ML19329F921

Steam Generator Water Hammer Demonstration Test.

ML19329F921
Person / Time
Site:	North Anna
Issue date:	06/24/1980
From:	Virginia Power (Virginia Electric & Power Co)
To:
Shared Package
ML19329F920	List: ML19329F919 ML19329F921
References
2-ST-22, NUDOCS 8007110365
Download: ML19329F921 (23)

Text

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2-ST-22 Page 1 of 10 06-24-80 VIRGINIA ELECTRIC AND 00WE2 CG.'IPAST NORTH ANNA POWER STATION L' NIT 2 STEAM GENERATOR WATER H. M R DEMONSTRATION TEST REFEPINCES 1 Stane and Webster System Description 5 - 13.

2 Stone and Webster D awings 12050 FM-18A, FM-74A.

3 Vepco Operating Procedures 2-OP-31.2 and 2-OP-32.0.

4 FSAR Coment 10.21.

5 Technical Specifications, Section 8.1.

6. STREG-0053, Supplement No. 10 to the Safety Ev.>1uation Report, Appendix B, B-5, A-1 Water Hanner.

1.0 Purpose 1.) To demonst ate that the possibility of water hammer, following steam generator feed ring recovery, has been eliminated by J-tube installation and feedwater loop seal arrorgements.

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2-ST-22 Page 2 of 10 1 06-24-80 Initials 2.0 Initial Conditions

. (6.1) (4.2) i 2.1 . I= mediately prior to the performance of this test the Test ,

Engineer has reviewed the latest revisions of the applicable i

references in order to improve his familiacity of thic proce-dure and insure that it is still valid for perform ..ee of the test.

2.2 Auxiliary Feedwater System is operable as per 2-0P-31.2.

i 2.3 Steam Generator Blowdown System is operable and ready to receive maximum blowdown flow from 2-RC-E-1A per 2-OP-32.0.

i 2.4 The Emergency Condensate Storage Tank, 2 CN-TK-1, contains l l

water that meets chemistry specifications. l

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2.5 Special Test exceptions from Technical Specifications have been l i

received (see precaution 3.2).

2.6 Reactor Coolant Systez is borated to cold shutdown condition.

A 2.7 The' pressure transmitts: (?g) and associated strip chart recorder -

) are installed.per Attach =ent 6.3.

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2.8 A six pen strip char recorder (or equivalent) is installed to monitor auxiliary scea= generator flow (FI-FW 200A), steam generator narrow range level (LI 2472), steam generator pressure

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(PI 2474), .and auxiliary steam generator pressure- (PI-FW2013).

i 2.9 All test equipment is operational and in calibration and recorded on Attachment 6.1.

2.10 . All three reactor; coolant pumps are operating.

'2.11 The shutdown banks are fully withdrawn.

i 2.12 A jumper has been installed between the following contacts:

SSP-A LOGIC CABINET (2-EI-CB-47C) TB 516-5 TO o

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SSP-3 LCGIC CABINET (2-EI-CS-47D) T3 516-5 TO 6

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2-ST-22 Page 3 of 10 09.-24-80 Initials 2.0 Initial Cenditions (6.1) (4.2)

NOTE: This jumper will block safety injectier. actuation from high steam line differential pressu-e and au.<tliary feedwater pump automatic start on low steam generator level from steam generator 2-RC-E-1A.

NOTE: Permissive status light "RCL 1A STOP VVS CLOSED 2ER.M CHI" will illuminate when jumpers are installed.

CAUTION: See Precaution 3.2.

2.13 Record initial position and align valves in Test Position pec Attachment 6.4.

2.14 The NRC has been notified of the impending test. (Notify 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to test performance, see T.S. 8.1).

2.15 Notify Quality Control of the i=pending test.

2.16 Notify the Shift Supervisor on duty of the inpending test and coordinate its performance through him.

i 2-ST-22 Page 4 of 10 06-24-S0 Initials 3.0 Precautions and Limitations (6.1) (4.2) 3.1 The NRC has reviewed this procedure as required by T.S. 8.1.

3.2 Special test exceptions to t.he following Technical Specifica-tion. have been granted by the NRC:

3/4.3.2.1 ESFAS Instrumentation, Differential Pressure Between Steam Lines - High

, 3/4.3.2.1 ESFAS Instrumentation, Table 3.3-4, 6. Auxilia ry Feedwater Pump Start Low Steam Generator ~evel.

3/4.5.2 ECCS Req 2 ired T gyg B350 F.

3/4.7.1.2 Auxiliary Feedwater System 3/4.7.1.3 Emergency Condensate Storage Tank 3.3 If water is addeo to the emergency condensate storage tank, have Chemistry Depart =ent verify chemistry.

3.4 Do not allow steam generator narrow range level to d:op below 0% during this test.

3.5 Do not allow steam generator differential pressure to exceed 100 psid. If pressure differential approaches 100 psid termi-nate the test (secure auxilia ry feedwater flow) .

3.6 If steam generator differential pressure exceeds 100 psid in an uncontrolled manner, manually initiate Safety Injection.

3.7 Maintain 60,000 gallons of <ater in the Emergency Condensate Storage Tank.

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2-SI-22 Page 5 of 10 06-24-80 Initials-4.0 Instructions 4.1 Demonstration tests at RCS Average Te:perature 45015 F.

7 (steam pressure approximately 400 psig).

4.1.1 Reactor coolant system pressure is below 2000 psig and the low pressure SI actuation is blocked.

4.1.2 Steam generators 2-RC-E-13 and 2-RC-E-lc levels are between 40 to 50% (narrow range).

4.1.3 While maintaining 2-RC-E-1A level above the feed ~ ring, record initial position and align valves in Test Posi-tion per Attachment 6.5.

NOTE: This isolates the secondary side o f 2-RC-E-1A with reactor coolant circulating in the primary side at 450 4

5' F. Regulate steam dump from 2-RC-E-13 and IC to maintain RCS temperature at 450 : 5' F.

4.1.4 Initiate steam generator b'cudown per 2-OP-32 until level in 2-RC-E-1A is 3% - 5% of the narrow range (level below bott:= of feed ring). Then secure steam generator blevdown.

4.1.5 Wait 30 =inutes.

NOTE: Steam generator 2-EC-E-1A level cust be . held below feed ring (3% - 5%) fr: the entire 30 zinutes.

4.1.6 Record RCS average temperature, tL=e of feed ring drainage (from step 4.1.5), and steam pressure en Attachment 6.2.

4.1.7' .Close/ verify close HCV-FW-200A and start 2-?W-P-3A per i 2-0P-31.2.

4.1.8 Start both . rip chart recorders .menitoring Py , AFV flow and steam generator-level.

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2-ST-22 Page 6 of 10 06-24-80 Initials 4.0 Instructions (cont.)

4.1.9 Initiate AFW flew to 2-RC-E-1A by opening HCV-FW-200A.

Regulate fisw at 220 gpm. Locally observe feedwater piping.

CAUTION: Expect an RCS temperature drop of up to 5 F and pres-sure drop of up to 50 psi.

4.1.10 Secure auxiliary feed water flow and secure strip chart recorders af ter two minutes.

4.1.11 Initiate steam generator blowdown per 2-OP-32 until level in 2-RC-E-1A is 3% - 5% of the narrow range (level below bottom of feed ring). Then secure steam generator blewdown.

4.1.12 Wait 30 minutes.

FOTE: Steam gererator 2-RC-E-1A level must be held belcw feed r!ng (3% - 5%) for the entire 30 minutes.

4.1.13 Record RCS average temperature, ti=e of feed ring drainage (fros step 4.1.12), and steam pressure on Attachment 6.2.

4.1.14 Close/ verify close ECV-FW-200A and start 2-FW-P-3A per 2-OP-31.2.

4.1.15 Start both strip chart recorders conitoring P ,7 AIV flow and steam generator level. Initiate AFW flow to 2-RC-E-1A by opening HCV-FW-200A. Regulate flow at 350 gps. Locally observe feedwater piping.

CAUTION: Expect an RCS temperatuce drop of up to 5 F and pres-sure drop of up to 50 psi.

4.1.16 Secure strip chart recorders af te r two minutes .

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2-ST-22 Page 7 of 19 06-24-80 Initials 4.O' Instructions (cont.)

4.1.17 Return steam -generator 2-RC-E-1A level to nor=al. Use main feed or auxiliary feedwater.

, 4.1.18 Secure auxiliary feedwater as required.

1 4.1.19 Remove jumpers installed in step 2.12.

a 4.1.20 Refill Emergency Condensate Storage Tank (ECST) to greater than 110,000 gallons.

4.1.21 Return valves to initial positions per Attach =ent 6.4 and 5.5.

4.1.22 Notify Shift Supervisor that this section of the test i

is conplete.

4.2 Demonstration test at RCS average te=perature 547 t 5* F and

i steam pressure 1005 psig.

4.2.1 Initial conditions are satisfied.

4.2.2 Precautions are noted.

i. 4.2.3 Reactor Coolant System pressure is being controlled at 2235 psig.

4.2.4 Steam generator 2-2C-E-1B and 2-RC-E1C levels are between 40 to 50% (narrow range).

f 4.2.5 While . maintaining 2-RC-E-1A level from the feed ring, record initial position and align valves in Test Posi-tion per Attachment 6.5.

I NOTE: This isolated the. secondary side of 1-RC-E-1A with  !

reactor coolant circulating in the primary side at 547 F t 5* F. Regulate . steam dump from 2RC-E-13 and IC. to

maintain RCS temperature at 547*F' t ~ 5' F.

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4.2.6 Initiate steam generator blowdown valve per 2-OP-32

. until level in 2-RC-E-1A is 3% - 5% of the narrow range (level below bottom of feed ring). Then secure steam generator blowdown.

4.2.7 Wait two hours.

NOT5: Steam genrator 2-RC-E-1A level must be held below feed ring (2% - 5%) for the entire two hours.

4.2.8 Record RCS average temperature, time of feed 2Ing drainage (from step 4.2.7) and steam pressure, Attach-ment 6.2.

4.2.9 Close/ verify closed ECV-FW-200A and start 2-FW-P-3A per 2-0P-31.2.

k 4.2.10 . Start both strip chart recorders monitoring F 7, A2V flow and steam generator level.

4.2.11 Initiate AFW flow to 2-RC-E-1A by opening HCV-FW-200A.

Regulate flow at 350 gpa. Eocally observe feedwater piping.

CAUTION: E:cpect an RCS temperature drop of up to 5* F and pres-sure drop of up to 50 psi.

CAJTION: If RCS pressure drops to 2000 psig, ter=inate the test (secure auxiliary feedwater).

4.2.12 Secure strip recorders after two minutes.

14.2.13 . Return- steam generator 2-RC-I-1A level to normal.

Secure AFW as required.

4.2.14 J Remove jumpers lnstalled in step 2.12.

.4.2.15 Refill ECST and notify chemistr/ to sample ECST af ter fill is complete.

2-ST-22 Page 9 of 10 06-24-80 Initials 4.0 Instructions (cont.)

4.2.16 Return valves to initial Position per Attachments 6.4 and 6.5.

4.2.17 Remove strip chart recorders.

4.2.18 Notify the Shift Supervisor that this test is complete.

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2-ST-22 Page 10 of 10 06-24-30 5.0 Acceptance Criteria 5.1 No abnor= ally large pressure pulse occurred during f eed ring re cove ry. Pressure pulse registered by the presrure transducer is less than or equ&l to 495 psi.

5.2 No unusual audible noise attributable to water hammer occurred during feed ring recovery.

6.0 Attachments 6.1 Test Equipment Data Sheet 6.2 Data Sheet 6.3 Pressure Transmitter Location 6.4 Valve Line-Up 6.5 Valve Line-Up 6.6 Steam Generator /AFW Flow Test Line-Up I

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2-ST-22 Attachment 6.1 Page 1 of 1 06-24-80 TEST EQUIPMEST DATA SEET f

TEST EQUIPMEST DESCRIPTION *  !!ODEL NL?!3ER VEPCO QC NUM3ER W

• NOTE: This applies caly to te=porarily installed test equipment or instrumer.tation.

Per-.anent instrumentation which is part of the system and shown on drawings should not be included.

Completed By.

Date:

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2-ST-22 Attach =ent 6.2 Page 1 of 1 06-24-80 DATA SIEET RCS AVE. STEAM PRESSUP2 FEED RING DRAIN RUN NDtBER ' TDtP. (F*) (PSIG) TIME (SEC.) COMPLETED BY /DATE I /

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2-ST-22 Attachment 6.3 Page 1 of 1 06-24-80 PPISSL"RE TRf;5MITTER LOCATION

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Feedwater Pipe 2-RC-E-1A I

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} I Strip Chart 4 ^'

Racceder (P7 ) Pressure Transmitter

i 2-ST-22 Attachment 6.4 '

Page 1 of 2 06-24-80 SECTION 4.1 VALVE LINE UP SHEET (RCS 450 1 5* F)

RETUPSED INITIAL TEST TO INITIAL VALVE NO. DESCRIPTION POSITION PO9ITION POSITION 2-W-56* Isol Valve Downstream of HCV-200A OPE'i 2-W-130* Isol Valve Downstream of HCV-200C CLOSED 2-FW-12S* Isol Valve Downstream of MOV-200C OPEN MOV-?W200D Aux FV Hdr :10V to 2-RC-E-1A CLOSED HCV- W200C HCV on AFW Hdr to 2-RC-E-1C LLOSEL MOV-FW-200C MCV on AFW Hdr to 2-RC-E-1C OPEN TV-MS211A Steam Supply to 2-FV-P-2 CL SED 4

TV-MS2118 Steam Supply to 2-FV-P-2 CLOSED 1

• Nor= ally Locked Valves Val'.es Placed in Test By:

Date:

Valves Re urned to Initial Position By: _ _

Date:

Independent Verificatien of Valves Returned to Initial Position By:

Date:

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' Attachment 6.4 Page 2 of 2 06-24-80 SECTION 4.2 VALVE LINE UP SHEET (RCS 547 5 F)

RETL~4NED INITIAL TEST TO INITIAE VALVE No. DESCRIPTION POSITION POSITION POSITION 2-FV-66* Isol Valve Downstress of HCV-200A OPI21 2-FW-130* Isol Valve Downstrexm of HCV-200C CECSED 2-FW-128* Isol Valve Downstream of MOV-200C OPEN MOV-FW200D Aux FW Edr MOV to 2-RC-E-1A CICSED HCV-FW200C HCV on AFW Hdr to 2-RC-E-IC CLOSED HCV-FW200C MOV on AFW Hdr to 2-RC-E-1C CPEN TV-MS211A Steau Supply to 2-FW-P-2 _

CLOSED TV-MS211B Steam Supply to 2-FW-P-2 CLOSED

• Normally Locked Valves Valves Placed in Test By:

Date:

Valves Returned to Initial Position Ey:

Date:

Independent Verification of Valves Returned to Initial Position Sy:

Date:

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2-ST-22 Attach =ent 6.5 Page 1 of 2 06-24-80 SECTION 4.1 VALVE LINE UP SH"ET'

-(RCS 450* 1 5' F)

RETURNED INITIAL TEST TO INITIAL VALVE NO. DESCRIPTION POSITION POSITION POSITION FCV-2478 FW Reg. Valve CLOSED

.FCV-2479 FW Bypass Valve CLOSED TV-MS210A Main Steam Trip Valve CLOSED TV-MS213A Bypass Trip Valve CLOSED NRV-MS201A Main Steam NRV CLOSED NRV-MS203A Bypass NRV CLOSED __

i Valves Plar_d in Test By:

1 Date:

Valves Returned to Initial Position By:

Date:

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, Attachment 6.5 Page 2 of 2 06-24-80

' SECTION 4.2 VALVE LIE UP SEET (RCS 547* t 5 F)

RETURED INITIAL TEST TO INITIAL VALVE NO. DE'.iCRIPTION POSITION POSITION POSITION FCV-2478 W Reg. Valve CLOSED I

FCV-2479 W Bypass Valve CLOSED

{. TV-MS210A Main Steam Trip Valve CLOSEL TV-MS213A Bypass Trip Valve CLJSED NRV-MS201A Main Steam NRV- CLOSED NRV-MS203A Bypass NRV CLOSED I

Valves Placed in Test By:,

s Date:

Valves Returned to Initial Position By:

Date:

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, 2 -ST-22 Attach.::ent 6.6 Page 1 of 1 06-24-SO STEAM GENERATOR /AI~w' FLOW TEST LINE UP 2-RC-E-1A 2-RC-E-1B 2-RC-E-IC A 4 A HCV-P4-200A MOV-FW-2003 MOV-FW-200C A A A I

2-F4-P-3A 2-F4-P-33

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'THIS DOCUMEN P00R QUAUTY PAGES '

1.0 INTRODUCTIO1-AND

SUMMARY

The circumstances associated with severe feedwater line noise and pipe movement in some PWR plants show consistently that the steam genera-tor feed ring was uncovered and drained prior to recovering with cold

'feedwater. For this reason, an investigation was made to determine the J

. likely- mechanism for the phenomenon. This investigation showed .that steam water slugging in the feed ring and horizontal sections of the adjacent feed line pipe is the most plausible cause for the acise and pipe movement.

To avoid. water hammer events that are a result of steam bubble collapse following initiation of auxiliary feedwater flow, all steata i

generators have been modified by plugging the orifice holes on the bottom side of'the sparger and installing J-tubes on the upper side. The following Unit.2 special' test for water hammer is similar to r.he accepted x

Unit 1 preoperational test.

2.0 DESCRIPTION

OF TEST Objective - To :demonst' rate that the possibility of water hammer, following '

steam generator feed ring re::very, has been eliminated by J-tube installa-i

tion and. feedwater loop seal arrangements.

Method - With the reactor shutdown and borated for cold shutdown conditions j .the steam. generator level in one steam-generator is lowered to approximately

'. 3% .' After a time period at reduced level, auxiliary ' feed water flow is '

. fed to_'the steam generator. Testing is done at various steam generator 7

. pressures and auxiliary:feedwater flow. rates.

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3.0 IMPACT ON PLANT TECHNICAL SPECIFICATIONS Evaluation of the proposed test has determined that five technical specifications will be violated, and thus require exceptions, during the performance of the test. The following notes delineate the technical specifications that will require exceptions, the reasons for the excep-tions, and the basis for continued operation during the test.

3.1 IMPACT

SUMMARY

3.1.1 T.S. 3.3.2.1 ESFAS Instrumentation, Differential Pressure Between Steam Lines - High Steam line differential pressure input to the ESFAS from the steam generator being tested will be defeated to prevent inadvertent safety injection and to allow performance of the special test.

Indication of steam line differential pressure will still be available and the test will be terminated if a substantial diffaren-tial pressure equal to or greater than 100 psid is detected.

3.1.2 T.S. 3.3.2.1 ESFAS Instrumentation, Table 3.3-4, 6. Auxiliary Feedwater Pump Start Low-Low Steam Generator Level .

Low level input to the ESFAS from the steam generator being tested will be defeated to prevent starting the auxiliary feedwater pumps when level in the steam generator is reduced belcw the a , m level setpoint. The steam generator level of the steam generats being tested above will be maintained above 0% narrow ranle duri.a the test. Other steam generator level trips will not be defeated.

3.1.3 T.S. 3.5.2 ECCS Required TAVG > 3500F.

This technical specification will not be met, since a portion of the differential pressure between steam lines will be disaoled.

All ot.her ECCS will be operable.

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3.1.4 T.S. 3.7.1.2 Auxiliary Feedwater System This technical specification will not be met since the automatic ,

start of the AFW System in response to a signal from the steam generator being tested will not occur. This applies to the low-low level setpoint of one steam generator only.

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3.1.5 T.S. 3.7.1.3 Emergency Condensate Storage Tank During this test the steam generator will be feed from the ECST; this may rec'uce the level below the minimum allowable. Minimum allowable level during the special test will be 60,000 gallons of water. This will be sufficient for the test program because of the minimal amount of decay heat.

3.2 OPERATIONAL SAFETY CRITERIA

1. During the performance of this special test, testing will be aborted if L the following conditions are encountered, a) Steam Generator level (for steam generator < 0% Narrow range span being tested) ,

b) Steam Generator differential pressure > 100 psid (for steam generator being tested) c) ECST level < 60,000 gallons

2. During the performance of this special test, safety injection must be

. manually initiated if the following condition is encountered.

a) Steam Generator differential pressure > 100 psid in an (for steam generator being tested) uncontrolled manner 1:'

d 4.0 SAFETY EVALUATIO'i The safety effects of conditions which are outside the bounds of ccndi-tions assumed in the FSAR are evaluated below. The interaction of these conditions with the transient analyses in the FSAR are discussed.

4.1 Evaluation of Transients 4.1.1 CONDITION II FAULTS OF !!ODERATE FREQUEi!CY 4.1.1.1 Accidental Depressurization of i!ain Steam System The FSAR analysis for accidental steam system depres-surization indicates that if the transient starts ac hot shutdown conditions with the worst RCCA stuck out of the core, the negative reactivity introduced by Safety Injectier prevents the core from gaing critical.

Because of the borated condition of the core, which will exist during the test period, the reactor would remain subtritical even if it were cocled to rcom temperature without Safety Injection. Thus, the assumptions made in the FSAR analysis are met by the bcron cca:entration in the RCS, rather than a barition by SI.

4.1.1. 2 Scuricus Ocaration of t! e Safety Injection System at_

Power In order to reduce the possibility of unnecessary thermal fatigue cycling of the reactor coolant system components, the automatic actua tion of safety injection from the steam line differential pressure signal of the steam generator be#19 tested ,ill be disabled. Thus, this likely source of spurious Safety Injection has been eliminated.

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4.1.2 C0HDITIO:1 III - IMFREnUEUT FAULTS 4.1.2.1 Minor Secondary System Pi;e Creaks The small coderator temperature reactivity coefficient and primary boron concentration assures the reactor vould ren.ain subcritical assuming the worst RCCA s tuck out of the core, even if it were cooled to room tempera-ture without Safety Injection.

4.1.3 CONDITIO.1 IV - LIi1ITING FAULTS d.l.3.1 Major Secondary System Pioe Ruoture The small moderator temperature reactivity ccafficient and primary boron concentration assures the reactor would remain subcritical, assuming the worst RCCA stuck out of the core, even if it were cooled to rocm tempera-ture without Safety Injection.

4.2 Additional Considerations 4.2.1 Fuel Clad Integrity In all cases t was concluded by comparison with previously analyz-ed FSAR conditions that fuel clad integrity would be caintained without need for cperator mitigating action.

4.2.2 Auxiliary Feedwater Lire Uo During the perfornance of this test, the e.uxiliary feedwater system will be lined up in a manner different from that described in the FSAR. This is acceptable for the dLration of the test since the other motor driven auxiliary feedwater will be availa-ble for either of the other two non-tested steam generators.

This condition is bounded by the FSAR since only one AFW pump is necessary under accident conditions.