R&D in support to the ALFRED Project - SNETP · ALFRED Project R&D in support to the ALFRED Project...

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


► 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


► 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


► 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


► 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


► 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


► 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


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)


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

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


• 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


Section 3: 60 mm upstream the upper spacer grid

Section 1: 20 mm upstream the middle spacer grid

TCs OD 0.5 mm 29


30


''

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


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


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


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


35


36


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


39

Oxygen sensors (Pt-air) “strengthened” with heating element inside have been developed


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


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


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


57.29 [bar]

26.77 [bar]24.65 [bar]

57.29 [bar]

26.77 [bar] 24.65 [bar]

46


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


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

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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


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)

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Natural Circulation Tests

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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

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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).

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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

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59


60


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.

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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)


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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


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Remarks

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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 (?).

R&D in support to the ALFRED Project - SNETP · ALFRED Project R&D in support to the ALFRED Project...

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