SSTAR LEAD-COOLED, SMALL MODULAR FAST REACTOR …

A U.S. Department of EnergyOffice of Science LaboratoryOperated by The University of Chicago

Argonne National Laboratory

Office of ScienceU.S. Department of Energy

SSTAR LEAD-COOLED, SMALL MODULAR FAST REACTORWITH NITRIDE FUEL

byJ. J. Sienicki, A. V. Moisseytsev, P. A. Pfeiffer,

W. S. Yang, M. A. Smith, S. J. Kim.Y. D. Bodnar, D. C. Wade, and L. L. Leibowitz

Argonne National [email protected]

Workshop on Advanced Reactors with Innovative FuelsARWIF-2005Oak Ridge,

February 16-18,2005

mailto:[email protected]

Small Secure Transportable Autonomous Reactor (SSTAR)

• Mission - Electricity generation to match needs of developing nations and especially remote communities or industrial operations (e.g., mining) without major electrical grid connections- Alaska, Hawaii, island nations of the Pacific Basin (e.g., Indonesia),

and elsewhere- Possible niche market within which costs per unit energy that are

higher than those for large-scale nuclear power plants may be competitive

• “Report to Congress on Small Modular Reactors,” U.S. Department of Energy, May 2001- “In considering possible replacement power plants, it appears that

units less than 50 MWe would represent the majority of Alaskan generating capability, with units of 10 MWe or less being the most widely applicable.”

Nuclear Engineering Division

2

Pioneering Science andTechnology

Office of ScienceU.S. Department

of Energy


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• Current concept under investigation is 20 MWe (45 MWt)• Proliferation resistance

- Core lifetime/refueling interval of 20 years- Core is a single cassette and is not composed of individual

removable fuel assemblies- Restrict access to fuel during core lifetime- Refueling equipment present at site only during refueling- Transuranic fuel – Self protective in the safeguards sense

• Molten lead (Pb) primary coolant – Nitride fuel- Passive safety- Potential to operate at higher temperatures than traditional

LMRs• Fissile self-sufficiency

- Conversion ratio near unity- Realization of sustainable closed fuel cycle


• Autonomous operation- Core power adjusts itself to heat removal from reactor system

due to large inherent reactivity feedbacks of fast spectrum corewithout operator motion of control rods

- Active adjustment of shutdown rods for startup and shutdown, and compensation/control rods for burnup compensation

• Utilizes supercritical carbon dioxide (S-CO2) gas turbine Brayton cycle power converter- Higher plant efficiency than Rankine saturated steam cycle- Reduce balance of plant footprint, costs, and staffing

requirements• Natural circulation primary coolant heat transport

- Eliminate main coolant pumps and loss-of-flow accidents


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• Factory fabrication- All reactor and balance of plant components including reactor

and guard vessels- Reduced costs and improved quality control

• Factory assembly of components into transportable modules- Short modular installation and assembly times at site

• Full transportability by barge or rail, or possibly by road- Overland transport to remote sites

• Flexibility to be adapted to generate other energy products- Desalinated water- Hydrogen


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Illustration of Lead-Cooled Fast Reactor


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Pb Coolant


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• Enhanced passive safety- Chemically inert – Does not react chemically with CO2 working fluid

above ~ 250 °C. Does not react vigorously with air or water/steam- High boiling temperature of 1740 °C for Pb (1670 °C for Pb-Bi

eutectic) - Core and heat exchangers remain covered by ambient pressure single-phase primary coolant and single-phase natural circulation removes core power under all operational and postulated accident conditions

• Potential to operate at higher temperatures than traditional liquid metal-cooled fast reactors- Goal of peak cladding temperature of ~ 650 °C

• Two lead-bismuth eutectic (LBE)-cooled land prototypes and ten submarine reactors were operated in Russia providing about 80 reactor years experience- Development of coolant technology and control of structural material

corrosion

Nitride Fuel• Enhanced passive safety

- High melting temperature (> 2600 °C for UN)- High temperature for significant decomposition of nitride (> 1400 °C)- High thermal conductivity that together with Pb bond between fuel

and cladding reduces the fuel-coolant temperature difference• Compatible with fast neutron spectrum

- High atom density- Nitrogen is enriched in N15 to eliminate parasitic reactions in N14 and

waste disposal problems associated with C14 production• Compatible with ferritic/martensitic stainless steel cladding and

Pb coolant- Nitrogen is insoluble in Pb- Bonded to cladding by molten Pb

• Low irradiation-induced swelling and fission gas release


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Maximum Smeared Density for Nitride Fuel


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0.89

0.9

0.91

0.92

0.93

0.94

0.95

0.96

0.97

0.98

0.99

1

1.01

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16

PEAK BURN-UP (at %)

MA

XIM

UM

SM

EAR

DEN

SITY

Predicted maximumsmear densityLower Limit

Upper limit

• Smeared density of 90 % is selected for SSTAR for which the peak burnup = 12 atom %

Analysis of Nitride Fuel Thermal Decomposition


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• Calculate mass of each vapor species assuming ideal behavior, measured vapor pressures versus temperature, and fission gas plenum volume per fuel volume− Need to extrapolate vapor pressures outside of

measurement range• Assuming stoichiometry, vapor species mass is

subtracted from decomposed condensed phase mass• Fission gas plenum-to-active core height ratio = 3.5

− Conservative assumption based on F/M stainless steel cladding with HT9 mechanical behavior

• Fuel smeared density = 90 %• Ignore suppression of vaporization by nitrogen fill gas

Analysis of (U0.8, Pu0.2)N Thermal Decomposition


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Vapor Pressure of N2(g), U(g), and Pu(g) Over (U0.8, Pu0.2)N

1.0E-271.0E-251.0E-231.0E-211.0E-191.0E-171.0E-151.0E-131.0E-111.0E-091.0E-071.0E-051.0E-031.0E-01

600 850 1100 1350 1600 1850 2100 2350 2600 2850

TEMPERATURE, ºC

PRES

SURE

, at

mos

pher

es

N2(g)U(g)Pu(g)

Analysis of (U0.8, Pu0.2)N Thermal Decomposition


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Mass Fraction of Species from Thermal Decomposition of (U0.8, Pu0.2)N

1.0E-301.0E-281.0E-261.0E-241.0E-221.0E-201.0E-181.0E-161.0E-141.0E-121.0E-101.0E-081.0E-061.0E-041.0E-021.0E+001.0E+02

600 850 1100 1350 1600 1850 2100 2350 2600 2850TEMPERATURE, ºC

MAS

S FR

ACT

ION

U(g)U(l)Pu(g)Pu(l)N2(g)

20 MWe (45 MWt) – Neutronics Analysis• Assume 20-year core life, fixed fuel volume fraction of 0.55, fuel

smeared density of 90 %, and core height-to-diameter ratio of 0.8• Transuranic (TRU) fuel feed from LWR spent fuel following 25-year

cooling time - Allows for decay of Pu241 isotope• Low enrichment central region to reduce burnup reactivity swing• Seek to minimize burnup reactivity swing to less than one dollar• Fuel volume fraction of 0.55 is low enough to facilitate natural

circulation heat transport from core to Pb-to-CO2 heat exchangers• Burnup reactivity swing exhibits minimum at an active core

diameter of about 1.0 m• For this diameter, maximum power limited to 45 to 50 MWt by peak

fast neutron fluence limit of 4.0 x 1023 n/cm2 on ferritic-martensiticstainless steel


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Average Discharge Burnup and BurnupReactivity Swing versus Core Size


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0

50000

100000

150000

200000

250000

300000

0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4

Core Diameter (m)

Bur

nup

(MW

D/M

T)

-40.00

-35.00

-30.00

-25.00

-20.00

-15.00

-10.00

-5.00

0.00

5.00

Bur

nup

Swin

g (%

) {da

shed

}

25 MWt30 MWt35 MWt40 MWt45 MWt50 MWt25 MWt30 MWt35 MWt40 MWt45 MWt50 MWt

Average Discharge Burnup and Peak Fluenceversus Core Size


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0

25000

50000

75000

100000

125000

150000

175000

200000

225000

250000

275000

0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4

Core Diameter (m)

Bur

nup

(MW

D/M

T)

0.00

1.00

2.00

3.00

4.00

5.00

6.00

7.00

8.00

9.00

10.00

11.00

12.00

13.00

Peak

Flu

ence

(1E2

3) {d

ashe

d}

25 MWt30 MWt35 MWt40 MWt45 MWt50 MWt25 MWt30 MWt35 MWt40 MWt45 MWt50 MWt

45 MWt SSTAR Neutronics Conditions

Core Diameter, m 1.02Active Core Height, cm 80Fuel Smear Density, % 90Fuel Volume Fraction 0.55Cladding Volume Fraction 0.16Bond Volume Fraction 0.10Coolant Volume Fraction 0.18


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45 MWt SSTAR Neutronics Performance

Average Power Density, W/cm3 69Specific Power, KW/KgHM 10Peak Power Density, W/cm3 119Average Discharge Burnup, MWd/Kg 72Peak Discharge Burnup, MWd/Kg 120Peak Fast Fluence, 1023 n/cm2 3.6BOC to EOC Burnup Swing, % delta rho 0.13Maximum Burnup Swing, % delta rho 0.36BOC to EOC Burnup Swing, $ 0.35Maximum Burnup Swing, $ 0.96


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45 MWt SSTAR Reactivity Feedback Coefficients

BOC POC~ 13 years

EOC

Delayed Neutron Fraction 0.0035 0.0034 0.0034

Prompt Neutron Lifetime, s 1.8 X 10-07 1.8 X 10-07 1.8 X 10-07

Coolant Density, cents/°C -0.002 0.003 0.002

Core Radial Expansion, cents/°C

-0.16 -0.16 -0.16

Axial Expansion, cents/°C -0.06 -0.06 -0.06

Fuel Doppler, cents/°C -0.12 -0.12 -0.11

Coolant Void Worth, $ -0.99 -0.45 -0.71


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Fuel Pin Cladding Mechanical Analyses• LFR coolant technology database indicates that Si-

enhanced ferritic/martensitic stainless steel might resist attack by flowing molten Pb up to temperatures of about 650 C

• SSTAR goal is peak cladding temperature = 650 C• HT9 stainless steel taken as representative of Si-

enhanced F/M SSt for mechanical behavior• Assume HT9 mechanical properties to investigate

viability for application to SSTAR cladding


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Assumed Design Criteria for HT9 Cladding• Thermal component of plastic hoop strain, εTHN <0.01 = 1.0 %

- Thermal creep is what damages fuel pin cladding- Thermal creep strain consists of primary, steady state, and tertiary

components• Cumulative damage function,

- Value of 0.05 is specified to limit fuel pin failures in a statistical sense

• Radially averaged hoop stress, σH < 150 MPa- Purpose is to preclude unstable plastic deformation- Strain rates increase rapidly when cladding is subjected to stresses

above this level

0.05

t

orupture

dtCDF t= <∫


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Fuel Pin Cladding Mechanical Analyses• Analyze most severely stressed pin over 20-year core

lifetime• Location of power peak/hot channel varies with time

– Moves from outer part of core to core center• Assume hot channel pin at EOC has average channel

conditions for first 90 % of core life followed by hot channel conditions for last 10 % of core life

• Assume peak cladding temperature at middle of wall thickness equals 589 C for first 18 years and 638 C for final 2 years

• Assume that pressure loading from fission gas release increases linearly with time to final value


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Fuel Pin Cladding Mechanical Analyses• Total thermal creep strain criterion is most restrictive and limits

cladding hoop stress to 15 MPa


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TOTAL THERMAL CREEP STRAIN

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

1.E+09

1.E+10

0 20 40 60 80 100 120 140 160

EQUIVALENT STRESS, MPa

TH

ERM

AL

CR

EEP

STR

AIN

, %

Selection of Fission Gas Plenum Height• 1.0 % thermal strain requirement limits cladding hoop stress to ~

15 MPa for peak cladding temperature criterion of 650 C• To limit hoop stress to such a small value requires a large fission

gas plenum volume/height of ~ 2.4 m or greater• Fission gas plenum pressure depends upon gas release from

nitride fuel• Fraction of gas released assumed given by correlation of

Rogozkin, Steppennova, and Proshkin (Atomic Energy, Vol. 95, No. 3, p. 624, 2003)- F=3.05 B1.92 exp (-4130/RTfuel)- B = burnup in atom %, R = 1.98 cal/(mol-K), Tfuel = fuel centerline

temperature• For Tfuel = 953 C, release fraction, F = 0.28


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Selection of Fission Gas Plenum Height• Fuel pin cladding hoop stress versus fission gas plenum height for

28 % fission gas release and 595 C hot channel outlet temperature

Plenum Height, m Internal Pressure, MPa

Cladding Hoop Stress, MPa

0.8 3.23 38.71.0 2.58 31.01.6 1.61 19.42.8 0.92 11.13.2 0.81 9.683.6 0.12 8.61


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• Fission gas plenum height of 2.8 m selected for SSTAR


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Time Dependent Thermal Creep Strain for Hot Channel Fuel Pin Cladding at End-of-Core Life

• Total creep strain limited to 0.12 % but large creep rates are calculated during final 10 % of lifetime – Undesirable regime

THERMAL CREEP

-0.02

0

0.02

0.04

0.06

0.08

0.1

0.12

0.14

0.16

0.18

0.2

0 5 10 15 20

TIME, years

CREE

P ST

RAIN

, %

Thermal CreepSteady-StateTertiaryPrimary

Possible Solution – Oxide Dispersion Strengthened (ODS) Cladding Material

• ODS materials offer prospect of greater strength at temperatures above 600 C

• Examples (Klueh et. al., Journal of Nuclear Materials, Vol. 307-311, p. 773, 2002)- Fe-12Cr-2.5W-0.35Ti-0.25Y2O3 (Designated

12YWT)- Fe-12Cr-0.25Y2O3 (Designated 12Y1)

• Assumption of 12YWT mechanical properties results in CDF of 1 X 10-4 for cladding hoop stress of 150 MPa- Would enable short height fission gas plenum (e.g.,

one-fourth of active core height)


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Comparison of ODS Materials


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• Cumulative damage function for hot channel fuel pin cladding at EOC

CUMULATIVE DAMAGE FUNCTION

1.E-211.E-201.E-191.E-181.E-171.E-161.E-151.E-141.E-131.E-121.E-111.E-101.E-091.E-081.E-071.E-061.E-051.E-041.E-031.E-021.E-011.E+001.E+011.E+021.E+03

0 20 40 60 80 100 120 140 160

HOOP STRESS, MPa

CU

MU

LATI

VE D

AM

AG

E FU

NCTIO

N

HT9

9Cr-WMoVNb

12Y1

MA956/957

12YWT

System Thermal Hydraulic Analyses - What Determines the Reactor Vessel Size?


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• Transportability by rail and possibly by road assumed as a goal- Assumed size limitations are 18.9 m (62 feet) length by 6.1 m width

(20 feet)• Need to fit core and other components inside of vessel diameter

- 1.02 m active core diameter- 0.297 m reflector thickness- 2.54 cm core shroud thickness interior to downcomer- 5.72 cm thick gap between reactor vessel inner surface and 1.27

cm thick cylindrical liner to provide escape path to Pb free surface for CO2 void, in the event of HX tube rupture

- 5.08 cm thick reactor vessel- Kidney-shaped Pb-to-CO2 heat exchangers must fit inside of

annulus between shroud and cylindrical liner, and provide sufficient heat exchange performance to realize a significant Brayton cycle efficiency

Optimization of Fuel Pin Diameter and HX Tube Dimensions

• Determine optimal fuel pin diameter that minimizes the peak cladding temperature for a fixed fuel volume fraction of 0.55, fuel smeared density of 90 %, and fixed Pb core inlet temperature- Assume fixed cladding thickness = 1.0 mm

• A unique relation exists between the fuel pin diameter and the triangular pitch-to-diameter ratio for fixed fuel volume fraction and smeared density

• Optimal fuel pin diameter determined to be 2.5 cm• Fuel pin pitch-to-diameter ratio = 1.121• Determine HX tube height and pitch-to-diameter ratio that

maximize S-CO2 Brayton cycle efficiency for fixed tube diameter and thickness

• HX tube height determined to be 6 m and p/d = 1.242• Core inlet temperature = 420 °C provides PCT = 650 °C


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Relationship Between Fuel Pin Diameter and Pitch-to-Diameter Ratio


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PITCH-TO-DIAMETER RATIO AND HYDRAULIC DIAMETER VERSUS FUEL PIN DIAMETER

(FVF = 0.55; rhosmeared = 0.90)

1.001.011.021.031.041.051.061.071.081.091.101.111.121.131.141.151.16

1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0 3.2 3.4 3.6

FUEL PIN DIAMETER, cm

PITC

H-T

O-D

IAM

ETER

RA

TIO

0.00.10.20.30.40.50.60.70.80.91.01.11.21.31.41.51.6

CO

RE

HYD

RA

ULI

C D

IAM

ETER

,cm


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Optimization of Fuel Pin Diameter• Select fuel pin diameter = 2.5 cm that minimizes the peak cladding

temperature

PEAK CLADDING TEMPERATURE

590

2 2.5 3 3.5

FUEL PIN DIAMETER, cm

TEM

PER

AT

UR

E, o

C

375400425

Tinlet, C

HX:

L = 6 m

p/d = 1.3

Selection of Core Inlet Temperature• Select core inlet temperature = 420 °C that provides a peak

cladding temperature of 650 °C


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PEAK CLADDING TEMPERATURE

400

450

500

550

600

650

700

750

200 250 300 350 400 450

CORE INLET TEMPERATURE, oC

PEAK

CLA

DD

ING

TEM

P, o

C

22.533.5

FUEL PIN OD, cm

HX:

L = 6 m

p/d = 1.3

Peak Nitride Fuel Centerline Temperature• Peak fuel temperature = 953 °C


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PEAK FUEL TEMPERATURE

600

700

800

900

1000

1100

1200

1300

1400

200 250 300 350 400 450

CORE INLET TEMPERATURE, oC

PEA

K FU

EL T

EMPE

RA

TUR

E, o

C

22.533.5

FUEL PIN OD, cm

HX:

L = 6 m

p/d = 1.3

Schematic of SSTAR Coupled to S-CO2 Brayton Cycle Showing Nominal Conditions

100 % Power

552.7 31.2 19.8CO2 19.96 242.1 Kg/s 435.8

7.548402.119.96

Pb 180.61 atm Eff = 44.4 % 19.99565.8 44.9 Net = 44.0 % 188.3

19.99 186.8Air 7.463RVACS 176.8

418.7 0.1 T, C T,C 5.7 19.99Q,MW P,MPa

420.245

31.25 85.2 88.87.400 20.00 7.409

Ave Peak420.0 420.0 495.0 625.4

541.5 650.0 4.8565.9 682.3 0.13 30.0 35.8

2125 Kg/s 672.0 953 Kg/s 0.218 0.10167%

33%

SSTAR TEMPERATURES AND PRESSURES

CORE temperatures

Coolant

24.3

Fuel

28.4

68.5

CladdingBond 1,000

TURBINEHTR

CORE

RHX

REACTOR VESSEL

LTR

COMP. #1

COMP. #2

COOLER


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Autonomous Load Following Behavior• System temperatures following an autonomous power change

from nominal power to a new steady state at End-of-Life (EOC)CORE INLET/OUTLET AND PEAK CLADDING AND FUEL

TEMPERATURES (EOC)

350400450500550600650700750800850900950

100010501100

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2

FRACTION OF NOMINAL CORE POWER

TEM

PER

ATU

RE,

oC

T fuel

T clad

T out

T in


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Autonomous Load Following Behavior• Comparison of contributions from individual reactivity feedbacks

REACTIVITY FEEDBACKS (EOC)

-0.15

-0.1

-0.05

0

0.05

0.1

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2

FRACTION OF NOMINAL CORE POWER

REA

CTIV

ITY

, do

llars

Coolant Radial exp.

Axial exp. Doppler

Control rods Total

20 MWe (45 MWt) SSTAR Operating Conditions


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• Power 19.8 MWe (45 MWt)• Reactor vessel height 18.3 m (60.0 feet)• Reactor vessel outer diameter 3.23 m (10.6 feet)• Active core diameter 1.02 m (3.35 feet)• Active core height 0.80 m (2.62 feet)• Active core height-to-diameter ratio 0.8• Fuel volume fraction 0.55• Fuel smeared density 90 %• Fuel pin outer diameter 2.5 cm• Fuel pin pitch-to-diameter ratio 1.121• Core hydraulic diameter 0.964 cm• Fuel pin cladding material HT9• Cladding thickness 1.0 mm• Fission gas plenum height 2.8 m


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20 MWe (45 MWt) SSTAR Operating Conditions• HX tube height 6.0 m• HX tube outer diameter 1.4 cm• HX tube inner diameter 1.0 cm• HX tube pitch-to-diameter ratio 1.242• HX hydraulic diameter for Pb flow 0.983 cm• HX-core thermal centers separation height 12.2 m• Peak cladding temperature 650 C• Core outlet temperature 566 C• Maximum S-CO2 temperature 553 C• Core inlet temperature 420 C• Core coolant velocity 0.896 m/s• Pb coolant flowrate 2125 Kg/s• CO2 flowrate 242 Kg/s• Brayton cycle efficiency 44.4 %• Net plant efficiency 44.0 %


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Summary of SSTAR• Results of preconceptual core neutronics, fuel pin

cladding mechanical, and system thermal hydraulics calculations indicate that a single-phase natural circulation SSTAR small modular fast reactor concept with a 20-year core lifetime, good core reactor physics performance, good system thermal hydraulics performance, and a high S-CO2 gas turbine Braytoncycle efficiency of 44 % may be viable at an electrical power level of 20 MWe (45 MWt)- Average discharge burnup = 72 MWd/Kg HM- Maximum burnup reactivity swing during 20 year core

lifetime is less than one dollar- Mean core temperature rise is 146 C while the peak

cladding structural temperature is limited to 650 C

Summary of SSTAR


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• Key contributors to achievement of SSTAR goals are features of transuranic nitride fuel- High atom density- Low swelling enabling 90 % smeared density- Low fission gas release reducing fission gas plenum

pressurization- High melting and thermal decomposition temperatures

enabling high system temperatures including 650 C peak cladding temperature

- Compatibility with F/M stainless steel cladding and Pbbond

- High thermal conductivity and Pb bond limiting peak fuel centerline temperature to 953 C


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Summary of SSTAR• F/M stainless steel fuel pin cladding similar to HT9 has

insufficient strength above ~ 600 C• For F/M SSt with HT9 properties and 650 C peak

cladding temperature, the total thermal creep strain can be limited to less than 1.0 % but this requires significant enlargement of the fission gas plenum height or thicker cladding wall to accommodate thermal creep over 20 years resulting from internal fission gas pressurization

• High HT9 creep rates at end of core life are undesirable• Development of fuel pin cladding materials and

structural materials having greater strength and resistance to Pb attack at high temperatures is indicated

Summary of SSTAR• One path forward may be oxide dispersion

strengthened (ODS) Si-enhanced F/M SSt –Will require development, compatibility testing with flowing Pb, and code cases

• Steady state and transient irradiations of nitride fuel with selected cladding and Pb bonding including high burnups will be required


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