 plant 

renovations and safety 

upgrades 720 EFPD 

lifetime extension to support 

French long-lived radioactive 

waste management 

programme 

2003 – 2009 satisfactory/better than satisfactory operation broke 

own grid connection time in 2007, operating cycle length, & electrical 

production run without shutdown records 

2009 – 2010 end-of-life tests at low and zero power (core physics, 

thermal hydraulics, fuel issues, investigations of the 1989/1990 negative 

reactivity transients) V&V&Q of FR data, codes, methodologies 



Superphénix 

7 Sep 1985 first criticality 

14 Jan 1986 first grid connection 

Storage drum sodium leak, first reactor shutdown period May 1987 – 

January 1989 

Air intake in argon circuit causing primary sodium oxidation beyond 

allowable limits, second reactor shutdown period July 1990 – July 1994 

20 April 1998 shutdown by government order 

8.3 TW el ×h produced 

53 months of normal operation (disregarding unplanned outages due to 

incidents) 

– 8 months start-up tests 

– 34 months of power production 

– 11 months of planned outages for refuelling, in-service inspection 

and maintenance 



Superphénix, cont’d 



FBTR 

In operation since Oct 1985 over 330 GWh (th)/10 GWh (el) 

Efforts to attain design thermal power target (40 MW for original MOX 

fuelled core) P th increased from 11 MW [initial (Pu-U)C core 

rating] to 19 MW [(Pu-U)C with lower Pu/U] 

165 GWd/t Mark I fuel peak BU, no failure 

Several system upgrades and modifications additional 12 full 

power years expected 

FBTR’s mission 

– Irradiation tests for Indian Pressurized Heavy Water Reactors 

(PHWRs) and Prototype Fast Breeder Reactor (PFBR) 

– Physics and engineering experiments V&V&Q of FR data, codes, 

methodologies 



14

Joyo 

2006 ANS Nuclear Historic Landmark Award" 

30 th anniversary on 24 April 2007 

Joyo’s mission sodium cooled FR development various 

irradiation tests and PIE successfully conducted 

During 15 th periodical inspection (2007) obstacle found in the in-vessel 

storage rack 

Restoration works: 

– Replacement of upper core structure 

– Retrieval of irradiation test device 

“MARICO-2” Development of 

sodium cooled FR in-vessel 

inspection and repair techniques 



MONJU 

First criticality April 1994 

8 December 1995 

sodium-leak incident in 

the sec. coolant system, 

shutdown 

8 May 2010 restart after 

more than 14 yrs and 

comprehensive reviews of 

the validity of fast breeder 

reactor development in 

Japan 

-Verification of entire plant safety and soundness 

-Plant modification work against sodium leakage 

-Improvement of the anti-seismic margins 



BOR-60 

1968: start of operation 

2009: second lifetime extension till 2014 

Small number of unscheduled shutdowns (1 -2 events per year) 

Bor-60 experimental and power reactor 

– Fuel and material irradiation tests 

– Safety, I&C, and sodium technology experimental programs 



BN-350 

750 MW th (200 MW el ) and water desalination 

First criticality 29 November 1972 

Power operation 16 July 1973 

Initially, derated operation (design thermal rating 1000 MW) due to 

unsatisfactory operation of the steam generators 

Excessive heat rating of the fuel rods and considerable shape variation 

of core structural components core modernization & new 

structural materials solved problems 

Availability of stand-by components (six loops) & ability to operate with 

3-5 loops ensured stable operation of the plant repair and 

replacement work on steam generators carried out without shutting 

down the reactor 

Shutdown June 1994 



BN-600 

26 February 1980 first criticality 

8 April 1980 grid connection 

December 1981 full power (600 MW el ) 

As of 31 December 2009 

– Power operation exceeded 206,000 hrs 

– Over 112.5 TW el ×h produced 

Lifetime extended to 31 March 2020 

>30 years successful operation 

Sodium cooled FR technology attained industrial maturity 

Replacement and repair of sodium components, including pumps and 

steam generators mastered 

Once personnel is trained to master specific features of sodium 

technology, these features have no significant influence on safety and 

operational parameters of the power plant 



19

EBR-II 

November 1963 first criticality 

August 1964 grid connection 

Demonstration sodium cooled FR power plant (62.5 MW th / 19 MW el ) 

Demonstrate closed fuel cycle for U alloy fuel 

Fuels and materials irradiation tests 

Safety experiments (decay heat removal, operation in transient 

overpower conditions, and run-beyond-cladding-breach modes) 

Integral Fast Reactor (IFR) test bed (U-Pu alloy fuel) 1984 – 1994 

Actinide irradiation tests of actinide burning 

Shutdown 1994 due to the termination of the DOE FR development 

program 



FFTF 

Sodium cooled loop-type experimental FR (400 MW th ) to develop and 

test advanced fuels and materials 

First criticality 9 February 1980 

Full power operation 21 December 1980 

Operated safely and successfully as an irradiation facility till late 

1992 

Driver assemblies routinely reached peak burnups of 90 – 100 

000 MWd/t with only one fuel failure out of ~40 000 fuel pins 

1987 – 1992 core demonstration test (6 blanket and 10 driver 

assemblies), cladding and wrapper made from low-swelling alloy HT- 

9, 238 000 MWd/t peak burnups without failure 

Endurance tests to obtain cladding breaches: pins easily identified 

(using EBR-II gas-tag method) 



21

FFTF, cont’d 

Plant tests to study passive safety characteristics of oxide-fuelled sodium 

cooled FR (static and dynamic tests over a broad range of power, flow 

and temperature conditions, including demonstration of stable operation 

at low power under natural circulation cooling conditions) 

Development and testing of passive safety enhancement device GEM 

(gas expansion module) to offset the excess reactivity in ULOF 

conditions 

GEM effectiveness demonstrated: primary pumps stopped at 50% power 

(200 MWt) normal control-rod scram response disabled GEMs 

and inherent core reactivity feedback effects took the reactor sub-critical 

with only a short 85°C rise in sodium outlet temperature 

Primary & secondary pumps operated for over 200’000 hrs, no major 

maintenance required 

IHX thermal performance unchanged from initial values 

Electromagnetic pumps required little maintenance except for one that 

developed a pin-hole leak (flow tube deformation due to sodium 

freeze/thaw cycles & cavitation/erosion at high sodium flow; promptly & 

easily replaced 



PFR 

250 MW el (600 MW th ), sodium cooled, MOX fuelled, stainless steel 

clads, mechanical centrifugal pumps, highly-rated tube-in-shell SGs 

3 March 1974 first criticality 

31 March 1994 final shutdown 

Operating history of PFR conveniently divided into two decades: 

- 1974 – 1984: electrical output limited, mainly because of a series of 

leaks in the tube-to-tubeplate welds of the SGs (12% highest load 

factor in 1978) 

- 1984 – 1994: SG weld problems solved, plant performance 

improvement (56.5% load factor in the final year of operation); 18- 

month outage in 1991/92 due to oil leakage from a primary pump 

bearing into primary sodium 



23

Looking Ahead: Fast Reactor R&D&D 

Renewed interest in nuclear energy (?) 

Sustainability ⇒ spent fuel utilization and breeding returning to centre stage 

⇒ fast reactor necessary linchpin 

Fast reactor deployment likely to be accelerated 

– Restart in May 2010 of the industrial prototype Monju (Japan) 

– Commissioning, at the time horizon 2011 – 2023, of power fast reactors 

in India (7 PFBRs, 3.5 GWe total capacity) 

– SVBR-100 (Pb-Bi cooled) commissioning planned ~2023 (joint venture 

between Rosatom and RUSAL) 

– Planned construction ~2020 of the prototype fast reactor ASTRID 

(France) 

– Construction projects in China, India, Russia, Japan and the Republic of 

Korea 

Experimental fast reactor construction projects 

– Belgian government approval for MYRRHA construction 

– Construction of MBIR replacing BOR-60 in Dimitrovgrad 



24

First Criticality of China’s 25 Mwe Experimental 

Fast Reactor (CEFR) 26 July 2010 



25

PFBR, Kalpakkam: Erection of Vessels 

Safety vessel into reactor vault 

Main vessel into safety vessel 

Inner vessel into main vessel 

Handling of SV by heavy duty crane 

Grid plate into main vessel 

Thermal baffle into main vessel 



PFBR, Kalpakkam: Erection of Roof Slab 



Russia’s BN-800, Beloyarsk Site, Sept 2008 

Commissioning Planned ~2013 



Requirements for Next (Advanced) 

LMFR Prototype (e.g. ASTRID) 

Sustainability 

– Efficient uranium utilization (closed fuel cycle) 

– Improve ultimate radioactive waste form (decay heat, radiotoxic inventory…) 

– Enhanced non-proliferation 

Robust safety demonstration matching at least Gen III safety criteria 

– Enhanced prevention of severe core damage 

– Practical exclusion of large energy release in hypothetical core meltdown 

– Contain the risk due to sodium’s chemical reactivity 

– Robustness to external hazards 

Economic competitiveness 

– Reduction of capital cost and investment risk to about Gen III level 

– Improved plant operability (ISI&R, target of 90% availability, …) 

– Long lifetime (60 years) 



29

NEW GENERATION OF SFR 

ASTRID (~2020) 

– 600 MWe 

– Test-bed for innovation 

(systems, technologies) 

– Capability for materials 

and fuel testing 

– Advanced recycle modes 

demonstrations 

(U/Pu, MA) 

– Improved U utilization 

– Improved waste form 



30

GEN IV SFR: JAEA SODIUM COOLED FAST 

REACTOR (JSFR) 

Power 

Parameter 

Specifications 

3570 MWt / 1500 MWe 

Number of loops 2 

Primary sodium outlet 

temperature and flow rate 

550 / 395 °C 

3.24 x 10 7 kg/h/loop 

Secondary 

pump 

SG 

Secondary sodium 

temperature and flow rate 

Main steam temperature 

and pressure 

Feed water temperature 

and flow rate 

520 / 335 ° C 

2.70 x 10 7 kg/h/loop 

497 °C 

19.2 MPa 

240 °C 

5.77 x 10 6 kg/h 

Plant efficiency ~42% 

Fuel type 

TRU-MOX 

Breeding ratio Break even (1.03), 1.1, 1.2 

Cycle length 

26 months or less, 4 batches 

Integrated 

Pump-IHX 

Reactor Vessel 

Reactor Core 

Courtesy H. Negishi, JAEA 



GEN IV SFRs: Small (SMFR) and 

Medium Size (KALIMER) 

SMFR 

KALIMER 

AHX 

Chimney 

PDRC 

piping 

IHTS 

piping 

Steam 

Generator 

IHX 

DHX 

PHTS 

pump 

Reactor 

core 

IHTS 

pump 

Source: Dohee Hahn, KAERI 

In-vessel core 

catcher 



GEN IV SFR Design Data 

Design Parameters JSFR KALIMER SMFR 

Power Rating, MWe 1,500 600 50 

Thermal Power, MWth 3,570 1,525 125 

Plant Efficiency, % 42 42 ~38 

Core outlet coolant temperature, o C 550 545 ~510 

Core inlet coolant temperature, o C 395 370 ~355 

Main steam temperature, o C 503 495 480 

Main steam Pressure, MPa 16.7 16.5 20 

Cycle length, years 1.5-2.2 1.5 30 

Fuel reload batch, batches 4 4 1 

Core Diameter, m 5.1 3.5 1.75 

Core Height, m 1.0 0.9 1.0 

Fuel Type 

MOX 

(TRU bearing) 

Metal 

(U-TRU-10%Zr 

Alloy) 

Metal 

(U-TRU- 

10%Zr Alloy) 

Cladding Material ODS HT9M HT9 

Pu enrichment (Pu/HM), % 13.8 14.9 15.0 

Burn-up, GWd/t 150 79 ~87 

Breeding ratio 1.0–1.2 1.0 1.0 

Source: Dohee Hahn, KAERI 



33

GEN IV LFR Reactor Concepts 

SSTAR (10 – 100 MWe) ELSY (600 MWe) 

CLOSURE HEAD 

CO2 OUTLET NOZZLE 

(1 OF 8) 

CO 2 INLET NOZZLE 

(1 OF 4) 

Pb-TO-CO 2 HEAT 

EXCHANGER (1 OF 4) 

CONTROL 

ROD 

DRIVES 

CONTROL 

ROD GUIDE 

TUBES AND 

DRIVELINES 

THERMAL 

BAFFLE 

FLOW SHROUD 

RADIAL REFLECTOR 

ACTIVE CORE AND 

FISSION GAS PLENUM 

FLOW DISTRIBUTOR 

HEAD 

GUARD 

VESSEL 

REACTOR 

VESSEL 

Source: A. Alemberti, AnsaldoNucleare 



34

MYRRHA - Accelerator Driven System 

Accelerator 

(600 MeV - 4 mA proton) 

Reactor 

• Subcritical or Critical modes 

• 65 to 100 MWth 

Spallation Source 

Multipurpose 

Flexible 

Irradiation 

Facility 

Source: Hamid-Aït Abderrahim, SCK˃CEN 

Fast 

Neutron 

Source 

Lead-Bismuth 

coolant 



MYRRHA Reactor Layout 




MYRRHA Reactor Layout (Core) 

• k eff ≈0.95 (ADS mode) 

• 30-35 % MOX fuel 

• 7 In-Pile Section (IPS) 

positions 

Spallation window 

IPS 


Fuel Assemblies 



37

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