Super LWR (SCWR)
Conceptual Design
Yoshiaki Oka
Professor
The University of Tokyo
UT-UCB Internet seminar December 8, 2006
Generation IV Reactor Concepts
SCWR
(Super LWR / Super FR)
VHTR
SFR
GFR
LFR
MSR
3
Features of Super LWR/Super FR
• Simple & compact plant systems
– No water/steam separation
– Low flow rate, high enthalpy coolant
• High temperature & thermal efficiency (500C, ～44%）
• Utilizations of current LWR and Supercritical FPP
technologies
– Major components are used within the temperature range of
past experiences
BWR
PWR
Supercritical FPP Super LWR/
（once-through boiler） Super FR
Evolution of boilers
and
supercritical fossil fired power
plant technologies
Circular
Boiler
LWR
Water tube boiler
Once-through
boiler
Super LWR (SCWR)
Evolution of boilers
Supercritical Water in the Power Industry
Coal-fired SC plants in the world and their performance
Country /
Region
U.S.A.
Japan
Eastern Europe
Western Europe
Other Countries
TOTAL
Number of SC
Units
149
108
123
53
29
462
Installed MW
106,454
67,900
51,810
29,310
13,520
268,994
Year
1993
1994
1995
1996
1997
Subcritical
82.0
83.8
83.7
86.6
88.5
Supercritical
89.8
83.0
84.7
79.5
90.3
Source: World Bank Organization
Most new coal-fired power plants are supercritical.
SC turbines are proven technology
Major vendors of SCW
components include
GE, Toshiba, Hitachi,
MHI, B&W, Siemens
Toshiba: 700 MWe
(24MPa, 593/593°C)
MHI: 1000 MWe
(24.5MPa, 600/600°C)
Density and specific heat
of supercritical water (25 MPa)
Super LWR
• Super LWR: Supercritical-Water-Cooled and
Moderated Reactor developed at Univ. of Tokyo
• Once-through direct cycle thermal reactor
•
•
•
•
Pressure: 25 MPa
Inlet: 280℃
Outlet (average): 500℃
Flow rate: 1/8 of BWR
Control rods
Supercritical water
500℃
Turbine Generator
280℃
Core
Condenser
Reactor
Heat sink
Pump
Scope of studies and Computer codes
1.Fuel and core
Single channel thermal hydraulics (SPROD), 3D coupled
core
neutronic/thermal-hydraulic
(SRAC-SPROD),
Coupled sub-channel analysis, Statistical thermal design
method, Fuel rod behavior (FEMAXI-6), Data base of
heat transfer coefficients of supercritical water
2. Plant system; Plant heat balance and thermal efficiency
3. Plant control
4. Safety; Transient and accident analysis at supercriticaland subcritical pressure, ATWS analysis, LOCA analysis
(SCRELA)
5. Start-up (sliding-pressure and constant-pressure)
6. Stability (TH and core stabilities at supercritical and
subcritical-pressure)
7. Probabilistic safety assessment
Fuel and Core design
Core design criteria
Thermal design criteria
 Maximum linear heat generation rate (MLHGR) at
rated power ≦ 39kW/m
 Maximum cladding surface temperature at rated
power ≦ 650C for Stainless Steel cladding
 Moderator temperature in water rods ≦ 384C (pseudo
critical temperature at 25MPa)
Neutronic design criteria
 Positive water density reactivity coefficient (negative
void reactivity coefficient)
 Core shutdown margin ≧ 1.0%ΔK/K
Fuel assembly design
Design requirements
Solution
Low flow rate per unit power (< 1/8 of LWR) due
to large ⊿T of once-through system
Narrow gap between fuel rods to
keep high mass flux
Thermal spectrum core
Many/Large water rods
Moderator temperature below pseudo-critical
Reduction of thermal stress in water rod wall
Uniform moderation
Control rod
guide tube
Insulation of water rod wall
Uniform fuel rod arrangement
ZrO2
Stainless Steel
UO2 fuel rod
UO2 + Gd2O3
fuel rod
Water rod
Kamei, et al., ICAPP’05, Paper 5527
Fuel enrichment
Fuel enrichment is divided into two regions to prevent top
axial power peak


Average fuel enrichment 6.11wt%
1.26m 5.9wt%UO2
4.2m
6.2wt%UO2
2.94m
(a) UO2 fuel rod
1.26m
5.9wt%UO2
+4.0wt%Gd2O3
6.2wt%UO2
4.2m
2.94m
+4.0wt%Gd2O3
(b) UO2 + Gd2O3 fuel rod
Coolant flow scheme
Flow directions
Coolant
Moderator
Inner FA
Upward
Downward
Outer FA
Downward
Downward
CR guide tube
To keep high average coolant outlet temperature
Outlet:
Inlet:
Inner
Kamei, et al., ICAPP’05, Paper 5527 FA
Mix
Outer
FA
3D Coupled Core calculation

Neutronic calculation: SRAC code system (JAERI)

Thermal-hydraulic calculation: SPROD (Univ. Tokyo)
3D Core burnup calculation
（1/4 core）
FA
FA segment
2-D FA burnup calculation
（1/4 FA）
SRAC and ASMBURN
Macro
cross
section
N -TH coupling
Single channel thermalhydraulic analysis（SPROD)
coolant
qc(i)
Calculation
mesh
Average
condition
Water rod wall
qw(i)
pellet
cladding
moderator
3-D N-T Coupled Core Calculation
• T-H calculation based on
single channel model
• Neutronic calculation; SRAC
Core consists of homogenized
fuel elements
3-D core calculation
Homogenized
Fuel
element
Single channel T-H model
Coolant
qc(i)
pellet
Cladding
qw(i)
Moderator
Water rod
wall
1/4
Fuel
Single channel
symmetric
assembly T-H analyses
core
Fuel load and reload pattern

120 FAs of 1st ,2nd and 3rd cycle fuels and one 4th cycle FA
3rd cycle FAs which have lowest reactivity are loaded at the
peripheral region of the core to reduce the neutron leakage

This low leakage core is possible by downward flow
cooling in peripheral FAs

1st cycle fuel
2nd cycle fuel
3rd cycle fuel
4th cycle fuel
(a) 1st → 2nd cycle
(b) 2nd → 3rd cycle
¼ symmetric core
(c) 3rd → 4th cycle
Coolant flow rate distribution
FA with
ascending flow
cooling
Flow rate to each FA is
adjusted by an inlet
orifice to match radial
power distribution

48 out of 121FAs are
cooled with descending
flow
0.4
0.4
FA with
descending flow
cooling
1.02 0.8
0.8
0.5
1.02 0.84
1.13
0.8
0.7
1.02 1.08
1.02
1.13
0.8
0.5
1.08 0.84
1.08
1.02
1.13
0.8
0.95 0.95
0.84
1.08
0.84
0.8
0.4
0.76 0.95
1.08
1.02
1.02
1.02
0.4

Relative coolant flow distribution
(1/4 core)
Control rod patterns

X : withdrawn rate (X/40)
Blank box : complete withdrawal (X=40)
At the EOC, some CRs are slightly inserted to prevent a high axial
power peak near the top of the core
Prevent a high MCST

12
32
16
32
24
24
24
32 0
32
3212
3216
24
32
24
24
32 0
32
0.0GWd/t
32
36 28 28
32 28
4 28
3632
4.4GWd/t
24
32
28
24
32 0
32
3224
0.22GWd/t
1.1GWd/t
24 36 32
28 36
4 24
20
28
36
28
28
32 0
32
28
3224
2.2GWd/t
32 4
32
3628
3.3GWd/t
36
36
5.5GWd/t
32
20
20
4
28
4
32
20
20
4
28
6.6GWd/t
28
32
7.7GWd/t
20
8.8GWd/t
36
36
24
28
24
24
16
28
24
24
32
24
4
28
9.9GWd/t
32
28
36
11.0GWd/t
32
32
28
12.1GWd/t
36
36
28
28
36
36
32
32
13.2GWd/t
32
28
36
36
32
14.3GWd/t
Power distribution
Axial power peaking factor is
1.25-1.45

Core averaged axial power
Radial power peaking factor is
1.24-1.39 (Only Upward flow
cooling region 1.12-1.22)

1.6
1.4
1.2
1.0
0.8
0.6
BOC(2.2～3.3Gwd/t)
MOC(7.7～8.8Gwd/t)
EOC(13.2～14.3Gwd/t)
0.4
0.2
0.0
0
1
2
3
4
Core height　(m)
Axial power distribution
MOC
BOC
2.2～3.3GWd/t
7.7～8.8GWd/t
EOC
13.2～14.3GWd/t
Radial power distribution (¼ core)
0
0.1800
0.8000
0.9000
1.000
1.150
1.250
1.350
1.500
Coolant core outlet temperature and Maximum
cladding surface temperature distribution

Coolant temperature of inner FA is 420-570C (average 500C)

Coolant temperature of peripheral FA is 350-530C
BOC
MOC
EOC
100.0
300.0
330.0
384.0
430.0
470.0
500.0
520.0
550.0
610.0
650.0
(a) Coolant outlet temperature distribution (1/4 core)
BOC
MOC
(b) Maximum cladding surface temperature distribution (1/4
EOC
100.0
300.0
330.0
384.0
430.0
470.0
500.0
520.0
550.0
610.0
650.0
MLHGR and MCST
MLHGR and MCST are kept below 39kW/m
and 650C throughout a cycle respectively
Thermal design criteria are satisfied
660
Maximum linear heat
generation rate (kW/m)
39
38
37
36
35
34
33
32
31
30
0
2
4
6
8
10
Burnups (GWd/t)
(a) MLHGR
12
14
Maximum cladding surface
temperature (C)
40
650
640
630
620
610
600
0
2
4
6
8
10
Burnups (GWd/t)
(b) MCST
12
14
Water density reactivity
coefficient (⊿K/K/(g/cc))
Water density reactivity coefficient and
Shutdown margin
Water density reactivity
coefficient is positive (negative
void reactivity coefficient)


Shutdown margin is 1.27 %dk/k
1
0.1
0GWd/t
15GWd/t
45GWd/t
0.01
0.0
0.2
0.4
0.6
0.8
Water density (g/cc)

All CR clusters are inserted except the maximum worth cluster

Fuel and coolant temperature are 30C

No Xe or other FP in the core
Neutronic design criteria are satisfied
1.0
Super LWR characteristics summary
Core
Super LWR
Core pressure [MPa]
25
Core thermal/electrical power [MW]
27441200
Coolant inlet/outlet temperature [C]
280/500
Thermal efficiency [%]
43.8
Core flow rate [kg/s]
1418
Number of all FA/FA with descending flow cooling
121/48
Fuel enrichment bottom/top/average [wt%]
6.2/5.9/6.11
Active height/equivalent diameter [m]
4.2/3.73
FA average discharged burnup [GWd/t]
45
MLHGR/ALHGR [kW/m]
38.9/18.0
Average power density [kW/l]
59.9
Fuel rod diameter/Cladding thickness (material) [mm]
10.2/0.63 (Stainless
Steel)
Thermal insulation thickness (material) [mm]
2.0 (ZrO2)
Sub-channel analysis coupled with
3D core calculation
Principle for Preventing Cladding Failures
• Super LWR: no boiling, limit cladding temperature
BWR, PWR
Normal
Sufficient
operation margin to BT
Abnormal
transient
No BT
Super LWR
No creep rupture1)
(Design limit temperature for normal
operation)
No plastic strain & no buckling collapse2)
(Design limit temperature for abnormal
transient)
Accurate evaluation of the peak cladding
temperature is essential
1) A. Yamaji, Y. Oka, J. Yang, et al., “Design and Integrity Analyses of the Super LWR
Fuel Rod.,” Proc. Global2005, Tsukuba, Japan (2005)
2) A. Yamaji, Y. Oka, Y. Ishiwatari, et al., “Rationalization of the Fuel Integrity and
Transient Criteria for Super LWR,” Proc. ICAPP’05, Seoul, Korea (2005)
Characteristics of supercritical water cooling
• Large axial Δρ ⇒ induces flow distribution within FA
• High sensitivity of (ΔT/ΔH) near the core outlet
Subchannel analysis is required to accurately determine
Maximum Cladding Surface Temperature (MCST)
–Underestimated MCST
Temperature [℃]
–T-H calculation based on
single channel model
Inlet
Outlet
600
800
Temperature
700
550
Density
600
500
500
450
400
400
300
350
200
300
100
250
0
1.0 1.5 2.0 2.5 3.0 3.5
Specific enthalpy [104J/kg・K]
Density [kg/m3]
•3-D core calculation:
Evaluation of Max. peak cladding temperature
DNBR 1.0 (95/95) → 1.17
Failure limit
Margin
DNBR 1.3
Overpower/
Over temperature
(abnormal transients)
DNBR 1.72
Engineering
uncertainties
Applicable local
flux factor
Applicable radial
and axial
flux factor
Limit for design
transients (95/95)
Maximum peak steady
state condition (95/95)
Nominal peak
steady state condition
Nominal steady state
core average condition
25MPa, outlet 500℃... etc
Plant safety
analyses
Statistical
thermal design
Subchannel
analyses
650℃
3-D core
calculations
Yamaji, et al., Global 2005, Paper 556
Method: Calculation Flow
3-D core calculations
(Homogenized FA model)
Core power, burnup,
density, distribution
CR patterns
N&T-H
coupled subchannel analyses
Pin power distribution
f(burnup history,
density, CR insertion)
Reconstructed pin power distribution
Subchannel analysis
Peak cladding surface temperature
Subchannel Analysis Code
• Subchannel analysis code for Super LWR FA
developed at the Univ. of Tokyo1)
• 3 type of subchannels
– Type A: corner of WR
– Type B: side of WR
– Type C: corner of FA
• WR thermally insulated (ZrO2)
• Heat transfer correlation: Watts
• 2.5% neutron
heating assumed
(1) T. Tanabe, S. Koshizuka, Y. Oka, et al., “A subchannel analysis code for Supercritical-pressure
LWR with Downward-flowing water rods,” Proc. ICAPP’04, Pittsburgh, PA, USA (2004)
Basic Characteristics of Super LWR FA
3 types of subchannels and fuel rods
Heat transfer to WR
(Wetted length/ Heated length)
A
Low
Small (1.43)
B
High
Large (1.70)
C
Mid
V.Large (2.53)
Type “B”
Type “A”
Uniform
outlet
temperature
Power
Temperature
Fuel rod number
Coolant temperature
Type “C”
Relative fuel rod power
Type
Pin power
(Thermal neutron
flux)
Reconstruction of pin power distributions
Core power distributions
(3-D core calculations)
Coupled subchannel analyses
Pin power distribution
f(burnup history,
density, CR insertion)
Height [m]
Homogenized
FA
Normalized power
Reconstructed pin power distribution
Mass Flux and MCST Distributions
[kg/m2sec]
1120
1060
1000
947
889
832
[kg/m2sec
]
1130
1110
1080
1050
1020
[kg/m2sec
]
990
964
938
912
887
861
992
Mass flux distributions
[℃]
732
676
621
565
509
[℃]
709
665
622
578
[℃]
730
680
631
535
491
531
481
453
FA-a2
(Large gradients)
MCST distributions
FA-b
(Gadolinia rods)
581
FA-c
(CR withdrawal)
Comparison of Single Channel
Analysis and Subchannel Analysis
• MCST predicted by subchannel analysis is about 58℃
higher than that predicted by single channel analysis
• Large power gradient inside FA should be avoided by
design
MCST ΔT from single channel analysis to subchannel analysis
FA-a1
(large power
gradient)
44℃
FA-a2
(large power
gradient)
58℃
FA-b
(Gadolinia
rods)
49℃
FA-c
(CR withdrawal
at EOC)
46℃
Peak Cladding Surface Temperature
Peak cladding
surface temperature
Failure limit
Limit for
design transients
Maximum peak
steady state condition
Nominal peak
steady state condition
Nominal peak steady
state condition
(Homogenized FA)
Nominal steady state
core average condition
Criterion: ?
Plant safety
(ΔT4= ? )
analyses
Statistical
(ΔT3=?)
thermal
design
Subchannel
(ΔT2=58℃)
analyses
?
?
708℃
650℃
3-D core
calculations
(ΔT1=150℃)
Ave. outlet:500℃
Statistical Thermal Design
• Taking uncertainties into evaluation of peak
cladding temperature
Methods to evaluate the engineering uncertainty
 Classification:
(1) The direct method:
All uncertainties are set at their worst values and occur at the
same location and at the same time.
Traditional and conservative.
(2) The traditional way by using hot spot and hot channel factors:
(a) The deterministic method by using factors.
(b) The statistical method by using factors.
(c) The semi-statistical method:
Two groups of uncertainties: direct and statistical factors.
The factors are evaluated separately and combined statistically.
(3) The statistical thermal design method:
System parameters uncertainties are combined statistically.
Uncertainties of nuclear hot factors are considered statistically.
Engineering hot spot factors are used in a statistical way.
Uncertainties considered in statistical design procedure
system parameter
uncertainties
Inlet coolant temperature uncertainty
Core power uncertainty
Inlet coolant flow rate uncertainty
Core pressure uncertainty
Uncertainty of the ratio of the flow rate in
water rods to the whole flow rate
hot factor uncertainties
Nuclear hot factor
Engineering hot spot factor
Heat transfer correlation uncertainty
Nuclear and engineering hot factors
Nuclear hot factors are used to consider the power distribution in
the hot assembly, in the hot channel and at the hot spot.
Factors
Definition
f Rn
The radial nuclear hot assembly factor. The ratio of the hot
assembly power to the average assembly power
The axial nuclear hot assembly factor. The Ratio of the max
planar power to the average planar power in the hot assembly
The nuclear hot spot factor. The ratio of the maximum
linear heat flux to the average linear heat flux in the core
f Zn
f Pn
Engineering hot spot factors are used to consider some engineering
uncertainties, such as manufacturing tolerances, data uncertainties, etc.
Factors
Definition
fl e
The coolant temperature rise engineering hot spot factor
f cse
The cladding surface temperature rise engineering hot spot factor
Engineering sub-factors
Sub-factors
Uncertainties explained by sub-factors
Nuclear data
Nuclear properties of the fuel rods
Power distribution
Shapes of power distribution in hot assembly
Fissile fuel content tolerances
Enrichment and amount of fissile material
Inlet Flow maldistribution
Assembly hydraulic resistance and orifice
uncertainties
Flow distribution calculation
Intra-assembly flow maldistribution
Subchannel flow area
Geometric tolerances of fuel rod diameter and pitch
on sub-channel flow area
Pellet-cladding eccentricity
Eccentric position of fuel pellet within cladding
Coolant properties
Data of coolant properties
Cladding properties
Thickness and thermal conductivity of cladding
Gap properties
Conductivity of gap between fuel and cladding

Monte Carlo statistical procedure
Step 1: To sample all parameters and hot factors uncertainties
following their distributions and to combine them into groups.
Step 2: For each group of samples, sub-channel code is used
to calculate the MCST of the core.
Step 3: The MCST distribution is analyzed to get the standard
deviation of system parameters and hot factors  PF . The
correlation uncertainty  C is evaluated. Total uncertainty is:  T
2
 T2   PF
  C2
Step 4: The engineering uncertainty is evaluated as: kσ T
Monte
 PF
Carlo statistical procedure
C
2
 T2   PF
  C2
Engineering uncertainty
is evaluated as: kσT
k=1.645 is to ensure 95/95
limit.
Sub-channel area uncertainty
- Fabrication tolerance, Cladding strain, etc.
Monte Carlo Process
5000 groups of samples of
locations and diameters of all
fuel rods in the 1/8 hot assembly
Calculation
conditions
Results of MCST distributions
by Monte Carlo procedure
Distributions of MCST of case 1 (Left) and case 2 (right)
Case 1: system parameters are sampled as normal distributions
Case 2: system parameters are sampled as uniform distributions
Statistical characteristics of MCST distributions
Case 1: system parameters are sampled as normal distributions
Case 2: system parameters are sampled as uniform distributions
Thermal margin for engineering uncertainty
Standard deviation of system parameter uncertainty
and hot factor uncertainty
 PF  18.32C
Standard deviation of correlation uncertainty
 C  6.33C
Engineering uncertainty:
1.645 
2
PF
   31.88C
2
C
Peak Cladding Surface Temperature
Peak cladding
surface temperature
Failure limit
Limit for
design transients
Maximum peak
steady state condition
Nominal peak
steady state condition
Nominal peak steady
state condition
(Homogenized FA)
Nominal steady state
core average condition
Criterion: ? ℃
Plant safety
(ΔT4= ? ℃)
analyses
Statistical
(ΔT3=32℃)
thermal
design
Subchannel
(ΔT2=58℃)
analyses
?℃
740℃
708℃
650℃
3-D core
calculations
(ΔT1=150℃)
Ave. outlet:500℃
Mechanical strength requirements for
fuel cladding
at normal operating condition
FEMAXI-6 Fuel Analysis Code Models
• Fuel analysis code for LWR
• Thermal and mechanical coupled analysis
• Temperature, FP gas release, FEM calculations for each segment
• SUS304 for
cladding
• Irradiation
creep and
corrosion by
oxidation are
not
calculated
Basic Fuel Rod Behavior
103
600℃
650℃
700℃
102
10
PNC1520 PNC316
600℃
650℃
700℃
750℃
10
102
103
750℃
104
Cladding primary
membran stress [MPa]
Creep rupture strength [MPa]
• Peak cladding centerline temperature: 757℃ at Segment No. 9
(peak cladding surface temperature: 740℃)
• Creep rupture strength becomes most limiting at high temperature
• Primary membrane stress needs to be evaluated
• Compressive (BOL) ⇒ tensile (EOL) due to PCMI
105
80
60
40
20
0
-20
-40
-60
-80
0
Segment no. 8
Segment no. 9
Segment no. 10
10 20 30 40 50 60
Fuel rod average burnup[GWd/t]
Time to rupture [h]
Creep rupture strength of advanced SS (JNC)
Primary membrane stress on cladding
Yamaji, et al., Global 2005, Paper 556
Fuel rod designs and cladding hoop stresses
• Placing gas plenum in lower part of fuel rod is effective
in reducing stress (required gas plenum volume reduced
to 1/5)
Constraints on Gas plenum
Pinternal
position
Less than
Pcoolant
Upper part
Lower part
Upper part
No
constraints
Lower part
Gas plenum /
fuel vol. ratio
Clad hoop stress (at
seg. No. 9) [MPa]
0.1
-140～+129
0.5
-118～+100
0.1
-96～+100
0.5
-75～+59
0.1
-93～+97
0.5
-66～+68
0.1
-64～+71
0.5
-34～+36
(“Zero PCMI” design)
SCWR IEM Tsukuba, Oct.12, 2005
Mechanical Strength Requirements for Fuel
Rod Cladding (at normal operation)
• Requirements:
– Life time：36,000～48,000 hours（500days×3～4）
– Peak temperature： 757℃（centerline）/ 740℃ (surface)
Primary membrane stress on the cladding
(plenum/fuel) Plenum position Constraint on Pinternal
vol. ratio
0.1
Bottom
No constraints
0.1
Bottom
Less than Pcoolant
Stress
-64 to +71
-96 to +100
• Requirements exceed mechanical strengths of commercially available
stainless steels (e.g. SUS316)
• Advanced Austenitic Stainless Steel of JNC（PNC1520）almost meets
the requirement
- Creep rupture strength of approx. 90MPa (750℃, 10,000hours)
Safety
Super LWR design features
Once-through cooling
⇒
No circulation
Low flow rate (large ΔT)
Supercritical-pressure
⇒
No phase change
No boiling transition
No water level
Downward-flow water rods ⇒ Large volume fraction
Heat sink
Large coolant in top dome
water level circulation
outlet
inlet
BWR
PWR
Super LWR
Safety principle of Super LWR
• Keeping coolant inventory is not suitable due to no water level and
large density change.
• Coolant inventory is not important due to no circulation.
• No natural circulation
Safety principle is keeping core coolant flow rate.
Coolant supply (main coolant flow rate)
Coolant outlet (pressure)
BWR
Requirement RPV inventory
Monitoring
RPV water
level
PWR
Super LWR
PCS inventory
Core flow rate
Pressurizer
water level
Main coolant flow rate,
Pressure
Summary of transient analysis
Peak pressure [MPa]
O
Peak cladding temperature [ C]
29
850
800
Criterion
750
700
650
600
300
250
Criterion
28
27
26
25
1
1
2
3
4
5
6
7
8
9 10 11
Transient number
1. Loss of feedwater heating
2. Inadvartent startup of AFS
3. Partial loss of main coolant flow
4. Loss of offsite power
5. Loss of load with turbine bypass
6. Loss of load without turbine bypass
7. MSIV closure
8. CR withdrawal from hot-standby
9. CR withdrawal from full power
10. Main coolant flow control system failure
11. Pressure control system failure
2
3
4
5
6
7
8
9
10 11
Transient number
Highest cladding temperature: 730℃
= initial temp. + 90℃
Highest pressure: 26.8MPa
= 25MPa×1.07
(ABWR: 7.07MPa×1.12)
No plastic deformation of cladding
Very shot duration of high cladding temp.
over 700℃: < 1s, over 670℃: < 5s
O
Peak cladding temperature [ C]
Summary of accident analysis
1200
Criterion
1. Total loss of reactor coolant flow
2. Reactpr coolant pump seizure
3. CR ejection at full power
4. Large LOCA
5. Small LOCA
1000
800
960
830
790
780
700
600
1
2
3
4
5
Accident number
“Containment pressure high” signal at Small LOCA (12% cold-leg break):
Not considered (like ABWR) ⇒ late ADS (7.6 s) ⇒ PCT= 960℃
Considered (like PWR) ⇒ early ADS (1.0 s) ⇒ PCT< 800℃
LOCA cladding temperature sensitive to ADS timing (ADS delay)
1300
1200
1 Loss of offsite power
2 Partial loss of reactor coolant flow
3 Loss of load without turbine bypass
4 Loss of load with turbine bypass
5 MSIV closure
6 CR withdrawal at normal operation
7 Main coolant flow control system failure
8 Pressure control system failure
900
800
700
600
1
2
3
4
5
6
Event number
7
8
Peak pressure [MPa]
Criterion: 1260 C
No depressurization
Depressurization (5s after scram signal)
Without scram
With scram
31
O
O
Peak cladding temperature [ C]
Summary of ATWS analysis
Criterion 30.3 MPa
1 Loss of offsite power
2 Partial loss of reactor coolant flow
3 Loss of load without turbine bypass
4 Loss of load with turbine bypass
5 MSIV closure
6 CR withdrawal at normal operation
7 Main coolant flow control system failure
8 Pressure control system failure
30
29
28
27
26
25
1
2
3
4
5
6
Event number
Highest cladding temperature: 890℃ (initial temp. + 250℃)
Highest pressure: 26.8MPa = 25.0MPa × 1.07
(ABWR: 7.07MPa × 1.27)
Maximum reactor power: 150% (SCLWR-H)
450% (ABWR)
7
8
Start up
Sliding Pressure Startup System
Sliding pressure supercritical water-cooled reactor
The reactor starts at a subcritical pressure, which increases
with load.
100
500
90
450
80
core outlet temperature
400
Ratio (%)
70
60
50
40
350
feedwater temperature
feedwater
flow rate
30
300
250
main steam
pressure
20
10
200
150
100
reactor power
50
0
start of start of turbine pressurization
feedwater nuclear startup
pump
heating
0
line
temperature
switching raising
power
raising
Temperature o(C) / Pressure (bar)
Sliding Pressure Startup Curve
Stability
Thermal-hydraulic stability
Coupled N-TH stability
supercritical and sub-critical
pressure (start-up)
Plant System
CR drive
Upper dome
Upper
plenum
RPV
Turbine control valve
Turbine bypass valve
Exit valve
Active core
Fuel channel
Water rod
Downcomer
Lower dome
Feedwater pump
HP heaters
LP heaters
Stability Criteria
The same stability criteria for
Decay ratio = y2/y1
y(t)
y1
BWR are used for Super LWR.
steady-state
t1
0
y2
t2
t
time (t)
Normal operating conditions
All operating conditions
Thermal-hydraulic
stability
Decay ratio  0.5
(damping ratio  0.11)
Decay ratio < 1.0
(damping ratio > 0)
Coupled neutronic
thermal-hydraulic
stability
Decay ratio  0.25
(damping ratio  0.22)
Decay ratio < 1.0
(damping ratio > 0)
Sliding pressure startup curve
(Thermal criteria only)
(Both Thermal and Stability
criteria)
500
100
500
90
450
90
450
core outlet temperature
400
Ratio (%)
70
60
50
40
350
feedwater temperature
feedwater
flow rate
30
300
250
main steam
pressure
20
10
200
150
100
reactor power
50
0
start of start of turbine pressurization
feedwater nuclear startup
pump heating
0
line
temperature
switching raising
power
raising
80
main steam temperature
400
70
Ratio (%)
80
Temperature o(C) / Pressure (bar)
100
60
50
350
feedwater temperature
feedwater flow rate
40
300
250
main steam
pressure
200
30
150
20
100
10
core power
50
0
start of start of turbine pressurization line
feedwater nuclear startup
switching
pump
heating
0
power-raising
Temperature o(C) / Pressure (bar)
Sliding pressure startup curve
Economic potential
1
43m
SCLWR-H(1700MWe)
ABWR(1350MWe)
PWR(1100MWe)
PWR(1100MWe)
Comparison of containments
Improvement of 1700MWe Super
LWR from 1350MWe ABWR
Super LWR ABWR improvement in %
Thermal efficiency, %
44.0
34.5
28%
RPV weight, t
750
910
18%
CV volume, m3
54%
7900
17000
Steam line number
2
4
50%
Turbine speed, rpm
3000*
1500*
50%
2
3
33%
Condenser
*3600rpm and 1800rpm in the western Japan
Super fast reactor (FR)
• Once-through coolant cycle is compatible
with tight lattice fast core because of high
head pump and low core flow rate. (Less
design limitation than high conversion
LWR)
• Super LWR can be a fast reactor with the
same plant system.
Research and Development
Status and Collaboration and Related Researches
1989: Started at The University of Tokyo ; Conceptual design
1998: JSPS-Monbusho funding (University of Tokyo) ; Pulse radiolysis and heat transfer
1998: US DOE-NERI funding (ANL); water-chemistry (pulse-radiolysis) experiments
2000: HPLWR program by EC (FZK, CEA, Framatome, VTT, PSI, KFKI and Univ. of Tokyo)
2001:Japanese NERI of METI funding (Toshiba, Hitachi, Univ. of Tokyo, Kyushu Univ. Hokkaido
Univ.) Thermal hydraulics, materials screening, plant concept
2001:US DOE-NERI funding, 3 programs; (INEEL, WH, Univ. Michigan, Univ. of Tokyo, Univ. of
Wisconsin, GNF, SRI International, VTT) Corrosion, thermal hydraulics, design studies
2002: Japanese NERI of MEXT (Univ. of Tokyo, CRIEPI, JAERI, Toshiba, Hitachi) Water Chemistry
2002: US Generation 4 reactor (only one among water cooled reactors)
2004: I-NERI between Japan and USA material/corrosion
2005: Japanese NERI of MEXT (Univ. of Tokyo, JAEA, Kyushu Univ. TEPCO) Super Fast Reactor
2006: HPLWR 2nd phase (Europe), Heat transfer (KAERI)
References of Univ. Tokyo studies
• SCR2000 symposium, Univ. Tokyo,(2000)
first symposium of SCR
• J.Nucl. Sci.Technol. pp.1081-,(2001),AESJ
summary of design
• ICAPP’02; review of past concepts
• GENES4/ANP2003; 8 papers; overview,
fuel/core, safety, LOCA, control, start-up, stability
• Global’03; 4 papers; overview, core, ATWS and stability
• ICAPP’03, ICAPP’05 papers
• J. Nucl.Sci. Technol. papers