ML17212B292

Emergency Procedure 2-0120041,Revision 0, Steam Generator Tube Rupture.

ML17212B292
Person / Time
Site:	Saint Lucie
Issue date:	10/13/1981
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17212B290	List: ML17212B289 ML17212B291 ML17212B292 ML17212B293 ML17273A036
References
2-0120041, 2-120041, NUDOCS 8201130429
Download: ML17212B292 (31)

Text

EMERGENCY PROC ED URE 2-0120041 REV 0 SGTR FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UNIT 2 EMERGENCY PROCEDURE NUMBER 2-0120041 REVISION 0 October 13, 1981 STEAM GENERATOR TUBE RUPZURE REV FRG APPROVAL PLT MGR TOTAL NO OF PAGES 20 820ii30429'20108 0O000389 PDR aDOCV

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SGTR EMERGENCY EROCEDURE NRfBER 2<<0120041 REVISION 0 1~ 0 SCOPE:

This procedure provides operator instruction for two conditions:

(A) S/G tube leak less than charging pump capacity (B) S/G tube leak greater than charging pump capacity. This requires use of procedure 2-0030130 (RT/TT) concurrent with these ins true tions.

2. 0 SYMPZ Pi1S:

2.1 Unique to this incident: 2. 1 Radiation monitoring system 2.1-1 S/G Blowdown Monitor Alarm 2.1.2 Condenser Air Effector Alarm NOTE: Any or all of the following may be evident due to a tube failure.

2.2 Decreasing PRZR level 2.2 Indications BU Charging pump start PRZR heaters de-energize 2.2 Alarms H-17,H-18,H-25,H-26,H-29,H-30 2.3 Decreasing PRZR pressure: 2.3 Indications BU heaters energize 2.3 Alarms H-9,H-10,H-1,H~2,H-3,H-4

SGTR EHERGENCY PROCEDURE NIMBER 2-0120041 REVISION

2. 0 SYMPTOMS: (Cont. )

2.4 Initial increase in affected S/G 2.4 Dependent on size of level followed by return to tube leak programaed level LR-9011, 9012 2.5 Feed flow less than steam flow on 2.5 Dependent on size of affected S/G tube leak FR-8011/9011 FR8021/9021 2.6 Decreasing letdown flow 2.6 Caused by decreasing PRZR level FIA<<2202 2.7 Increasing charging flow ~ 2~ 7 Will cause VCT level to decrease LIC-2226 FIA-2212

2. 7 Alarms M-3

4S SGTR EMERGENCY PROCEDURE NRfBER 2-0120041 REVISION 0 3.,0 "

AUTOMATIC ACTIONS:

3o I LEAK ( CHARGING PUMP CAPACITY 3.1.1 PRZR level controls close to minimum .

3.1.2 S/G blowdown and sample valves close on high radiation 3.2 LEAK > CHARGING PMP CAPACITY 3.2.1 Reactor trip from TM/LP (var iab le) 3.2.2 SIAS when RCS pressure ~

decreases to 1600 PSIA 30 2 ~ 3 CIS from initiation of SIAS

3. 2.4 Turbine trip from reactor trip 3.2-4.1 FW Reg valves close and 15X bypass valves open to 5X flow position

3. 2.5 PRZR level controls close to minimum 3.2.6 PRZR heaters de-energize on low lov level 30 2 ~ 7 S/G blowdown and sample valves close on high radiation 3~ 2.8 PORV's open at'2400 PSIA

SGTR EMERGENCY PROCEDURE NIMBER 2-0120041 REVISION 4 0 IMMEDIATE OPERATOR ACTIONS:

4. 1 LEAK < CAPACITY OF CEQCING PUMPS 4 1.1 Ensure all required auto-matic actions have occurred 4.1.2 Start additional charging NOTE: Backup charging pumps as necessary do not auto start on Unit P2 4.1.3 IF EITHER RCS leakage exceeds 1 GPM OR Specific activity of s,econdazy system is

> ~ 1 MCI/GM dose equivalent I-131 Notify system dispatcher NOTE: Tech Specs require:

of impending load Be in Hot Standby within reduction and reactor 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Cold Shutdown shutd own within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> 4.1.4 Determine affected S/G:

4. 1.4.1 By comparing 4.1.4 1 Increasing level S/G levels indicates S/G tube leak 4.1.4.2 By Comparing 4.1.4-2 Steam Flow > Feed Flow S/G Steam Flaw/ indicates S/G tube Feed Flow leak 4-1.4.3 By comparing 4.1.4-3 High secondary S/G radiation radiation levels monitors indicate S/G tube leak

L SGTR EMERGENCY PROCEDURE NRIBER 2-0120041 REVISION 0 4.0 IMMEDIATE OPERATOR ACTIONS: (Cont.)

4. 1 (Cont. )

4. 1.5 Take manual control and 4.1.5 PIC-08-1A for A S/G close atmospheric steam PIC-08-1B for B S/G dump on a ffect ed S/G.

4.1.6 Ensure condenser air NOTE: Plant conditions may ej e cto r ven t is aligned necessitate implementation to plant vent of the emergency plan 4~2 LEAK > CAPACITY OF CKQCING PIMPS 4-2.1 Ensure all required auto-matic actions have occurred

4. 2.2 Start additional charging NOTE: Backup charging pumps pumps do not auto start on Unit 82 4.2.3 Notify system dispatcher NOTE: Tech specs require:

of load reduction and Be in hot standby within 6 reactor shutdown and hours and cold shutdown start reducing turbine within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> load 4.2.4 Determine affected S/G:

4.2.4.1 By comparing 4 2.4. 1 Increasing level S/G levels indicates S/G tube leak 4.2 4.2 By comparing 4 2-4-2 Steam Flow > Feed S/G Steam Flow indicates S/G Flow/Feed Flow tube leak 4.2.4.3 By canparing S/G 4-2.4-3 High secondary r adia t ion moni to rs radiation levels indicate S/G tube leak 4.2.5 Take manual control and 4-2.5 PIC-08-lA for A S/G close atmospheric steam PIC-08-1B for B S/G dump on affected S/G

EMERGENCY PROCEDURE NE1BER 2-0120041 REVISION 0 4.0 IMMEDIATE OPERATOR ACTIONS: (Cont. )

4. 2 (Cont- )

4.2.6 Ensure condenser air effector vent is aligned with plant vent

,4.2e7 Ensure reactor trip at NOTE: The NPS is responsible TM/LP setpoint and for implementing the perform Immediate Emergency Plan Operation Actions for

~

Reactor Trip/Turbine Trip, 2-0030130 5.0 SUBS UENT ACTIONS:

5.1 LEAK < CHAEGING PIMP CAPACITY CHECK 5.1.1 Refer to Reactor Trip/ Turbine Trip, 2-0030130 and ensure that all subsequent actions (Section 5) have been or are being performed 5.1.2 Commence turbine shutdown per OP 2-0030125 5.1.3 Comaence reactor shutdown per OP 2-0030128 5 1.4 Comaence reactor cooldown using manual control of steam dump to the condenser 5.1.5 Maintain no-load S/G level with feedwater control 5.1.6 At ~Arux 900 PSPA, RCS pressure:

5.1.6.1 Close HCV-08-1A or HCV-08-1B on affected generator 5.1.6.2 Ensure bypass valve MV-08-1A or MV-08-1B closed 5-1.6.3 Isolate feedwater to the affected S/G

SGTR EMERGENCY PROCEDURE NUMBER 2-0120041 REVISION 0 5.0 SUBS UENT ACTIONS: (Cont. )

5. 1 (Cont. )

CHECK 5.1.7 Continue reactor cooldown to ccld shutdown condition using the unaffected S/G c higher than Th may be observed in idle loop, due to small amount of reverse flow 5.1 8 Verify cold shutdown Boron Concentration 5.1.9 Take condensate system to determine activity level 5.1.10 Take air particulate and gaseous samples for:

5-1.10.1 Air e5ector after condenser and gland steam condenser combined vent 5.1.10.2 Steam driven aux feed pump exhaust 5.1.11 Conduct radiation surveys and post radiation areas as necessary 5~ 2 LEAK > CHABGIHG PQKP CAPACITY 5.2.1 Refer to Reactor Trip/Turbine Trip, 2-0030130 and ensure that all subsequent actions (Sectin 5) have been or are being performed 5.2.2 IF: Safety Infection is caused by low RCS pressure:

THEN: Verify CEA's inserted for > 5 seconds AND stop running RCP s 5 2.3 Ensure HPSI flow to the core when RCS pressure is < pump shutoff head (1250 PSIA)

SGTR EMERGENCY PROCEDURE NIMBER 2-0120041 REVISION 0 5.0 SUBSE UENT ACTIONS: (Cont.)

5. 2 (Cont. )

CHECK 5.2.4 Check the ESFAS bypass status board for malfunctioning equipment Verify equipment operation per Table I, Safety Injection Actuation and Table II, Containment Isolation 5.2.5 Stop any unnecessary running equipment including the emergency diesels if offsite power is available 5.2.6 Maintain no load S/G levels with AFW to the non-affected S/G 5.2.7 If steam driven pump is used, ensure steam supply is from the non-affected S/G NOTE: Use S/G levels, AFW header flow rate indicators to and recorders to verify feedwater flow 5.2.8 Ensure HPSI pumps and charging pumps restore PRZR level and pressure NOTE: PRZR pressure should stabilize 9 approx 1175 PSIA and level 8 approx 10X 5.2.9 Restore ICW to TCW heat exchangers, OPEN MV-21-2 and MV-2 1-3 5.2.10 Close condenser hotwell reject LCV-12-5 5.2.11 Comnence RCS cooldown using SBCS. If not available, use atmospheric steam dump on non-affected S/G CAUTION: Do not exceed 75o/HR cooldown rate.

4$

SGTR EMERGENCY PROCEDURE NUMBER 2-0120041 REVISION 0

5. 0 SUBS UENT ACTIONS: (Cont. )

5. 2 (Cont. )

CHECK 5.2.12 If RCP operation is not possible, refer to Appendix A, Natural Circulation Cooldown 5.2.13 Stabilize RCS, T g 505op NOTE: This will ensure adequate NPSH to allow Pour (4) pump operation with RCS pressure 9 900 PSIA.

5.2.14 When PRZR level is > 30X, energize PRZR htrs 5.2.15 Throttle HPSI pp- discharge valves to maintain PRZR level 8 approx 35X 5.2.16 When RCS is 9 900 PSIA, which can be achieved via PRZR spray, isolate affected S/G. Close applicable MSIV and bypass.

5.2.17 Ensure feedwater to the affected S/G is isolated NOTE: When a S/G has been isolated, Tc h may be observed in the idle loop-This is due to a small amount of reverse flow.

~ 5.2.18 Continue RCS cooldown using SBCS or Atmospheric steam dump on non-affected S/G 5.2.19 Block MSIS 8 600 PSIA S/G Press 5.2.20 Establish and maintain 50op sub cooling in the RCS. Use all available indications for this determination.

5.2.21 Sample each S/G for activity

SGTR EMERGENCY PROC ED URE NUMBER 2-0120041 REVISION 0 5.0 SUBSE VENT ACTIONS: (Cont.)

5. 2 (Cont. )

CHECK 5.2.22 Sample RCS to determine fuel failure and verify shutdown Boron Concentration 5.2.23 Sample condensate system for radioactivity levels 5.2.24 Conduct radiation surveys and post areas as required, being especially mindful of any steam exhausts to atmosphere

SGTR EMERGENCY PROCEDURE NIHBER 2-0120041 REVISION 0 TABLE I SAFETY INJECTION ACTUATION SYSTEM (SIAS)

CONDITION CHECK (2) CCW PPS 2A, 2B, or 2C ON (2) CCW to Fuel Pool HX Isolation Valves MV-14-17, MV-14-18 Closed (4) CCW Hdr Non>>essential Isolation Valves HCV-14-8A, HCV-14-8B, HCV-14-9, HCV-14-1 0 Closed (2) CCW Outlet from Shutdown HX 2A", HX-2B, Valves, HCV-1 4-3A, HCV-1 4-3B ~Oen (2) LPSI Pumps (2) HPSI'umps On (4) LPSI Disch to Loops HCV-3615, HCV-3625, HCV-3635, HCV- 3645 OEen

'(8) HPSI Disch to Loops HCV-3617, HCV-3627, HCV-3637, HCV-3647 A Header HCV-3616, HCV-3626, HCV<<3636, HCV-3646 - B Header ~Oen (2) HPSI Hot Leg Leak Drain V3572, V3571 Closed (2) SI Test to RWT I<>21 Closed (2) Shield Bldg. Vent Isolation Valves FCV-25-32, FCV-25-33 ~Oee (2) Fuel Bldg. Emerg ~ Vent Isolation Valves FCV-25-30, FCV-25-31 Closed e

(6) Containment Sample Isolation Valves (on RTGB 206) FCV-26>>2, FCV-26%, FCV-26-6, FCV-26-1, FCV-26-3, FCV-26-5 Closed

EHERGENCY PROCEDURE NIMBER 2-0120041 REVISION 0

6. 0 PURPOS E/DISCUSSION:

6.1 The purpose of this procedure is to list the indications that will enable the operator to identify a Steam Generator Tube failure and to provide the action to be taken to control the accident and minimize radioactive release to the environment.

This procedure provides instructions for two cases, "Leak Within the Capacity of the Charging Pumps" and "Tube failure" (exceeds Charging Pump Capacity).

6.2 Discussion

A Steam Generator Tube Failure causes leakage of reactor coolant into the steam system. If the leakage exceeds the capacity of the charging pumps, pressurizer pressure will 'decrease rapidly, causing a thermal margin/low pressure trip. The subsequent cooldown following the reactor trip combined with the continued leakage of reactor coolant into the Steam Generator will cause a further reduction in pressurizer pressure and level, resulting in initiation of safety injection and containment isolation. The tube rupture will cause a reduction in reactor coolant system volume and due to reactor coolant leakage into the steam generator, the affected steam generator level will continue to increase after the feedwater block valves are closed by SIAS. The resulting decrease in RCS pressure and volume will result in the RCS briefly being at saturation conditions. The possibility the ex1sts for void formation in the reactor coolant system ~

Operator action should be directed toward prompt isolation of the affected Steam Generator, to minimize contamination of the steam system and prevent possible radioactive release to the environment. With the exception of a compound accident 1n which loss of power accanpanies the Steam Generator Tube failure, steam is dumped to the condenser, rather than the atmosphere, to .prevent gross release of contaminat1on to the environment. Action must be taken to identify the affected Steam Generator as soon as possible and to isolate its feedwater flow to prevent water slugging the steam lines.

The Steam Generator Tube failure accident is most severe when it occurs at low power levels, due to the low inventory of water initially present in the pressurizer.

SGTR EMERGENCY PROCEDURE NRiBER 2-0120041 REVISION 0

6. 0 PURPOSE/DISCUSSION: (Cont. )

6- 2 (Cont. )

If a controlled reactor shutdown is commenced, reduce plant load at a maximum rate which will not of itself cause a plant'rip; a power reduction of approximately 5X/minute is recomnanded ~

Minimize the use of the atmospheric steam dump valves. Any necessary releases that are made must receive appropriate authorization.

Use motor driven emergency or normal feedwater pumps to reduce the release of potentially radioactive steam from turbine driven pump exhausts.

intain reactor coolant pressure approximately equal to the affected steam Do not exceed a maximum cooldown rate of approximately 750F/HR Condensate activity may be transferred to the service water circulating system by way of condenser tube leakage. The condenser should be isolated if vacuum is lost and the condenser is not being used for reactor plant cooldown.

To facilitate cooldown a main steam isolation signal may be avoided by bypassing the signal setpoint on each safety channel.

If reactor coolant pressure control is maintained, a safety injection actuation signal may be avoided by bypassing the signal setpoint on each pressurizer pressure safety channel, thus facilitating cooldown and depressurization.

Maintain the affected steam generator level below the maximum indicatable level by draining by way of the blowdown or sampling systems to the radioactive waste system.

Although it is possible in the long term to note an increasing steam generator level, automatic feedwater modulation keeps the steam generator level approximately constant during the short term.

After the faulted steam generator has been isolated and the cooldown is proceeding via natural circulation an inverted T (i.e., Tc hi h than T ) may be observed in the idle loop. This is due to a small amount of reverse heat transfer in the isolated steam generator and will have no effect on natural circulation flow in the intact steam generator.

SGTR EMERGENCY PROCEDURE NIMBER 2-0120041 REVISION 0 APPENDIX "A" LOSS OP H)RCED COOLING CHECK Cooldown using natural circulation which can be verified by observing:

(1) Loop T is < full power T (440p) is constant or decreasing h is stable, not steadily increasing (4) No abnormal differences between Th-RTD's and core exit thermocouples When possible, restart (1) RCP in each loop to establish cooling in an isolated S/G. Use the following criteria.

(1) Unaffected S/G is removing heat from RCS (2) PRZR. level 6 pressure are responding to contol (3) RCS is > 200 p Subcooled (4) NPSH pressure is available to RCP's An alternate method of enhancing natural circulation is to fill and drain (to the WMS) the isolated S/G During RCS depressurization monitor for void formation as indicated by:

a) PRZR. level increase > expected b) PRZR level decrease when charging pumps operated c) With PRZR level control in "AUTO",

unanticipated letdown flmr > charging flow

AEGIR EMERGENCY MOCEDURE NUMBER 2-0120041 REVISION 0 APPENDIX "A" (Cone ~ )

LOSS OP H)RCE COOLING CHECK If voiding in the RCS is indicated (1) Isolate letdown (2) Stop depressurization operations (3) Stop cooldown (4) Repressurize RCS to climate voids by operating (a) Pressur izer heaters.

(b) HPSI pumps.

(c) Charging pumps.

'I SGTR EMERGENCY PROCEDURE NlMBER 2-0120041 REVISION 0 7e 0

REFERENCES:

7.1 Instruction Manual - Steam Generators, St. Lucie Unit No. 1 8770-5008.

7. 2 EBASCO Prints 8770-G-0 79, 080.

7.3 Accident Analysis, FSAR, Section 15.4.4, Steam Generator Tube FaQ.ur e.

7.4 CEN 117, Inadequate Cooling ~

8.0 Records Re uired:

8.1 Normal Log Entries.

8.2 Applicable Transient Recorder Charts 9 0 Approval:

Reviewed by: Plant Nuclear Safety Committee Approved by: For/Plant Manager Revision Reviewed by FRG Approved by: Plant Manager "LAST PAGE" Emergency Procedure 2-0120041 Rev. 0 TOTAL NO. OF PAGES 20