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ALFRED Project
R&D in support to the ALFRED Project M. Tarantino on behalf of Falcon Consortium ([email protected])
Biennal Conference, 17-18-19 March 2015, Brussels, Belgium
Overview
ALFRED Development: Main Issues Studies Ongoing Experimental Activities Ongoing Remarks
2
ALFRED Development: Main Issues
Structural Material
Coolant and Cover Gas Chemistry
Thermal Hydraulics
Reactor Components
Instrumentation
Fuel/Advanced Fuel
Neutronics
3
► Fuel assembly with extended stem (feasibility, structural integrity, support structure from the top or from the bottom of the FA, buoyancy, failed fuel detection).
► In-gas fuel handling machine (reliability, accuracy, shielding, fuel loading procedure).
► Spent fuel element cooling system (active and passive, reliability, behavior with blocked FA).
► Core arrangement integrity (core compaction prevention, radial expansion of core & feed back on reactivity).
► Control and shut-down rods (insertion from the top or from the bottom, reliability, actuation, speed of insertion).
► Cylindrical Inner Vessel (support, replacement, resistance to lead sloshing loads).
► Pump (Pressure head, reliability, inertia, bearings…)
Reactor Components
4
ALFRED Development: Main Issues
► Steam Generator (Design validation, material selection, component
behavior in forced and natural convection, tube rupture/leakage detection, tube rupture mitigation, reliability and performance assessment, replacement, SG type: spiral-tube, helical-tube, bayonet-tube).
► Reactor vessel (seismic assessment). ► Decay heat removal systems - Reactor Vessel Air-Cooling System (RVACS) (Functional design, sensitivity
to atmospheric conditions). - Dip Coolers (Water inventory, containment function). - Condensers on the main steam lines (Water inventory, reliability, need of
reliable discrimination of the SG with ruptured tube).
5
Reactor Components
ALFRED Development: Main Issues
► Corrosion in Lead (oxidation, dissolution). ► Embrittlement in Lead (reduction of fracture toughness and ductility, Fatigue, Creep,
standardization of LME tests). ► Irradiation effects on materials in lead (irradiation performance of candidate materials,
corrosion in Lead under irradiation, Irradiation embrittlement, irradiation creep, swelling).
► Material for short term deployment (austenitic steel such as AISI 316L and 15-15 Ti, DS stainless steel, ferritic martensitic such as grade 91 steels).
► Coatings development (FeAl, FeCrAlY, Tantalum coatings) ► High temperature materials for long term deployment (700-800°C, ODS steels, “MAX”
phase materials). ► Main components issues: - Fuel Cladding (irradiation effects, compatibility with coolant, material workability, component fabricability, oxide formation thickness, coating compatibility). - Material for pump impeller (high relative coolant velocity). - SG and HX (oxide formation thickness) - Reactor vessel (long lifetime, seismic loads) - Seismic isolators (qualification)
Structural Material
6
ALFRED Development: Main Issues
► Manufacturing constrains (Pellet specifications - density, grain size, homogeneity, porosity, pellet straightness, pellet mechanical stability, roughness - Clad specifications - wall thickness, wall thickness variation, straightness, toughness, roughness, cleanliness, corrosion resistance, heat treatment)
► Fuel pellets (FP release, fuel restructuring, densification and swelling) ► Fuel cladding (swelling, embrittlement, corrosion (coolant and fuel side), clad-FP interaction,
fatigue, creep) ► Fuel operating conditions and failure margin assessment (normal operation, operational
transients, accidental conditions) ► Failed fuel pin behaviour (release of gaseous FP, dissolution of FP in the coolant, loss of fuel
particles chemical reaction between fuel and coolant) ► Corium-coolant interaction (Fuel dispersion in coolant, chemical stability of corium)
Fuel/Advanced Fuel
For the existing fuel can be considered the experience gained in SFR (Phénix, Monju, Superphénix, etc..). For the new advanced fuels (MA mixed MOX fuel, nitride fuel, carbide fuel) all aspects related to fuel need to be re-assessed.
7
ALFRED Development: Main Issues
► Phenomena: - HLM pool thermal-hydraulics (flow patterns in forced convection, mixing & stratification, surface level oscillations, transition to natural circulation driven flow, natural circulation flow, etc..) - Water injection in lead (SGTR accident, rupture of DHR tubes) - Fuel Pin - coolant interaction (Flow induced vibrations) - FA blockage (Fuel Assembly damage) - Gas entrainment (two phase flow) - Coolant freezing (flow blockage) ► Modeling: - System thermal-hydraulic codes (V&V) - STH and neutronics (V&V) - CFD and mechanical code coupling (V&V) - STH code and CFD code coupling (V&V)
Thermal hydraulics
8
ALFRED Development: Main Issues
► Experiments: - Integral tests (including natural circulation and decay heat removal to
support the licensing process) - Fuel bundle (sub-channel analysis, heat transfer, cross-flow, pressure loss,
etc..) - Steam Generator ( functional tests, SGTR accident) - Lead pumps (functional tests) - FA cooling system at refueling (performance test) - Core cooling with total inlet flow blockage (SGs or downcomer frozen)
Thermal hydraulics
9
ALFRED Development: Main Issues
► Coolant control & purification (oxygen control, oxygen sensor reliability, coolant filtering, lead purification, lead cleaning from components)
► Cover gas control (radiotoxicity assessment of different elements, migration flow path into cover gas, removal and gettering)
► Temperature-level-pressure-flow-neutron flux measurements (improvement of reliability, failure mode investigation, irradiation effects)
► In service Inspection (ultrasound imaging & inspection, transducers development, image reconstruction technologies, sensors for in vessel inspection, sensor for reactor vessel inspection, inspection strategies, )
Instrumentation and chemistry control
10
ALFRED Development: Main Issues
Neutronics and reactor operation/control
11
► Validation measurements for nuclear data improvement (MA cross-sections, uncertainty reduction on cross-sections as Pb-MA-241Pu, 242Pu, etc..).
► Determination of flux gradients in fast spectrum. ► Reactivity effects (local-coolant void reactivity worth, density coefficients,
Doppler coefficient, dimension coefficients). ► Fuel neutronic performances (fuel utilization, spectrum evolution with burn-up,
delayed neutron fraction). ► Absorbers neutronic performances (boron carbide, europium…) ► Neutronic shielding (absorbing materials, moderating materials) ► Reactivity control (control rods). ► Shut down systems (active and passive, additional protection measures). ► Trip parameters (first and second) ► Neutron code development (Improvement & extension MCNPX, Deterministic
code set-up & validation) ► Neutron code validation experiments (Nuclear data improvement, integral
measurements & comparison with calculations)
ALFRED Development: Main Issues
ALFRED Development: Main Issues Studies Ongoing Experimental Activities Ongoing Remarks
Studies ongoing
Core Design Thermal Power 300 MW FAs 171
INN 57 OUT 114
Control Rods 12 Safety Rods 4 Dummies 110 Inner Vessel 1.65 m
13
Enrichment Inner core 21.8% wt% Enrichment Outer core 27.9% wt%
14
Lattice Triangular FA wrapped Pitch 13.86 mm Ranks 7 Pins 127 p/D 1.32 Diameter 10.5 mm
Core Design
Studies ongoing
Flow Blockage
Studies ongoing
• ANSYS CFX 13 • SST k- Menter (1994) • y+1 at the wall • Nnodes22·106 (160 axial) • t 1 ms (CFL 1) • qwall=1 MW/m2
qwall
15
Flow Blockage
Studies ongoing
velocity
Temperature
Unperturbed condition (case 0)
16
Central Blockage, Nblock=61, β=0.48, case 13, stationary Flow Blockage
Studies ongoing
Basic phenomenology addressed: recirculation vortex downstream the blockage Local phenomena dominant for ‘high’ blockage parameter β
Results indicate that a blockage of ~15% leads to a maximum clad temperature around 800 °C, and this condition is reached in a characteristic time of 3-4 s without overshoot. Local clad temperatures around 1000 °C can be reached for blockages of 30% or more. CFD simulations indicate that Blockages >15% could be detected by putting thermocouples in the plenum region of the FA.
17
high quality coatings custom process: bottom-up approach process at room temperature
FeCrAlY bonding layer + fully dense and compact topcoat
Pulsed Laser Deposition
18
Studies ongoing
Radiation tolerant nanoceramic coatings for LFR nuclear cladding
19
as-deposited cycled cycled
AISI 316L + 500 nm FeCrAlY + 1,3 μm bp-Al2O3
AISI 316L + 500 nm RF-FeCrAlY
+ 1,3 μm bp-Al2O3
AISI 316L + 500 nm PLD-FeCrAlY
+ 1,3 μm bp-Al2O3
25 cycles 350 °C – 650 °C – 350 °C @ 6°C/min
Thermal cycling
Studies ongoing
20
Corrosion: 500 h and 1000 h exposure to
oxidizing stagnant Pb @550 & 600 °C
F. Garcia Ferre et al. Corros Sci 77 (2013) 375
oxidizing Pb
Studies ongoing
21
Validation of the coupled calculations between RELAP5 STH code and ANSYS FLUENT CFD code
Studies ongoing
EXECUTE 1 TIME STEP OF
FLUENT TRANSIENT CALCULATIONS
END OF THE FLUENT
TIME STEP
WRITE FLUENT RESULTS
NEEDED AS B.C. FOR RELAP5
EXECUTE RELAP5 TRANSIENT
CALCULATIONS FOR 1 TIME STEP
TO RELAP
END OF
TRANSIENT ?
WRITE RELAP5 RESULTS
NEEDED AS B.C. FOR FLUENT
END OF RUN
Yes
No
START RELAP5 CALCULATION TO FIND
INITIAL STEADY STATE CONDITIONS Time step
Simulation
… 2012 Development of a Coupling tool between RELAP5 and FLUENT CFD codes.
… 2013 Coupled Numerical Simulations of Forced Circulation Tests performed in the framework of the NACIE experimental campaign.
… 2014 Improvements of the coupling tool and Code Validation.
160
Expansion Vessel
210100
110
120
125
130
146
148
Sep 150
156
152
170
172
180
510
TDV-500(Water in)
HX
200
206
1.5 m
5.7 m
2.35 m
0.765 m
7.5 m
1 m
TDJ-505
TDJ-405
TDV-520(Water out)
TDV-320(Argon out)
TDV-400 (Argon in)
5 m
0.89 mFPS
Section simulated by Fluent code 1.1 m
22
Studies ongoing
Experimental Mass Flow Rate overestimated by less than 2%
FPS Inlet and Outlet temperatures
LBE Mass Flow Rate
0 0.1 0.2 0.3 0.4 0.5 0.6 0.70
1
2
3
4
5
6
7
8
9
10LBE mass flow rate
Time [h]
LB
E m
ass flo
w r
ate
[kg
/s]
Balance Eq.
RELAP5
RELAP5+FLUENT 2D
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7250
260
270
280
290
300
310
320
330
340
350T FPS
Time [h]T
em
pe
ratu
re [°C
]
Tout
Exp
Tout
RELAP5
Tout
RELAP5+FLUENT 2D
Tin
Exp
Tin
RELAP5
Tin
RELAP5+FLUENT 2D
TEST 301-NC RELAP5 stand-alone simulation Coupled simulation (CFD 2D axialsymmetric domain)
23
Studies ongoing
Implicit Coupling scheme Explicit coupling scheme
24
Studies ongoing
24
0 1 2 3 4 5 6 7 80
2
4
6
8
10
12
14
16
18
20
Time [h]
LB
E M
ass F
low
ra
te [kg
/s]
Experimental
RELAP5 Stand Alone
RELAP5+FLUENT 2D Explicit Scheme
RELAP5+FLUENT 2D Implicit Scheme
• Differences in the LBE mass flow rate obtained using the Explicit and the Implicit coupling tool are lower than 1%.
• Results of the coupled
simulations overestimate results of RELAP5 stand alone calculation by less than 5%.
• A sensitivity analysis for the Time step value shows that the Implicit scheme allows the use of a time step 5x without loosing accuracy.
Implicit coupled simulation performed adopting the same time step of the Explicit coupled simulation were conducted to verify the new scheme (Test 206-FC)
Computational time reduced by the use of the implicit scheme
ALFRED Development: Main Issues Studies Ongoing Experimental Activities Ongoing Remarks
Parameters Value Outside Diameter 1200 mm Wall Thickness 15 mm Material AISI 316L Max LBE Inventory 90000 kg Electrical Heating 47 kW Cooling Air Flow Rate 3 Nm3/s Temperature Range 200 to 550 °C Operating Pressure 15 kPa (gauge) Design Pressure 450 kPa (gauge) Argon Flow Rate 15 Nl/s Argon Injection Pressure 600 kPa (gauge)
Pool Thermal-Hydraulic
Electrical Power (Pin Bundle) 1 MW
26
Heat Exchanger
Feeding Conduit
Separator
Riser
Dead Volume
Fitting Volume
FPS
Flow Meter
Heat Exchanger
Feeding Conduit
Separator
Riser
Dead Volume
Fitting Volume
FPS
Flow Meter
27
Pool Thermal-Hydraulic
• Thermal Power: 800 - 925 kW
• LBE TAV: 300 - 350 °C
• Core ΔT: 100 °C
• Core velocity: 1.0 m/s
• LBE Flow Rate: 55.2 kg/s
Assembly: Hexagonal
d: 8.2 mm
p/d: 1.8
Lact: 1000 mm
NPin: 37
q”: 100 W/cm2
QPin: 26 kW
28
Pool Thermal-Hydraulic
Section 3: 60 mm upstream the upper spacer grid
Section 1: 20 mm upstream the middle spacer grid
TCs OD 0.5 mm 29
Pool Thermal-Hydraulic
30
Pool Thermal-Hydraulic
''
eq
eqc b
HTC DNu
Kq D
T TNu K
234
4
2
eq
p dD d
Steady State condition maintained at least for 15 min
521<Pe<2916
31
Pool Thermal-Hydraulic
0 0.05 0.1 0.15 0.2 0.2560
62
64
66
68
70
72
74
76
78
80
Time [h]
Tout
FP
S -
Tin F
PS
0 0.05 0.1 0.15 0.2 0.250
10
20
30
40
50
60
70
LB
E M
ass F
low
Ra
te [kg
/s]
Time [h]
0 0.05 0.1 0.15 0.2 0.250
5
10
15
20
25
30
35
40
Nu
sse
lt n
um
be
r [-
]
Time [h]
Nu Section 1 Central Sub channel
27
1 7
2
0 0.05 0.1 0.15 0.2 0.25300
310
320
330
340
350
360
370
380
Te
mp
era
ture
[°C
]
Time [h]
Section 1, Temperatures Central Sub channel 1-2-7
T bulk
T Pin 7
T Pin 1
LBE
ṁ
[kg/s]
Pe
[-]
SECTION 1 SECTION 3
T clad
[°C]
T bulk
[°C]
Nu
[-]
T clad
[°C]
T bulk
[°C]
Nu
[-]
1-FC 70 2916 365 312 27 413 355 35
2-FC 65 2754 363 311 26 410 353 33
3-FC 60 2577 351 300 25 398 343 31
4-FC 55 2357 352 304 24 399 348 31
5-FC 49 2117 343 298 23 387 340 27
6-FC 43 1876 336 291 21 381 335 25
7-FC 40 1761 325 285 21 369 326 24
1-NC 25 1009 435 372 16 522 464 20 2-NC 23 925 429 376 15.6 510 460 21 3-NC 21 815 452 409 15.48 522 483 20 4-NC 19 746 431 399 15.41 490 459 18 5-NC 14 585 364 341 14.77 420 398 17 6-NC 12 521 321 309 14 354 342 17
32
Pool Thermal-Hydraulic
Section 1 Section 3
2 0.56 0.19 /7.55 / 20 / 0.041 / valid for 1.2 p/d 2 and for 1 Pe 4000
p dNu p d p d p d Pe
3.8 ( 1) 0.77 0.047 1 250
valid for 1.1 p/d 1.95 adn for 30 Pe 5000
xNu e Pe
0 500 1000 1500 2000 2500 30000
5
10
15
20
25
30
35
40
Pe
Nu
Section 1
Ushakov
Mikityuk
Nu-csc +15%
-15%
0 500 1000 1500 2000 2500 30000
5
10
15
20
25
30
35
40
PeN
u
Section 3
Ushakov
Mikityuk
Nu-csc
-15%
+15%
33
Pool Thermal-Hydraulic
The experimental results agree very well each other General trend slightly underestimates values obtained by Ushakov and Mikityuk correlations CFD calculations well agree with experimental results The uncertainty of the obtained Nu is within ±20%, while the uncertainty of the Pe is within ±12%
0 500 1000 1500 2000 2500 3000 35000
5
10
15
20
25
30
35
40
Pe [-]
Nu
[-]
Ushakov
Mikityuk
Nu-csc Sez-1
Nu-csc Sez-3
Nu 37 pin SST
Nu 37 pin RSM
34
Pool Thermal-Hydraulic
35
Pool Thermal-Hydraulic
36
Pool Thermal-Hydraulic
0 1 2 3 4 5 6 7280
290
300
310
320
330
340
350
360
Position [m]
Te
mp
era
ture
[°C
]
Time = 0.3 h
Line A
Line H
Line I
Line B
Line C
Line D
Line E
Line F
0 1 2 3 4 5 6 7280
290
300
310
320
330
340
350
360
Position [m]
Te
mp
era
ture
[°C
]
Time = 5.6 h
Line A
Line H
Line I
Line B
Line C
Line D
Line E
Line F
0 1 2 3 4 5 6 7270
280
290
300
310
320
330
340
350
360
Position [m]
Te
mp
era
ture
[°C
]
Time = 7.8 h
Line A
Line H
Line I
Line B
Line C
Line D
Line E
Line F
0 1 2 3 4 5 6 7280
290
300
310
320
330
340
350
360
370
Position [m]
Te
mp
era
ture
[°C
]
Time = 47.8 h
Line A
Line H
Line I
Line B
Line C
Line D
Line E
Line F
The refurbishment of the CIRCE facility is completed
New main heat exchanger (HERO - DWBT @ 180 bar)
Filtering section at the outlet section of the HX (coldest zone, with DP)
Filtering section at the outlet section of the riser (hottest zone, NO DP –
free level)
Air- DHR (cold trap)
Mass exchanger
Oxygen sensors in cover gas
Ar/H2 injection system into the pool (NO Oxygen getter)
Oxygen sensor to be installed (600 mm, 4500 mm, 7500 mm)
Oxygen Control System
37
Stainless Steel filter : cartridge type (Riser Outlet)
38 Stainless Steel filter : cartridge type (SG outlet)
Multilayer (5 layer) Sintered Filter Media (thickness 2,4 mm), meshing 150 micron
Oxygen Control System
39
Oxygen sensors (Pt-air) “strengthened” with heating element inside have been developed
Oxygen Control System
40
Long Term Tests
41
Long Term Tests
42
SGTR - Separate Effect Test
S4 – Storage tankLBE
LBE
S2Water
S1 - Interaction vessel
V24
Ar supply
V1
Wa. ½’’
S3
Pb – LBE 2'’
Drainage
D1
Gas / LBE 3'’
V13
V11
V3
V4 V14
V5
V6
V16
V15
V9
V7
TC-S4V-01
PC-S4V-01
Discharge
V20
V22
Ar gas
PC-S1V-01 LC-S1V-01
V21
Argon Argon[High P]
MT-S2L-01
Drainage
coriolis flow meter
LevelGauge
Housing
LIFUS5/Mod2THINS configuration 2012
V23
Discharge
Vacuum pumpAir circuit
PC-AUX-01
V19
Compressor PC-AUX-02
Wa. 2’’
Wa. 1/2’’
LT-S2V-01
PC-S2V-01
PC-S2V-02
PT-S2L-07
TC-S2L-01
TC-S2V-01
PT-S2V-06
TC-S2V-02
PC-S3V-01
Water
TC-S3L-01
Discharge
Discharge
PT-S1V-05
PT-S1V-03 PT-S1V-02
TC-S1V-02
SG-S1V-03 SG-S1V-02
TC-S1V-01
V25
Discharge
V26
PManometro
Argon[Low P]
V27½’’
¼’’
TC-S2L-02
S4 – Storage tankLBE
LBE
S2Water
S1 - Interaction vessel
V24
Ar supply
V1
Wa. ½’’
S3
Pb – LBE 2'’
Drainage
D1
Gas / LBE 3'’
V13
V11
V3
V4 V14
V5
V6
V16
V15
V9
V7
TC-S4V-01
PC-S4V-01
Discharge
V20
V22
Ar gas
PC-S1V-01 LC-S1V-01
V21
Argon Argon
MT-S2L-01
Drainage
coriolis flow meter
LevelGauge
Housing
LIFUS5/Mod2THINS configuration 2012
V23
Discharge
Vacuum pumpAir circuit
PC-AUX-01
V19
Compressor PC-AUX-02
Wa. 2’’
Wa. 1/2’’
LT-S2V-01
PC-S2V-01
PC-S2V-02
PT-S2L-07
TC-S2L-01
TC-S2V-01
PT-S2V-06
TC-S2V-02
PC-S3V-01
Water
TC-S3L-01
Discharge
Discharge
PT-S1V-05
PT-S1V-03 PT-S1V-02
TC-S1V-02
SG-S1V-03 SG-S1V-02
TC-S1V-01
V25
Discharge
XX-YYY-NNMST type MST position MST No.
ID MST position Position
FNS Test section in S1: Frame Nord Sud FEW Test section in S1: Frame Est West S1V Interaction Vessel S1 S2V Water tank S2 S2L Water injection line S3V Dump system S3 tank S3L Dump system S3 Line AUX Auxiliary systems
ID MST type MST device No. MST
installed LC Level gauge (control) 1 LT Level gauge (acquisition) 1 MT Flow meter 1 PC Absolute pressure transducer 6 PT Dynamic pressure transducer 7 TC Thermocouple for acquisition and control 106 TR Thermocouple for regulation 35 SG Strain Gage 6
TC-S2L-02
PT-S1V-08
PT-S2L-07
TC-S2L-02
S1 - main vessel 100 litres
S2 - water tank 15 liters
S3 - safety volume 2000 liters
S4 - LBE storage tank
650 litres
Water injection line
• 188 tubes 43
• cylindrical shape • height of 400 mm • Internal radius 155 mm
• 200 holes of 15 mm of diameter • two closing flanges each one equal to 20 mm of thickness Water injection line
Parameter ELSY SG Number of tubes 218 length of tubes (m) 55 Outer diameter of the tubes (mm) 22.22 Thickness of tubes (mm) 2.5 Radial pitch (mm) 24 Axial pitch (mm) 24 Height (coils only) (mm) 2620
Single Tube configuration
ELSY Steam Generator
SGTR - Separate Effect Test
44
188 tubes:
• 12 pressurized tubes
• 128 holed dummy tubes
• 48 closed dummy tubes
• ID 18 mm
• 20 thermocouples with a diameter of 0.5 mm
• 50 thermocouples with a diameter of 1 mm
• 7 strain gauge
• pitch 19.8 mm
• triangular lattice
thermocouples
Strain gauge
Lev.A
SGTR - Separate Effect Test
S1
S2
S4
S3
45
T#B1. (0.1 A)
Injection Valve
LBE temperature
[°C]
Water pressure
[bar]
Water temperature
[°C]
Test section tubes pressure
[bar]
Injection time [s]
Injector orifice
diameter [mm]
1 V4 400 180 260 180 2 4 2 V4 400 180 270 180 2 4 3 V4 400 180 270 180 2 4
SGTR - Separate Effect Test
57.29 [bar]
26.77 [bar]24.65 [bar]
57.29 [bar]
26.77 [bar] 24.65 [bar]
46
SGTR - Separate Effect Test
47
NUMERICAL ANALISYS
V14
V4
RELAP5/MOD3.3 SIMMER-III
• Sensitivity analysis of LIFUS5/Mod2 water injection line (THINS configuration);
• Simulation of the first part of test B1.1 transient.
• Simulation of the second part of test B1.1 transient.
RELAP5/MOD3.3 ANALYSIS # T vlv opening
[s] K
[V4],[V14] K
[Coriolis] A Coriolis
[m2] a 0.25 7 0.5 9.06613e-05
q 0.25 10 2.5 4.53e-05
0.00
5.00
10.00
15.00
20.00
25.00
30.00
35.00
40.00
45.00
0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 4.50
Pressure [bar]
Time [s]
Test A1.2_2
experimental data P [Pa] caso_q
Test A1.2_2
P RELAP5 case_q Experimental data
water injection line pressure trend
Energy loss coefficient for the valves and for the Coriolis (K) Valve opening/closing change rate Coriolis tubes area
0.00
5.00
10.00
15.00
20.00
25.00
30.00
35.00
40.00
45.00
0.00 0.50 1.00 1.50 2.00 2.50 3.00 3.50 4.00 4.50
Pressure [bar]
Time [s]
experimental data case_q
Test A1.2_1
Experimental data P RELAP5 case_q
15 simulations for two different tests
LBE
Argon/Air
Steel/Can wall
No calcolation region
Water
S3
S1
S2
water injection line
V4
LEADER test section
Coriolis flow meter
48
SIMMER-III ANALYSIS (TEST B1.1)
MODELING
S1
top flange
tube
hole
bottom flange
LBE Argon/AirSteel/Can wall No calcolation region
RELAP5 condition
RESULTS
49
The SGTR scenario needs to be analysed, in the integrated pool type LFR reactor
configuration, aiming to predict the hazardous consequences of the SGTR taking
place in the Lead pool (pressure wave propagation and cover gas pressurization,
domino effect, steam dragged into the core, primary system pollution and slug
formation).
Tube rupture propagation into the SG-tube bundle
Pressure waves propagation into the SG-tube bundle and damping effect of the SG-shell
towards the surrounding structures
Assessment and performance evaluation of the safety-guard devices (rupture disk and
fast valves) aiming to mitigate the effects of the SGTR event
Steam trapping in the main coolant flow path and dragging towards the core inlet region
Investigation on the solid impurities formation after the SGTR event, accompanied by a
quantitative qualification of filtering performance in the pool
Investigation of the oxygen concentration variation due to the SGTR event
SGTR – Large Scale Assesment
Operating conditions Unit Value
LBE temperature °C 350
LBE mass flow rate (one SGTR-TS) kg/s ~80
H2O inlet temperature °C 200
H2O outlet temperature °C 201.4
H2O pressure bar 16
H2O mass flow rate (one SGTR-TS) kg/s ~0.07
CIRCE cover gas pressure (abs) bar 1
CIRCE Ar cover gas volume m³ ~2.2
CIRCE LBE pool volume m³ ~7
Ar monitoring tubes pressure bar 16
Tubes length mm ~5200
O
Q
C3
C2C1
B B1
C
B2
D
B B
A A
H3H2
EM
L1
A1
A
F
G
HH
H1
H4
I
LBE level
H5
K
L2
L1
R S
Sez. A - A Sez. B - B
A
A1
G
HH
A
N
HH
H H
H H
E
CIRCE vessel
2 rupture positions Bottom (2 ruptures) Middle (2 ruptures)
4 SGTR-TSs one rupture
in each SGTR-TS
B
B
M
M
50
SGTR – Large Scale Assesment
O
Q
C3
C2C1
B B1
C
B2
D
B B
A A
H3H2
EM
L1
A1
A
F
G
HH
H1
H4
I
LBE level
H5
K
L2
L1
R S
Tube notch SGTR-TS
SPACER GRID
LBE INLET WINDOWS
LBE OUTLET OPENING
SGTR-TS
51
SGTR – Large Scale Assesment
SGSG
SG
SG
SG
SG
higher livel (+ 500 mm)
middle livel(+ 250 mm)
lower livel (+ 0 mm)
+ 0 mm
+ 250 mm
+ 500 mm
10
00
mm
spacer grid
lower tube plate
6 Strain Gages in the bottom SGTR-TS rupture scenario
At the rupture level Three SGs on three tubes simmetrically arranged (120°) around the water tubet
250 mm above the rupture level One SG on a gas tube One SG on the 6’’ outer surface
500 mm above the rupture level One SG on a gas tubet
X 30 overall SGs 28 SGs on the SGTR-TS
2 spare SGs on the separator 52
SGTR – Large Scale Assesment
O
Q
C3
C2C1
B B1
C
B2
D
B B
A A
H3H2
EM
L1
A1
A
F
G
HH
H1
H4
I
LBE level
H5
K
L2
L1
R S
Detailed SIMMER 3D model of the whole CIRCE vessel and implemented Test Section
Highly detailed SIMMER 3D model of the SGTR-TS region in which the ruptures occur
RELAP5 model of the feedwater and two-phase discharge line
Test section construction is ongoing. (due date may 2015)
53
SGTR – Large Scale Assesment
Natural Circulation Tests
54
Features: • 19 pins (triangular lattice); • d = 6.55 mm, Lactive = 600 mm; • Lentry > 600 mm (> 2 pitches); • p/d = 1.28; • q” = 1 MW/m2 , uniform (≈ 235 kW); • wire spaced (d = 1.75 mm). • Each pin can be fed separately
2·l = 45.434 mm
55
Natural Circulation Tests
2·l = 45.434 mm
Instrumentation • 11 pins instrumented; • 5 subchannels instrumented; • data on:
1. wall temperature; 2. bulk temperature; 3. heat transfer coefficient; 4. thermal stratification; 5. (thermal entrance length); 6. eventual hot spots.
• pins instrumented with: a) Embedded-wall TCs 0.35 (52); b) subchannel TCs 0.5 mm (15).
56
Natural Circulation Tests
57
Natural Circulation Tests
HELENA is a Heavy Liquid Metal Experimental Loop for Advanced Nuclear Applications
The aim of the facility is to:
Investigate the thermal hydraulic behavior of heavy liquid metal coolant
(i.e. heat exchange under forced convection)
Characterize structural material when working in lead (i.e. AISI 316L, T91, 15-15 Ti, coated steels)
Qualify prototypical components (i.e. centrifugal pump, ball valve)
Develop and test suitable instrumentations for HLM applications
Support the CFD and system code validation when employed in lead cooled system
Forced Circulation Tests
58
59
Forced Circulation Tests
60
Forced Circulation Tests
Assessment of the SGBT by RELAP-5
HERO Test Section (CIRCE)
0
0.2
0.4
0.6
0.8
1
1.2
250
300
350
400
450
500
550
0 1 2 3 4 5 6 7 8 9 10
Void
frac
tion [
/]
Tem
pera
ture
[°C]
Tube length [m]
Annular riser Feedwater tubeLead subchannel Annular riser void fraction
The maximum steam temperature is predicted in the range 438-456 °C depending on the diamond conductivity.
Superheated steam is always predicted, a void fraction close to 1.0 is reached within the first 3 m of the annular riser.
The lead temperature drop is about 80 °C.
61
HERO Test Section (CIRCE)
Central byonet tube
1500
2700
2400
2100
1800
3000
3300
3600
3900
4200
7016
6000
6800
0
D-D120°
0°
240°
C-C120°
240°
B-B120°
240°
A-A120°
240°
0°
0°
0°
A-A
B-B
C-C
D-D
Tube 0TC-C0-I00
TC-C0-I01
TC-C0-O15
TC-C0-O18
TC-C0-O21
TC-C0-O24
TC-C0-O27
TC-C0-O30
TC-C0-O33
TC-C0-O36
TC-C0-O39
TC-C0-O42
TC-C0-O60
TC-C0-O70
TC-W0-P15
TC-W0-P30
TC-W0-P42
TC-W0-L10
TC-W0-L30
TC-W0-L40
TC-W0-L60
TC-W0-L10
TC-W0-P60
TC-W0-L11
TC-W0-L12
TC-W0-P15TC-C0-O15
TC-W0-L30
TC-W0-L31
TC-W0-L32
TC-W0-P30
TC-C0-O30
TC-W0-L40
TC-W0-L42
TC-W0-P42
TC-C0-O42
TC-W0-L60
TC-W0-L61
TC-W0-L62
TC-W0-P60
TC-C0-O60 TC-W0-W68
Central tube Fluid flow: 1 TC at inlet, 1 TC at
the end of the descendent tube, 1 TC at the end of the active height and 1 TC its outlet. These TC are located at the center of the channel.
Boiling height: 10 TC in the center of the riser (pitch 300mm).
Condensation?: 1 TC at the riser outlet located in the descendent tube outer surface.
Si-C side: 4 TC at four heights. Lead side: 12 TC at the same
heights of Si-C side and 3 azimuth. In total 31 TH (19 TC-0.5mm, 12-
1mm)
HERO Test Section (CIRCE)
63
Pressure drop and mass flow 7 Mass-flow meters, one at each tube inlet. 1 Mass-flow meter, at the FW collector
inlet. 1 pressure transducer at FW collector inlet
at one at Steam collector outlet. Delta p in the central tube: overall,
descending, ascending. 6 delta p overall in the remaining tubes.
Tube 1 Tube 2 Tube 3 Tube 4 Tube 5 Tube 6Tube 0
DP-C0-W00 DP-C1-W00 DP-C2-W00 DP-C3-W00 DP-C4-W00 DP-C5-W00 DP-C6-W00
DP-C0-W01 DP-C1-W01 DP-C2-W01DP-C1-W02 DP-C2-W02DP-C0-W02
MF-C1-I00MF-C0-I00 MF-C2-I00 MF-C3-I00 MF-C4-I00 MF-C5-I00 MF-C6-I00
HERO Test Section (CIRCE)
64
Remarks
65
R&D program in support of the ALFRED Project is quite large and comprehensive but not full addressing. For the whole project implementation more efforts are needed, in term of budget, human resources, and research infrastructure availability. The in-kind contribution of FALCON Consortium will speed-up the ongoing activities, widening the R&D program. Other countries are involved on the LFR technology development (i.e. Russia and China) with large R&D programs. Synergies could be implemented (?).