Safety Analysis Results of the DEC Transients of ALFRED LEADER Lead-cooled European Advanced DEmonstration Reactor G. Bandini (ENEA), E. Bubelis, M. Schikorr.

Safety Analysis Results of the DEC Transients of ALFRED

LEADERLead-cooled European Advanced DEmonstration Reactor

G. Bandini (ENEA), E. Bubelis, M. Schikorr (KIT), A. Lazaro, K. Tucek (JRC-IET)P. Kudinov, K. Kööp, M. Jeltsov (KTH), M. H. Stempnievicz (NRG), Z. Youpeng, K. Mikityuk (PSI)

Technical Workshop to Review Safety and Design Aspects of ALFRED, ELFR and ELECTRA

JRC-IET, Petten, 27-28 February 2013

2

Outline

Introduction The ALFRED reactor DEC transients for ALFRED DEC transient results Conclusions

3

Introduction

One of the main objectives of the LEADER EU project was the evaluation of the safety aspects of the lead-cooled demonstrator reactor ALFRED

Both Design Basis Conditions (DBC) and Design Extensions Conditions (DEC) have been considered in the safety analysis of ALFRED

The DEC accident scenarios are very low probability events which include the failure of prevention or mitigating systems

The main objective of DEC transient analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the plant

More representative DEC events for ALFRED have been analysed by several research organizations using different system codes

4

ALFRED: Reactor block

Vertical section

Horizontal section

Pool-type reactor of 300 MWth power 171 fuel assemblies in the core 8 pump-bayonet tube SG connected to

the 8 secondary circuits

5

ALFRED: Secondary circuits

DHR System (4 x 2 IC loops)

In-water pool isolation condenser (IC)

Valve

Water

Hot Lead

Cold Lead

Steam

Water

Hot Lead

Cold Lead

Steam

SG

Feedwater

Steam

From DHR system

To DHR system

Steam lines

Feedwater lines

6

Steady-state at nominal power (EOC)

Parameter Unit ALFRED RELAP5 CATHARE SIM-LFR

Reactor thermal power MW 300 300 300 300

Total primary flow rate kg/s 25980 25250 25460 25682

Total ΔP in the primary circuit bar 1.5 1.5 1.5 1.5

ΔP through the core bar < 1.0 1.0 1.0 1.0

Core inlet temperature °C 400 400 400 400

Upper plenum temperature °C 400 480 480 480

Max core outlet temperature (*) °C - 483 483 487

Peak clad temperature °C ~550 508 518 514

Peak fuel temperature °C ~2000 1991 1985 2064

Feedwater temperature °C 335 335 335 335

Feedwater flow rate kg/s 192.8 192.8 196.6 193.6

Steam temperature °C 450 450 451 450

Steam pressure bar 180 180 180 180

(*) Hottest FA flow rate is ~120% of average FA flow rate

7

Analysis of DEC transients

Organizations and codes: ENEA (RELAP5, CATHARE), KIT (SIM-LFR), JRC-IET (TRACE, SIMMER), KTH (RELAP5),

NRG (SPECTRA), PSI (TRACE/FRED)

TRANSIENT Initiating Event Reactor scram

Primary pump trip

MHX FW trip

MSIV closure

DHR startup

TR-4: UTOP Insertion of 250 pcm in 10 s

No No No No No

TDEC-1: ULOF All primary pumps coastdown

No 0 s No No No

TDEC-3: ULOHS All MHX feedwater trip No No 0 s 1 s DHR-1 at 2 s (3 IC loops)

T-DEC4: ULOHS+ULOF All primary pumps and MHXs feedwater trip

No 0 s 0 s 1 s DHR-1 at 2 s (3 IC loops)

T-DEC5: Partial block. in the hottest FA

10% to 97.5% blockage at the hottest FA inlet

No No No No No

TO-3: All prim. pumps stop + reduction of FW temperature

T-fw: 335330°C in 1s + all p. pumps stop

2 s, low pump speed

0 s 2 s 2 s DHR-1 at 3 s (4 IC loops)

TO-6: All prim. pumps stop + increase of FW flow rate

FW-flow +20% in 25 s + all p. pumps stop

2 s, low pump speed

0 s 2 s 2 s DHR-1 at 3 s (4 IC loops)

T-DEC6: SCS failure Depressurization of all secondary circuits

2 s, low sec. pressure

No 2 s No No

UN

PRO

TECT

EDPR

OTE

CTED

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DEC: Unprotected transients

Objective: Verify the intrinsic safety behaviour of the ALFRED plant and its response to more unlikely accidental events

Analysed transients without reactor scram: UTOP: Reactivity insertion of 25 pcm in 10 s

(core compaction, core voiding following SGTR, etc.) ULOF: Loss of all primary pumps ULOHS: Loss of feedwater to all MHXs ULOHS + ULOF: Loss of feedwater to all MHXs + loss of

all primary pumps Partial FA blockage verify the maximum acceptable

flow are blockage without fuel rod damage

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Reactivity feedbacks at EOC

 REACTIVITY COEFFICIENT Unit Ref. Temperature Value

Control rod differential expansion (*) pcm/K T upper plenum -0.218

Coolant expansion (**) pcm/K Average T-core -0.268

Axial clad expansion pcm/K Average T-clad 0.039

Axial wrapper tube expansion pcm/K Average T-wrapper 0.023

Radial clad expansion pcm/K Average T-clad 0.011

Radial wrapper tube expansion pcm/K Average T-wrapper 0.003

Diagrid radial core expansion pcm/K T-core inlet -0.152

Pad radial core expansion pcm/K T-core outlet -0.430

Axial fuel expansion: free pcm/K Average T-fuel -0.155

Axial fuel expansion: linked pcm/K Average T-clad -0.242

Doppler constant pcm Average T-fuel -566

(*) Prompt response (the delayed response has been neglected)(**) Calculated on the whole height of the fuel assembly (the other feedbacks are calculated only in the fissile zone)

UTOP transient (1/4)

Insertion of 250 pcm in 10 s without reactor scram No feedwater control on secondary side Codes used: TRACE, SIM-LFR, RELAP5, CATHARE, TRACE/FRED, SPECTRA

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Total reactivity and feedbacks Core and MHX powers

RELAP5 Results

Total

Inserted

Doppler

Fuel exp.

Core power

MHX power

Maximum net reactivity insertion of 85 pcm Initial core power peak of 680 MW

UTOP transient (2/4)

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Core temperatures Max clad and fuel temperatures

RELAP5 Results

Core outlet

MHX inlet

MHX outlet

Core inlet

Max fuel

Max clad

Maximum clad temperature remains below 650 °C Maximum fuel temperature of ~2930 °C at t = 50 s (hottest pin, middle core plane,

fuel pellet centre) exceeds the MOX melting point (~2700 °C) only local fuel melting

Core outlet

Max clad

Core inlet

UTOP transient (3/4)

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Differences in fuel expansion reactivity feedback (free/linked effects) and fuel rod gap dynamic modelling

Only local fuel melting in the hottest pin is confirmed by all codes

Peak power and max fuel temperatures: RELAP5: 679 MW and 2930 °C SIM-LFR: 656 MW and 2996 °C CATHARE: 735 MW and 2866 °C TRACE/FRED: 642 MW and 2779 °C

UTOP transient (4/4)

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SIM-LFR:Minimum clad failure time >> 1.0E+7 s

Different heat transfer correlations used by RELAP5 and CATHARE for fuel rod bundle

Maximum clad temperature is below 650 °C

1E+00

1E+02

1E+04

1E+06

1E+08

1E+10

1E+12

1E+14

0 20 40 60 80 100 120 140 160 180 200Time [sec]

Cla

d F

ailu

re T

ime

[se

c]

19.5

20.0

20.5

21.0

21.5

22.0

Fis

sio

n G

as

Pre

ssu

re

[ba

r]

Clad Failure Time [sec]

Fission Gas Pressure [bar]30 min

14

ULOF transient (1/4)

All primary pumps coastdown without reactor scram No feedwater control on secondary side Codes used: RELAP5, SIM-LFR, CATHARE, TRACE, TRACE/FRED, SPECTRA

Active core flowrate Core and MHX powers

RELAP5 Results

Core power

MHX power

Natural circulation in the primary circuit stabilizes at 23% of nominal value Core power reduces down to about 200 MW due to negative reactivity feedbacks

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ULOF transient (2/4)

Core temperatures

Core temperatures

RELAP5 Results Initial clad peak temperature of 764 °C Max clad temp. stabilizes below 650 °C Positive Doppler and fuel exp. effects are

mainly counterbalanced by negative radial core exp. (Pad + Diag.), control rods and coolant exp. effects

Total reactivity and feedbacks

Max clad

Max lead

Core inlet

Max clad

Core inlet

Max fuel

Doppler

Fuel exp.

C. Rods

Pad + Diag.

Cool. exp.

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ULOF transient (3/4)

Slight deviations in the initial core flow rate transient, but good agreement in stabilized natural circulation flow rate in the primary circuit

Core power at t = 200 s is slightly under predicted by SIM-LFR (P = 177 MW) and TRACE/FRED (P = 180 MW) with respect to RELAP5 (P = 195 MW) and CATHARE (P = 198 MW)

17

ULOF transient (4/4)

SIM-LFR:Minimum clad failure time >1.0E+5 s

The initial clad peak temperature is calculated in the range 730° C–764°C

Maximum clad temperature predicted by the codes at t = 200 s is around 650 °C

No clad failure is expected under ULOF in the short and long term

No vessel wall temperature increase (Tw < 400 ° C during ULOF transient)

1E+00

1E+02

1E+04

1E+06

1E+08

1E+10

1E+12

1E+14

0 50 100 150 200 250 300Time [sec]

Cla

d F

ailu

re T

ime

[se

c]

19.2

19.4

19.6

19.8

20.0

20.2

20.4

Fis

sio

n G

as

Pre

ssu

re

[ba

r]

Clad Failure Time [sec]

Fission Gas Pressure [bar]

30 min

Min. Pin Clad Failure Time = 1.6E+5 sec at transient time t = 17.8 sec

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ULOHS transient (1/4)

Loss of feedwater to all MHXs without reactor scram Startup of DHR-1 (3 out of 4 IC loops are in service) Codes used: RELAP5, SIM-LFR, CATHARE, TRACE, TRACE/FRED, SPECTRA

Core power progressively reduces down towards decay level removed by DHR-1 Maximum clad and vessel temperatures rise up to ~700 °C after about one hour

Core and MHX powers Core and vessel temperatures

Core power

MHX power

Max vessel

Max clad

CATHARE Results

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ULOHS transient (2/4)

Total reactivity and feedbacksCore temperatures

Fuel temperature reduces down close to clad temperature Positive Doppler and fuel and clad expansion effects are mainly counterbalanced by

negative radial core expansion (Pad + Diag.), coolant expansion and control rods effects

Max clad

Core inlet

Max fuel

Doppler

Fuel exp.

Clad exp.

C. Rods

Cool. exp.

Pad + Diag.

CATHARE Results

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ULOHS transient (3/4)

Vessel wall temperature is over predicted by RELAP5 and CATHARE (no heat losses from the external wall surface) with respect to SIM-LFR

Maximum vessel temperature rises over about 650 °C in 30 minutes no vessel failure is expected in the medium term vessel integrity is not guaranteed in the long term

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ULOHS transient (4/4)

SIM-LFR:Minimum clad failure time > 1.0E+6 s

Maximum clad temperature stabilizes around 700 °C after one hour transient

No clad failure is calculated by SIM-LFR code in the short and long term

1E+00

1E+02

1E+04

1E+06

1E+08

1E+10

1E+12

1E+14

0 500 1000 1500 2000 2500 3000 3500Time [sec]

Cla

d F

ailu

re T

ime

[se

c]

0

5

10

15

20

25

30

Fis

sio

n G

as

Pre

ssu

re

[ba

r]

Clad Failure Time [sec]

Fission Gas Pressure [bar]30 min

22

ULOHS+ULOF transient (1/4)

Loss of feedwater to all MHXs and all primary pumps without reactor scram Startup of DHR-1 (3 out of 4 IC loops are in service) Codes used: SIM-LFR, RELAP5, CATHARE, SPECTRA

0.0

0.2

0.4

0.6

0.8

1.0

0 500 1000 1500 2000 2500 3000 3500Time [sec]

rel.

un

its [

fr]

Power_th

Flow_Cool

200

700

1200

1700

2200

0 500 1000 1500 2000 2500 3000 3500Time [sec]

Te

mp

era

ture

[°C

] Fuelc_peak Clad_peakCool_out Cool_inT_wall

Core flow rate and power Core and vessel temperatures

Max fuel

Max lead, cladCore inlet, vessel

Power

Flow rate

SIM-LFR Results Sharp decrease of core power and flow rate in the initial transient phase and then their

progressive decrease Core flow rate/power ratio is ~1/3 of nominal value Large ΔT through the core

Maximum clad temperature rises up to ~800 °C in 30 minutes

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ULOHS+ULOF transient (2/4)

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ULOHS+ULOF transient (3/4)

Similar evolution of core flow rate and core power is calculated by the codes

Calculated vessel wall temperature is in the range 440 °C - 520 °C after 30 min.

Vessel integrity is guaranteed in the medium term and likely also in the long term according to RELAP5 results

25

ULOHS+ULOF transient (4/4)

SIM-LFR:Minimum clad failure time > 1.0E+4 s

The maximum clad temperature is over predicted of 25° C – 30 °C by RELAP5 and CATHARE with respect to SIM-LFR

The minimum clad failure time predicted by SIM-LFR is of about 3 hours

1E+00

1E+02

1E+04

1E+06

1E+08

1E+10

1E+12

1E+14

0 500 1000 1500 2000 2500 3000 3500Time [sec]

Cla

d F

ailu

re T

ime

[se

c]

0

5

10

15

20

25

30

Fis

sio

n G

as

Pre

ssu

re

[ba

r]

Clad Failure Time [sec]

Fission Gas Pressure [bar]

30 min

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Partial FA blockage (RELAP5 results)

Code used: RELAP5, SIM-LFR, SIMMER

RELAP5 assumptions: Total ΔP over the FA = 1.0 bar ΔP at FA inlet = 0.22 bar Flow area blockage at FA inlet No heat exchange with

surrounding FAs

MAIN RESULTS: 75% FA flow area blockage 50% FA

flowrate reduction 85% blockage T-max clad = 700 °C No clad melting if area blockage < 95% Fuel melting if area blockage > 97.5% 50% inlet flow area blockage can be

detected by TCs at FA outlet

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SCS Failure (1/2) (RELAP5 results)

Secondary pressure

Core and MHX powers

Primary lead temperatures

Depressurization of all secondary circuits at t = 0 s (no availability of the DHR)

Reactor scram at t = 2 s on low secondary pressure

Initial MHX power increase up to 850 MW no risk for lead freezing

MHXs

Core

MHX outlet

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SCS Failure (2/2) (RELAP5 results)

Core decay and MHX powers Core and vessel temperatures

No risk for lead freezing in the initial transient phase Slow primary temperature increase due to large thermal inertia of the primary

system large grace time for the operator to take opportune corrective actions

Core

MHX

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Conclusions (1/2)

In all simulated transients there is a very large margin to coolant boiling since the coolant is always at least 900 °C below the lead boiling point (1740 °C)

Clad failure is not predicted in all simulated transients except for: Undetected FA blockage greater than ~85% which might be excluded by design

(many orifices at the FA inlet) The very unlikely ULOHS+ULOF event, when the time-to-failure reduces down

to few hours, but still leaving enough grace time for corrective operator actions Fuel melting is excluded in all simulated transients except for local fuel

melting in the hottest pins in case of UTOP transient The vessel integrity seems guaranteed in the long term in all simulated

transients except for the ULOHS transient, but even in this case there is enough grace time for corrective operator actions

No relevant safety issues have been identified for ALFRED in case of representative DEC events – In particular the ULOF transient can be accommodated without the need of corrective operator actions

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Conclusions (2/2)

The analysis of DEC transients with various codes has highlighted the very good intrinsic safety features of ALFRED design thanks to:

Benign characteristics of the coolant Good natural convection in the primary circuit Large thermal inertia to slow down the transients Prevalent negative reactivity feedbacks to limit power

excursions In all analyzed unprotected transients there is no risk for

significant core damage and then for transient evolution towards severe accidents enough grace time is left to the operator to take the opportune corrective actions and bring the plant in safe conditions in the medium and long term