pipe break accident analysis of stella-2 using mars-lmr

Pipe Break Accident Analysis of STELLA-2 using MARS-LMR

Jewhan LEE, Yong-Bum LEE

Korea Atomic Energy Research Institute, 989-111Dadeok-daero, Yuseong-gu, Daejeon, Korea *Corresponding author: [email protected]

1. Introduction

One of the Design Basis Events(DBE) of pool-type

Sodium-cooled Fast Reactor (SFR) is “Pipe Break” and

it is more precisely the primary pump discharge pipe

connected to the inlet plenum where the core inlet starts.

Unlike the LWR flow piping, the system is under

atmospheric pressure and thus there is very little chance

of guillotine break. However, for conservatism, the

safety analysis assumes the total break of the discharge

pipe and the coolant flows out to the cold pool.

In the case of pipe break accident, the appropriate flow

path for natural circulation inside the pool cannot be set

and due to leakage point and eventually the coolant

flow rate cannot be developed enough to remove the

core decay heat. Therefore, it must be addressed during

the safety analysis with the consideration of various

conditions.

In Korea Atomic Energy Research Institute (KAERI),

there is a large-scale integral effect test facility to

support the SFR development, namely the STELLA-2

facility established under the Sodium Integral Effect

Test Loop for Safety Simulation and Assessment

(STELLA) program[1] and it is capable of simulating

the Pipe Break Accident. The facility has a dedicated

sub-system consisting of a very special valve and flow

path to simulate the Pipe Break.

In this study, various transient behavior under pipe

break events with diverse decay heat removing

combinations were analyzed using MARS-LMR. The

scope of this analysis bounds within the preliminary

code calculation results. The comparison with the

experiment data and verification will be continued next

year.

2. STELLA-2 Facility

The STELLA-2 is a down-scaled test facility to verify

the performance of DHRS of the reference reactor. At

the same time, the experiment database is used for

V&V of safety analysis code[2,3].

The STELLA-2 includes all the major components of

the reference reactor except the nuclear fuel core, the

steam generator, and the mechanical pump. Instead, the

electric core simulator[4], the straight tube-type sodium

to air heat exchanger and the EMP replaces each

component. In the STELLA-2, there are four lines of

DHRS. Two loops are for the passive heat exchanger

and the other two loops are for the active heat

exchanger. All four heat exchangers are of same

capacity.

The facility was designed to conserve the characteristic

and transient behavior of the reference reactor and it

was evaluated at several stages using various means

and tools including CFD and system code.

Fig. 1 STELLA-2 Facility Layout

Fig. 2 STELLA-2 Installation and Control System

3. MARS-LMR Analysis

3.1 Pipe Break Accident

The various conditions for the pipe break events are

listed in Table 1 for STELLA-2 facility. Based on this

test matrix the experiments are going to be conducted

and will be compared with the preliminary analysis

result.

For the MARS-LMR analysis, the event progress with

time is as follows.

1. Pipe break valve open: 4.47s

2. Pump trip: 4.47s

3. UHX blower off: 4.47s

4. Rx trip: 5.81s

5. DHRS starts to operate: 8.26s

Transactions of the Korean Nuclear Society Virtual Autumn Meeting

October 21-22, 2021

6. (1) 1 PDHRS + 1 ADHRS working

(2) 2 PDHRS + 2 ADHRS working



The time is set to satisfy the time scale ratio between

the reference reactor and the STELLA-2. Therefore, the

assumption of the event start time in real scale is 10sec

and this corresponds to 4.47sec in STELLA-2.

Table 1 STELLA-2 Pipe Break Test Matrix

PHTS

Pump

Discharge

Pipe Break

- Rx Trip

- with LOOP

- 1 line Break

- IHTS Na is not

considered

- PHTS pump 1&2 stops

- Break Simulation Valve On

- IHTS pump 1&2 stops

- DHRS working condition:

· 2 passive + 2 active

- Rx Trip

- with LOOP

- 1 line Break

- IHTS Na is

considered

- IHTS sodium inventory

consideration:

· SG F/W drayout simulation

using UHX blower

PHTS

Pump

Discharge

Pipe Break

+ DHRS 1

loop fail

- Rx Trip

- with LOOP

- 1 line Break

- IHTS Na is not

considered







- Rx Trip

- with LOOP

- 1 line Break

- IHTS Na is

considered


consideration:


using UHX blower

PHTS

Pump

Discharge

Pipe Break

+ DHRS 2

loops fail

- Rx Trip

- with LOOP

- 1 line Break

- IHTS Na is not

considered





· 2 passive

· 2 active


- Rx Trip

- with LOOP

- 1 line Break

- IHTS Na is

considered


consideration:


using UHX blower

3.2 Node layout and Assumptions

The node layout is shown in Fig. 3. Based on the heat

balance of the reference reactor design, the steady-state

point was set to match the temperature distribution

inside the pool and the transient started with the

opening of the valve. The primary and intermediate

loop pumps stop and the core heater starts to follow the

decay heat curve.

Fig. 3 Node Diagram

4. Results and Discussion

The representative case result of 1 passive + 1 active

DHRS working condition is described as follows. This

case is selected because this condition is the least heat

removing condition and is the worst case for the core.

The other cases are not included in this paper for it

showed similar trend with small difference in heat

removing capacity.

The pipe break occurs at the Pump Discharge line 4 and

connects to the cold pool directly. The calculation time

of analysis was upto 10,000 seconds.

4.1 Flowrate Trend

PHTS Pump Inlet 1: 1.25kg/s (0.0s) → 0.053kg/s

(200s, min flowrate at early stage) → 0.097kg/s

(540s, max flowrate at early stage) → continuous

decrease → small back flow starts at 890s →

1,380s flow recovers direction and increase →

0.29kg/s(3,190s, max flowrate and decrease) →

0.1kg/s(10,000s)

PHTS Pump Inlet 2: 1.25kg/s (0.0s) → 0.043kg/s

(190s, min flowrate at early stage) → 0.085kg/s

(550s, max flowrate at early stage) → continuous

decrease → small back flow starts at 880s →

1,280s flow recovers direction and increase →

0.29kg/s(3,160s, max flowrate and decrease) →

0.1kg/s(10,000s)

Pump Discharge 1&2: the trend is same as the

PHTS Pump Inlet 1 at the same timeline

Pump Discharge 3: most of sodium back flow from

Discharge line 4 goes through Discharge line 3 and

to the Inlet Plenum(Core Inlet). The trend is

symmetrical to Discharge line 4 (opposite direction

and same flowrate).

Pump Discharge 4: flowrate decreases rapidly after

19s → back flow from cold pool to inlet plenum

starts at 23s → -0.64kg/s (44s max back flowrate

at early stage) → -0.45kg/s (320s, min back

flowrate at early stage) → -0.68kg/s (max back

flowrate) → 1,890s flow recovers direction →

0.11kg/s (3,180s, max flowrate) → 0.00023kg/s

(10,000s)

190

02

01

100

02

01

100

06

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03

02

100

01

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05

01

02

09

0317

01

03

193

170

926

01

153

913

944

933

925

948

945

DHX

202

208

IHX

242

02183

01

01

04

06

07

02

03

06

06

946

562

570

Cold Leg - 540

Hot Leg - 526

590

586

592

AHX

248

282

288

280

273

291

01

404

412

UHX

Pump

Cold Leg - 750

Hot Leg - 726

292

294

Hot Leg - 275

Cold Leg - 290

FHX

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RVCS


October 21-22, 2021

Inlet Plenum to Cold Pool line: the trend is similar

to that of Pump Discharge line 4 but with different

flowrate.

Through Core:

flowrate decrease after 5s →

1.13kg/s (290s, min flowrate at early stage) →

1.69kg/s (640s, max flowrate) →

0.5kg/s (10,000s)

Passive DHX shell side:

sudden increase after 140s

and the max flowrate(0.38kg/s) at 250s → 0.25kg/s

(10,000s)

Active DHX shell side:

sudden increase after 110s

and the max flowrate(0.41kg/s) at 140s → 0.22kg/s

(10,000s)

AHX tube side:

starts to increase after 30s and

0.5kg/s of max flowrate at 210s → slowly

decreases and 0.37kg/s at 10,000s

FHX tube side: starts to increase after 30s and

0.51kg/s of max flowrate at 120s → slowly

decreases and 0.32kg/s at 10,000s

Fig. 4 Core and Pump Inlet Sodium Flowrate

Fig. 5 Pump Discharge Line Sodium Flowrate

Fig. 6 DHX Shell-side Sodium Flowrate

Fig. 7 AHX Tube-side Sodium Flowrate

Fig. 8 FHX Tube-side Sodium Flowrate

4.2 Temperature Trend

In early stage, the natural circulation flow through

the core develops sufficiently to decrease the core

outlet temperature → the flowrate reaches the max

at 640s and slowly decreases → the min core outlet

temperature 443℃ at 680s → temperature slowly

increases and reaches 506℃ at 10,000s

1 10 100 1000 10000

0.0

0.5

1.0

1.5

2.0

2.5

Flo

wra

te, kg

/s

Time, s

Core inlet

2 PSLS

1 10 100 1000 10000-1.0

-0.5

0.0

0.5

1.0

1.5

Flo

wra

te, kg

/s

Time, s

Discharge-1 Discharge-2

Discharge-3 Discharge-4

Discharge-5

1 10 100 1000 10000-0.1

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

Flo

wra

te, kg

/s

Time, s

DHX-1 Shell DHX-2 Shell

DHX-3 Shell DHX-4 Shell

1 10 100 1000 10000-0.1

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

Flo

wra

te, kg

/s

Time, s

AHX-1 Tube

AHX-2 Tube

1 10 100 1000 10000-0.1

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

Flo

wra

te, kg

/s

Time, s

FHX-1 Tube

FHX-2 Tube


October 21-22, 2021

Core inlet temperature starts to increase after

3,000s and reaches 430℃ at 10,000s

AHX tube-side inlet sodium temperature maintains

497℃ upto 200s and starts to decrease to 238℃ at

10,000s. AHX tube-side outlet sodium temperature

rapidly decreases until 140s to 304℃ and then

increases to 360℃ at 290s. It reaches 165℃ at

10,000s

FHX tube-side inlet sodium temperature maintains

499℃ upto 140s and starts to decrease to 243℃ at

10,000s. FHX tube-side outlet sodium temperature

rapidly decreases until 110s to 315℃ and then

increases to 379℃ at 170s. It reaches 174℃ at

10,000s

Fig. 9 Core In/Out Sodium Temperature

Fig. 10 DHX Shell-side In/Out Sodium Temperature

Fig. 11 AHX Tube-side In/Out Sodium Temperature

Fig. 12 FHX Tube-side In/Out Sodium Temperature

4.3 Heat Removal Trend

At the end of calculation time (10,000s), the core

produces 48.4kW

PDHRS and AHX removes 35.9kW

ADHRS and FHX removes 29.0kW

Fig. 13 Core Heat Transfer Rate

1 10 100 1000 10000250

300

350

400

450

500

550

600

650

Tem

pe

ratu

re, x()

x(2

103

)x()

Time, s

Core Inlet

Core Outlet

1 10 100 1000 10000100

200

300

400

500

600

700

x()

Tem

pe

ratu

re, x()

x(2

103

)x()

x()

Time, s

DHX-1 Shell In DHX-1 Shell out




1 10 100 1000 100000

100

200

300

400

500

600

700

x()

Tem

pe

ratu

re, x()

x(2

103

)x()

x()

Time, s

AHX-1 Tube In AHX-1 Tube Out

AHX-2 Tube In AHX-2 Tube Out

1 10 100 1000 100000

100

200

300

400

500

600

700

x()

Tem

pe

ratu

re, x()

x(2

103

)x()

x()

Time, s

FHX-1 Tube In FHX-1 Tube Out

FHX-2 Tube In FHX-2 Tube Out

1 10 100 1000 100000

100

200

300

400

500

He

at T

ran

sfe

r R

ate

, kW

Time, s

Core


October 21-22, 2021

Fig. 14 AHX Heat Removal Rate

Fig. 15 FHX Heat Removal Rate

4.4 Summary

The developed natural circulation flow at 10,000 for

(1) Core: 0.5kg/s

(2) PDHX Shell-side: 0.25kg/s

(3) ADHX Shell-side: 0.22kg/s

The natural circulation flow through DHX shell-side is

a local path flow developed within cold pool and is

almost same as the main path(core–hot pool–cold pool–

core) flow (Approximately 94%).

In early stage of the accident, the decay heat from the

core is larger than the heat removed by 1 AHX and 1

FHX but it balances after 190s and reverses. At 1,360s,

the difference is at max and the difference slowly

decreases upto 10,000s.

The flow peak occurs at about 600s in Fig. 4 due to the

incoming sodium flow from cold pool through the

breakage point to the core. This flow develops because

the pressure drop is smaller than the normal path.

5. Conclusion

In this study, the pipe break events of STELLA-2

facility with various conditions were analyzed with

MARS-LMR. As expected, the back flow from cold

pool to the inlet plenum occurs through the pump

discharge pipe but the effect of decay heat removal was

not significantly large at early stage of the event.

However, the long-term behavior of heat removal is

negative and there should be a technical solution to

secure the long-term coolability of the core. The further

study of comparison with experiment data will show

more realistic analysis results and hopefully provide

feedback to the safety design of the reference reactor.

ACKNOWLEDGEMENT

This work was supported by the National Research

Foundation of Korea (NRF) grant funded by the Korea

government (MSIT). (No. 2021M2E2A2081063 & No.

2021M2D1A1084836)

REFERENCES

[1] J. Eoh et al., “Computer Codes V&V Tests with a

Large-Scale Sodium Thermal-Hydraulic Test Facility

(STELLA),” American Nuclear Society 2016 Annual

Meeting, New Orleans, US, June 12-16, 2016.

[2] J. Eoh, “Engineering Design of Sodium Thermal-

hydraulic Integral Effect Test Facility (STELLA-2)”,

KAERI SFR Design Report, SFR-720-TF-462-

002Rev.00, 2015.

[3] J. Lee et al., “Design Evaluation of Large-scale

Sodium Integral Effect Test Facility (STELLA-2) using

MARS-LMR”, Annals of Nuclear Energy, Vol. 120,

pp.845-856, 2018.

[4] J. Lee et al., “Design of Electric Core Simulator

System (ECSS) for Large-scale Integral Effect Test

Facility”, Annals of Nuclear Energy, Vol. 133, pp.762-

776, 2019.

1 10 100 1000 10000

0

20

40

60

80

100

120

He

at T

ran

sfe

r R

ate

, kW

Time, s

AHX-1 Tube AHX-2 Tube

AHX-1 Shell AHX-2 Shell

1 10 100 1000 10000

0

20

40

60

80

100

120

He

at T

ran

sfe

r R

ate

, kW

Time, s

FHX-1 Tube FHX-2 Tube

FHX-1 Shell FHX-2 Shell


October 21-22, 2021

pipe break accident analysis of stella-2 using mars-lmr

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