“Design and safety analysis of ALFRED” Accident analysis overview G. Bandini ENEA UTFISSM-SICSIS...

“Design and safety analysis of ALFRED”

Accident analysis overview

G. BandiniENEA UTFISSM-SICSIS

3rd LEADER International Workshop Bologna, 6th - 7th September

Introduction ALFRED design features Codes and reactor modelling Steady-state results Analyzed DBC and DEC transients Preliminary results from transient analysis Preliminary Conclusions

2

Outline

3

Introduction

The conceptual design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is under development within the LEADER project to meet the safety objectives of GEN-IV nuclear energy systems

For the safety analysis of ALFRED representative accident initiators for Design Basis Conditions (DBC) and Design Extension Conditions (DEC) have been identified by the application of a simplified line-of-defence strategy and on the basis of the design solutions adopted for the ALFRED reactor

The identified event initiators have been categorized according to their frequency and the more representative for each category have been selected for safety analysis to be carried out within the LEADER project

Preliminary results of the analysis of some selected accidental transients within DBC and DEC performed with the RELAP5 and CATHARE codes are presented

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

System codes used

The ALFRED accident analysis is performed by various organizations using different codes: KIT-G (SIM-LFR), NRG (SPECTRA), KTH (CFD, RELAP5), JRC (SIMMER, TRACE), PSI (TRACE/FRED), CIRTEN (SIMMER), ENEA (RELAP5, CATHARE), CEA(CATHARE)

RELAP5 and CATHARE (CEA collaboration) codes are used by ENEA for the analysis of selected DBC and DEC transients RELAP5 (developed in USA) and CATHARE (developed in France) are system

codes for thermal-hydraulic transient analysis of light water reactors The RELAP5 code has been modified by ENEA, Ansaldo and Univ. of Pisa for

LFR transient analysis by the implementation of LBE and lead thermal properties - Code validation on CHEOPE, NACIE and CIRCE experiments performed at ENEA/Brasimone

LBE and lead thermal properties have been recently implemented in CATHARE in the frame of an ENEA-CEA collaboration – Code validation on TALL experiments at KTH/Stockholm, NACIE experiments (ENEA/Brasimone) and HELIOS Korean loop experimental data (LACANES Benchmark)

7

ALFRED: Reactor modelling

Feedwater

Steam

521-8

531-8

551-8 561-8

151-

8

Feedwater

Steam

521-8

531-8

551-8 561-8

151-

8

841- 4

851- 8441-8

801- 4

811-4831- 4

815

401-8

841- 4

-

801- 4

811-4831- 4

815

841- 4

-

801- 4

811-4831- 4

815

611- 8

841- 4

-

801- 4

811-4831- 4

815

611-

711-

731-

741- 4

751- 8

761- 4

621- 4

641- 4

771

781- 8

601- 4

661- 4

611-

711- 8

731- 8

741- 4

751-

761- 4

621- 4

641- 4

771- 8

781-

601- 4

661- 4

611-

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771-

781-

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771-

781

601- 4

661- 4

841- 8

-

801- 4

811-4831- 4

815

841-

-

801- 8

811-8831- 8

815

611-

841-

-

801-

811-831-

411-8

611-

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771

781

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771-

781-

601- 8

661- 8

611-

711-

731-

741- 8

751-

761- 8

621- 8

641- 8

841- 4

851- 8441-8

801- 4

811-4831- 4

815

401-8

841- 4

-

801- 4

811-4831- 4

815

841- 4

-

801- 4

811-4831- 4

815

611- 8

841- 4

-

801- 4

811-4831- 4

815

611-

711-

731-

741- 4

751- 8

761- 4

621- 4

641- 4

771

781- 8

601- 4

661- 4

611-

711- 8

731- 8

741- 4

751-

761- 4

621- 4

641- 4

771- 8

781-

601- 4

661- 4

611-

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771-

781-

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771-

781

601- 4

661- 4

841- 8

-

801- 4

811-4831- 4

815

841-

-

801- 8

811-8831- 8

815

611-

841-

-

801-

811-831-

411-8

611-

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771

781

711-

731-

741- 4

751-

761- 4

621- 4

641- 4

771-

781-

601- 8

661- 8

611-

711-

731-

741- 8

751-

761- 8

621- 8

641- 8

ALFRED Nodalization scheme with RELAP5 (2 secondary loops with weight = 4 with CATHARE)

8 SGs (2 x 4)

8 Secondary loops (2 x 4)

Primary circuit

8 IC loops (2 x 4)

Steam line

Feedwater line

100

101102109

110

115

060061-8 070

050

020

200 151-8

121-8

131-8

141-8

220

230

210

100

101102109

110

115

060061-8 070

050

020

200 151-8

121-8

131-8

141-8

220

230

210

100

101102109

110

115

060061-8 070

050

020

200 151-8

121-8

131-8

141-8

220

230

210

8

ALFRED: Steady-state at EOC(RELAP5 and CATHARE preliminary results)

Parameter Unit CATHARE RELAP5 Note

Reactor power (thermal) MW 300 300

Primary flow rate kg/s 25670 25280 To get average core T = 80 °C

Active core flow rate kg/s 24080 23745

Core bypass flow rate kg/s 1305 1270 About 5% of primary flow rate

Inner vessel bypass kg/s 285 255 About 1% of primary flow rate

Total primary circuit P (P core) bar 1.43 (0.82) 1.40 (0.80) Higher CATHARE primary flow rate

Core inlet temperature °C 400 400

Upper plenum temperature °C 480 480

Hot FA outlet temperature °C 489 489 Flow rate +18% of average FA

Hot FA peak clad temperature °C 522 510 Different heat transfer correlation

Hot FA peak fuel temperature °C 1942 1931 No fuel rod gap dynamic model

Average fuel temperature °C 1126 1120

Primary lead mass kg 3685000 3569000 Higher CATHARE lead free level

SG feedwater flow rate kg/s 196 192.8 To get Tout steam = 450 °C

SG feedwater temperature °C 335 335

SG steam outlet temperature °C 450 450

SG outlet pressure bar 180 180

400

420

440

460

480

500

520

540

0.0 0.1 0.2 0.3 0.4 0.5 0.6

Elevation (m)

Tem

pera

ture

(°C

)

T lead

T ext clad

9

Maximum core temperature (1/2)(RELAP5 and CATHARE preliminary results)

400

420

440

460

480

500

520

540

0.0 0.1 0.2 0.3 0.4 0.5 0.6

Elevation (m)

Tem

pera

ture

(°C

)

T lead

T ext clad

RELAP5 CATHARE

Maximum clad temperature is below the safety limit of 550 °C for normal operation ΔT lead-clad is over predicted by CATHARE due to different correlations used for the

calculation of heat transfer inside the fuel rod bundle: Seban-Simazaki in CATHARE and Ushakov in RELAP5

Hot FA: Lead and external clad temperatures

10

0

400

800

1200

1600

2000

0.0 0.1 0.2 0.3 0.4 0.5 0.6Elevation (m)

Tem

pera

ture

(°C

)

T lead

T ext clad

T int fuel

Maximum core temperature (2/2)(RELAP5 and CATHARE preliminary results)

0

400

800

1200

1600

2000

0.0 0.1 0.2 0.3 0.4 0.5 0.6Elevation (m)

Tem

pera

ture

(°C

)

T lead

T ext clad

T int fuel

CATHARERELAP5

Hot FA: Lead, external clad and internal fuel temperatures

Maximum fuel temperature is below 2000 °C – Large margin to fuel melting (approximately 730 °C)

Axial distribution of fuel temperature is strongly influence by fuel rod gap dynamic behaviour (fuel swelling and thermal dilatation) not taken into account in this preliminary analysis constant gap size along the height and during transient

320

340

360

380

400

420

440

460

480

0.0 1.0 2.0 3.0 4.0 5.0 6.0Elevation (m)

Tem

pera

ture

(°C

) T h2o liq

T h2o gas

T int wall1

T ext wall1

T int wall2

T ext wall2

T lead

11

SG: Axial temperature profile(RELAP5 and CATHARE preliminary results)

SteamLead

SG bayonet tube

- HTC on lead side by Seban-Simazaki in CATHARE and Ushakov in RELAP5 - SG heat transfer surface + 14.5% with CATHARE due to reduced heat transfer capability with respect to RELAP5

CATHARE

320

340

360

380

400

420

440

460

480

0.0 1.0 2.0 3.0 4.0 5.0 6.0Elevation (m)

Tem

pera

ture

(°C

) T h2o liq

T h2o gas

T int wall1

T ext wall1

T int wall2

T ext wall2

T lead

RELAP5H2O

Gap

12

ALFRED: DBC and DEC transients

Transient Initiating Event Category Preliminary Results TR-1 Drop (or insertion) of one control rod DBC2 TR-2 Spurious withdrawal of the most reactive control rod DBC2 TR-3 Reactivity insertion (due to FA errors) DBC3 TR-4 Reactivity insertion (enveloping SGTR, flow blockage, core compaction) DEC RELAP5 & CATHARE TD-1 Spurious reactor trip DBC2 TD-2 Turbine trip (TxT) DBC2 TD-3 Loss of AC power DBC2 RELAP5 TD-4 Loss of Normal feedwater DBC2 TD-5 Loss of one primary pump (with AC power available) DBC2 TD-6 Loss of one primary pump(with AC power not available) DBC2 TD-7 Loss of all primary pumps DBC3 RELAP5 & CATHARE TD-8 Partial blockage in the hottest fuel assembly DBC4 TO-1 Reduction of FW temperature DBC2 RELAP5 TO-2 Reduction of FW temperature + one pump stop DEC TO-3 Reduction of FW temperature + all pumps stop DEC TO-4 Increase of FW flowrate DBC2 RELAP5 TO-5 Increase of FW flowrate + one pump stop DEC TO-6 Increase of FW flowrate + all pumps stop DEC

TRB-1 Steam system piping break DBC4 TRB-2 Cover gas piping break DBC4

T-DEC1 Complete loss of forced flow + Reactor trip fails (total ULOF) DEC RELAP5 & CATHARE T-DEC2 Loss of one primary pump + Reactor trip fails (partial ULOF) DEC T-DEC3 Loss of SCS + Reactor trip fails (ULOHS) DEC T-DEC4 Loss of off-site power (LOOP) + Reactor trip fails (ULOHS + ULOF) DEC RELAP5 & CATHARE T-DEC5 Partial blockage in the hottest fuel assembly DEC T-DEC6 Loss of SCS + Reactor trip fails (ULOHS) + DHR totally fails DEC

20 transients (12 DBC and 8 DEC) of a total of 26 transients will be analyzed

13

Safety analysis

The main objective of the analysis of the DBC transients is to verify that in all foreseen design basis accident conditions the protection system by reactor scram and startup of the DHR system is able to bring and maintain the reactor in safe conditions assuring:

The core decay heat removal in the short and long term That fuel rod and vessel temperature limits for each category (DBC1 –

DBC4) are not exceeded The DEC transients are events with very low frequency which include the failure of

prevention and mitigation systems like the reactor scram in the so-called Unprotected transients

One of the main objectives of the Unprotected transient analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the ALFRED reactor

4 DEC (Protected) transients and 3 DEC (Unprotected) transients have been calculated in this preliminary analysis with RELAP5 and CATHARE codes

14

ALFRED: Preliminary transient analysis

Main events and reactor scram threshold

RELAP5 and CATHARE (CEA collaboration) codes are used by ENEA for ALFRED DBC and DEC transient analysis

Preliminary RELAP5 and CATHARE results are presented. These results will be updated after comparison with other partner calculations (largest uncertainties for UTOP transient results due to the lack of a fuel rod gap dynamic model)

15

DBC (PROTECTED) TRANSIENTS

0

5000

10000

15000

20000

25000

0 50 100 150 200 250

Ma

ss fl

ow

ra

te (

kg/s

)

Time (s)

Core flow

0

5000

10000

15000

20000

25000

0 50 100 150 200 250M

ass

flo

w r

ate

(kg

/s)

Time (s)

Core flow

16

PLOF: Loss of primary flow (1/3)(RELAP5 and CATHARE preliminary results)

RELAP5 CATHARE

Core mass flow rate

IE: Coast-down of all primary pumps at t = 0 s (pump speed halving time = 2 s) Reactor scram on low primary pump speed after 3 s FW and MSIV isolation

on secondary circuits – Startup of DHR system (3 out of 4 IC loops) Core flow rate reduces down to about 15% in 20 s and then progressively to

about 4% after 120-150 s

0

50

100

150

200

250

300

0 50 100 150 200 250

Po

we

r (M

W)

Time (s)

Core power

IC power

0

50

100

150

200

250

300

0 50 100 150 200 250P

ow

er (

MW

)

Time (s)

Core power

IC power

17


Core and IC powers

RELAP5 CATHARE

After reactor scram at 3 s the core power reduces down to decay level (calculated by the code – higher core decay power level with CATHARE)

DHR system is promptly operational - Heat removal by 3 IC loops is about 4 MW with RELAP5 and about 8 MW with CATHARE (much enhanced steam condensation on the inner side of IC tubes)

350

400

450

500

550

600

0 50 100 150 200 250

Tem

per

atu

re (

°C)

Time (s)

T core in

T core out max

T clad peak

350

400

450

500

550

600

0 50 100 150 200 250

Tem

pe

ratu

re (°

C)

Time (s)

T core in

T core out max

T clad peak

18


Core inlet/outlet and clad peak temperatures

RELAP5 CATHARE

Initial temperature peak at core outlet due to core flow rate reduction with delayed reactor scram (3 s)

Maximum value of clad peak temperature calculated by RELAP5 (584 °C at 10 s) is well within the safety limits

0

50

100

150

200

250

300

0 50 100 150 200 250P

ow

er (

MW

)Time (s)

Core power

IC power

19

PLOF-PLOHS: Loss of AC power (1/3)(RELAP5 preliminary results)

0

5000

10000

15000

20000

25000

0 50 100 150 200 250

Ma

ss fl

ow

ra

te (

kg/s

)

Time (s)

Core flow

Core mass flow rate Core and IC powers

RELAP5 RELAP5

IE: loss all primary and FW pumps with reactor scram at t = 0 s FW and MSIV isolation on secondary circuits – Startup of DHR system (3 out of 4 IC

loops) Core flow rate reduction like in PLOF and instantaneous transition to core decay

level

20


300

350

400

450

500

550

0 40 80 120 160 200

Tem

pe

ratu

re (°

C)

Time (s)

T core in

T up plenum

T in SG ps

T out SG ps

Primary lead temperatures

350

400

450

500

550

0 50 100 150 200 250Te

mpe

ratu

re (

°C)

Time (s)

T core in

T core out max

T clad peak

Core in/out and clad peak temp.

RELAP5 RELAP5

No significant increase in primary lead temperatures No risk for lead freezing (T > 327 °C) at SG outlet in the initial part of the transient

after DHR startup with injection of cold water from IC loop Clad peak temperature is limited to 534 °C at t = 10 s, within the normal operation

limit of 550 °C

0

4

8

12

16

20

0 3000 6000 9000 12000 15000

Pow

er (

MW

)

Time (s)

Core power

IC power

21


300

340

380

420

460

500

540

0 3000 6000 9000 12000 15000

Tem

pe

ratu

re (°

C)

Time (s)

T core in

T up plenum

T in SG ps

T out SG ps

Core, SG and IC powers Primary lead temperatures

RELAP5 RELAP5

Maximum power removed by 3 IC loops is around 5.4 MW (1.8 MW per IC loop) Core decay power is efficiently removed by the 3 IC loops after about t = 2500 s When the IC removed power exceeds the core decay power the primary lead

temperatures start to reduce – The minimum lead temperature is calculated at the SG outlet – lead freezing point (327 °C) is reached after about t = 13000 s (3.6 h)

0

50

100

150

200

250

300

0 50 100 150 200 250

Po

we

r (M

W)

Time (s)

Core power

IC power

22

POVC: Loss of FW pre-heaters (1/2)(RELAP5 preliminary results)

350

400

450

500

550

0 50 100 150 200 250Te

mpe

ratu

re (

°C)

Time (s)

T core in

T core out max

T clad peak

RELAP5 RELAP5

Core, SG and IC powers Core in/out and clad peak temp.

IE: loss of FW pre-heaters with FW temperature (335 °C) down to 300 °C in 1 s Reactor scram on low FW temperature after 2 s FW and MSIV isolation on

secondary circuits – Startup of DHR with 4 IC loops to evaluate the risk of lead freezing

Core power reduces to decay level after reactor scram - Clad temperatures quickly reduce since the primary pumps remain in operation at nominal speed

300

340

380

420

460

500

0 3000 6000 9000 12000 15000 18000Time (s)

Tem

pera

ture

(°C

)

T core in

T up plenum

T in SG ps

T out SG ps

23

POVC: Loss of FW pre-heaters (2/2)(RELAP5 preliminary results)

RELAP5 RELAP5


Maximum power removed by 4 IC loops is around 7.2 MW (1.8 MW per IC loop) Core decay power is efficiently removed by the 4 IC loops after about t = 700 s When the IC power exceeds the core decay power the primary lead temperatures start

to reduce – Lead freezing point (327 °C) at SG outlet is reached after t = 17000 s (4.7 h)

0

4

8

12

16

20

0 3000 6000 9000 12000 15000 18000Time (s)

Pow

er (

MW

)

Core power

IC power

RELAP5

24

PTOC: SG feedwater flow +20% (1/2)(RELAP5 preliminary results)

0

50

100

150

200

250

300

350

0 100 200 300 400 500 600

Po

we

r (M

W)

Time (s)

Core power

SG power

IC power

360

380

400

420

440

460

480

500

0 100 200 300 400 500 600

Tem

pe

ratu

re (°

C)

Time (s)

T core in

T up plenum

T in SG ps

T out SG ps

RELAP5 RELAP5


IE: SG FW mass flow rate +20% in 1 s (over cooling of primary side) No reactor scram since the scram threshold set-points are not reached Increase in SG heat removal capability and core power balances at 313 MW power

level after about t = 300 s Maximum core temperature decrease at core inlet is of 14 °C

25

PTOC: SG feedwater flow +20% (2/2)(RELAP5 preliminary results)

0

400

800

1200

1600

2000

0 100 200 300 400 500 600

Tem

pe

ratu

re (°

C)

Time (s)

T fuel peak

T fuel average

T clad peak

-6

-4

-2

0

2

4

6

0 100 200 300 400 500 600R

eact

ivity

(pc

m)

Time (s)

Rea doppler

Rea fuel exp

Rea clad exp

Rea cool exp

Rea diagrid

Rea pads

Rea c.rods

Rea total

RELAP5 RELAP5

Clad peak and fuel temperatures Total reactivity and feedbacks

No significant fuel peak temperature increase Clad peak temperature reduces by 10 °C Core power evolution is determined by total reactivity behaviour – Negative

reactivity feedbacks mainly by doppler and fuel expansion - Positive reactivity feedbacks by radial core and coolant expansion

26

DEC (UNPROTECTED) TRANSIENTS

27

ULOF: Loss of primary flow (1/6)(RELAP5 and CATHARE preliminary results)

0

5000

10000

15000

20000

25000

0 300 600 900 1200 1500 1800

Mas

s flo

w r

ate

(kg

/s)

Time (s)

Core flow

0

5000

10000

15000

20000

25000

0 300 600 900 1200 1500 1800

Mas

s flo

w r

ate

(kg

/s)

Time (s)

Core flow

RELAP5 CATHARE

Core mass flow rate

IE: Coastdown of all primary pumps without reactor scram The secondary circuits remain in operation in forced circulation After an initial small core flow rate undershot natural circulation stabilizes in the

primary circuit - RELAP5 and CATHARE codes predict similar stable natural circulation flow values RELAP5 = 23.7% and CATHARE = 23.2% of nominal flow rate

28


0

50

100

150

200

250

300

0 300 600 900 1200 1500 1800

Pow

er (

MW

)

Time (s)

Core power

SG power

IC power

0

50

100

150

200

250

300

0 300 600 900 1200 1500 1800

Pow

er (

MW

)Time (s)

Core power

SG power

IC power

RELAP5 CATHARE

Core, SG and IC powers

The core power initially reduces due to negative reactivity feedbacks and then stabilizes at 205 (CATHARE) – 210 (RELAP5) MW in equilibrium with SG power

The SG power initially decreases due to reduced primary flow and then increases according with lead temperature increase at SG inlet

29


300

350

400

450

500

550

600

650

700

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T core in

T core out max

T core out ave

300

350

400

450

500

550

600

650

700

0 300 600 900 1200 1500 1800Te

mp

erat

ure

(°C

)Time (s)

T core in

T core out max

T core out ave

RELAP5 CATHARE

Core inlet and outlet temperatures

Initial lead temperature increase at core outlet max calculated value near 700 °C by RELAP5 at 15 s

Max core outlet temperature stabilizes just above 600 °C The core inlet temperature progressively decreases and then stabilizes at 344 °C

30


350

400

450

500

550

600

650

700

750

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T clad peak

T vessel

350

400

450

500

550

600

650

700

750

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T clad peak

T vessel

RELAP5 CATHARE

Clad peak and max vessel temperatures

The initial clad peak temperature increase is below 750 °C max calculated value is 738 °C by RELAP5 at 12 s

Clad peak temperature stabilizes below 650 °C – highest value calculated by CATHARE due to different heat transfer correlations used by the codes

No safety concern for clad and vessel wall temperatures

31


400

800

1200

1600

2000

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T fuel peak

T fuel average

400

800

1200

1600

2000

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T fuel peak

T fuel average

RELAP5 CATHARE

Fuel average and peak temperatures

Peak and average fuel temperatures reduce according to the decrease of core power level

32


-120

-80

-40

0

40

80

120

0 300 600 900 1200 1500 1800

Rea

ctiv

ity (

pcm

)

Time (s)

Rea doppler

Rea fuel exp

Rea clad exp

Rea cool exp

Rea diagrid

Rea pads

Rea c.rods

Rea total

-120

-80

-40

0

40

80

120

0 300 600 900 1200 1500 1800

Rea

ctiv

ity (

pcm

)Time (s)

Rea doppler

Rea fuel exp

Rea clad exp

Rea cool exp

Rea diagrid

Rea pads

Rea c.rods

Rea total

RELAP5 CATHARE

Total reactivity and feedbacks

The negative control rod and core radial expansion (pads at core top) feedbacks induced by temperature increase at core outlet are mainly counterbalanced by positive Doppler and fuel expansion feedbacks (fuel temperature decrease)

33

ULOF+ULOHS: All SGs and PP trip (1/5)(RELAP5 and CATHARE preliminary results)

0

5000

10000

15000

20000

25000

0 600 1200 1800 2400 3000 3600

Mas

s flo

w r

ate

(kg

/s)

Time (s)

Core flow

0

5000

10000

15000

20000

25000

0 600 1200 1800 2400 3000 3600

Mas

s flo

w r

ate

(kg

/s)

Time (s)

Core flow

RELAP5 CATHARE

Core mass flow rate

IE: Loss of offsite power (all SG FW and PP trip) without reactor scram Startup of DHR system on the secondary side (4 IC loops) The core mass flow rate initially reduces down to 20% of nominal value and then

progressively reduces down to a residual flow rate of about 1.0-1.6%

34


0

50

100

150

200

250

300

0 600 1200 1800 2400 3000 3600

Pow

er (

MW

)

Time (s)

Core power

SG power

IC power

0

50

100

150

200

250

300

0 600 1200 1800 2400 3000 3600

Pow

er (

MW

)

Time (s)

Core power

SG power

IC power

RELAP5 CATHARE


The core power reduces down due to negative reactivity feedbacks induced by core temperature increase – The power suddenly reduces down to about 170 MW and then progressively towards IC power level

Significant thermal inertia of secondary circuit contributes to remove power from the primary system in the initial phase of the transient

35


300

400

500

600

700

800

900

0 600 1200 1800 2400 3000 3600

Tem

per

atu

re (

°C)

Time (s)

T core in

T core out max

T core out ave

300

400

500

600

700

800

900

0 600 1200 1800 2400 3000 3600

Tem

per

atu

re (

°C)

Time (s)

T core in

T core out max

T core out ave

RELAP5 CATHARE


Initial core outlet temperature peak at about 700 °C and then progressive temperature increase up to about 800 °C

Temperature increase at core inlet is limited below 500 °C after 3600 s due to very low natural circulation flow rate in the primary circuit and large primary system thermal inertia

36


300

400

500

600

700

800

900

0 600 1200 1800 2400 3000 3600

Tem

per

atu

re (

°C)

Time (s)

T clad peak

T vessel

300

400

500

600

700

800

900

0 600 1200 1800 2400 3000 3600

Tem

per

atu

re (

°C)

Time (s)

T clad peak

T vessel

RELAP5 CATHARE


Initial clad peak temperature increase below 750 °C and then progressive temperature increase up to about 800 °C – Clad rupture may occur (to be verified)

Max vessel temperature increase is limited below 500 °C after 3600 s due to very low natural circulation flow rate which stabilizes in the primary circuit and the large primary system thermal inertia

37


-150

-100

-50

0

50

100

150

200

250

0 600 1200 1800 2400 3000 3600

Rea

ctiv

ity (

pcm

)

Time (s)

Rea doppler

Rea fuel exp

Rea clad exp

Rea cool exp

Rea diagrid

Rea pads

Rea c.rods

Rea total

-150

-100

-50

0

50

100

150

200

250

0 600 1200 1800 2400 3000 3600

Rea

ctiv

ity (

pcm

)Time (s)

Rea doppler

Rea fuel exp

Rea clad exp

Rea cool exp

Rea diagrid

Rea pads

Rea c.rods

Rea total

RELAP5 CATHARE


The negative control rod, core radial expansion (pads) and coolant expansion feedbacks induced by temperature increase at core outlet are mainly counterbalanced by positive Doppler and fuel expansion feedbacks (fuel temperature reduction with decreasing core power)

38

UTOP: Reactivity insertion (1/6)(RELAP5 and CATHARE preliminary results)

-150

-100

-50

0

50

100

150

200

0 5 10 15 20 25 30

Rea

ctiv

ity (

pcm

)Time (s)

Rea doppler

Rea fuel exp

Rea clad exp

Rea cool exp

Rea diagrid

Rea pads

Rea c.rods

Rea total

-150

-100

-50

0

50

100

150

200

0 5 10 15 20 25 30

Rea

ctiv

ity (

pcm

)

Time (s)

Rea doppler

Rea fuel exp

Rea clad exp

Rea cool exp

Rea diagrid

Rea pads

Rea c.rods

Rea total

RELAP5 CATHARE


IE: Insertion of 250 pcm in 2 s without reactor scram (beta = 335 pcm) The secondary circuits remain in operation in forced circulation The inserted reactivity is mainly counterbalanced by negative Doppler and fuel

expansion feedbacks induced by fuel temperature increase Total reactivity reaches a maximum of about 175 pcm at 2 s and then reduces

according to negative feedbacks

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0

200

400

600

800

1000

0 5 10 15 20 25 30

Pow

er (

MW

)

Time (s)

Core power

0

200

400

600

800

1000

0 5 10 15 20 25 30

Po

we

r (M

W)

Time (s)

Core power

RELAP5 CATHARE

Core power

The core power increases up to 870 MW (about 300%) in 2 s and then quickly reduces down to about 450 MW (150%) at t = 10 s

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0

200

400

600

800

1000

0 300 600 900 1200 1500 1800

Pow

er (

MW

)

Time (s)

Core power

SG power

IC power

0

200

400

600

800

1000

0 300 600 900 1200 1500 1800

Pow

er (

MW

)Time (s)

Core power

SG power

IC power

RELAP5 CATHARE


After the initial transient the core power progressively reduces and stabilizes at about 380 MW in equilibrium with SG removed power

SG power increases according to temperature increase at SG inlet on primary side and consequent steam outlet temperature increase on the secondary side (constant FW flow rate)

41


350

400

450

500

550

600

650

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T core in

T core out max

T core out ave

350

400

450

500

550

600

650

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T core in

T core out max

T core out ave

RELAP5 CATHARE


After an initial jump of about 40 °C the core outlet temperature progressively increases according to core temperature increase at core inlet

The max core outlet temperature stabilizes at about 600 °C

42


350

400

450

500

550

600

650

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T clad peak

T vessel

350

400

450

500

550

600

650

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T clad peak

T vessel

RELAP5 CATHARE


After an initial jump of about 60 °C the clad peak temperature progressively increases and stabilizes below 650 °C

The max vessel temperature remains below 500 °C There is no safety concern for maximum clad and vessel temperatures

43


1000

1400

1800

2200

2600

3000

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T fuel peak

T fuel average

1000

1400

1800

2200

2600

3000

0 300 600 900 1200 1500 1800

Tem

per

atu

re (

°C)

Time (s)

T fuel peak

T fuel average

RELAP5 CATHARE

Fuel average and peak temperatures

The fuel peak temperature reaches a maximum value of 2600 °C in the initial part of the transient and then progressively reduces around 2400 °C

Fuel melting seems excluded (MOX melting point ~ 2673 °C) Fuel rod gap dynamic behavior (not modeled) may significantly affect the UTOP

transient results further confirmation is needed using a more realistic gap model

44

Preliminary conclusions (1/2)

The preliminary accident analysis for ALFRED has confirmed the good inherent safety futures of the design that mainly rely on: Low pressure drops in the primary system with enhanced natural

circulation after primary pump trip

Large primary system thermal inertia for slowing down the transients

Redundant systems working in natural circulation for core decay heat removal

Significant negative reactivity feedbacks for limiting the core power and temperature increase during transients

45

In particular the preliminary transient analysis for ALFRED has confirmed that: In case of DBC (Protected Accidents) the prompt reactor scram

actuation by the protection system and the startup of the decay heat removal system is able to maintain the core and vessel temperatures within the safety limits with adequate margin

In case of DEC (Unprotected Accidents) the core degradation and vessel failure is excluded and a large grace time is left to the operator to take the opportune corrective actions for bringing the plant in safe conditions in the medium and long term

The preliminary results must to be confirmed by further analysis, taking into account the fuel rod gap dynamic behaviour and enlarging the analysis to the whole set of representative accident initiators for DBC and DEC

Preliminary conclusions (2/2)

“Design and safety analysis of ALFRED” Accident analysis overview G. Bandini ENEA UTFISSM-SICSIS...

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Transcript of “Design and safety analysis of ALFRED” Accident analysis overview G. Bandini ENEA UTFISSM-SICSIS...