Research and Development of Supercritical-pressure light water cooled reactors, Super LWR and Super FR Yoshiaki Oka Professor Department of Nuclear Engineering and Management, School of Engineering, University of Tokyo, Japan Presentation includes the results of “Research and Development of the Super Fast Reactor” entrusted to the University of Tokyo by the Ministry of Education, Culture, Sports, Science at Technology of Japan (MEXT). IAEA International Conference on Opportunities and Challenges for Water Cooled Reactor in the 21st Century , Vienna, Austria, October 27-30, 2009 IAEA-CN-164-18KS
Transcript
Research and Development of Supercritical-pressure light water
cooled reactors, Super LWR and Super FR
Yoshiaki Oka
ProfessorDepartment of Nuclear Engineering and Management,
School of Engineering, University of Tokyo, Japan
Presentation includes the results of “Research and Development of the Super Fast Reactor” entrusted to the University of Tokyo by the Ministry of Education, Culture, Sports, Science at Technology of Japan (MEXT).
IAEA International Conference on Opportunities and Challenges for Water Cooled Reactor in the 21st Century , Vienna, Austria, October 27-30, 2009
IAEA-CN-164-18KS
2
Outline1. Introduction2. Fuel and core design3. Safety4. Fast reactor5. R&D
3
Change of density and specific heat of water with temperature at supercritical pressure (25 MPa)
4
Control rods
Supercritical water
Turbine Generator
Condenser
PumpHeat sink
Reactor
Core280℃
500℃
Super LWR• Super LWR: Supercritical-pressure light 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
5
Circular Boiler
Water tube boiler
Once-through boiler
LWR
Super LWR, Super FR (SCWR)
Evolution of boilers
Water level
Water level
Supercritical fossil-fired power plantsOnce-through boilersNumber of units are larger than that of LWRs.Proven technologies; turbines, pumps, piping etc.USA; developed in 1950’s, Largest unit is 1300MWe. Japan; deployed in 1960’s and constantly improved.Many plants in Russia and Europe.
Compact SC turbine (700MWe,31.0MPa,566℃)
7
Features of Super LWR/Super FR•
Compact & simple plant systems; Capital cost reduction–
No steam/water separation and no SGs: Coolant enthalpy inside CV is small.
–
High specific enthalpy & low flow rate: Compact components
•
High temperature & thermal efficiency (500C, ～ 44%）
•
Utilize LWR and Supercritical FPP technologies:-
Temperatures of major components below the experiences
•
Same plant system between thermal and fast reactor Supercritical FPP
（once-through boiler）Super LWR/ Super FRBWR PWR
8
Fuel and core design
9
Core design criteriaThermal design criteria
Maximum linear heat generation rate (MLHGR) at rated power ≦ 39kW/mMaximum cladding surface temperature at rated power ≦ 650C for Stainless Steel claddingModerator temperature in water rods ≦ 384C (pseudo critical temperature at 25MPa)
Neutronic
design criteriaPositive water density reactivity coefficient (negative void reactivity coefficient)Core shutdown margin ≧ 1.0%ΔK/K
Capacity:AFS TD 3 units: 50kg/s/unit (4%)* at 25MPaLPCI/RHR MD 3 units: 300kg/s/unit (25%)* at 1MPaSRV/ADS 8 units: 240kg/s/unit (20%)* at 25MPa
Configuration: AFSLPCI
AFS AFSLPCI LPCI
Safety system design
*100% corresponds to rated flow rate
31
Water rods mitigate loss-of-flow events.
Under loss-of-flow condition:Heat conduction to water rods increases. →
“Heat sink” effect
Water rods supply their inventory to fuel channels due to thermal expansion. →
“Water source” effect
Water rod
- 80
- 60
- 40
- 20
0
20
40
60
80
100
0 10 20 30 400
100
200
300
400
500C r it er ion
A verage c hannel in le t f low rat e
Incr
ease
of
tem
pera
ture
fro
m in
itia
l val
ue (
℃)
Inc rease o f ho t t es t c ladding t em perat ure
T im e [ s ]
Wat er r od average dens it y
Wat er r od bo t t om f low rat e
M ain c oo lant +A FS f low rat eH ot c hanne l in le t f low rat e
Wat er r od t op f low rat e
P ow er
Pow
er, flow rate and density (% of initial value)
Total loss of reactor coolant flow
⊿MCST≃250℃
32
Alternative action is not necessary under ATWS conditions (Super LWR)
- 0 .04
- 0 .02
0.00
0.02
0 100 200 300 400 500 600 700- 100
0
100
200
300
400
500
Wat er r od f low rat e (t op)
C r it e r ion o f t em perat ure inc rease
M ain c oo lan t + A FS f low rat e
Hot c hanne l in le t f low rat e
Inc rease o f ho t t estc ladd ing t em per at ure
P ow er
T im e [ s ]
Incr
ease
of
tem
pera
ture
fro
m in
itia
l val
ue [
℃]
Pow
er, f
low
rat
e an
d de
nsit
y [%
]
Reac t iv it y o f dens it y feedbac k
Reac t iv it y o f D oppler feedbac k
Net reac t iv it y
Reactivity [dk/k]
- 0 .0015
- 0.0010
- 0.0005
0.0000
0.0005
24
26
28
30
0 1 2 3 4 50
20
40
60
80
100
120
A verage c hannelin le t f low r at e
P ow er
T im e [ s ]
Pow
er, f
low
rat
e an
d de
nsit
y [%
]
C r it er ion o f p ressure
P ressure
Pressure [M
Pa]
Reac t iv it y o f densit y f eedbac k
Reac t iv it y o fD oppler feedbac k
Net reac t iv it y
Reactivity [dk/k] - 0 .010
- 0.005
0.000
0.005
0.010
0.015
0 50 100 150 200 2500
20
40
60
80
100
120
140
160
Inc rease o f ho t t es tc ladd ing t em perat ureM
ain c
oo lant f
low ra
t e
P ow er
T im e [ s ]
Incr
ease
of
tem
pera
ture
fro
m in
itia
l val
ue [
℃]
Pow
er, f
low
rat
e an
d de
nsit
y [%
]
C R r eac
t ivit y
Reac t iv it y o f dens it y f eedbac k
Reac t iv it y o f D oppler f eedbac k
Net r eac t iv it y
Reactivity [dk/k]
Analysis results for ATWS events without an alternative action
Loss of offsite power Loss of turbine load without bypass
Uncontrolled CR withdrawal at normal
operation
33
Good inherent safety characteristics of Super LWRWhy ATWS is mild?1. Small power increase by valve closure.
• flow stagnation mitigates density increase
• no void collapse2. Power decreases with core flow rate due
to density feedback.Good ATWS behavior without alternative
action inserting negative reactivity
34
Summary of safety analysis results
0
100
200
300
400
500
600
T rans ien t s A c c ident s A T WS w it hout alt ernat ive ac t ion A T WS w it h alt ernat ive ac t ion (A D S )
1 2 3 4 9 1 2 3 5 6 1 2 3 4 6 7 9
C r it e r ion fo r t r ans ien t
C r it e r ion fo r ac c iden t and A T WS
Ev ent num ber
Incr
ease
of
tem
pera
ture
fro
m in
itia
l val
ue [
℃]
100
120
140
160
180
200
8 3 7 6 9
C r it e r ion fo r pow err is ing rat e o f 0 .1 - 1%
C r it e r ion fo r pow er r is ing rat e o f 1 - 10%
C r it er ion fo r pow err is ing rat e of ov re 10%
T r ans ient num ber
Pea
k po
wer
[%]
25
26
27
28
29
30
31
2 3 4 8 9
T rans ient s A c c iden t s A T WS w it hout alt e rnat ive ac t ion A T WS w it h alt e rnat ive ac t ion (A D S)
3 4 2 3 4 7 8 9
C r it e r ion for t r ans ien t
C r it er ion fo r ac c iden t and A T WS
Event num ber
Pea
k pr
essu
re [
MP
a]
Transients Accidents1. Partial loss of reactor coolant flow2. Loss of offsite power3. Loss of turbine load4. Isolation of main steam line5. Pressure control system failure6. Loss of feedwater heating 7. Inadvertent startup of AFS8. Reactor coolant flow control system failure9. Uncontrolled CR withdrawal at normal operation10. Uncontrolled CR withdrawal at startup
1. Total loss of reactor coolant flow
2. Reactor coolant pump seizure3. CR ejection at full power 4. CR ejection at hot standby5. Large LOCA 6. Small LOCA
35
35
ΔMSCT for abnormal events
Nominal steady statecore average condition
Maximum peak steady state condition 3-D core design
Subchannel
analysisStatistical thermal design
MarginCriterion for transients
Failure limit for transient
240℃
520℃
Ave. outlet:500℃
850℃
1260℃
740℃
Criterion for accident
Failure limit for accidentMargin
60℃
Large LOCA
Small LOCA
Loss-of-flow
ATWS
380℃330℃
220℃250℃
110℃Transient
36
Summary of safety characteristics of Super LWR
• Core cooling by depressurization• Top dome and water rods serve as an “in-
vessel accumulator”• Loss of flow mitigated by water rods• Short period of high cladding temperature at
transients• Mild behavior at transients, accidents and
ATWS• Simple safety principle (keeping flow rate) due
to once-through cooling cycle
37
Super fast reactorTight fuel lattice Supercritical-pressure light water cooled fast reactor Same plant system as Super LWR
Control Rods
RPV
Turbine Bypass Valve
Turbine Control Valve
Condenser
LP FW Heaters
HP FWHeaters
Reactor Coolant Pump(Main Feedwater Pump)
Turbine
Containment
Deaerator
Condensate Pump
Booster Pump
MSIV
Plant system of Super LWR and Super FR
38
Advantages of Super Fast Reactor
Low reactor coolant flow rate due to high enthalpy rise High head pumps of the once-through direct cycle plant
Compatible with tight fuel lattice core of Super FR, a light water cooled fast reactor
No pumping power increase and instability problems of high conversion LWR
Same plant system as Super LWR, the thermal reactorFast reactors have higher power densities than thermal reactors due to no moderator necessary.
Making capital cost of Super FR lower than LWRs(Capital cost; Super FR< Super LWR< LWRs)
39R&D of Super Fast ReactorUniversity of Tokyo, JAEA, Kyusyu Univ. and TEPCOentrusted by MEXT as one of the Japanese NERI, 5 years, Dec. 2005-March 2010
Fuel pellet density 95%TDRod OD[mm] 7.0 5.5Pitch/ OD 1.16 1.19Cladding Material SUS304Thickness [mm] 0.43 0.4Effective heating length [cm] 300 200
CoreNo. of seed fuel assemblies 126 162
No. of blanket fuel assemblies 73
Pitch of FA 14.2 11.6
Core characteristics (700MWe)
42
42Thermal hydraulic experimentsKyusyu University ;HCFC22 (Freon) JAEA Naka-lab; Supercritical Water
Single rod 7-rod bundle
Heater rods and spacers
(1) single tube and 7-rod bundle
(2) critical heat flux near critical pressure
(3) critical flow and condensation
43
Wall temperature and heat transfer coefficient of 7-rod bundle test
Maximum wall temperature at critical heat flux
0
50
100
150
200
Wal
l tem
pera
ture
Tw° C
Tb
Tpc
hpc
HCFC22 Upward flow
P = 5.5 MPaG = 1000 kg/(m2 ·s)q = 40 kW/m2
200 250 300 350 400 450 5000
2
4
6
8
Hea
t tra
nsfe
r coe
ffici
ent α
kW/(m
2 ·K)
hpc
Dittus-Boelter
Bulk fluid enthalpy hb kJ/kg
-BundleⅠBundleⅡ
3.0 3.5 4.0 4.5 5.0 5.560
80
100
120
140
0.7 0.8 0.9 1.0 1.1
Max
imum
wal
l tem
pera
ture
T wm
ax°C
Reduced pressure P / Pc
HCFC22 Upward flow
G = 400 kg/(m2·s) q = 15 kW/m2
PcTsat
Dec. Inc.
Dec. Inc.
Pressure P MPa
BundleⅠ
BundleⅡ
Grid spacers
Bundle Ⅰ BundleⅡ
Grid spacer effect on heat transfer coefficients and critical heat flux
Experimental results; HCFC22(Freon)
44
0 10 20 30 40 50 60- 80
- 60
- 40
- 20
0
20
40
60
80
Pri
mar
y m
embr
ane
stre
ss [
MP
a]
F ue l r od ave. burnup [ GWd/ t ]
S egm ent no . 8 S egm ent no . 9 S egm ent no . 10
Time to rupture [h]
Cre
ep ru
ptur
e st
reng
th [M
Pa]
750℃700℃650℃600℃
10
102
103
10 102 103 104 105
600℃650℃700℃750℃
PNC1520PNC316
Need for Developing High Creep Strength Clad•
Max. stress on clad at peak T (700-750℃): 70-100MPa–
Exceed creep strength of SS for LWR (SUS316L)
–
Advanced SS for LMFBR (PNC1520) almost satisfies the requirement but SCC susceptibility, corrosion and neutron absorption properties need to be improved
•
High creep strength clad needs to be developed for Super FR
Creep rupture strength of advanced SSFuel rod analysis results
(Super LWR)
700-
750℃
45
Developed Good Thermal Insulator Yttria
stabilzed
zirconia
(YSZ)
•
Large ΔT (～250℃) •
Thermal insulator is required for:–
reduction of thermal stress
–
maintaining coolant temperature
0 100 200 300 400 500 600 700
100
200
300
400
500
σ ＞ S u
(1/ 2× S u)＜σ ＜ S u
σ ＜ (1/ 2× S u)
M id w all t em perat ure [ ℃ ]
ΔT
(℃
)
Thermal stress on the wall
No thermal insulation
Thermally insulated
Max. thermal stress
Insulated No insulation
Stainless steelHot
ColdT Hot
ColdT
Su: tensile strength
0
0.05
0.1
0.15
0.2
0 200 400 600 800 1000Th
erm
al c
ondu
ctiv
ity (W
/mK
) Temperature (oC)
Thermal conductivity of YSZ
~1/20 of Zirconia
46Experimental devices
Elution decreases with temperature (at 25 MPa)
0
0.05
0.1
0.15
0.2
0.25
0 100 200 300 400 500Time [hour]
300℃、25MPa Flow rate:1 L/h
Air free
Elu
tion
effic
ienc
y [g
/M2 ]
[O2 ]=200pb
[O2 ]=４00pb
Elution depends on O2
Elution of structural material in SC water
47
4747SCWR R&D in the worldJapan: University of Tokyo; Super LWR concept (since 1989), Super FR R&D (2005-2010). Toshiba; SCPR R&D, Consortium for GIF R&D
China; Shanghai JTU (8 organizations) SCWR R&D (2007-2012), CGNPC announced the plan of constructing an experimental SCWR from 2016.
EU; HPLWR phase 1 (FZK, 2000-2), phase 2 (FZK, 10 organizations of 8 countries 2006-9), planning of phase 3
Canada: pressure tube type SCWR R&D：NSERC/NRCan/AECL-Universities program
Korea: thermal hydraulics (KEARI)
Russia: SC thermal hydraulic loops of IPPE, WS at NIKIET in 2008
USA: TH and materials at Univ. Wisconsin and Univ. Michigan (finished)
GIF SCWR OECD/NEA (Canada, EU, Japan and other countries) phase 2
IAEA: CRP of supercritical thermal hydraulics
SCR symposiums; 1st and 2nd at University of Tokyo in 2000 and 2003, 3rd at Shanghai JTU in 2007 and 4th in Heidelberg in 2009
Sliding Pressure Startup Procedure(1) Start of Nuclear Heating(2) Turbine Startup(3) Pressurization to 25 MPa(4) Line Switching(5) Temperature Raising(6 ) Power Raising
54/124
Pressurization phase
8 10 12 14 16 18 20 22 240
100
200
300
400
500
600
700
800
0
10
20
30
40
50
60
70
80
90
100
Inlet flow rate
Core power
Inlet temperature
Average out let temperature
MCST
Tem
pera
ture
[℃
]
Pressure [MPa]
Rat
io (
%)
BT
•
Sliding pressure startup system (nuclear heating starts at subcritical pressure)
•
Clad temperature increase in pressurization phase is due to BT
•
Power / flow region is limited by CHF•
CHF may be increased by grid spacers
8 10 12 14 16 18 20 22 240
5
10
15
20
25
30
35
40
O perable region dur ing pressur izat ion phase
F low rat e = 35%In le t t em perat ure = 280 oC
M ax. allow able pow er M in. r equired pow er
Pow
er [
％]
P r essure [ M P a]
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Sliding Pressure Startup Curve
0
10
20
30
40
50
60
70
80
90
100
0
50
100
150
200
250
300
350
400
450
500
powerrais ing
t emperat ureraising
lineswit c hing
pressur izat iont urbinest ar t up
st ar t ofnuc learheat ing
st ar t offeedwat erpump
R
atio
(%)
r eac t or power
feedwat erf low rat e
main st eampressure
feedwat er t emperat ure
c ore out let t emperat ure
Tem
pera
ture
(o C
) /
Pre
ssur
e (b
ar)
(By Thermal-Hydraulic Analysis)
56/124
Linear Stability Analysis (for Supercritical Pressure)
