close

Вход

Забыли?

вход по аккаунту

?

ICONE25-66832

код для вставкиСкачать
Proceedings of the 2017 25th International Conference on Nuclear Engineering
ICONE25
July 2-6, 2017, Shanghai, China
ICONE25-66832
CONTROL SYSTEM DESIGN OF SUPERCRITICAL CO2 DIRECT CYCLE GAS FAST
REACTOR
Xianshan Zhang
School of Energy and Power Engineering,
Xian Jiaotong University,
Xian, Shaanxi, China
Peiwei Sun
School of Energy and Power Engineering,
Xian Jiaotong University,
Xian, Shaanxi, China
ABSTRACT
The Supercritical CO2 Direct Cycle Gas Fast Reactor(SCGFR) is a Generation IV reactor. The thermal properties
of supercritical carbon dioxide are different from helium gas.
Therefore, it is necessary to develop its dynamic model and study
its control system. A mathematical model is developed for the
SC-GFR. One prompt neutron group point kinetics equations
with six groups of delayed neutrons is applied in the model.
The core consists of an assembly of hexagonal fuel elements:
the innovative Tube-in-Duct(TID) fuel assembly. Steady-state
calculation is performed and the results are compared with the
design data. The transient results are analyzed and the responses
are reasonable in theory. The transient results show that the
model can properly predict the SC-GFR dynamics. The study
showed that it is not only feasible to build a numerical model of
the SC-GFR, but also the control system can satisfy the design
purposes.
m mass (kg)
n
neutron flux (n/(cm2 ·s))
n
in Eq.(8): number of fuel assemblies
N
filter coefficient
Nu Nusselt number
P
pressure (MPa)
Pr Prandtl number
Q heat flux per unit length (W/m)
Re Reynolds numbers
t
time (s)
T
K or ◦ C
Greek letters
α
heat transfer coefficient (W/m·K)
β
delayed neutron fraction
ζ
pressure loss coefficient
κ
Thermal conductivity W/(m·K)
λ
decay constant
Λ
the average neutron life time (s)
µ
dynamic viscosity (Pa·s)
ρ
fluid density (kg/m3 ))
Subscripts
b
bulk fluid
c
carbon dioxide coolant
d
cladding
f
fuel
i
node number
i
in Eq.(13–14): group number
s
inter surface of clad
w
wall
NOMENCLATURE
A
area of the flow cross-section (m2 )
C
specific heat (J/(kg·K))
D diameter (m)
f
fanning friction factor
g
gravitational acceleration
G fluid mass flow rate (kg/s)
h
fluid enthalpy per unit mass (J/kg)
k
gain parameter
l
length (m)
1
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
TABLE 1. SC-GFR plant key parameters [3, 4]
1 INTRODUCTION
[Matlab & Simulink] Six kinds of Generation IV nuclear reactor concept are proposed in 2000 [1]. In which the Gas-cooled
Fast Reactor is a fast-neutron reactor with some advantages
other fast-neutron reactors have. In addition, it also operates
within a closed fuel cycle which leads to a better utilization of
uranium resources, generated little low-level radioactive waste
and compliance with nuclear non-proliferation requirements.
Various refrigerants may be used in GFR as the coolant, typically
helium and carbon dioxide. In order to improve the Breeding
Ratio (BR) and prevent positive void coefficient, the neutron
capture cross sections need to be low.
As a type of the GFR, the Supercritical CO2 Direct Cycle
Gas Fast Reactor (SC-GFR) uses the recuperate Brayton cycle of
supercritical carbon dioxide, which runs more efficient than the
Rankine cycle or recuperate gas Brayton cycles working with
Helium.
The main advantages of this concept are as follows [2]:
⋄ High thermal efficiency with relatively low reactor outlet
coolant temperatures;
⋄ Compact, economical power conversion system;
⋄ Endurable core with a closed fuel cycle;
⋄ Small void reactivity worth from loss of coolant;
⋄ The coolant is stable, non-toxic and easy to obtain;
⋄ Natural convection decay heat removal;
⋄ Prevent nuclear proliferation.
A mathematical model is developed for the SC-GFR which
implemented with the Matlab & Simulink. One prompt neutron group point kinetics equations are applied in the primary
reactor kinetics model. Two reactivity feedback mechanisms
are considered. The reactor thermal-hydraulics is developed
based on the fundamental conservation of fluid mass, energy and
momentum. The mathematical model is built though all these
Partial Differential Equations(PDEs) and some other equations.
The heat transfer coefficient is derived from the immediate inner
wall temperature, the coolant pressure and temperature because
the particularity of coolant. By look up the table, all physical
and thermodynamic properties are calculated with the local
temperature and pressure. The core consists of an assembly of
hexagonal fuel elements: the innovative Tube-in-Duct(TID) fuel
assembly, and its temperature distribution has been numerically
calculated by the PDE Toolbox in Simulink. The property of
the clad and fuel, especially the thermal conductivity is also
calculated with a higher accuracy.
A mathematical model is developed for the SC-GFR which
implemented with the Matlab & Simulink. Point kinetics with
one prompt neutron group with six groups of delayed neutrons
is applied in the reactor neutron dynamics. Two reactivity
feedback mechanisms are considered. Based on the first law
of thermo-dynamics, developed the reactor thermal-hydraulics
model by the balances of mass, energy and momentum. The
Parameter
Value
Core Thermal Output
2400 MWth
Specific Power
20.7 kW/kgHM
Power Density
85.4 kW/L
Active Core Diameter
1.54 m
Effective Core Diameter
4.81 m
Reflector
Ti(axial),S-CO2 (radial)
Core Flow
1.1708 × 104 kg/s
Plant Electrical Output
1200 MWe
Coolant Inlet Temperature
485.5 ◦ C
Coolant Outlet Temperature
650 ◦ C
Coolant volume fraction
25%
Peak Coolant Pressure
20 MPa
Fuel emissivity
0.79
Plant Lifetime
60 years
mathematical model is built though all these Partial Differential
Equations(PDEs) and some other equations. Because of the
particularity of the coolant, the heat transfer coefficient is derived
from real-time inner wall temperature, the coolant temperature,
pressure and flow rate. By looking up the table, all physical and
thermodynamic properties are calculated by the local temperature and pressure in real time. The core consists of a new type
of assembly: the innovative Tube-in-Duct(TID) fuel assembly,
and its temperature constant distribution has been numerically
calculated by the PDE Toolbox in Simulink. The property of
the clad and fuel, especially the thermal conductivity is also
calculated with a higher accuracy.
2
S-CO2 GAS-COOLED FAST REACTOR
There are two main design directions in the international
research on SC-GFR.
First, the research has been done by MIT mainly in the
USA, and is pretty deep. This research has adopted the CO2
re-condensation direct loop which has no CO2 condensation.
Using the columnar core component, and cancel the plutonium
proliferation that traditional fast reactor has, turn to employ a
new “Tube-in-Duct”(TID) assembly to enlarge the portion of
fissionable material.
2
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
FIGURE 2.
FIGURE 1.
Simplified Plant Layout (Top view).
Simplified Plant Layout (Front view).
Second, the Tokyo Institute of Technology’s Advanced Gas
Reactor represent another direction of SC-GFR, in the core it
uses Partial pre-cooling direct cycle, which can reduce the heat
taken by condensation, in the primary design they think pebble
or rod type fuel element could be used in S-CO2 cooled AGR
and receive fairly good effect.
To obtain the dynamic characteristics of SC-GFR, a MIT
concept [5] is chosen in this study, and some supercritical carbon
dioxide Brayton cycle and design is refereed to work done by
Dostal [6, 7] and by Moisseytsev [8].
While the research presented in this work deals mainly with
the simulation of plant and its control system, it is important
to understand the larger framework into which it fits. Figure
1 and Fig.2 show the layout of the plant. Figure 1 shows two
parallel 600 MWe turbo-machinery sets (turbine, re-compressor,
main compressor, and electric generator) each served by four
heat exchanger trains (high and low temperature recuperators,
and pre-cooler) in two pairs on both upper and lower floors,
straddling the shaft.
The Fig.1 is a top-view drawing, due to the angle, only the
two upper floor trains per shaft can be seen. Turbine capacity
is limited at about 600 MWe, based on keep in the affordable
limit of turbine blades and shaft rpm to 1800, a standard electric
generator value. And the diameter of pipe limits to about
1 meter to conform the attempted practice, therefore the heat
exchanger train capacity limit to about 320 MW, corresponding
to about 150 MWe. Finally, the Printed Circuit Heat Exchangers
(PCHE) produced by Heatric™meet the needs of design, which
FIGURE 3. 2D and 3D Rendering of a Hex-nut fuel Pellet of TID
assembly.
offers a compact solution, the ability of working with a very
large pressure difference between the hot side and the cold side
in the heat exchanger, a low pressure drop and high thermal
effectiveness [9]. The plant layout can fit inside a diameter of
54m containment, able to remove decay heat after LOCA.
The reactor core uses a new type of fuel assemblies called
Tube-in Duct (TID). The TID fuel assembly consists of a
hexagonal outer tank having a cylinder coolant channel placed
in the fuel center, between the fuel layer and coolant is the
clad, which manufactured by INCOLOY alloy MA956, whose
melting point is 1480 ◦ C, density is 7.25 g/cm3 . This metal
obtained by mechanical alloying, allowing the clad has excellent
anti-oxidation and prevent carbon corrosion characteristics, has
been selected as lining material of the aero engine combustion
chamber. The TID fuel assemblies is displayed at Fig.3, note the
fuel and coolant have switch places, compared to original design.
The selection of TID fuel assembly rather than traditional
3
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
PIN type fuel assembly, it is a compensation measures, and
it cancels the external blankets. Compare to same pitch to
diameter ratio, this design allows a higher fuel volume fraction
than PIN type, it is favorable for numerous neutronic reasons.
It has advantages in terms of anti-nuclear proliferation, because
the isotopes which could produce nuclear weapons(e.g., 239 Pu)
would mixed with other fission products, make it difficult to
extract.
Compared to PIN type, TID assembly also has advantage
in thermal-hydraulics, TID type have a lower fuel temperature
and less pressure drop than PIN type at a certain fuel volume
fraction. Besides, TID fuel assembly need no spacer grid either,
which reduce further pressure drop. PIN type would result in
less favorable thermal hydraulic performance to achieve the same
neutronics target.
Much work has gone into development of a S-CO2 Brayton
Re-compression Cycle for use as a nuclear plant both in USA and
Japan, the main feature of the cycle which distinguished it from
the original Brayton cycle is that its compression operates near
the critical point, where density is very high, and it uses another
compressor called re-compressing compressor, which sets before
the pre-cooler. It different from the traditional Brayton cycle
where there is typically only one recuperator as well [6]. This
arrangement can avoid the pinch point in the recuperators which
usually happens when a simple Brayton cycle was adopted.
In order to get the accurate physical properties such as specific heat capacity and density, used Look-up block of Simulink.
As mentioned in the above chapter, set the main compressor
operated near the critical point of CO2 , where density is very
high, to improve its efficiency. It is worth noting that because of
the difference in physical properties, the start-up period is quite
different from helium-cooled fast reactor.
balance:
dm
= Gin − Gout
dt
(1)
d(mh)
= Gin hin − Gout hout + Q
dt
(2)
Energy balance:
Momentum balance:
pout = pin − ζ
G2in
ρ
(3)
In the reactor core, the fission energy is released into the
fuel assemblies in the form of heat, then transferred into the clad
through the gas gap by conduction and to the middle coolant
channel afterwards by convection. This process can be described
as Eq.(4) in each fuel node:
d
(mf,i hf,i ) = P − (hA)fc,i (Tf,i − Tc,i )
dt
(4)
And the variations in temperature of carbon dioxide coolant
satisfied Eq.(5):
d
(mc,i hc,i ) = (hA)fc,i (Tf,i − Tc,i ) + Gi−1 hi−1 − Gi hi
dt
(5)
The subscripts number i − 1 in Eq.(4) and Eq.(5) turn into
“inlet” when the i equals 1. Outlet fluid mass flow rate and the
variations of fluid density satisfied Eq.(6)
3
GOVERNING EQUATIONS
This chapter includes the derivation of the simplified dynamic models for each of the considered operating units. A
model constructed based on the combination of mass, energy and
momentum balance.
The following simplifying modeling assumptions are considered:
Vi
⋄ The reactor is regarded as a united system with 5 operating
nodes of core and two reflectors as two operating nodes,
and the properties such as pressure, temperature and flow
in every nodes are uniform;
⋄ The initial reactivity is considered as 0;
⋄ One dimensional fluid flow model is utilized.
dρ
= Gi−1 − Gi
dt
(6)
3.2 Heat transfer
For each node, the thermal power can be related to the
temperature drop from fuel to coolant via the overall heat transfer
resistance as:
(hA)fc =
First law of thermo-dynamics
Considered the principle of conservation, the reactor model
is bounded by the balance of mass, energy and momentum: Mass
P
Tf − Tc
(7)
3.1
And the heat transfer resistance can be described by the
summation of thermal resistance of the fuel, the gas gap, the
4
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
3.4 Neutron dynamics
The neutrons behavior in the SC-GFR core can be described
with a point kinetics model, consisting of six energy groups.
cladding and the convective heat transfer resistance in carbon
dioxide coolant. Hence, the (hA)fc can be formulated as:
(hA) f c =
n
1
4πκf,i
+ 2π r
1
f,i hg,i
+
rf +tg +Tc
1
1
2πκf,i ln( rf +tg ) + π dhs
(8)
The gap heat transfer coefficient hg is calculated by:
κg
hg =
+
tg
1
εc
σ
+ ε1 − 1
f
(
4
Tf,inner
− Td4
(
κw
κb
)0.22 (
µw
µb
)−1.13
(9)
TABLE 2. Fast neutron fission parameter for 235U in SC-GFR
3.3
Pressure drop
Based on the assumptions mentioned above, momentum
balance equations can be summed up as a summation of frictional
pressure drop and form-loss pressure drop.
L
G2
( fh + 1.2)
2ρ
D
]
[
8qw
1 ∂ρ
fh = f +
− ( ) m
Gc p
ρ ∂t
0.316Re−1/4
for Re ≤ 2 × 104
0.184Re−1/5
for Re ≥ 2 × 104
βi
λi /s−1
τi
1
0.000251
0.0127
78.74
2
0.001406
0.0317
31.55
3
0.001241
0.115
8.70
4
0.002687
0.311
3.22
5
0.000845
1.40
0.714
6
0.000172
3.87
0.258
Summation
β =0.00660
β /λ =0.084
Regarded the initial steady state as 0, the temperature
variation of fuel and coolant results in a reactivity change, which
would feedback to stabilize the system by itself, and complied
the thermal-hydraulics and the neutronics,
ρ = ρext + αf ∆Tf + αc ∆Tc
(11)
(15)
Where the αf = −0.45 ¢/K and αc = −0.38 ¢/K are the
temperature coefficient of the TID assemblies and Supercritical
CO2 at the Beginning of Cycle (BOC) [3].
For turbulent flow in smooth tube, the friction factor ( f ) is
given by:
f=
Group
(10)
The frictional pressure drop is calculated by Eq.(10), Where
Hydraulic drag ( fh ) is given by [11]:
{
(14)
Where the i = 1, 2, . . . , 6, noted that the SC-GFR is different
from typical PWR, it is a fast reactor with a different coolant.
The parameter of neutron in the SC-GFR are shown in Tab.2 [9].
Guptas correlation was developed for Super Critical CO2
by the experimental data within a range of thermodynamic
parameterspressures belongs to 7.4–8.8 MPa, mass fluxes 900–
3000 kg/m2 ·s and inlet fluid temperatures 20–40 ◦ C. Due to the
Dittus-Boelter correlation was better applied to sub-critical CO2 ,
and CO2 is very sensitive to temperature and pressure within
the pseudo-critical and critical ranges, therefore chose the Eq.9
rather than Dittus-Boelter correlation.
pout − pin =
dci
βi
= n − λiCi
dt
Λ
Tf,inner − Td
)0.93 (
ρw
ρb
(13)
)
And the heat transfer coefficient hs can be calculated by the
work of Gupta [10]:
Nu = 0.01Re0.89 Pr−0.14
6
dni
ρ − βi
=
n + ∑ λi ci
dt
Λ
i=1
(12)
4 DYNAMIC MODEL OF SC-GFR
4.1 Dynamics model of SC-GFR without controller
The thermal power of reactor can be calculated by the neutron flux, firstly, rearrange the Eq.(4) and Eq.(5), then Laplace
Above equation is valid for supercritical CO2 within the
range of Re ≤ 1.5 × 105 .
5
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
FIGURE 4.
control.
Simulink representation of reactor model without PID
FIGURE 5.
control.
Simulink representation of reactor model with two PID
transform is used to combine outlet and input.
hf,i =
1
[P − (hA)fc,i (Tf,i − Tc,i )]
mf,i s
constant control strategy, to avoid the influence from coolant
temperature continuous rise, to avoid the thermal stress-cracking
in TID assemblies and structures in vessel. However, it drew
higher demand in turbine.
In the primary control system design, considered two individual mechanisms: to by the control rod and to by the valve
opening. Adopted a combination of two PID controllers, through
the PID controllers export two signals, passed to the stepper
motor, to adjust the control rod and valve opening, to change the
reactivity and inlet coolant flow rate. And to avoid the frequent
action of control rod, after the calculation of outlet temperature
and target temperature, there is a dead zone allowing a certain
error. This model with two PID controller is shown in Fig.5.
The PID controllers are weighted according to the independent gain parameter kP , kI , and kD [12]. The transfer function of
a PID in Simulink is shown in Eq.(18):
(16)
And the variations in temperature of carbon dioxide coolant
satisfied Eq.(5):
hc,i =
1
[(hA)fc,i (Tf,i − Tc,i ) + Gi−1 hi−1 − Gi hi ]
mc,i s
(17)
The thermal properties of supercritical carbon dioxide
change due to coolant temperature and pressure at the moment,
hence, use Look-up block to obtain properties of coolant. As
Eq.(8) has shown, the (hA)fc would change with the power, flow
rate and temperature of fuel and coolant, therefore, during the
model founding, it could be calculated by dynamic properties.
In order to get the accurate thermal resistant, use the reactor
thermal properties such as power, flow rate and temperature of a
unit time step before the current time.
The model is using a Matlab ODE solver called ODE45.
The equations above can be solved directly via the Matlab &
Simulink platform. The structure of the SC-GFR dynamic model
(without control system) is shown in Fig.4.
1
N
Gc (s) = kP + kI + kD
s
1 + N 1s
(18)
5 SIMULATION RESULTS
There are four inputs in the model: the temperature (Tc,in ),
pressure (pin ) and flow rate (Gin ) of inlet coolant, the reactivity
introduced by outside ((ρext )). The ρext can be considered as
disturbance. set it to 0, and the other inputs set as a constant
conform with design, the outputs values are shown in Tab.3,
outputs are corresponded with design propose.
To investigate the dynamic characteristics of the reactor
model without any control system, for the comparisons to show
the control system effect. In each case, one input variable
4.2
Control system based on dynamics analysis
In order to adjust the cycle in such a way that the plant
produces electricity as expected, chose the reactor parameters
that can describe the difference between power level and the
grid demand, there chose a plan of average coolant temperature
6
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
TABLE 3. Steady-state Values and Design values
P/MW
Gout /kg·s−1
pout /MPa
Tc,out /◦ C
Design values
2400
1.78×104
18.2
650
Model values
2400
1.779×104
18.223
649.45
is change from its design values, while other inputs are kept
unchanged.
5.1
A step increase in reactivity without PID
The responses to the step increase in reactivity from 0 to
5¢at 50s without PID controller have shown in Fig.6-Fig.8, as a
result, the final reactor power increases because of the increase
of neutron flux, then the temperature of fuel increases which
lead to the increase of coolant by thermal transfer, and because
of the reactivity-feedback from both fuel and coolant as Eq.15
described, the reactor power decrease slightly and stabilizes at
102.13% FP (Full Power equals 2400WM). During this process,
the fuel average temperature finally increases 17.6◦ C to 827.6◦ C,
the coolant average temperature increases 3.85◦ C. These results
change trend has same shape with similar HTR model [13].
FIGURE 7. Fuel temperature responses to a step increase in reactivity.
FIGURE 8.
reactivity.
Coolant temperature responses to a step increase in
TABLE 4.
FIGURE 6.
Power responses to a step increase in reactivity.
Parameters of two PID controllers.
Number
kP
kI
kD
N
1 Pset
-1.27×10−3
-9.442×10−3
-1.627×10−5
55.41
2 Tc,ave
80.66
43.57
0
100
The responses to the 5% FP step increase in power set have
shown in Fig.10-Fig.12 at the beginning, the reactor stabilizes at
the design value; after the step increase in power set introduces
at 10s, the PID controllers pass the difference value between the
power and power set to the stepper motor, then changes the valve
opening and the control rod, which leads to changes in reactivity
5.2
A step increase in power set
The parameters of two PID controllers are designed with
the help of Optimization block of Simulink toolbox, there are
the optimize process in Fig.9. The final two PID controllers
parameters set values are shown in Tab.4.
7
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
FIGURE 9.
A process of Optimization.
FIGURE 11. Fuel temperature response to a step increase in Power
set
and inlet flow rate, finally the reactor power stabilizes at 105% FP
due to the reactivity feedback and PID controller, and the valve
opening decreases, to balance the increase of coolant temperature
caused increase in pass through flow rate. The fuel temperature
finally increases by about 10◦ C.
FIGURE 12.
Power set.
FIGURE 10.
Coolant temperature response to a step increase in
in temperature set pass an increase of valve opening, it increases
the inlet coolant flow rate, make the coolant temperature promote
more through the reactor core, with higher temperature in reactor
core, it drives the negative reactivity feedback mechanisms
thereby resulting in the reactor power stabilizes at 100% FP as
set after the slightly shake, and the fuel temperature increases
by 10.4◦ C, average coolant temperature increases at the new set
point of 572.75◦ C.
Power response to a step increase in Power set.
5.3
A step increase in coolant temperature set
As discussed in chapter 4, in this model there used a plan
of average coolant temperature constant control, so the average
coolant temperature set is another control object. The responses
to the 5◦ C step increase in average coolant temperature set values
are shown in Fig.13-Fig.15, at the beginning, the reactor stabilizes at the design value because there is no disturbance; after the
step increase in temperature set introduces at 10s, the increase
6 CONCLUSIONS
Steady-state calculation is performed and the results are
compared with the design data. The errors are small and the
dynamic model is verified. In order to study the dynamic behaviors of the reactor system, step disturbances are introduced. The
8
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
FIGURE 13. Power response to a 5◦ C step increase in Tc,ave set.
FIGURE 15.
Tc,ave set.
Coolant temperature response to a 5◦ C step increase in
and the Project-sponsored by SRF for ROCS, SEM.
REFERENCES
[1] A technology roadmap for generation iv nuclear energy
systems. Report GIF-002-00, US DOE Nuclear Energy
Research Advisory Committe and the Generation IV
International Forum, 2002.
[2] Parma, E. J., Wright, S. A., Vernon, M. E., Fleming, D. D.,
Rochau, G. E., Suo-Anttila, A. J., Rashdan, A. A., and
Tsvetkov, P. V., 2011. “Supercritical co2 direct cycle gas
fast reactor (sc-gfr) concept”. SANDIA REPORT.
[3] Handwerk, C. S., 2007.
“Optimized core design
of a supercritical carbon dioxide-cooled fast reactor”.
Department of Nuclear Science and Engineering.
[4] Pope, M. A., 2006. “Thermal hydraulic design of a 2400
mw direct supercritical co-cooled fast reactor”. Thesis.
[5] Gibbs, J. P., Hejzlar, P., Gong, Y., and Driscoll, M. J.,
2007. “Plant layout for a 1200 mwe direct brayton cycle
gfr”. Transactions of the American Nuclear Society, 96,
pp. 793–794.
[6] Dostal, V., Driscoll, M. J., and Hejzlar, P., 2004. A
supercritical carbon dioxide cycle for next generation
nuclear reactors. Report.
[7] Dostal, V., Hejzlar, P., and Driscoll, M. J., 2006. “Highperformance supercritical carbon dioxide cycle for nextgeneration nuclear reactors”. Nuclear Technology, 154(3),
pp. 265–282.
[8] Moisseytsev, A., Kulesza, K. P., Sienicki, J. J., Division,
N. E., and Univ., O. S., 2009.
Control system
options and strategies for supercritical co2 cycles. Report ANL-GENIV-081; TRN: US1000457 United States
10.2172/958037 TRN: US1000457 ANL ENGLISH, ;
FIGURE 14. Fuel temperature response to a 5◦ C step increase in
Tc,ave set.
transient results are analyzed and the responses are reasonable in
theory. The transient results show that the model can properly
predict the SC-GFR dynamics. After the lumped parameter
modeling, two PID controllers are designed with the help form
the Simulink Design Optimization block in Simulink to yield
desired steady-state error and a faster response. The study shows
that it is not only feasible to build a numerical model of the SCGFR using the Matlab & Simulink, and also the control system
can satisfy the design purposes.
ACKNOWLEDGMENT
This project is supported by the National Natural Science
Foundation of China (Grant No.11405126), the China Postdoctoral Science Foundation (Grant No.2014M552455), the Fundamental Research Funds for the Central Universities (xjj2014040)
9
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
Argonne National Laboratory (ANL).
[9] Gezelius, K., 2004. “Design of compact intermediate heat
exchangers for gas cooled fast reactors”. Thesis.
[10] Gupta, S., Saltanov, E., Mokry, S. J., Pioro, I., Trevani,
L., and McGillivray, D., 2013. “Developing empirical
heat-transfer correlations for supercritical co2 flowing in
vertical bare tubes”. Nuclear Engineering and Design,
261, pp. 116–131.
[11] Sarkar, J., Bhattacharyya, S., and Ramgopal, M., 2008.
“Pressure drop for in-tube supercritical co2 cooling:
Comparison of correlations and validation”. 19th National
& 8th ISHMT-ASME Heat and Mass Transfer Conference.
[12] Li, Y., Ang, K. H., and Chong, G. C., 2006. “Pid control
system analysis and design”. IEEE Control Systems, 26(1),
pp. 32–41.
[13] Li, H., Huang, X., and Zhang, L., 2008.
“A
simplified mathematical dynamic model of the htr-10 high
temperature gas-cooled reactor with control system design
purposes”. Annals of Nuclear Energy, 35(9), pp. 1642–
1651.
10
Copyright © 2017 ASME
Downloaded From: http://proceedings.asmedigitalcollection.asme.org/ on 10/25/2017 Terms of Use: http://www.asme.org/about-asme/terms-of-use
Документ
Категория
Без категории
Просмотров
0
Размер файла
747 Кб
Теги
icone25, 66832
1/--страниц
Пожаловаться на содержимое документа