GENES4/ANP2003, Sep. 15-19, Kyoto, JAPAN
Paper 1160
LOCA analysis of high temperature reactor cooled
and moderated by supercritical light water
Yuki Ishiwatari1*, Yoshiaki Oka1 and Seiichi Koshizuka1
Nuclear Engineering Research Laboratory, the University of Tokyo, Tokai-mura, Ibaraki, 319-1188, Japan
Tel: +81-29-287-8442, Fax: +81-29-287-8438, Email: [email protected]
1
Large break LOCA of a high temperature thermal reactor (SCLWR-H) with descending flow water rods is analyzed. The
criterion of the Ni-alloy cladding temperature is set to 1260°C. The safety system of the SCLWR-H is similar to that of
BWR. The safety principles of the SCLWR-H are (1) “feedwater from cold-leg is kept” and (2) “coolant outlet is opened at
hot-leg”. To prevent closure of the hot-leg, automatic depressurization valves (ADS) are actuated without delay by the
same signal as that of main steam isolation valves (MSIV) such as “Flow rate low level 3” or “Core pressure low level 2”
or “Containment pressure high”, while in ABWR the ADS is actuated by “Water level 1” and “Containment pressure
high” with 30 s delay. The low pressure core injection system (LPCI) is actuated by the same signal as that of the ADS
with 30 s delay due to the starting of the emergency diesel generators.
In the case of 100% cold-leg break LOCA, the core flow rate decreases and then the cladding temperature increases up
to about 720°C during blowdown phase. After the ADS are opened, the core flow rate is recovered and the cladding
temperature decreases. The core reflooding begins 78 s after the break. The peak cladding temperature in the reflooding
phase is about 790°C. In the case of 100% hot-leg break LOCA, the core coolability is much higher than that in cold-leg
break. The peak cladding temperature does not exceed that of the normal operation. In the reflooding phase, since the
cold-leg loops are isolated by check valves, only upward flow in the core is established. It is a kind of forced cooling by the
LPCI. The reflooding phase is also less severe than that of the cold-leg break LOCA. Thus, reflooding phase of hot-leg
break LOCA is not calculated in this study.
In summary, large break LOCA of the SCLWR-H is mitigated by MSIV, ADS and LPCI. The flow system of the
“descending flow water rods” makes the water inventory large and contributes to the high coolability. This is an
advantage. But it should be noticed that the core coolability is small before the ADS are opened at cold-leg break LOCA.
KEYWORDS: supercritical water reactor, descending flow water rods, large break LOCA, blowdown, reflooding
I. Introduction
Supercritical pressure light water cooled reactor (SCR)
has been conceptually designed in the University of
Tokyo1)2). It adopts technologies of light water reactors
(LWR) and supercritical fossil-fired power plants (FPP).
Once through cooling system makes the reactor pressure
vessel (RPV) and the containment vessel (CV) small. High
specific enthalpy of main steam makes the balance of plant
(BOP) compact and thermal efficiency high. The SCR has
a potential of a large cost reduction from LWRs.
In the past study3), a LOCA analysis code SCRELA was
developed. It consists of the “blowdown estimation
module” and the “reflood estimation module”. These
modules were validated by the REFLA-TRAC code, which
was developed in the Japan Atomic Energy Research
Institute (JAERI) based on TRAC-PF1. Cold-leg and
hot-leg break LOCAs of a low temperature thermal reactor
(SCLWR) were analyzed using the SCRELA. After that a
high temperature thermal reactor (SCLWR-H) with
descending flow water rods has been designed. But the
SCRELA does not contain the water rod model. A plant
transient analysis code SPRAT-DOWN has been developed
containing the descending flow water rod model. But this
code can calculate only supercritical pressure region. The
purposes of this study are to modify the SPRAT-DOWN for
LOCA analysis based on the SCRELA, to analyze large
break LOCA of the SCLWR-H, and to clarify its
characteristics.
II. Reactor core with descending flow water rod
The fuel assembly of the SCLWR-H is shown in Fig. 1.
It contains many water rods for neutron moderation. The
coolant flows downward in the water rods. The flow path
at normal operation is shown in Fig. 2. Part of the
feedwater is led to the top dome and cools it. Then it flows
downward in the control rod guide tubes and the water rods.
In the bottom of the fuel assemblies it is mixed with the
rest coolant which has descended in the downcomer.
This concept has advantages. Since there is no mixing of
hot and cold coolant at the upper plenum, the coolant outlet
temperature is kept high. Since the difference in the
average water density is small between upper and lower
parts of the core, the axial distribution of the neutron
moderation does not change substantially.
*
Corresponding author, Tel. +81-29-287-8442, Fax. +81-29-287-8488, E-mail: [email protected]
1
29.22cm
Water rod
UO2 ＋ Gd2O3
fuel rod
Control rod
guide tube
Capacity:
RCIC(AFS)
AFS
ADS
LPCI
TD 1 unit: 4%/unit
TD 2 units: 4%/unit
8 units: 20%/unit at 25MPa
MD 3 units: 300kg/s/unit at 1.0MPa
Configuration:
TD- RCIC
LPCI/RHR
UO2 fuel rod
TD-AFS
LPCI/RHR
RCIC: reactor core isolation cooling system
RHR: residual heat removal system
TD: turbine driven
MD motor driven
Fig. 1: Fuel assembly
Top dome
TD-AFS
LPCI/RHR
Control rod drive
Fig. 4: Capacity and configuration of safety system
Control rod guide tube
Hot leg
IV. Safety principle
The SCLWR-H has no water level. To keep the core
coolability, flow rate should be watched and maintained
while water level is kept in BWR. Decrease in the
feedwater flow rate is directly followed by decrease in the
core flow rate because the SCLWR-H has no recirculation.
Closure of the coolant outlet is also followed by decrease
in the core flow rate. The safety principles of the
SCLWR-H are “to keep feedwater from cold leg” and “to
keep coolant outlet open at hot leg”. The system pressure
should be also watched. The relation between abnormal
levels and actuations of the safety system is shown in Fig.
5. In the LOCA analysis, actuation of the high pressure
ECCS (AFS) is neglected as in BWR.
Hot leg
Cold leg
Cold leg
Upper plenum
Down comer
Active core
Water rod
Bottom dome
Fig. 2: Flow path at normal operation
III. Safety system
The plant system of the SCLWR-H is shown in Fig. 3.
The safety system is similar to that of BWR. It consists of
reactor scram, high-pressure auxiliary feedwater system
(AFS), low-pressure core injection system (LPCI), main
steam isolation valves (MSIV), safety relief valves (SRV)
and automatic depressurization valves (ADS). Capacity
and configuration of the safety system are shown in Fig. 4.
RPV
CV
Turbine control valve
Turbine bypass valve
CR
Suppression
chamber
LPCI
Condensate water
storage tank
Turbine
Condenser
Reactor scram
AFS
MSIV/ADS/LPCI system
Reactor scram
MSIV/ADS/LPCI system
Reactor scram
SRV
AFS: auxiliary feedwater system
MSIV: main steam isolation valve
ADS: automatic depressurization system
LPCI: low pressure core injection system
AFS
SRV/ADS
Flow rate low
Level 1
Level 2
Level 3
Pressure low
Level 1
Level 2
Pressure high
Level 1
Level 2
Fig. 5: Abnormal level and actuation of safety system
LPCI
High pressure
feedwater
heaters
Fig. 3: Plant system
Main feedwater
pumps
Low pressure
feedwater heaters
V. LOCA analysis code
1. Blowdown phase
Since water rods are not modeled in the “blowdown
estimation module” of SCRELA3), a plant transient
analysis code SPRAT-DOWN, which includes “descending
2
flow water rods” model but can calculate only supercritical
pressure region, is modified to calculate blowdown phase
of the SCLWR-H. The calculation model of the modified
SPRAT-DOWN is shown in Fig. 6. The hottest single
channel is divided into 20 nodes. A water rod channel is
also divided into 20 nodes. Heat transfer between these 2
channels is considered. The main feedwater lines are
divided into 10 nodes. The downcomer is divided into 20
nodes including the bottom dome. The upper plenum is
divided into 20 nodes including the main steam lines. The
top dome is divided into 10 nodes including the CR guide
tubes. Since the fuel channel and the water rod is modeled
as single channels, the volumes of the top dome, the
downcomer, the upper plenum and the main feedwater
lines are divided by the total number of the fuel rods. The
mass and energy conservations are calculated.
∂ρ ∂G
=0
+
∂t ∂Z
∂ ( ρH ) ∂ (GH )
+
= Q' ' '
∂t
∂Z
where
(1)
(2)
t: time
ρ: density
G: mass flow rate
Z: position
H: specific enthalpy
Q’’’: heat generation rate peer unit volume
In the nodes of tow-phase flow, the average density and the
average specific enthalpy are determined as:
A = Al (1 − x ) + Av x
(3)
where
A: density ρ or specific enthalpy h
l: liquid
v: vapor
x: void fraction
The boundaries are the main feedwater pumps, the break
point, the MSIV and the ADS. Since the pressure drop in
the main feedwater lines and the RPV is much smaller than
that at the break, the pressure is assumed as constant in
them.
The decreasing rate of the pressure in the blowdown
phase is governed by the flow rate at the break. The
calculation module of the break flow is the same as that in
the SCRELA. At subcritical pressure, three correlations of
the break flow in superheated vapor, sub-cooled water, and
two phase are used. Critical flow in the supercritical
pressure is not known. But the pressure at the break is
subcritical even if the stagnation pressure is supercritical.
The correlations in the subcritical pressure are also used in
the supercritical pressure. If the stagnation temperature is
below or equal to the pseudo-critical temperature, the
critical flow is treated as in the sub-cooled region. If the
temperature is higher than the pseudo-critical temperature,
the superheated vapor region is assumed. Fig. 7 shows the
critical mass fluxes at various pressures.
The heat transfer coefficient is evaluated by
Dittus-Boelter’s correlation in supercritical, sub-cooled and
superheated regions. In the film boiling region,
Dougal-Rhosenow’s correlation of film boiling is used. It
is also conservatively used in the nucleate boiling region.
The radiation heat transfer is involved. Heat capacities of
the RPV and other structures are neglected.
The axial power distribution is cosine. The reactor
power is calculated by the point-kinetics equation with six
delayed neutron groups while the decay heat is calculated
using a two-group approximation of 120% of the ANS
evaluation4). Doppler and coolant density feedbacks are
considered. Reactor scram is completed 2.8 s after the
signal including 0.55 s delay. The reactivity worth is
10 %dk/k. The reactivity curve shown in Fig. 8 is the same
as that of PWR. When “LOCA” is detected, the main
feedwater pumps are assumed to trip. The pump
coast-down time is assumed to be 5 s and the flow rate
decreases linearly. The flow chart of the calculation is
shown in Fig. 9.
In the past study3), the “blowdown estimation module”
of the SCRELA was validated by comparing with the
REFLA-TRAC code, which was developed in the Japan
Atomic Energy Research Institute (JAERI) based on
TRAC-PF1. The calculation started at a core pressure 17
MPa in the REFLA-TRAC code, since this code can not
treat supercritical pressure. A low temperature thermal
reactor (SCLWR) analyzed in another past study5) was used
for the validation calculation. In this study, the modified
SPRAT-DOWN is also compared with these 2 calculations.
Fig. 10 shows the pressures of 100% hot-leg LOCA.
Top dome
Main steam line
CR guide tube
ADS
MSIV
Break (hot leg)
Break (cold leg)
Check
valve
ADS
Main
feedwater line
line
Main
feedwater
pump
Upper plenum
Fuel channel
Water rod
LPCI
Down comer
Bottom
dome
Fig. 6: Calculation model of blowdown phase
3
250
Pressure(bar)
200
REFLA-TRAC
SCRELA
150
100
SPRAT-DOWN
50
0
0
10
20
30
40
50
Time(sec)
Fig. 7: Critical max flux as a function of stagnation
enthalpy and pressure
1.0
Reactivity ratio
0.8
0.6
0.4
0.2
0.0
0.0
0.5
1.0
1.5
2.0
2.5
3.0
Fig. 10: Pressure trend in 100% hot-leg break LOCA
2. Reflooding phase
The “reflood estimation module” of the SCRELA is used.
Water rods are not modeled in this code. But a reflooding
calculation without considering water rods is conservative
because quench front goes up slower and heat transfer from
fuel channels to water rods is neglected. It includes
“System momentum calculation”, “Thermal equilibrium
relative velocity correlation” and “Quench front velocity
correlation”. Various heat transfer correlations are prepared
according to the flow conditions such as single-phase
liquid, saturated two-phase, transient, dispersed and
superheated steam flow. The flow chart of the calculation is
shown in Fig. 11. This code was also validated by
comparing with the REFLA-TRAC code. The detail of this
code is explained in Ref. 3).
Time [s]
Start
Fig. 8: Scram reactivity curve
Reading calculation condition and blowdown output data
Downcomer water level and quench front level
Start
Input
Momentum conservation in RPV
Pressure assumption
Mass & Energy Conserv. in core nodes
Heat transfer
Intact Loop, ECCS
Flow rate and enthalpy at core top
Break flow
Cladding temperature
∆P change
Mass & Energy Conserv.
Flow balance check
System momentum equation – System (Upper plenum) pressure
No
Yes
Heat transfer
(Clad - Coolant, Fuel channel - Water rod)
Heat conduction in fuel rod
Time step change
No
Pressure at core water level
Time step change
No
Finish of Reflood
Yes
End
Reactivity and Power
Finish of Blowdown
Yes
Fig. 11: Calculation flow chart of reflood phase
End
VI. Sequence
Actuation conditions of the safety system are compared
with those of ABWR and PWR in Table 1. The reactor is
assumed to be tripped by “Flow rate low level 1” or “Core
Fig. 9: Calculation flow chart of blowdown phase
4
100
80
Ratio [%]
pressure low level 1” or “Containment pressure high”. The
high pressure AFS are assumed to fail. The characteristic of
the MSIV, which is the same as that of ABWR, is shown in
Fig. 12. If the MSIV are closed and the ADS are not
opened at cold-leg break LOCA, the coolant outlet at the
hot leg is closed. It means that one of the safety principles
“to keep coolant outlet open at hot leg” is not satisfied and
therefore the core flow rate and the core coolability are
significantly small. Thus, the ADS are opened without
delay by the same signal as that of the MSIV such as
“Flow rate low level 3” or “Core pressure low (level 2)” or
“Containment pressure high”, while in ABWR the ADS are
opened by “Water level 1” and “Containment pressure
high” with 30 s delay. It is the most important
characteristic of the LOCA sequence of the SCLWR-H.
The LPCI are actuated by the same signal as that of the
MSIV with 30 s delay due to the starting of the emergency
diesel generators. Two out of three LPCI units are assumed
to start at a pressure 0.8 MPa which is conservatively less
than the design pressure 1.0 MPa.
60
40
Closed
MSIV signal
20
0
0.0
0.5
1.0
1.5
2.0
2.5
3.0
Time [s]
Fig. 12 Characteristic of MSIV
Table 1: Actuation conditions of safety system in LOCA analysis
(Conditions with underlines are actually used in analysis.)
Safety system
ABWR
Reactor scram
Containment pressure high, or
Rapid decrease in core flow rate
Accumulator
High-pressure
ECCS
Low-pressure
ECCS
MSIV
ADS
Fail
Water level 1, or
Containment pressure high
(30 s delay)
PWR
Core pressure low, or
ECCS startup
Core pressure below 4 MPa
Containment pressure high, or
Core pressure low and pressurizer
water level low, or
Core pressure abnormally low
(32 s delay)
Containment pressure high, or
Core pressure low and pressurizer
water level low, or
Core pressure abnormally low
(32 s delay)
Water level 1.5
(no delay)
Water level 1, and
Containment pressure high
(30 s delay)
VII. LOCA analysis
The characteristics of the SCLWR-H analyzed here are
shown in Table 2. The plant parameters of the initial
condition (normal operation) are as follows:
a)
b)
c)
d)
e)
f)
core power 100% (2300 MWt)
core pressure 25.0 MPa
feedwater flow rate 100% (1190 kg/s)
main steam temperature 500°C
maximum cladding temperature 643°C
maximum linear power 39 kW/m
SCLWR-H
Flow rate low level 1, or
Core pressure low level 1, or
Containment pressure high
Fail
Flow rate low level 3, or
Pressure low level 2, or
Containment pressure high
(30 s delay)
Flow rate low level 3, or
Pressure low level 2, or
Containment pressure high
(no delay)
The limitation of the cladding temperature of LWR is
1260°C for stainless steels, which was obtained
considering metal-water reaction. In this study, the
limitation of the Ni-alloy cladding temperature is
determined to be the same as that of stainless steels. The
oxidation characteristics of Ni-alloy should be subject for
future study.
5
Table 2: Characteristics of SCLWR-H
Core
Core diameter / height [m]
Number of fuel assemblies
Coolant inlet / outlet temperature [°C]
Coolant density coefficient [dk/k/(g/cm3)]
Doppler coefficient [dk/k/°C]
Maximum linear power [kW/m]
3.6 / 4.2
96
280 / 500
0.2
-1.2×10-5
39
RPV and Main loop
Inner diameter / wall thickness / total height [m]
Volume of top dome / upper plenum / bottom dome / down comer [m]
Inner diameter of main feedwater line / main steam line [m]
Length of main feedwater line / main steam line [m] (1 loop)
Number of main loops
Fuel assembly
Fuel rod diameter / pitch [mm]
Cladding material / thickness [mm]
Water rod wall material / thickness [mm]
Number of fuel rods / water rods
Mass flux in fuel channel / water rod [kg/s/m2]
System
Core pressure [MPa]
Thermal / electric power [MW]
Thermal efficiency [%]
Feedwater flow rate [kg/s]
25.0
2300 / 1000
43.5
1190
cladding temperature is not so sensitive to the LPCI
capacity because the highest water level in the downcomer
is constant.
Table 3: Time sequence of 100% cold-leg break LOCA
0 s 100 % Cold leg break
0.1 Scram signal by “Pressure low level 1 (24.0MPa)”
MSIV/ADS/LPCI signal
by “Pressure low level 2 (23.5MPa)”
0.2 ADS opened
2.9 Scram completed
3.1 MSIV closed
42 Pressure 0.8 MPa
LPCI actuated
78 Start of reflooding phase
255 Highest cladding temperature 792°C (Reflooding)
500 Complete of reflooding phase
800
Pressure [MPa]
25
pressure
peak cladding temperature
20
700
15
600
10
500
5
400
o
(b) Reflooding phase
The reflooding phase starts from 78 s. The quench front
level, the downcomer level, the peak cladding temperature
and its axial position are shown in Fig. 15. The downcomer
is filled with the water from the LPCI at 112 s. The axial
position of the peak cladding temperature goes up with the
quench front level. The highest cladding temperature is
792°C at 255 s. The quench front reaches the core at about
500 s.
The effect of the LPCI capacity is shown in Table 5. If it
is smaller, the reflooding phase begins later. But the peak
10.2 / 11.2
Ni-alloy / 0.63
Ni-alloy / 0.20
300 / 36
1161 / 45
Temperature [ C]
1. Cold leg break LOCA
(a) Blowdown phase
The time sequence is shown in Table 3. The pressure
and the peak cladding temperature are shown in Fig. 13.
The flow rates and the reactor power are shown in Fig. 14.
Before the ADS are opened, the core flow rate is small
because a large quantity of high-density water in the top
dome and the water rods flows to the break without passing
through the core. After the ADS are opened, the core flow
rate is recovered and the cladding temperature decreases.
The reactor power is promptly decreased by density
feedback and scram. When the LPCI flow from the
suppression chamber reaches the core bottom at 78 s, the
blowdown calculation is finished. The highest cladding
temperature is 721°C.
The influences of various parameters are shown in Table
4. The peak cladding temperature is sensitive to the ADS
parameters such as the time delay and the number of valves
opened. But it still has a good margin compared with
1260°C even if the delay is a little longer and some of the
valves are not opened.
4.34 / 0.35 / 15
55 / 24 / 21 / 26
0.27 / 0.46
20 / 20
2
0
0
20
40
60
300
80
Time [s]
Fig. 13: Pressure and peak cladding temperature at 100%
cold leg break LOCA (blowdown phase)
6
hot-leg break LOCA is expected to be much less severe
than that of cold-leg break LOCA. Thus, only blowdown
phase is analyzed in this study.
The time sequence of 100% hot-leg break LOCA is
shown in Table 6. The pressure and the peak cladding
temperature are shown in Fig. 16. The flow rates and the
reactor power are shown in Fig. 17. The core flow rate is
significantly increased because the high-density water in
the RPV and the main feedwater lines flows through the
core to the break and the ADS. At the beginning the core
power is temporally increased by density feedback. But the
core flow rate is much larger. Thus, the cladding
temperature does not exceed that of normal operation. The
blowdown calculation is finished at 66 s.
200
150
Ratio [%]
100
50
0
Flow rate at core top
Flow rate at core bottom
Flow rate at water rod top
Reactor power
-50
-100
-150
0
2
4
6
8
10
Time [s]
Fig. 14: Flow rate and reactor power at 100% cold-leg
break LOCA (blowdown phase)
Table 4: Sensitivity analysis of 100% cold-leg break LOCA
(blowdown phase)
Break size [%] 30 50
PCT [oC]
705 708
PCT [oC]
0.1
0.5 1.0
2.0
3.0
5.0
100
721
10
15
30
721 778 861 955 999 1053 1136 1180 1284
Number of ADS opened
PCT [oC]
4
8
16
1026 882 721
800
4
600
3
500
2
400
1
0
600
15
500
10
400
5
300
0
0
10
20
30
40
50
60
200
70
Time [s]
Fig. 16: Pressure and peak cladding temperature at 100%
hot-leg break LOCA (blowdown phase)
o
Axial position [m]
5
Temperature [ C]
700
Quench front level
Downcomer level
Position of PCT
PCT
20
o
6
pressure
peak cladding temperature
Temperature [ C]
PCT: Peck Cladding Temperature
7
700
25
Pressure [MPa]
ADS delay [s]
70
722
Table 6: Time sequence of 100% hot-leg break LOCA
0 s 100 % Hot leg break
0.1 Scram signal by “Pressure low level 1 (24.0MPa)”
MSIV/ADS/LPCI signal
by “Pressure low level 2 (23.5MPa)”
0.2 ADS opened
2.9 Scram completed
3.1 MSIV closed
31 LPCI actuated (Pressure 0.8 MPa)
66 LPCI flow reaches core bottom.
PCT: Peak cladding temperature
100
200
300
400
300
500
400
Time [s]
300
Table 5: Influence of LPCI capacity
LPCI capacity [kg/s/unit]
100 150 300 500
When reflooding starts [s] 149 113 78
64
PCT [°C]
985 892 792 720
PCT: Peak cladding temperature
800
56
679
Ratio [%]
Fig. 15: Reflooding phase of 100% break cold-leg LOCA
100
0
2. Hot leg break
In the case of hot-leg break LOCA, the reflooding phase
is different from that of cold-leg break LOCA. Since the
cold leg pipes are isolated by the check valves, the coolant
outlet is only at the break and the ADS lines. It is a kind of
forced cooling by the LPCI. Thus, the reflooding phase of
Flow rate at core top
Flow rate at core bottom
Flow rate at water rod top
Reactor power
200
0
2
4
6
8
10
Time [s]
Fig. 17: Flow rates and reactor power at 100% hot-leg
break LOCA (blowdown phase)
7
VIII. Discussion
At normal operation the coolant density in the top dome,
which has a large volume fraction in the RPV, is the same
as that of the feedwater. The density in the water rods is
also high. Thus, the water inventory in the RPV of the
SCLWR-H is much larger than that without water rods. It
contributes to the high coolability at large break LOCA.
Fig. 18 and Fig. 19 respectively show the cladding
temperatures in blowdown and reflooding phase of 100%
cold-leg break LOCA analyzed in the past study3) without
considering water rods. The peak cladding temperatures of
the blowdown and the reflooding phase were respectively
800°C and 980°C with a LPCI capacity 805 kg/s/unit. In
this study they are about 720°C and 790°C with a LPCI
capacity only 300 kg/s/unit. But the core coolability is
small if the ADS are not opened. That’s why the ADS are
opened by “or” logic of signals without delay while in
ABWR they are opened by “and” logic with 30 s delay.
References
1) Y. Oka and S. Koshizuka, “Design Concept of
One-Through Cycle Supercritical-Pressure Light
Water Cooled Reactors”, Proc. 1st Int. Symposium on
Supercritical Water-cooled Reactors, Design and
Technology, Tokyo, Japan, Nov. 6-9, 2000, 1-22
(2000)
2) Y. Oka, S. Koshizuka, Y. Ishiwatari and A. Yamaji,
“Elements of design Consideration of Once-Through
Cycle, Supercritical-Pressure Light Water Cooled
Reactor”, Proc. Int. Conf. on Advanced Nuclear Power
Plants (ICAPP), Hollywood, Florida, June 9-13, 2002,
Sec. 3.04 (2002)
3) J. H. Lee, Y. Oka and S. Koshizuka, “Development of
a LOCA Analysis Code for the Supercritical Pressure
Light Water Cooled Reactors”, Ann. Nucl. Energy, Vol.
25, No. 16, 1341-1361 (1998)
4) “Decay energy release rates following shutdown of
uranium-fueled thermal reactors”, proposed standard
ANS-5.1-1971, American Nuclear Society (1971)
5) S. Koshizuka et al., “Large Break Loss of Coolant
Accident Analysis of a Direct-Cycle Supercritical
Pressure Light Water Reactor”, Ann. Nucl. Energy, Vol.
21, No. 3, 177-187 (1994)
Fig. 18: Cladding temperatures in blowdown phase of
100% cold-leg break LOCA analyzed in past study
Fig. 19: Cladding temperatures in reflooding phase of
100% cold-leg break LOCA analyzed in past study
IX. Conclusion
Large break LOCA of the SCLWR-H is mitigated by
MSIV, ADS and LPCI. The RPV structure with descending
flow water rods gives a large water inventory and makes
the core coolability high at LOCA. The highest cladding
temperature is only 792°C even though the LPCI capacity
is 300 kg/s/unit (37% of that in the past study). But it
should be noticed that ADS actuation is important at
cold-leg break LOCA.
8