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Status of development of the Small Secure Transportable Autonomous Reactor (SSTAR) for worldwide sustainable nuclear energy supply

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  • Advanced Reactor Concepts

Abstract and Figures

The STAR is a viable concept for a small fast reactor converter for international deployment in partner states providing proliferation resistance, fissile self-sufficiency, autonomous load following, simplicity of operation and reliability, transportability, as well as a high level of passive safety. SSTAR combines primary circuit natural circulation heat transport, lead (Pb) primary coolant, and transuranic nitride fuel in a pool vessel configuration inside of a small shippable reactor vessel. Conversion of the core thermal energy into electricity is accomplished with a supercritical carbon dioxide Brayton cycle power converter; a goal has been to operate at peak cladding temperatures as high as 650°C at which the core outlet temperature is 567°C and the net plant efficiency is 44 % to take advantage of the efficiency benefits of the supercritical CO2 Brayton cycle with increasing turbine inlet temperature.
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Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
Status of Development of the Small Secure Transportable Autonomous
Reactor (SSTAR) for Worldwide Sustainable Nuclear Energy Supply
James J. Sienicki, Anton Moisseytsev, David C. Wade, and Anna Nikiforova1
Argonne National Laboratory
9700 South Cass Avenue, Argonne, Illinois 60439, USA
Tel: (630) 252-4848, Fax: (630) 252-4780, Email: sienicki@anl.gov
1Massachusetts Institute of Technology
Abstract –.The STAR is a viable concept for a small fast reactor converter for international
deployment in partner states providing proliferation resistance, fissile self-sufficiency, autonomous
load following, simplicity of operation and reliability, transportability, as well as a high level of
passive safety. SSTAR combines primary circuit natural circulation heat transport, lead (Pb)
primary coolant, and transuranic nitride fuel in a pool vessel configuration inside of a small
shippable reactor vessel. Conversion of the core thermal energy into electricity is accomplished
with a supercritical carbon dioxide Brayton cycle power converter; a goal has been to operate at
peak cladding temperatures as high as 650
°
C at which the core outlet temperature is 567
°
C and
the net plant efficiency is 44 % to take advantage of the efficiency benefits of the supercritical
CO2 Brayton cycle with increasing turbine inlet temperature.
I. INTRODUCTION
The Small Secure Transportable Autonomous Reactor
(SSTAR) here has been developed under the U.S.
Department of Energy Generation IV Nuclear Energy
Systems Initiative as a small reactor for international
deployment in partner states (i.e., non-fuel cycle states)
incorporating concepts that are reflected in the Global
Nuclear Energy Partnership (GNEP). The user-supplier
paradigm has been part of the Lead-Cooled Fast Reactor
(LFR) work in the U.S. since its inception; the focus has
been on Secure Transportable Autonomous Reactor
(STAR) fast reactor converters (Conversion Ratio ~ 1)
suitable for deployment in developing nations. STARs
offer an alternative approach to actinide management by
“storing” actinides in long core lifetime (e.g., 15- to 30-
year) fissile self-sufficient operating power reactors. Thus,
instead of burning minor actinides (MAs) in Advanced
Burner Reactors, the MAs are incorporated into a
comparable number of STARs which return the fissile
resources at the end of the core lifetime. Small fast reactor
converters for international deployment in partner states for
meeting future global energy demands have thus been part
of the U.S. LFR work since its inception.
The focus of LFR work under Generation IV has been
the assessment of viability and development of a pre-
conceptual design for the SSTAR LFR concept. SSTAR is
a small (20 MWe/45 MWt) natural circulation fast reactor
plant for international deployment incorporating
proliferation resistance for deployment in partner states,
fissile self-sufficiency for efficient utilization of uranium
resources, autonomous load following suitable for small or
immature electric grids, and a high degree of passive safety.
SSTAR has been developed for the future global energy
economy recognizing the desire to meet projected
worldwide energy demands during this century (e.g., 1000
to 2000 GWe by 2050) in a sustainable manner while
maintaining CO2 emissions at or below today’s level
without using up the world’s resources of fissile materials
(e.g., known plus speculative virgin uranium resources = 15
million tones). To meet such energy demands, fast reactors
shall need to be massively deployed in partner states.
Significant progress and improvements have been made on
development of a pre-conceptual design of SSTAR since it
was last reported on.1-3
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
II. DESCRIPTION OF SSTAR
SSTAR has been developed as a modular fast reactor
concept at the low end of the power spectrum at 19.8 MWe
(45 MWt) which can meet the electricity requirements of a
town with a population of ~ 25400. The SSTAR concept
can be scaled up to 181 MWe (400 MWt) for efficient
electricity production for a city of ~ 225000 (STAR-LM)
with optional potable water production using a portion of
the reject heat. Customers of SSTAR include: i) clients
looking for energy security at small capital outlay; ii) cities
in developing nations; and iii) deregulated independent
power producers in developed nations. The SSTAR pre-
conceptual design integrates three major features in a pool
vessel configuration: primary coolant natural circulation
heat transport; lead (Pb) coolant; and transuranic nitride
fuel.
The high Pb boiling temperature (1740 °C) exceeding
the temperatures at which traditional cladding and
structural materials lose their strength and melt offers a
pathway to reactor operation at higher temperatures not
limited by coolant boiling at which the benefits of more
efficient power conversion using the supercritical carbon
dioxide (S-CO2) Brayton cycle can be realized provided
that suitable cladding and structural materials are
developed for service in Pb at the system temperatures.
Lead has a low neutron absorption (e.g., relative to sodium)
permitting the core to be opened up by increasing the
coolant volume fraction without significant reactivity
penalty decreasing the core frictional pressure drop making
natural circulation more effective. Shippable LFRs with
significant power (181 MWe/400 MWt) can be designed
with natural circulation at power levels exceeding 100 %
nominal. Natural circulation eliminates main coolant
pumps and loss-of-flow accident initiators. The high Pb
density (ρPb = 10400 Kg/m3) limits void growth and
downward penetration following postulated in-vessel heat
exchanger (HX) tube rupture eliminating significant void
transport to the core4 such that HXs can be located inside
of the reactor vessel eliminating the need for an
intermediate coolant circuit between the primary coolant
and working fluid.
Lead is calculated not to react chemically with the CO2
working fluid above ~ 250 °C which is less than the 327 °C
Pb freezing temperature. Lead does not react vigorously
with air. Lead is also expected not to react vigorously with
water or steam making it possible to eliminate the
intermediate circuit when a Rankine steam cycle is utilized
as the power converter.
Experiments carried out at the Forschungszentrum
Karlsruhe have shown that iodine, cesium, and cesium-
iodide (i.e., fission products with low melting and boiling
points) are absorbed and immobilized by lead-bismuth
eutectic at temperatures of 400 and 600 °C. Cesium forms
inter-metallic compounds in LBE while iodine forms PbI2.
Transuranic nitride fuel offers a number of potential
benefits provided that it can be demonstrated to perform
suitably well in steady state and transient irradiation tests
and can be reliably manufactured to meet the performance
requirements. The high transuranic nitride fuel
decomposition temperature (estimated > 1350 °C) would
enable the fuel to maintain its integrity at temperatures
above those at which cladding and structural materials lose
their strength and melt further supporting the pathway to
operation at higher temperatures. The high transuranic
nitride atom density enables the fuel volume fraction to be
further reduced further contributing to increasing the
cooling volume fraction facilitating natural circulation.
Transuranic nitride has the potential for low volumetric
swelling and fission gas release per unit burnup allowing
the gap size between fuel pellets and cladding to be
reduced and reducing thermal creep of the cladding
resulting from internal pressurization by released fission
gas. Nitride is compatible with Pb coolant and bond as
well as ferritic/martensitic (F/M) steel cladding. The fast
neutron spectrum core with Pb and transuranic nitride has
strong reactivity feedbacks enabling autonomous load
following, inherent power shutdown to match the heat
removal from the reactor, and enabling core designs having
low burnup reactivity swing over long time.
Use of Pb coolant also introduces a number of
potential negatives that need to be considered. The high
coolant density requires greater vessel thicknesses and
costs to support the coolant weight and accommodate
seismic loadings. The effective reactor power level may be
limited by the need to accommodate seismic loads.
Reliance upon natural circulation may result in a taller
reactor vessel and internal components adding to cost.
There is a need to supply non-nuclear heating to assure that
the coolant doesn’t freeze. There is a need to deal with the
effects of accidental freezing of coolant and thawing of
frozen coolant. Although heavy liquid metal coolant
dissolves Fe, Cr, and Ni from unprotected steels at rates
which increase with temperature, active maintenance of the
dissolved oxygen potential in the coolant within a proper
concentration window has been well established as a means
for forming protective oxide layers (Fe3O4 at temperatures
below ~ 570 °C) significantly retarding the dissolution rate
and avoiding the formation of solid PbO particulate.5
Figure 1 illustrates SSTAR which is a pool-type
reactor; the Pb coolant is contained inside of a reactor
vessel surrounded by a guard vessel. Lead is chosen as the
coolant rather than lead-bismuth eutectic (LBE) to reduce
the amount of alpha-emitting 210Po isotope formed in the
coolant by two to three orders of magnitude relative to
LBE, and to eliminate dependency upon bismuth which
might be a limited resource. The Pb coolant flows through
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
a perforated flow distributor head located beneath the core;
this structure provides an essentially uniform pressure
boundary condition at the inlet to the core. The Pb flows
upward through the core and a chimney above the core
formed by a cylindrical shroud (Figs 1 and 2). The vessel
has a height-to-diameter ratio large enough to facilitate
natural circulation heat removal at all power levels up to
and exceeding 100 % nominal. The coolant flows through
openings near the top of the shroud and enters four modular
Pb-to-CO2 HXs located in the annulus between the reactor
vessel and the cylindrical shroud. Inside each HX (Figs. 3
and 4), the Pb flows downwards over the exterior of tubes
through which the CO2 flows upwards. The CO2 enters
each HX through a top entry nozzle which delivers the CO2
to a lower plenum region in which the CO2 enters each of
the vertical tubes. The CO2 is collected in an upper plenum
and exits the HX through two smaller diameter top entry
nozzles. The Pb exits the HXs and flows downward
through the annular downcomer to enter the flow openings
in the flow distributor head beneath the core.
SSTAR does not incorporate an intermediate heat
transport circuit. This is a simplification possible with Pb
coolant which is calculated not to react with the CO2
working fluid below ~ 250 °C. A safety grade passive
pressure relief system is provided on the reactor system to
vent CO2 from the reactor in the event of a heat exchanger
tube rupture. A deliberate escape channel located between
the HXs and the reactor vessel/thermal baffle enables CO2
to rise benignly to the Pb free surface.
Fig. 1. Small Secure Transportable Autonomous Reactor
(SSTAR).
Fig. 2. SSTAR Reactor and Internals.
Fig. 3. SSTAR Pb-to-CO2 Heat Exchanger.
Fig. 4. Tubesheet of SSTAR Pb-to-CO2 Heat Exchanger.
CLOSURE HEAD
CO
2
INLET NOZZLE
(1 OF 4)
CO2 OUTLET NOZZLE
(1 OF 8)
Pb-TO-CO
2
HEAT
EXCHANGER (1 OF 4)
ACTIVE CORE AND
FISSION GAS PLENUM
RADIAL REFLECTOR
FLOW DISTRIBUTOR
HEAD
FLOW SHROUD GUARD
VESSEL
REACTOR
VESSEL
CONTROL
ROD
DRIVES
CONTROL
ROD GUIDE
TUBES AND
DRIVELINES
THERMAL
BAFFLE
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
A thermal baffle is provided in the vicinity of the Pb
free surface. As shown in Figs. 5 and 6, the baffle consists
of a cylindrical shell that is welded to the reactor vessel at
the bottom of the baffle and extends upward to nearly the
bottom of the upper closure head. The annular space
between the baffle and reactor vessel is filled with argon
cover gas to thermally insulate the reactor vessel.
Horizontal plates containing hole perforations are provided
to maintain spacing between the baffle and the reactor
vessel. The insulating effect is necessary to protect the
reactor vessel from thermal stresses that would otherwise
result from exposure to the Pb during startup and shutdown
transients when the Pb temperature varies between the cold
shutdown temperature of 420 °C and the nominal core
outlet temperature of 567 °C.
The SSTAR core (Fig. 7) is an open lattice of large
diameter (2.5 cm OD) fuel pins on a triangular pitch (p/d =
1.185) that does not consist of removable fuel assemblies.
The fuel pins are permanently attached to an underlying
support plate. This configuration restricts access to fuel
and eliminates fuel assembly blockage accident initiators.
The compact active core which is 1.22 m in diameter by
0.976 m in height is removed as a single cassette during
refueling and replaced by a fresh core. The active core
diameter is selected to minimize the burnup reactivity
swing over the 30-year core lifetime. The power level of
45 MWt is conservatively chosen to limit the peak fluence
on the cladding to 4 Χ 1023 neutrons/cm2 which is the
maximum exposure for which HT9 cladding has been
irradiated. The core (Fig. 7) incorporates two low
enrichment zones in the core central region which helps to
reduce the burnup reactivity swing. Three driver
enrichment zones reduce the peak-to-average power.
Primary and secondary sets of control rods are uniformly
dispersed throughout the active core. The radial reflector
consists of an annular “box” containing 50 volume %
stainless steel rods and 50 volume % Pb with a small
flowrate of Pb to remove the small power deposition in the
reflector. Stainless steel is necessary to shield the reactor
vessel from neutrons reducing the fluence at the reactor
vessel.
The cladding incorporates Si-enhanced steel to retard
the oxidation rate of the cladding resulting from exposure
to dissolved oxygen in the coolant. A promising approach
for the cladding may be weldment of a surface layer of Si-
enhanced steel upon F/M steel to form a layered billet
which is co-extruded.6 The Si-enhanced layer provides
improved corrosion resistance but has poor irradiation
stability. The substrate provides structural strength and
irradiation stability. The fuel pellets are bonded to the
cladding by molten Pb to reduce the temperature difference
between the pellet outer surface and cladding inner surface.
THERMAL
BAFFLE
REACTOR
VESSEL
GUARD
VESSEL
Fig. 5. SSTAR Cover Gas-Filled Thermal Baffle in Pb Free
Surface Region.
Fig. 6. Detail of Thermal Baffle.
THERMAL
BAFFLE
REACTOR
VESSEL
GUARD
VESSEL
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
Fig. 7. SSTAR 30-Year Core with All Fuel Pins Are Shown
Primary Control Rod Locations Are Shown in Magenta and
Secondary Control Rod Locations in Blue.
The fuel pins are held by two levels of grid spacers
similar to the approach followed in Light Water Reactor
(LWR) fuel assemblies. The grid spacer configuration is
shown in Figs. 8 and 9. A grid spacer is welded to a control
rod guide tube; the grid spacer holds the surrounding fuel
pins by means of spring clips allowing for thermal
expansion of the fuel pins relative to the control rod guide
tube.
Fig. 8. SSTAR Fuel Pins Are Held by Grid Spacers Welded to
Control Rod Guide Tubes.
GRID SPACER CELL WITH
SPRING CLIPS
Fig. 9. Illustration of Grid Spacer Cell.
Two sets of control rods are provided for independence
and redundancy of scram. The control rods incorporate
boron carbide absorber with enhancement of the isotope,
10B. Small adjustments of the control rods are carried out
to compensate for small changes in the burnup reactivity
swing. Each control rod moves inside of a control rod
guide tube occupying a position in the triangular lattice
(Figure 10). Control rod drivelines are shrouded by guide
tubes above the core (distinct from the control rod guide
tubes inside the core) and are connected to control rod
drives located above the upper closure head (Figures 11
and 12). For a realistically sized drive, the uniform
distribution of control rods throughout the core lattice
results in a dense packing of the control rod drives above
the upper closure head as illustrated in Figures 11 and 12.
This dense packing is a consequence of providing a
separate drive for each individual control rod. An
alternative approach would involve clustering rods together
with each cluster moved by means of a single drive; this
approach would decrease the number of drives.
Fig. 10. SSTAR Control Rod, Control Rod Driveline, and Control
Rod Driveline Guide Tube.
STAINLESS STEEL PINS OF
RADIAL REFLECTOR
(SST AND Pb)
TWO INDEPENDENT GROUPS
OF CONTROL RODS
LOW ENRICHMENT CENTRAL
REGION (TWO
ENRICHMENT ZONES)
DRIVER (THREE ENRICHMENT
ZONES)
CONTROL ROD
GUIDE TUBE
FUEL PIN
CONTROL ROD
DRIVELINE
GUIDE TUBE
BORON
CARBIDE
ABSORBER
CONTROL
ROD GUIDE
TUBE IN
CORE
REGION
CONTROL ROD
DRIVELINE
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
Fig. 11. SSTAR Core, Reflector, Control Rod Drivelines, and
Control Rod Drives.
Fig. 12. SSTAR Upper Closure Head Region.
Normally, refueling equipment is not present on the
site. Refueling equipment including a crawler crane is
brought onsite only following the 30-year core lifetime.
The control rods are inserted into the core, “locked” to
prevent their withdrawal, and detached from the control rod
drivelines. The upper closure head for the guard and
reactor vessels is removed together with the control rod
drives and drivelines, the spent core is removed from the
vessel, the core is placed inside of a shipping cask and
transported to a fuel cycle support center located in a fuel
cycle state. A fresh core is installed in the reactor vessel,
the closure head is re-installed, drivelines are attached to
control rods, and the refueling equipment is removed from
the site. Additional proliferation resistance features of
SSTAR are discussed in Reference 7.
SSTAR incorporates two independent and redundant
safety grade active shutdown systems; Fig. 7 shows the
primary and secondary control rod locations. The low
burnup reactivity swing of the 30-year lifetime core
decreases the excess reactivity requirements reducing the
amount of reactivity insertion that would accompany the
unintended withdrawal of one or more control rods.
SSTAR currently incorporates a single safety grade
emergency heat removal system which is the Reactor
Vessel Auxiliary Cooling System (RVACS) for decay heat
removal should the normal heat removal path involving the
Pb-to-CO2 HXs be unavailable. The RVACS involves heat
removal from the outside of the guard vessel due to natural
circulation of air which is always in effect. Because the
RVACS represents only a single safety grade system, it
would be required to have a high reliability with respect to
seismic events or sabotage. To provide for greater
reliability of emergency decay heat removal beyond that
corresponding to a single RVACS system, it is planned to
incorporate multiple safety grade Direct Reactor Auxiliary
Cooling System (DRACS) HXs inside of the reactor vessel
to provide for independent and redundant means of
emergency heat removal.
Conversion of the core thermal energy to electricity is
accomplished using a supercritical carbon dioxide (S-CO2)
Brayton cycle energy converter providing higher plant
efficiencies and lower balance of plant costs than the
traditional Rankine steam cycle operating at the same
reactor core outlet temperature. The interest in higher plant
efficiencies has heretofore driven interest in operation of
SSTAR at higher Pb temperatures to take advantage of the
increase in plant efficiency with temperature of the S-CO2
Brayton cycle. The increase in cycle efficiency with
turbine outlet temperature calculated for SSTAR is shown
in Fig. 13.
SSTAR S-CO
2
versus Turbine Inlet Temperature
30
35
40
45
50
55
300 400 500 600 700 800 900
TURBINE INLET TEMPERATURE,
o
C
CYCLE EFFICIENCY, %
Fig. 13. Efficiency Advantage of S-CO2 Brayton Cycle.
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
A peak cladding inner surface temperature of 650 °C
has been adopted as a goal. This is comparable to the peak
cladding temperature of ~ 630 °C of the Russian BREST-
OD-300 design which reflects the optimism of the Russians
to operate at such temperatures with EP-823 Si-enhanced
F/M steel cladding. Reactor and S-CO2 Brayton cycle
conditions calculated for a 650 °C peak cladding
temperature are shown in Figure 14.
Efficiencies 100 % power
Cyc = 44.2 %
Net = 44.0 % 551.6 244.5 30.4 19.8
19.88 kg/s 438.1
CO
2
7.883
402.9
19.93
Pb 177.8
1 atm 19.98
567.1 44.7 85.89 178.5
T, C T,C 7.614 19.98 185.6
Air Q,MW P,MPa 7.833
RVACS 177.4
419.0 0.3 5.2 19.98
420.5
45
33.85 31.25 84.5 87.7
7.775 7.400 20.00 7.783
Ave Peak
420.0 420.0 497.5 627.7
538.3 650.0 4.7
558.5 668.9 750 0.02 30.0 37.5
2107 kg/s 629.2 841 kg/s 0.127 0.101
67%
33%
Temperature Distribution and Heat Balance in SSTAR system.
CORE temperatures
Coolant
23.6
Fuel
29.1
70.4
Cladding
Bond
TURBINE
HTR
CORE
RHX
COOLANT
MODULE VESSEL
LTR
COMP. #1
COMP. #2
COOLER
Fig. 14. Conditions Calculated for SSTAR with S-CO2 Brayton Cycle Energy Converter.
The coupled reactor-S-CO2 Brayton cycle plant system
thermal hydraulic pre-conceptual design has been
optimized to maximize the cycle efficiency. For a 650 °C
peak cladding temperature, a reactor core outlet
temperature of 567 °C is achieved resulting in a Brayton
cycle efficiency of 44.2 % and a net plant efficiency of 44.0
%. Remarkably small sizes are calculated for the
centrifugal (i.e., radial flow) compressors and the axial
flow turbine. The small turbomachinery sizes reflect the
high density of supercritical CO2 (e.g., relative to gaseous
Helium) such that the flow area for energy transport with
supercritical CO2 is much less than that required for a gas
such as He in a gas Brayton cycle or steam in a Rankine
cycle steam turbine. Centrifugal compressors have been
incorporated in the S-CO2 Brayton cycle energy converter
for SSTAR because they provide a wider operating range
than axial compressors as illustrated in Figure 15. This is
especially important for the recompression cycle having
two compressors because the main compressor
(Compressor No. 1 in Fig. 14) operates with inlet
conditions close to the CO2 critical point which could be
expected to limit the compressor operating range.
Fig. 15. Comparison of Operating Ranges Between Compressor
Stall and Choking for SSTAR Centrifugal versus Axial Main
Compressor Pre-Conceptual Designs.
Compressor #1 Performance Map
(Centrifugal vs. Axial)
0.7
1
1.3
1.6
1.9
2.2
2.5
2.8
3.1
3.4
3.7
0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8
Relative Flow Rate
Pressure Ratio
0
10
20
30
40
50
60
70
80
90
100
Efficiency, %
pr (axial)
pr (centr)
eff (axial)
eff (centr)
Fixed inlet conditions,
design speed
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
An autonomous load following capability is needed for
deployment on small or immature electric grids. A control
strategy has been developed for the S-CO2 Brayton cycle
energy converter whereby the cycle is automatically
controlled such that the heat removal from the Pb-to-CO2
HXs is adjusted such that the generator power matches the
load demand from the electric grid. The control strategy
enables autonomous load following by the reactor whereby
the core power automatically adjusts itself due to the strong
reactivity feedbacks of the fast spectrum core with Pb
coolant and transuranic nitride fuel such that the core
power matches the heat removal from the Pb-to-CO2 HXs.
In particular, it is not necessary for the reactor core power
to be changed through the motion of control rods except for
startup and shutdown operations.
The automatic control strategy involves turbine bypass
control over the full power range combined with inventory
control between 35 and 100 % nominal grid load.
(Possible control mechanisms for the S-CO2 Brayton cycle
are shown in Fig. 16.) Turbine bypass control is a fast
acting mechanism that reduces the work performed by the
turbine through opening of a bypass valve. Inventory
control is a slower mechanism that removes mass from the
circuit to inventory control tanks or adds mass to the circuit
from inventory control tanks tending to maintain system
temperatures and cycle efficiency but its range is limited by
the inventory control tank volume. In principle, the control
strategy enables autonomous load following over the
complete plant operating range between essentially 0 and
100 % nominal grid load. The incorporation of centrifugal
compressors instead of axial flow compressors was found
to increase the operating range over which turbine bypass
control can be used to cover the full range from 0 to 100 %
load. The steady state conditions calculated for selected S-
CO2 Brayton cycle parameters under the control strategy
are shown in Fig. 17 where the ability of inventory control
to tend to maintain the cycle efficiency is demonstrated.
Autonomous load following enabled by S-CO2 Brayton
cycle automatic control simplifies operator requirements
and enhances plant reliability.
1 – Reactor core
2 – Pb primary coolant
(natural circulation)
3 – Pb-to-CO
2
in-reactor
heat exchanger
4 – CO
2
turbine
5 – Generator
6,7 – High and low
temperature
recuperators
8 – Cooler
9,10 – Compressors
11 – Cooling circuit to
ultimate heat sink
12 – RVACS
13 – Atmosphere heat
sink
17 – IRHX
bypass valve
18 –
Turbine inlet valve
19 – Turbine bypass
valve
20 – Inventory control
21 – Flow split valve
1
2
3
4
5
6
7
8
9
10
11
12
12
17
18
19
20
21
3
Color coding: Blue – Heat sink
Black – Primary circuit Red – Shutdown and decay heat cooling
Green – Normal heat removal Orange – Control system components
13
13
Fig. 16. Possible Control Mechanisms for S-CO2 Brayton
Cycle Energy Converter Used with Natural Circulation LFR.
COMPRESSOR STALL CHECK
(STALL AT 1)
1
1.1
1.2
1.3
1.4
1.5
1.6
1.7
1.8
0% 10% 20% 30% 40% 50% 60% 70% 80% 90% 100%
GENERATOR LOAD
f_stall
Comp1
Comp2
EFFICIENCIES vs. GENERATOR OUTPUT
0
5
10
15
20
25
30
35
40
45
50
0 10 20 30 40 50 60 70 80 90 100
GENERATOR LOAD, %
CYCLE EFFICIENCY, %
Cycle
System
TURBINE AND COMPRESSOR CHOKE CHECK
(MAXIMUM MACH NUMBER)
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0% 10% 20% 30% 40% 50% 60% 70% 80% 90% 100%
GENERATOR LOAD
MAX MACH NUMBER
Comp1
Comp2
Turb
MINIMUM AND MAXIMUM PRESSURES IN CYCLE
10
12
14
16
18
20
22
HIGH PRESSUR E, MPa
p_max
6.0
6.5
7.0
7.5
0% 10% 20% 30% 40% 50% 60% 70% 80% 90% 100%
GENERATOR LOAD
LOW PRESSURE, MPa
p_min
p_crit
Fig. 17. SSTAR S-CO2 Brayton Cycle Control Strategy Steady State Performance.
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
Table I provides conditions for the SSTAR plant.
TABLE I
Conditions and Dimensions for SSTAR
LFR SSTAR (Small Secure
Transportable
Autonomous Reactor)
Power, MWe (MWt) 19.8 (45)
Client – Assume 4.0 tonnes
of Oil Equivalent per Capita
per year = 167 GJ per
Capita per year = 5.3 KWt-
year per Capita per year of
which ~ 1/3 is used for
electricity
Electricity for a Town
of ~ 25400
Coolant Pb
Fuel Transuranic Nitride
(TRUN) Enriched to
N15
Enrichment, % 1.7/3.5/17.2/19.0/20.7
TRU/HM, 5 Radial
Zones
Core Lifetime, years 30
Core Inlet/Outlet
Temperatures, °C
420 / 567
Coolant Flow Rate, Kg/s 2107
Power Density, W/cm3 42
Average (Peak) Discharge
Burnup, MWd/Kg HM
81 (131)
Burnup Reactivity Swing, $ < 1
Peak Fuel Temperature, °C 841
Cladding Si-Enhanced
Ferritic/Martensitic
Stainless Steel Bonded
to Fuel Pellets by Pb
Peak Cladding Temperature,
°C
650
Fuel/Coolant Volume
Fractions
0.45 / 0.35
Core Lifetime, years 30
Fuel Pin Diameter, cm 2.50
Fuel Pin Triangular Pitch-
to-Diameter Ratio
1.185
Active Core Dimensions
Height/Diameter, m
0.976 / 1.22
Core Hydraulic Diameter,
cm
1.371
Pb-to-CO2 HXs Type Shell-and-Tube
Number of Pb-to-CO2 HXs 4
HX Tube Length, m 4.0
HX Tube Inner/Outer
Diameters, cm
1.0 / 1.4
Number of Tubes (all HXs) 10,688
HX Tube Pitch-to-Diameter
Ratio
1.222
HX Pb Hydraulic Diameter,
cm
0.904
HX-Core Thermal Centers
Separation Height, m
6.80
Reactor Vessel Dimensions
Height/Diameter, m
12.0 / 3.23
Reactor Vessel Thickness,
cm
5.08
Gap Between Reactor
Vessel and Guard Vessel, cm
12.7
Gap Filling Material Air
Guard Vessel Thickness, cm 5.08
Air Channel Thickness, cm 15
Air Ambient Temperature,
°C
36
Working Fluid Supercritical CO2
CO2 Turbine Inlet
Temperature, °C
552
Minimum CO2 Temperature
in Cycle, °C
31.25
Max/Min CO2 Pressures in
Cycle, MPa
20 / 7.4
CO2 Flow Rate, Kg/s 245
Net Generator Output, MWe 19.8
Supercritical CO2 Brayton
Cycle Efficiency, %
44.2
Net Plant Efficiency, % 44.0
Proceedings of ICAPP 2007
Nice, France, May 13-18, 2007
Paper 7218
III. SUMMARY AND CLOSURE
The SSTAR LFR is a viable concept for a small fast
reactor converter for international deployment in partner
(non-fuel cycle) states providing proliferation resistance,
fissile self-sufficiency, autonomous load following,
simplicity of operation and reliability, transportability, as
well as a high level of passive safety. Interest in achieving
higher plant efficiencies has heretofore driven interest in
operation of SSTAR at higher Pb temperatures to take
advantage of the increase in plant efficiency with
temperature provided by the S-CO2 Brayton cycle power
converter. A peak cladding temperature of 650 °C has been
used as a goal; at this temperature, a reactor core outlet
temperature of 567 °C is achieved resulting in a net plant
efficiency of 44.0 %. It was always recognized that this
would require the development of cladding and structural
materials for long-term service in Pb coolant up to 650 °C
peak temperature with corrosion protection provided by
active maintenance and control of the dissolved oxygen
potential in the coolant giving rise to the formation of
protective oxide layers on the steel cladding and structures.
SSTAR development has been supported by testing in the
DELTA loop at Los Alamos National Laboratory of alloy
specimens with special treatments or coatings which might
enhance corrosion resistance at the temperatures at which
SSTAR operates.5
The LFR has suffered from the lack of operation of a
LFR test demonstration reactor in the West. The focus of
LFR development in the U.S. is now shifting towards the
investigation and development of a near-term deployable
LFR demonstration test reactor. The demo will implement
innovative engineering features that should allow lead to
show its economic potential and industrial attractiveness.
The demo would operate at lower temperatures enabling
the use of existing materials such as T91 or HT9 F/M steels
that have been shown to have corrosion resistance to lead-
bismuth eutectic with active oxygen control at temperatures
below ~ 550 °C in experiments carried out in the DELTA
loop and elsewhere.5
ACKNOWLEDGMENTS
Argonne National Laboratory’s work was supported by
the U.S. Department of Energy Generation IV Nuclear
Energy Systems Initiative. The authors are indebted to Dr.
Rob M. Versluis, U.S. Department of Energy, NE-33,
Generation IV Program Manager.
REFERENCES
1. J. J. SIENICKI and A. V. MOISSEYTSEV, “SSTAR
Lead-Cooled, Small Modular Fast Reactor for
Deployment at Remote Sites System Thermal
Hydraulic Development,” 2005 Congress on Advances
in Nuclear Power Plants (ICAPP 2005), Seoul, May
15-19, 2005, Paper 5426.
2. W. S. YANG, M. A. SMITH, S. J. KIM, A. V.
MOISSEYTSEV, J. J. SIENICKI, and D. C. WADE,
“Lead-Cooled, Long-Life Fast Reactor Concepts for
Remote Deployment,” 2005 Congress on Advances in
Nuclear Power Plants (ICAPP 2005), Seoul, May 15-
19, 2005, Paper 5102.
3. J. SIENICKI, D. WADE, A. MOISSEYTSEV, W. S.
YANG, S.-J. KIM, M. SMITH, G. ALIBERTI, R.
DOCTOR, and D. MATONIS, “STAR Performer,”
Nuclear Engineering International, July 2005, p. 24.
4. M. T. FARMER and J. J. SIENICKI, “Analysis of
Transient Coolant Void Formation During a Guillotine-
Type Tube Rupture Event in the STAR-LM System
Employing a Supercritical CO2 Brayton Cycle,” 12th
International Conference on Nuclear Engineering
(ICONE12), Arlington, Virginia, April 25-29, 2004,
Paper ICONE12-49227.
5. N. LI, “Lead-Alloy Coolant Technology and Materials
– Technology Readiness Level Evaluation,” The 2nd
COE-INES International Symposium on Innovative
Nuclear Energy Systems, INES-2, Yokohama, Japan,
November 26-30, 2006.
6. J. Y. LIM and R. G. BALLINGER, “Alloy
Development for Lead-Cooled Reactor Service,” MIT-
Tokyo Tech Symposium on Innovative Nuclear Energy
Systems, Massachusetts Institute of Technology,
Cambridge, Massachusetts, November 2-4, 2005.
7. J. J. SIENICKI and D. C. WADE, “Nonproliferation
Features of the Small Secure Transportable
Autonomous Reactor (SSTAR) for Worldwide
Sustainable Nuclear Energy Supply,” Transactions of
the American Nuclear Society, Vol. 93, p. 340 (2005).
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One proposed concept for the STAR-LM Lead Fast Reactor (LFR) incorporates a supercritical CO2 gas turbine Brayton cycle to achieve high cycle efficiency and reduced plant footprint. In this design, 100+% of core full power is transferred by natural circulation from the core, located at the bottom of the reactor vessel, to in-vessel heat exchangers (HXs) located at the top of the vessel in the annulus between the core shroud and vessel inner wall. Although this approach extremely simplifies the plant design, the presence of the HX’s within the vessel raises concerns regarding the potential rupture of a HX tube that would initiate a high-pressure blowdown of CO2 into the lead coolant. The principal issue is to what extent, if any, is void entrained downwards with the coolant and then upwards through the core where adverse reactivity effects or degraded heat removal could result. To address this question, a scoping analysis of transient void formation during a guillotine-type HX tube rupture event in the STAR-LM employing a supercritical CO2 Brayton cycle has been performed. The void formation process is evaluated by solving a coupled set of ordinary differential equations describing: i) the supercritical CO2 blowdown, ii) bubble center-of-mass trajectory, iii) bubble growth rate, iv) bubble gas internal energy, and v) discrete bubble formation rate due to Taylor instability at the bubble/coolant interface. The results indicate that for thermal hydraulic conditions consistent with the current STAR-LM design, the peak blowdown rate from a single tube rupture is ∼ 2.5 kg/sec. The void formation process is dominated by large coherent gas bubbles that penetrate minimally downwards into the coolant due to the large coolant density. Rather, the gas pockets are predicted to periodically rise due to buoyancy and vent to the core cover gas region, as opposed to being swept downwards with the coolant. Moreover, the total CO2 fraction that is rendered in the form of discrete bubbles during blowdown is found to be small (∼ 3%), and the bubbles are of fairly large diameter (≥ 0.7 cm). Thus, these discrete bubbles are also calculated to benignly rise to the cover gas region since the terminal rise velocity for the bubbles exceeds the average lead coolant down flow velocity below the HX.
SSTAR Lead-Cooled, Small Modular Fast Reactor for Deployment at Remote Sites -System Thermal Hydraulic Development
  • J J Sienicki
  • A V Moisseytsev
J. J. SIENICKI and A. V. MOISSEYTSEV, "SSTAR Lead-Cooled, Small Modular Fast Reactor for Deployment at Remote Sites -System Thermal Hydraulic Development," 2005 Congress on Advances in Nuclear Power Plants (ICAPP 2005), Seoul, May 15-19, 2005, Paper 5426.
Lead-Cooled, Long-Life Fast Reactor Concepts for Remote Deployment
  • W S Yang
  • M A Smith
  • S J Kim
  • A V Moisseytsev
  • J J Sienicki
  • D C Wade
W. S. YANG, M. A. SMITH, S. J. KIM, A. V. MOISSEYTSEV, J. J. SIENICKI, and D. C. WADE, "Lead-Cooled, Long-Life Fast Reactor Concepts for Remote Deployment," 2005 Congress on Advances in Nuclear Power Plants (ICAPP 2005), Seoul, May 15-19, 2005, Paper 5102.
Lead-Alloy Coolant Technology and Materials -Technology Readiness Level Evaluation
  • N Li
N. LI, "Lead-Alloy Coolant Technology and Materials -Technology Readiness Level Evaluation," The 2 nd COE-INES International Symposium on Innovative Nuclear Energy Systems, INES-2, Yokohama, Japan, November 26-30, 2006.
Alloy Development for Lead-Cooled Reactor Service
  • J Y Lim
  • R G Ballinger
J. Y. LIM and R. G. BALLINGER, "Alloy Development for Lead-Cooled Reactor Service," MIT-Tokyo Tech Symposium on Innovative Nuclear Energy Systems, Massachusetts Institute of Technology, Cambridge, Massachusetts, November 2-4, 2005.