Steam Generator Tube Rupturep

Chapter 4.6

Objectives1. Discuss why operator intervention is

necessary to limit or prevent radiological releases during a Steam Generator Tube Rupture (SGTR) event.

2. Discuss the primary-side and secondary-

2

side indications of an SGTR in the control room.

3. Discuss how the affected SG may be identified either prior to or following the reactor/turbine trip.

Objectives (Cont)4. List the initial actions taken by the operator

once the affected SG has been identified.

5. Discuss the actions required to stop the primary-secondary leakage.

6. Discuss the problems associated with the

3

6 scuss e p ob e s assoc a ed efollowing: Secondary-to-primary leakage, SG Overfill.

7. List the principal systems/components affected by a loss of offsite power (LOOP).

Objectives (Cont)8. Discuss how plant cooldown and pressure

control are accomplished with an SGTR and LOOP.

9. Discuss what affect the following events had on the SGTR transient at the Ginna Plant:

4

Tripping of the reactor Coolant Pumps,

Failure of pressurizer power-Operated relief valve,

Automatic operation of letdown valves,

Pressurizer relief tank failure, and

Steam generator Safety Valve Failure.

SGTR

• Most frequent occurring major accident.

• Provides a direct release path for primary coolant to environment via SG and its safety/relief valves (Containment Bypass).

5

y ( yp )

• Timely operator involvement is required to prevent SG overfill and limit radiological releases.

Figure 4.6-1 Closeup View of SGTR

Past Steam Generator Tube Rupture Accidents at Pressurized Water Reactors

PlantDate Leak Rate (gpm) Cause

Point Beach Unit 1 February 26, 1975 125 wastage

Surry Unit 2 September 15, 1976 330 PWSCC in U-bend

Doel Unit 2 June 25, 1979 135 PWSCC in U-bend

Prairie Island 1 October 2, 1979 390 loose parts

Ginna Unit 1 January 25, 1982 760 loose parts and tube wear

7

Fort Calhoun May 16, 1984 112 ODSCC at a crevice

North Anna Unit 1 July 15, 1987 637 high cycle fatigue in a U-bend

McGuire Unit 1 March 7, 1989 500 ODSCC in the free span

Mihama Unit 2 February 9, 1991 700 high cycle fatigue

Palo Verde Unit 2 March 14, 1993 240 ODSCC

Indian Point Unit 2 February 15, 2000 150 PWSCC in U-bend

Industry Improvements

• Secondary Side Inspections

• Improved Steam Generator Designs

• Water Chemistry Control

8

• More Reliable Eddy Current Techniques

Primary-Side Indications of SGTR• Decreasing Prz level

• Decreasing Prz Pressure.

• Increasing charging flow.

These indication would also occur for a LOCA,how could the operator differentiate between

9

how could the operator differentiate betweenthe two?• Lack of degraded containment conditions.• Abnormal secondary side indications.

1 tube

10 tubes

PR

ES

SU

RE

(PS

IG)

1000

1250

1500

1750

2000

2250

2500

RCS pressure drop is one indication of SGTR or LOCA

Trip from 0T T or If reducing Turb Load, Low Prz Press

10

RC

SP

0

250

500

750

0 2 4TIME (MINS)

6 8 10

Figure 4.6-2(a) Initial Pressurizer Pressure Response4.6-35

Trip Trip

indication of SGTR or LOCA.

40

60

80

100

1 tube

RE

SS

UR

EL

EV

EL

(%S

PA

N)

Prz level drop is another indication of SGTR or LOCA.

11

Figure 4.6-2(b) Initial Pressurizer Level Response4.6-35

8 10

0

2010 tubes

RC

SP

R

0 2 4

TIME (MINS)

6

LOCA.

T hot

T cold

MP

ER

AT

UR

E(°

F)

300

400

500

600

700

12

Figure 4.6-3 RCS Temperature Following Reactor Trip4.6-37

6 8 10

TE

TIME (MINS)

0 2 4

0

100

200

Secondary-Side Indications of SGTR

• Chemistry Samples.

• Condenser Off Gas RM.

• SGBD Rad Monitor Alarms.

13

• Main Steam Line RM’s

• SF/FF mismatch on affected SG.

• SG level rise on affected SG.

AFFECTED SG

INTACT SG AFFECTED SG

VE

L(%

NA

RR

OW

RA

NG

E)

100

80

60

40

TRIP

14

Figure 4.6-4(a) Steam Generator Response Following Reactor Trip4.6-39

TIME (MINS)SG

WA

TE

RL

EV 20

0

1086420

SG

PR

ES

SU

RE

(PS

IG)

250

500

750

1000

1250

15

TIME (MINS)

S

Figure 4.6-4(b) Steam Generator Response Following Reactor Trip4.6-39

2 4 6 8 10

0

250

0

AM

FL

OW

(%N

OM

INA

L)

INTACT SGAFFECTED SG

100

80

60

40

20

16

ST

EA

Figure 4.6-4(c) Steam Generator Response Following Reactor Trip4.6-39

20

0

1086420

TIME (MINS)

Initial Operator Actions

• Isolate the affected SG.

1. Isolate Feedwater & AFW.

2. Close MSIV.

• These actions help minimize radiological

17

releases & prevent SG overfill.

• Isolating the affected SG also allows use of the non-affected SG for cooling down RCS.

Recovery Actions

• Following isolation of the effected SG the intent is to match RCS pressure with affected SG pressure to stop the leak.

• 1st step cooldown RCS using intact SG’s.

2nd t d i RCS i P

18

• 2nd step depressurize RCS using Prz sprays or PORV if RCPs unavailable.

• 3rd Step terminate SI to prevent re-pressurization of RCS.

RE

SS

UR

E(P

SIG

)

1000

1200

1400

1600

1800

2000

2200

(1)

(2)

(4) (10)

BREAK FLOW(N) = # of FAILED TUBES

19

0 20 40 60 80 100 120 140

FLOW (LB/SEC)

160 180 200

RC

SP

Figure 4.6-5 Equilibrium Break Flow4.6-41

600

800 SI FLOW

0

200

400

500

600

700

MP

ER

AT

UR

E(°

F)

300

400

T hot

T cold

RCS COOLDOWN INITIATED

Cooldown RCS to less than SaturationTemp for affected SG pressure.

i.e. SG Press 800-899 Core Exit TC 470o

Saturation for 899# is ~532o

20

10 20 30 40 50 60TIME (MINS)

TE

Figure 4.6-6 RCS Response - Offsite Power Available4.6-43

0

0

100

200 This will allow depressurization of RCSto SG press and stop the leak with the RCS sub-cooled.

TRIP

Cooldown Initiated

Prz Spray Initiated

SI TerminatedSI Flow=Break Flow

21

Prz Spray Terminated

Break FlowStopped

100

125

150

75

W(L

BM

/SE

C)

SI FLOW

As Primary is depressurizedBreak flow decreases and SIFlow increases refilling Pzr.

SI Flow=Break Flow Cooldown Initiated

SI Termination

22

10 20 30 40

TIME (MINS)

50 60

Figure 4.6-8 SI Flow and Break Flow4.6-47

0

25

50FL

OW

BREAK FLOW

0

SI Flow Break Flow Cooldown Initiated

Cooldown Stopped

40

60

80

100

UR

IZE

RL

EV

EL

(%s

pan

)SERIES COOLDOWN

andDEPRESSURIZATION

CONCURRENTCOOLDOWN and

DEPRESSURIZATION

SI Termination

SI Termination:RCS Sub Cooling >25o

SG FF >720gpm or 1 Intact SG >7%

23

TIME(MINS)

40 50 60

20

PR

ES

SU

0

0 10 20 30

Figure 4.6-10 Pressurizer Level Response - RCS Cooldown and Depressurization4.6-51

gpRCS Press stable or increasingPzr level >5%

2000

22002400

AFFECTED SG PRESSURE

PR

ES

SU

RE

(PS

IG)

600

800

1000

1200

1400

1600

1800RCS PRESSURE

24

0 10 20 30 40 50 60TIME (MINS)

Figure 4.6-7(a) Multiple Tube Failure Response4.6-45

AFFECTED SG PRESSURE

0200

400

When several tubes fail RCS pressure can drop below SG pressure during recovery. This can result in secondary – to – primary leakage which will dilute RCS boron concentration thus effecting SDM.

80

100

AFFECTED SG

INTACT SG

LE

VE

L(%

NA

RR

OW

RA

NG

E)

20

40

60

25

Figure 4.6-79(b) Multiple Tube Failure Response4.6-45

0 10 20 30 40 50 60

TIME (MINS)

SG

WA

TE

R

0

20

40

60

80

100

LE

VE

L(%

Na

rro

wR

an

ge

)

AFFECTEDSG

INTACTSG

Affected SG level recovers faster andhigher for = feedflow

100% Narrow Range is not a Solid SG

26

50 60

0

20

SG

WA

TE

RL

Figure 4.6-12 Steam Generator Levels4.6-55

TIME (MINS)0 10 20 30 40

Break Flow Stopped7%

SGTR with LOOP• No RCPs – natural circulation cooldown

required, slows recovery.

• No Steam Dumps to condenser – must use SG PORVs.

• Prz Sprays unavailable must use Prz

27

• Prz Sprays unavailable – must use Prz PORVs or Aux Spray.

• Overall this results in increased potential for rad releases and SG overfill

500

750

1000

12501300

PR

ES

SU

RE

(psi

g)

OFFSITE POWER AVAILABLE

LOSS OF SITE POWER

Steam Dumps 1092#Higher SG pressure SG PORV (1125#), or Safeties (1st @ 1170#). Can results in rad release from affected SG

28

0

250

SG

P

Figure 4.6-13 Steam Generator Pressure Following Reactor TripWith and Without Offsite Power

4.6-57

4

TIME (MINS)

0 2 6 8 10

Higher no-load Tave (1125#)Slower Pressure Decrease

29

Trip

T HOT

T COLD

EM

PE

RA

TU

RE

(0F

)

400

300

700

600

500

RCPs Trip resulting in Natural Circulation

30

6 8 10TIME (mins)

T

Figure 4.6-15 RCS Temperature Following Reactor Trip Without Offsite Power4.6-61

200

100

0

0 2 4

40

60

80

100

110

PF

LO

W(%

No

min

al)

Without RCPs running upperhead becomes inactive.Stays hot and prone to voidingduring subsequent cooldown

31

10

TIME (min)

0

20

40

LO

OP

Figure 4.6-16 Natural Circulation Flow Following Loss of Offsite Power4.6-63

0 2 4 6 8

300

400

500

600

700

MP

ER

AT

UR

E(°

F)

THOT

TCOLD

Cooldown done with PORV on SG vs Steam Dumps

32

0

0

10 20 30 40 50 60

100

200

TE

M

TIME (MINS)

Figure 4.6-17 Intact RCS Temperature, Without Offsite Power4.6-65

When the affected SG is isolatedThat loop may stagnate.

CS

PR

ES

SU

RE

(PS

IG)

1000

1250

1500

1740

2000

2250

PRZR PORV OPENED

Cooldown SGPORV

SI Termination

33Figure 4.6-18 RCS Pressure Response, Without Offsite Power

4.6-67

RC

TIME (MINS)

0 10 20 30 40 50 60

0

250

500

750 PRZR PORV CLOSED

During DepressurizationHead may void causing rapidIncrease in pressurizer level.

Questions

34

Ginna SGTR

35

January 25, 1982

RX

M

STBY AUX FEED PUMP

MOTOR DRIVENAUX FEED PUMP

HHSI

LPSI

SG

STBY AUX FEED PUMP

MOTOR DRIVENAUX. FEED PUMP

SV

ATMOS

MSIV

4 SAFETYVALVES

STOPVALVES

GOVERNORVALVES

TURB/GEN

ATMOS

4 SAFETYVALVES

MSIVBLOCKVALVE

TOPRT

PORV

M

HHSI

CIRCULATINGWATERPUMPS

SGPZR

36

AUX. FEED PUMP

CHARGING

HI-STRENGTH CONCRETESTEEL LINED CONTAINMENT

cvcsRCP

MAIN FEEDWATERPUMP

TURBINE DRIVENAUX. FEED PUMP

MAIN FEEDWATERPUMP

Fig 4.6-19 GINNA OVERVIEW

PRESSURIZER

M RHR

SISTE

SEAL RETURNRELIEF

PZRSPRAYVALVESPT

LTACC.

M SIS

RUPTUREDISC

TE LT

DEMINERALIZEDWATER SPRAY

TE

PCV435

AUXILIARY SPRAY

LETDOWNRELIEF

VENTHEADER

PT

HEATER CONTROL

TE

M

PCV 431C

PCV 430

PCV434

M

TE

37

M

MM

RCPM

S/G

STEAM GENERATOR

CHARGING

TE

CVCSLETDOWN LINE SIS

M

LP SISACC.

MSIS

M

AUXILIARYCHARGING

TE

TE

PT

M LOOP A

RHR HEATREMOVAL LOOP

SISLP SIS

M

TE

LOOP BRCP

Figure 4.6-20 Ginna RCS Piping and Instrumentation

38

47% Normal Pzr Level

Increase Charging and Load reduction stabilize Press and Level decreaseThis indicates leak ~750gpm

R T i l

39

Tube Rupture at ~09:25

Plant Computer Data Available

Rx Trip on lowPress at 1900#

BREAK FLOW(B S/G at 1050 psig)

3 SAFETYINJECTIONPUMPS

SU

RE

(ps

ig)

1200

1300

1400

1500

1600

1700

40

0800

200 400 600

FLOW (gpm)

800 1000

900

1000

1100PR

ES

S

Figure 4.6-23 Ginna SGTR - SI and Break Flow4.6-77

SI pumps recover RCS pressure to about1350# and with S/G at about 1050 there is~400 gpm going into “B” SG

"B" S/GISOL.

RCPTRIP

RXTRIP

PZRPORV

LOOP "B" COLD LEGTEMPERATURE

LOOP "A" COLD LEGTEMPERATURE

ESTIMATED

600

560

520

480

440

400

360

320TE

MP

ER

AT

UR

E(0

F)

Cooling downon “A” SG

41

START"A" RCP

TIMEPM

9:00 10 20 30 40 50 10:00 10 20 30 40 50 11:00 10 20 30 40 50 12:00AMAMAM

280

240

Figure 4.6-24 Ginna SGTR - Cold-leg Temperature4.6-79

At 9:40 “B” SG is isolatedB coolant loop stagnates.B loop reverse flow out theRuptured tube

Attempt to depressurizeUsing PORV and it sticksOpen on 4th cycle and isIsolated. Rapidly filling Pzr.

SI Termination Criteria exists

Tube Leak increasing“B” S/G Pressure

2 Chg Pps no letdownMaintains RSC ~40#>SGBreak Flow continues

Break FlowReverses

42

Heat removal changed to PORV

Letdown relief open

43

LE

VE

L(%

)"B" S/G LEVEL

"B" S/G PRESSURE

100

90

80

70

60

50

40

30

20

1100

1000

900

800

700

600

500

400

300

PR

ES

SU

RE

(ps

ig)

Control break-flow by keeping RCS pressure ~25# <SG press

44

RCS PRESSURE

12:00NOON1/25/82

1 2 3 4 5 6 7 8 9 10 11 1 2 3 4 5 6 7 8 9 10 11 1 2 3 4 5 6 7 8 9 10 1112:00NOON1/26/82

12:00MIDNIGHT

12:00MIDNIGHT TIME

20

10

0

300

200

100

Figure 4.6-27 Ginna SGTR - Long-Term Cooldown and Depressurization4.6-85

Ginna Lessons Learned• Tripping RCPs slowed down and

complicated recovery actions.• PORV failure resulted in sec-to-pri leakage,

RCS steam void formation, and Prz overfill.• Opening of letdown iso valves resulted in

overpressure of PRT

45

overpressure of PRT.• PRT rupture disc burst resulting in release of

Reactor Coolant to Containment.• Failure of SG Safety resulted in Rad releases

& complicated recovery actions (lower SG pressure.

ObjectivesObj-1 Discuss why operator intervention is necessary to limit or prevent radiological releases during a Steam Generator Tube Rupture (SGTR) event.

Without operator action the affected SG and associated

46

Without operator action the affected SG and associated steam line will fill water solid. S/G PORVs and Safety valves are not designed to pass water and may fail open. The steam lines may fail due to the weight of water in the lines. These failures could result in a containment bypass event releasing raditation directly to the environment.

Obj-2 Discuss the primary-side and secondary-side indications of an SGTR in the control room.

• Primary– Decreasing pressurizer level

– Decreasing RCS pressure

47

• Secondary– Condenser Offgas rad-monitor

– SGBD rad-monitor

– Main Steam rad-monitor

– SG Steam flow / feed flow mismatch

– SG water level deviation alarms

Obj-3 Discuss how the affected SG may be identified either prior to or following the reactor/turbine trip.

• Feedflow / Steam Flow mismatch• Water level in affected SG returns on scale before

48

other SGs

• Water level in affected SG higher than other S/Gs

• SGBD rad-monitor (if not common)

• Main Steam Line rad-monitor

Obj-4 List the initial actions taken by the operator once the affected SG has been identified.

• Isolate the affected SG.1. Isolate Feedwater & AFW.2. Close MSIV.

49

2. Close MSIV.3. Raise Setpoint on Affected SG ASDV

• These actions help minimize radiological releases & prevent SG overfill.

• Isolating the affected SG also allows use of the non-affected SG for cooling down RCS.

Obj-5 Discuss the actions required to stop the primary-secondary leakage.Isolate the affected SG

• Cooldown RCS to saturation temp below affected SG pressure

50

• Depressurize RCS down to SG Pressure

• When pressurizer level is restored and SI termination criteria are meant stop SI pumps.

• Pressurizer level will slowly decrease as RCS depressurizes to SG Press and break flow stops

Obj-6 Discuss the problems associated with the following: Secondary-to-primary leakage, SG Overfill.

• Secondary to Primary Leakage– Dilution of primary (SD Margin)

51

• SG Overfill– Safety Valves and PORVs are not designed to

pass water and may fail.

– SG Lines fill with water to the MSIV and main steam line may fail.

Obj-7 List the principal systems & components affected by a loss of offsite power (LOOP).

• No RCPs – natural circulation cooldown required, slows recovery.

52

q , y

• No Steam Dumps to condenser – must use SG PORVs.

• Prz Sprays unavailable – must use Prz PORVs or Aux Spray (backup).

Obj-8 Discuss how plant cooldown and pressure control are accomplished with an SGTR and LOOP.

• Cooldown using SG PORV vs Stm Dumps

• Depressurize using PORV vs Normal or

53

p gAux Spray

OBJ-9 Discuss what affect the following events had on the SGTR transient at the Ginna Plant:

Tripping of the reactor Coolant Pumps,

Failure of pressurizer power-Operated relief valve,

Automatic operation of letdown valves,

Pressurizer relief tank failure, and

54

,

Steam generator Safety Valve Failure.

The End

55