ML20126D142

Draft Emergency Operating Instructions EOI-3, Steam Generator Tube Rupture, Revision 7

ML20126D142
Person / Time
Site:	Sequoyah
Issue date:	03/31/1980
From:	Record D, Jacqwan Walker, George Wilson TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20126D128	List: ML20126D126 ML20126D130 ML20126D134 ML20126D138 ML20126D142
References
EOI-3, NUDOCS 8004170234
Download: ML20126D142 (13)

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8 Ecquoyrh !uclear Plan

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DISTRIDUTIO" ESTEGENCY OPE?ATING INSTRl'CTICN 1C Plant t!as ter File Superintendent 1U Assistant Superintendent (Oper.)

EOI-3 1U Assistant S u p e rin t.e nd e n t (!!a i n t . )

Administrative Supervisor STEAM CENERATOR TUEE El'PT"rr Maintenance Supervisor (!!)

Assistant flaintenance Supervisor (!!)

!!aintenance Superv2 so r (E)

Units 1S2 Assistant Maintenance Superv.tsor (E)

Maintenance Su;'e rvisor (I)

Results Supervisor 1C Operations Supervisor Quality Assurance Supervisor e

_ Health Physics Public Safety Services Supv.

Chief Storekeeper Preop Test. Program Coordinator Outage Director Chemical Engineer (?ccults) ,

Radiochem Laboratory Instrument Shop Reactor Engineer (Results)

Instrument En;;ineer (Maint. (!))

Mechanical Engineer (Results)

Staf f Industrial En;;ineer (F1, sys )

1c Training Center Coordinator PSO - Chickamauga Engrg l'I.it - SNP Prepared By: Georce Wilsen Public Safety Services - SNP 1c Shift Engineer's Office Revised By J._R. Uniher _

1r Unit Control Room 4

Submitted By:/d'/!#dt.I.

//' ffMd6 / 7 QA&A Rep. - SNP Health Physics Labora. tory Sper"isor 1 11 Asst Dir NUC PR (Oper), 727 E3-C 1 17 Nuclear Document Control Unit, 606 EB-C PORC Review: 3/3l f(C 1p Superintendent, Wi'NP date Superintendent, D F."P l , Superintendent, EENP is ," N!'3, W9C174C-F.

Approved By: [D.hg V- . D t .\ \ \

• 4 Sunv., HPiiPS ROD MS Super [?!tendent in NRC-IE:II Power Security Officer, 620 CST 2-C l Date Approved: 3\h\hhI \

Nuclear Materials Coordinater-1410 CUBE I Manager. OP-QA&A St.aff 1r Resident NRC Inspector - SNP l in NSRS, 249A HEB-K ic Technical Support Center I

Rev. No. Date Revised Pages Rev. No. Date Revised Paees 5 2/9/S0 All' 6 2/14/80 2 7 J , .3 /, R) All The last page of this instruction is Number 12 -

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SQNP E0I'3 -' Units 1 & 2 Page 1 of 1 Rev. 7 >

STEAM GENERATOR TUBE RUPTURE PURPOSE The objectives of these instructions are as follows:

t l 1. To minimize the releas'e of radioactive material by identifying and isolating the faulted steam generator and by reducing reactor coolant steam pressure below the steam generator safety valve settings.

2. To establish capability to supply feedwater to all steam generators and to isolate feedwater to the faulted steam generator.

3. To maintain the ability to remove the necessary residual heat from the reactor through the intact steam generators via the steam dump valves or power operated relief valves.

4. To maintain the reactor coolant system in a subcooled state during the recovery.

5. To prevent overflooding of the faulty steam generator.

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SQNP E01 Units ~1 & 2 Page 1 of 9 Rev. 7 STEAM GENERATOR TUBE RUPTURE I. IMMEDIATE ACTIONS Refer to section on immediate actions or E01-0, Immediate Actions and Diagnostics, if not already performed.

II. SUBSEQUENT OPERATOR ACTIONS CAUTION: The diesels should not be operated at idle or minimum load for ex-tended periods of time. If the diesels are shut down, they should be prepared for restart.

NOTE: If at any time during the conduct of steps A through H the faulted

' steam generator is positively identified, immediately proceed to step I.

Following completion of this step, the remainder of the recovery must be accomplished from the last step of steps A through H which had been com-pleted prior to identifying the faulted steam generator.

Make arrangements to sample centainment vtmosphere and steam generators NOTE:

to identify presence of abnormal radioactivity.

NOTE: The process variables referred to in this instruction are typically monitored by more than one instrumentation channel. The redundant  !

channels should be checked for consistency while performing the steps of this instruction. ,

l NOTE: The pressuriner water level indication should always be used in con-junction with other reactor coolant system idications to evaluate  ;

system conditions and to initiate manual operator actions. l A. Verify that all pressurizer power operated relief valves are closed.

Verify the open status and availability of power to all pressurizer  !

power operated relief valve backup isolation valves.

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B. Stop all reactor coolant pumps after the high head safety injection pump I operation has been verified and when the wide range reactor coolant pres-sure decreases to 1550 psig.

CAUTION: If component cooling to the reactor coolant pumps is isolated on a containment Phase B isolation signal, all reactor coolant pumps are to be stopped within 5 minutes because of loss of motor bearing cooling.

CAUTION: If reactor coolant pumps are stopped, the seal injection flow should be maintained.

I NOTE: The conditions given above for stopping reactor coolant pumps should l be continuously monitored through Step J of this instruction.

l l l NOTE: See Table 1 or computer subcooling program for subcooling margin I

and see Appendix A for guidelines on natural circulation. i'

4 SQNP E0I Units 1 & 2 ,,

Page 2 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont. )

C. If of fiste power and the condenser are available, open bypass valves, equalize pressure across flISV(s), open any closed main steam line iso-lation valves to provide a flowpath to the condenser dump valves.

D. Establish power sources necessary to operate at least one pressurizer power operated relief valve, at least one steam generator power operated relief valves, and charging and letdown flowpaths.

NOTE: Ensure that containment isolation is maintained, i.e., not reset until such time as manual action is required on necessary process streams.

E. Stabilize the reactor coolant system at approximately no-load temperature by steam dump to the main condenser if offsite power and the cocMenser i are availlable. If offsite power or the condenser is not available, l utilize the steam generator power operated relief valves to stabilize the reactor coolant system at approximately no-load temperature.

F. Regulate the auxiliary feedwater flow to the steam generators to restore ,

and maintain steam generator water level in the narrow range span, or in the wide range span at a level sufficient to assure that the U-tubes

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l are covered (76%).

G. If reactor coolant system pressure is above the low head safety injection pump shut-off head, manually reset safety injection so that safeguards equipment can be controlled by manual action. Stop the low head safety injection pumps and place in the standby node and request performance of SI-268 to verify position of P-4 contacts (failure of F-4 may prevent resetting safety injection).

CAUTION: If the reactor coolant system pressure decreases uncontrollably below the low head safety injection shut-off head, the low head safety injection pumps must be manually restarted to de-liver fluid to the reactor coolant system.

CAUTION: Automatic reinitiation of safety injection will not occur since the reactor trip breakers are not reset.

CAUTION: Subsequent to this step, should loss of offsite power occur, manual action (e.g., manual safety injection initiation) will be required to load the safeguards equipment onto the diesel powered emergency busses.

H. Identify the faulted steam generator by one or more of the following methods:

1. An unexpected rise in one steam generator water level with auxiliary feedwater flow reduced or stopped.

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E0I Units 1&2 .'_..

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Page 3 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.)

H. (Cont.)

2. High radiation from any one steam generator blowdown line radiation monitor.

3. High radiation from any one steam generator blowdown line, as de-termined by analysis or radiation detector.

4. High radiation from any one steam generator main steam line.

I. When the faulted steam generator has be::. posi;is e , 2. 1_ 11ed, then:

1. Stop all feedwater flow to the faulted steam generator.

2. Close the main steam isolation valve and bypass valves associated with the faulted steam generator.

3. Verify the closure of all power operated relief valves associated with the faulted steam generator.

4. Verify the steam driven auxiliary feedwater pump is not .being supplied with steam from the faulted S/G.

CAUTION: Do not proceed to step J until the faulted steam gen-erator has been identified and isolated.

NOTE: With faulted S/G isolated @ RCS temperature of 547 F, the faulted S/G pressure will be G 1000 psig.

J. After the faulted steam generator has been identified and isolated, begin a cooldown of the reactor coolant system, the rate of cooldown should be relatively fast, but not so fast as to cause the UHI accumulators to dump on low pressure. Terminate the cooldown at 497 F on the RCS.

1. If offsite power and the condenser are available, dump steam to the main condenser from the intact steam generators by manual control of s l the steam header pressure controller.

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2. If offsite power is not available or the main condenser is not avail- ]

able, dump steam from the intact steam generators through the steam  !

. generator power operated relief valves.

K. Declare Gas Cond'ition I or in the event blackout ecnditions exist, delcare

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l Gas Condition II. {

Sound the plant radiological emergency siren to expedite assembly of per-

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L. l sonnel and to reduce onsite doses to, personnel. l l

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SQNP E0I Units 1 & 2 Page 4 of 9 - - -

Rev. 7 l

II. SUBSEQUENT OPERATOR ACTIONS (Cont. )

fl. Survey meteorological information and dispatch the shif t HP technician to survey the downwind sector at the plant boundary and request HP section perform survey of secondary site of plant.

N. Transfer NR-45 to 1 SR and 1 IR detector.

O. Af ter the reactor coolant system temperature has been reduced to 50 F below the no-load temperature, if necessary begin a de pressurization of the reactor coolant system to a value equal to the faulted steam gen-erator steam pressure while maintaining 50 F subcooling.

NOTE: With RCS temperature @ 497 F, 50 F subcooling will be maintained down to E 1000 psig on the RCS.

NOTE: During subsequent controlled reactor coolant system depressur-ization, the reactor coolant system pressure criteria for tripping the reactor coolant pumps established in step B DOES NOT APPLY.

If the RCP's are in service, use the pressurizer spray to reduce the pressure.

If offsite power is not available, or the RCP's are not in service open one pressurizer PORV to decrease pressure.

NOTE: Prior to the initiation of a controlled RCS depressurization, there may be no indicated PZR level. As the depressurization process proceeds, an increase in indicated PZR level is expected as liquid replaces steam in the PZR.

CAUTION: Monitor containment indications to verify that a loss of reactor coolant other than the steam generator tube rupture is not in progress. If recirculation sump level or a contaiwaent sample (if available at this time) are not in the normal pre-event range, further accident recovery must be directed according to Emergency Instruction E0I-1, Loss of Reactor Coolant, step

!! (Small LOCA).

P. As the reactor coolant system pressure decreases, due to the quenching of the steam by the pressurizer spary or due to the steam release through the pressurizer PORV, monitor the pressurizer water level indications and stop the depressurization operation:

1. If the indicated water level in the pressurizer rises above 50 per- <

cent of span OR

2. As soon as the reac' tor coolant system pressure decreases to a value I equal to the faulted steam generator steam pressure. l l

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SQ"P p~ -- -- "

E01 Units 1&2 . -  ; ,

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Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.)

Q. When the reactor coolant system pressure is reduced to 1500 prig, iso-late the upper head injection system as follows or verify isolated:

1. Close FCV-87-21 and GAG

2. Close FCV-87-22 and GAG

3. Close FCV-87-23 and GAG 4 Close FCV-87-24 and GAG R. Isolate the cold leg accumulators by closing the RCS pressure drops below 1000 psig if content of 4-_ >. .a.>._ .u t ct been dumped to RCS.

NOTE: Power will have to be placed on these. valves.

1. FCV-63-118

2. FCV-63-98

3. FCV-63-80

4. FCV-63-67 i S'. Af ter the depressurization operation has been verified to have been term-inated (using the pressurizer PORV stem-mounted position indicators or acoustic valve position monitoring system and spray valve demand signal),

continue to monitor the reactor coolant system pressure and the pressurizer water level.

1. If the pressurizer water level continues to rise or remains nearly constant concurrent with a reactor coolant system pressure decrease, suspect leakage from the pressurizer steam space. Monitor the pres-sure relief tank (PRT) pressure, temperature and level to identify continuously increasing conditions. Close the PORV isolation valves

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if a reactor coolant leak to the PRT is identified. Monitor PRT con-ditions to verify PRT integrity.

CAUTION: If pressurizer relief tank integrity is lost, abnormal con-tainment conditions could exist and may not be true indi-cations of a continued loss of reactor coolant. If con-ditions o'f step S1 pers.ist after closing the pressurizer PORV isolation valves, further recovery must be directed according to E0I-1, Loss of Reactor Coolant, step M. The conditions of step S2 must be satisfied before proceeding to step T.

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SQNP EDI Units 1 & 2 Page 6 of 9 '~

Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.)

S. (Cont.)

2. If the pressurizer water level subsequently continues to increase concurrent with a reactor coolant system pressure increase con-current with verified PRT integrity, the safety injection flow is greater than the leak.

Then, when reactor coolant system pressure has increased by at least 200 psi (after shutting the spray valve or verified closure of the pressurizer PORV) and an indicated water level has returned in the pressurizer, stop all operating safety injection pumps not needed for normal charging and reactor coolant pump seal injection flow.

CAUTION: The diesels :hould not be operated at idle or minimum load for extended period of time. If the diesels are shut down, they should be prepared for restart.

NOTE: Following termination of safety injection, pressurizer pressure should decrease to a value equal to the faulted steam generator steam pressure.

T. Place all safety injection pumps in a standby mode and maintain operable safety injection flow paths.

i U. Verify main control room ventilation isolation (See S0I-30.13).

V. Verify U-2 containment equipment hatch temporary doors closed (734' El.)

W. Verify fuel handling floor equipment transfer hatch cover closed (734 El.

to lower elevations).

X. Re-establish charging and letdown flows to maintain the pressurizer water level in the operating range (approximately 25 percent indicated 1cvel):

CAUTION: If, during subsequent recovery actions, pressurizer water level cannot be maintained above 20 percent indicated level, manually initiate safety injection flow to re-establish pressurizer water level in the operating range. If pressurizer water level cannot be established by this method, return to step 0 and proceed with the instruction from that point.

1. Close seal. injection water flow control valve FCV-62-89.

2. Open the charging pump normal suction valves FCV-62-132 and FCV-62-133 from the VCT. .

l 3. Close the charging pump suction valves FCV-62-135 and FCV-62-136 from the refueling water storage tank.

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SQNP E01 Units 1 & 2 1 l

Page 7 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.)

X. (Cont.)

4. Open the centrifugal charging pumps miniflow isolation valves ICV-62-98 and FCV-62-99.

5. Open the charging line isolation valves FCV-62-90 and FCV-62-91.

6. Open seal water heat exchanger inlet isolation valves FCV-62-61 and FCV-62-63.

7. Gradually open the seal injection water flow control valve FCV ^'

Wiust the seal water flow to 8 gpm per RCP.

8. Open letdown isolation valves FCV-62-69 and FCV-62-70.

9. Open the letdown line isolation valve FCV-62-77.

10. Open the 45 GPM letdown orifice isolation valve FCV-62-73.  !

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11. Position PCV-62-81 to control pressure at letdown orifices above I steam flash point.

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12. Close the BIT inlet isolation valves FCV-63-39 and FCV-63-40 and  !

outlet isolation valves FCV-62-25 and FCV-62-26.

NOTE: Flush the injection lines and reestablished BIT concentration per AOI-19, IV, L thru M.

Y. Re-establish the use of the PZR heaters to maintain the RCS pressure.

If offsite power is available, establish the required conditions for operation of a reactor coolant pump and start the pump in a non-faulted loop (preferably in loop 2 or if not available, in loop 1). If all RCP's are running, trip all but one RCP so as to maintain one pump operating in the loop connected to the PZR (loop 2), or if this is the faulted loop, in loop 1.

Z. If offsite power is available, begin a controlled cooldown of the RCS at a rate of about 50 F/hr. by use of the steam dump to the main condenser fron. the non-faulted S/G's. Control the water levels in the S/G's to maintain S/G water level in the narrow range span or in the .. a c <-..s-span at a level sufficient to assure that the U-tubes are cevered sE -s.

If offsite power'is not available, dump steam from the non-faulted S/G's through the S/G PORV's to provide a controlled cooldown of the reactor coolant system at a rate of about 50 F/hr.

SQNP E0I Units 1 & 2 Page 8 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.)

AA. Simultaneous with the cooldmen using the non-faulted S/C, slowly decrease the f aulted S/G pressure by opening the MSIV bypass valve to the con-j denser (if available), or using the S/G PCRV.

BB. As pressure is reduced in the faulted S/G, control the RCS pressure at a value approximately equal to the steam pressure in the faulted S/C to minimize the leakage flow. RCS pressure control should be accomplished by use of the PZR heaters and action of one of the following:

1. Normal PZR spray (if a RCP is in service)

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2. Use of PZR auxiliary spray (if spray is heated by letdown through the regen. HX.)

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3. Brief intermittant opening of one PZR PORV.

NOTE: Maintain RCS temperature and pressure within the limits of the normal cooldown curves in TI-28.

l CAUTION: If RCS pressure control is accomplished by use of the PZR  !

PORV, continuously monitor the PRT pressure, temperature, l and water level and take appropriate actions to verify and maintain PRT integrity. Verify PZR PORV closure using the PORV stem-mounted position indicators, ACOUSTIC VALVE POSITION MONITORING SYSTEM and PRT conditions. If a reactor coolant leak to the PRT is identified, close the PORV isolation valves.

CC. Periodically sample and analyze the reactor coolant boron concentration during the continuing cooldown. Borate as necessary to maintain the required shutdown margin at all times during the cooldown.

DD. Continue to cooldown and depressurize the reactor coolant system and faulted steam generator until the reactor coolant. hot leg temperatures are below 400 F in the non-faulted loops and the reactor coolant pres- l sure has reached about 400 psig (do not collapse the pressurizer steam bubble).

EE. Place the residu'al heat removal system.in operatD'n using Normal Cooldown Procedurcs.

NOTE: Throughout this c'ooldown procedure, maintain a steam bubble in the pressurizer. Solid water pressure control may not be effective.

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SQNP E0I Units 1 & 2 Page 9 of 9 Rev. 7 II. SUBSEQUENT OPERATOR ACTIONS (Cont.)

FF. Continue the plant cooldown in a normal mode except that after the RCP operation has been terminated, continue to simultaneously control the faulted steam generator steam pressure and reactor coolant pressure to minimize the leakage flow.

GG. When the reactor coolant system hot leg temperatures are reduced below 200 F, the pressure in the pressurizer may be reduced by using auxiliary I

spray until reactor coolant system pressure and the faulted nystem gen-erator pressure equilibrate.

IDI. Continue the operation of the residual heat removal system to remove the core residual heat and maintain the charging and letdown in service to control the pressurizer water level and provide a boration path, III. RECOVERY Following a steam generator tube rupture, the exact procedure will be planned by the Plant Operations Review Committee for repairing the affected tube or tubes and decontamination of the secondary system. The procedure for repairing the af fected steam generator will include necessary decantamination and expe:ure i precautions for maintenance personnel. Decontamination of the secondary system  ;

will be carried out after the extent and tyTe of radiation present is analyzed, I with protection to plant personnel being the mose important consideration.

IV. REFERENCES FSAR 15.4.3 l

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SQNP E0I Units 1 & 2 Table 1

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'i; ..iTION STE/J1 T/diIf (Tu +_cu ~res rounded to neare.st. rj Sat. 50 F. Sat. 50 F PSIG Temp F Subcooled PSIG Temp F Subcooled 300 422 372 1350 584 534 350 436 386 1400 588 538 400 448 398 1450 593 543 450 459 409 1500 597 547 500 470 420 1550 602 552 550 480 430 1600 606 556 600 4S9 439 1650 610 560 650 497 447 1700 614 564 700 505 455 1750 618 568 750 513 463 1800 622 572 800 520 470 1850 626 ,

576 850 527 477 1900 630 580 900 534 484 1950 633 583 I

950 540 490 2000 637 587 i 1000 546 496 2050 640 590 1050 552 502 2100 644 594 1100 558 508 2150 647 597 1150 563 513 2200 650 600 1200 569 519 2235 653 603 1250 574 524 .

2250 654 604 1

1300 579 529 Saturation temperatures may be read from hot leg temperature RTD's or incore T/C's.

SQNp E01 Units 1 & 2 Appendix A ,,_, . .. _ _ .

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Rev. 7 i

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NATL'RAL CIRCULATION A. The following are guidelines to determine if natural circulation is taking place in the primary system.

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1. Core AT as read on wide range RTD's (hot and cald) or an indicated AT between W.R. cold Icg and incore T/C's should be stable or dropping.

A relatively stable AT with values less than 55 F with a gradual de-crease, indicates natural circulation.

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2. Incore T/C's temperature indicating below saturation temperature for the existing primary system pressure.

3 liea t is being removed from the primary system by secondary system, i.e., S/G's steaming and water being added to S/G's and secondary system pressure near saturation pressure for the primary system temperature.

B. The following are guidelines to enhance natural circulation.

1. Keep S/G 1evels in narrow range (tubes covered).

2. Keep primary system pressure above saturation pressure for the existing hot let (W.R.) temperature or incore T/C temperature if possible.

3. Use condenser steam dump or S/G PORV's to steam off and cool primary system at desired rate.

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