ML17263A210

Proposed TS Table 3.6-1, Containment Isolation Valves

ML17263A210
Person / Time
Site:	Ginna
Issue date:	04/05/1993
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263A207	List: ML17263A206 ML17263A208 ML17263A210
References
NUDOCS 9304090209
Download: ML17263A210 (13)

Text

Attachment A

Revise the Technical Specification pages as follows:

Remove 3.6-4 3.6-6 3.6-7A 3.6-10 Insert.

3.6-4 3.6-6 3.6-7A 3.6-10 9'304090209 930405 PDR ADOCK,05000244 P'DR

TABLE 3.6-1 CONTAINMENT ISOLATION VALVES PENT.

NO.

29 IDENTIFICATION/DESCRIPTION Fuel Transfer tube 100 Charging line to "B" loop 101 SZ Pump 1B discharge 113 SZ Pump 1A discharge 120 Nitrogen to Accumulators 120 Pressurizer Relief Tank (PRT) to Gas Analyzer (GA) 102 Alternate charging to "A" cold leg 103 Construction Fire Service Water 105 Containment Spray Pump 1A 106 "A" Reactor Coolant Pump (RCP) seal water inlet 107 Sump A discharge to Waste Holdup Tank 108 RCP seal water out and excess letdown to VCT 109 Containment Spray Pump 1B 110 "B" RCP seal water inlet 110 SZ test line 111 RHR to "B" cold leg 112 Letdown to Non-regen.

Heat Exchanger ZSOLATZON BOUNDARY flange CV 370B CV 889B CV 870B CV 383B welded flange CV 862A CV 304A AOV 1728 MOV 313 CV 862B CV 304B MV 879 MOV 720 (20)

AOV 371 CV 889A CV 870A CV 8623 AOV 539 MAXIMUM ISOLATION TIME *(SEC)

NA NA NA NA NA NA NA NA 60 60 NA NA NA NA 60 NA NA NA 60 ISOLATION BOUNDARY (1)

(2)

(5)

(5)

(2)

MV 5129 (3)

(2)

AOV 1723 (4)

(3)

(2)

(5)

(6)

AOV 200A AOV 200B AOV 202 RV 203 (5)

(5)

AOV 846 MV 546 (7)

MAXIMUM/

ISOLATION TIME

• (SEC)

NA NA NA NA NA NA NA 60 NA NA NA NA NA 60 60 60 NA NA NA 60 NA

Pg PENT.

NO 141 142 g0 143 IDENTIFICATION/DESCRIPTION RHR-tl pump suction from Sump B

RHR-g2 pump suction from Sump B

RCDT pump suction ZSOLATION BOUNDARY MOV 850A (13)

MOV 850B (13)

AOV 1721 MAXIMUM ISOLATION TIME *(SEC)

NA NA 60 ISOLATION BOUNDARY MOV 851A (13)

MOV 851B (13)

AOV 1003A AOV 1003B MAXIMUM/

ISOLATION TIME

• (SEC)

NA NA 60 60 4P I

201 Reactor Compart. Cooling Unit A & B 202 "B" Hydrogen recombiner (pilot & main) 203 Contain. Press.

Transmitter PT-947

& 948 203 Post accident air sample to "B" fan 204 Shutdown Purge Supply Duct [Purge Supply Duct]

205 Hot leg loop sample 206 Przr. liquid space sample 206 "A" S/G sample 207 Przr.

Steam space sample 207 "B" S/G sample 209 Reactor Compartment.

Cooling Units A & B 210 Oxygen makeup to A & B recombiners MV 4757 (16)

MV 4636 (16)

MV 1076B MV 1084B PT 947 PT 948 MV 1563 MV 1566 flange (22)

[AOV 5870]

AOV 966C AOV 966B AOV 5735 AOV 966A AOV 5736 MV 4758 (16)

MV 4635 (16)

MV 1080A NA NA NA NA NA NA NA NA NA

[5]

60 60 60 60 60 NA NA NA SOV-IV-3B SOV<<ZV-5B MV 1819C MV 1819D MV 1565 MV 1568 AOV 5869 (22)

MV 956D (14)

MV 956E (14)

(17)

MV 956F (17)

SOV ZV-2A SOV IV-2B NA Normally Closed NA Normally Closed NA NA NA NA NA NA NA NA NA NA NA NA Normally Closed NA Normally Closed 0

0

PENT.

o'O.

O315 316 g0 317 318 319 320 IDENTIFICATION/DESCRIPTION Service Water from "C" fan cooler Service Water to "B" fan cooler Leakage test supply Dead weight tester (decomissioned)

Service Water from "A" fan cooler Service water to "C" fan cooler 4P I

321 A S/G Blowdown 322 B S/G Blowdown 323 Service Water from "D" fan cooler 324 Demineralized water to Containment 332 Cont. Press.

Trans.

PT-944, 949

& 950 ISOLATION BOUNDARY MV 4643 (16)

MV 4628 (16) flange welded shut MV 4629 (16)

MV 4647 (16)

AOV 5738 AOV 5737 MV 4644 (16)

CV 8419 PT 944 PT 949 PT 950 MAXIMUM ISOLATION TZME *(SEC)

NA NA NA NA NA NA 60 60 NA NA NA NA NA ISOLATION BOUNDARY (11)

(ll)

MOV 7443 welded shut (ll)

(11)

(17)

(17)

(11)

AOV 8418 MV 1819G MV 1819F MV 1819E MAXIMUM/

ZSOLATZON TIME

• (SEC)

NA NA Normally Closed NA NA NA NA NA NA NA NA NA NA 332 Leakage test and hydrogen monitor instrumentation lines MV 7448 MV 7452 MV 7456 SOV 921 SOV 922 SOV 923 SOV 924 NA NA NA NA NA NA NA cap cap cap (21)

(21)

(21)

(21)

NA NA NA NA NA NA m

~

0 0

(10)

(11)

(12)

(13)

(14)

(15)

(16)

(17)

The pressure transmitter provides a boundary.

Normally operating incoming and outgoing lines which are connected to closed systems inside containment and protected against missiles throughout their length, are provided with at least one manual isolation valve outside containment (FSAR 5.2.2 pg. 5.2.2-2).

The single remotely controlled containment isolation valve is normally open and motor operated.

The cooling water return line is not directly connected to the reactor coolant system and, should remain open while the coolant pump is running.

A second automatic isolation barrier is provided by the component cooling water loop, a closed system.

(FSAR 5.2.2 pg. 5.2.2-1a)

See FSAR Table 5.2.2-1 and Figure 5.2.2-2.

Sump lines are in operation and filled with fluid following an accident.

Containment leakage testing is not required.

The valves are subjected to RHR system hydrostatic test.

Normally operating outgoing lines connected to the Reactor Coolant System are provided with at least one automatically operated trip valve and one manual isolation valve in series located outside the containment.

In addition to the isolation

valves, each line connected to the Reactor Coolant System is provided with a remote operated root valve located near its connection to the Reactor Coolant System.

(FSAR 5.2.2 pg.

5.2.2-1)

See FSAR Table 5.2.2-1 and Figure 5.2.2-17.

The Serv'ice Water system operates-at a pressure higher than the containment accident pressure and is missile protected inside containment.

Therefore, these valves are used for flow control only and need not be leak tested.

The S/G tubes and secondary side provide a closed system inside containment.

(18) Fire Service Water will be used only to fight fires inside containment.

AOV 9227 is closed during power operation.

A containment isolation signal to automatically close this valve is not required because a spurious signal during a fire may be hazardous to personnel and may impede fire suppression activities.

(19)

(20)

See FSAR Table 5.2.2-1 and Figure 5.2.2-16.

Containment leakage testing is not, required per L. D. White, Jr. letter to Dennis L. Ziemann, USNRC dated September 21, 1978.

3.6-10 Proposed

Attachment B

Page 1 of 2 Attachment B

The purpose of this amendment is to revise several penetration listings in Technical Specification Table 3.6-1 to support an Integrated Leak Rate Test (ILRT) of the Ginna Station containment during the 1993 Refueling Outage.

As a result of preparing for this test, RG&E determined that a discrepancy existed between the Technical Specifications and the NRC approved In-Service Test (IST)

Program.

Currently, Technical Specification Table 3.6-1 identifies two manual valves (820 and 204A) as containment isolation valves (CIVs) for Penetration 112.

As part of the third 10 year IST program submittal for Ginna Station, RG&E identified that valves 820 and 204A would be removed from the IST program since their pressure and containment isolation functions were being performed by valves 200A, 200B, and 202 (Reference a).

The NRC accepted this change by Reference (b).

Therefore, RG&E requests that Table 3.6-1 be changed to replace valves 820 and 204A with AOVs 200A, 200B, and 202.

In addition, relief valve

203, which is located between containment isolation valves

200A, 200B, 202 and 371 for this penetration, will be added to the table.

All five valves are currently in the Ginna Station Appendix J testing program.

This change will enable Penetration 112 to meet the explicit requirements of GDC 56.

The current Table 3.6-1 Note 17 will also be deleted since it is no longer used.

Technical Specification Table 3.6-1 also identifies a manual valve in series with an air-operated valve as the CIVs for Penetrations

206, 207,

321, and 322.

These penetrations contain the steam generator (S/G) blowdown and blowdown sample piping.

RG&E requests that the manual valves be removed from Table 3.6-1 since the S/G tubes and secondary side provide one containment barrier.

These penetrations will then be treated analogous to the main steam and feedwater lines which are of similar configuration.

Table 3.6-1 Note 17 will then be revised and added to these four penetrations to reflect the use of the S/G tubes and secondary side as a

containment boundary.

Table 3.6-1 Note 7 will also be deleted from these four penetrations.

This note was originally applied to the S/G blowdown lines since they penetrate the missile shield inside containment in order to reach the S/Gs.

However, the NRC has approved the use of "leak-before-break" (LBB) with respect to large diameter Reactor Coolant System piping at Ginna Station (Reference c).

The use of LBB excludes the consideration of the dynamic effects associated with postulated pipe ruptures per 10 CFR 50, Appendix A, GDC 4.

In addition, plant walkdowns and a review of Systematic Evaluation Program (SEP) Topic III-S.A, Effects of Pipe Break on Structures,

Systems, and Components Inside Containment, (Reference d) confirms that even though these four lines penetrate the missile shield, they are not susceptible to any high energy line sources.

This includes the affects of pipe whip and jet impingement.

Therefore, the dynamic effects associated with any high energy line break inside containment does not require consideration with respect to the S/G closed system.

The blowdown lines inside containment are also Seismic Category I and Safety Class 2.

Consequently, the lines are considered missile protected based on their present location and have the necessary closed system design requirements.

Attachment B

Page 2 of 2 RG&E will also eliminate 10 CFR 50 Appendix J testing related to the four AOVs for Penetrations

206, 207,

321, and 322 since there are not any requirements to perform this testing.

Instead, RG&E will hydrostatically test these valves in accordance with 'the original requirements as outlined in the approved third 10 year IST program for Ginna Station (IWV-3421 through 3425) as a minimum.

This will provide the necessary assurance that the valves will perform their required isolation function.

These changes will be documented in the Ginna Station IST Program which will be submitted to the NRC following the conclusion of the 1993 Refueling Outage since RG&E expects to make several other unrelated IST Program changes at that time.

These changes will not require NRC approval prior to implementation.

In accordance with.10 CFR 50.91, these changes to the Technical Specifications have been evaluated to determine if the operation of the facility in accordance with the proposed amendment would:

1.

involve a

significant increase in the probability or consequences of an accident previously evaluated; or 2.

create the possibility of a new or different kind of accident previously evaluated; or 3.

involve a significant reduction in a margin of safety.

These proposed changes do not increase the probability or consequences of a previously evaluated accident or create a new or different type of accident.

Furthermore, there is no reduction in the margin of safety for any particular Technical Specification.

The detailed changes are described in Attachment C.

Therefore, Rochester Gas and Electric submits that the issues associated with this Amendment request are outside the criteria of 10 CFR 50.91; and a no significant hazards finding is warranted.

References:

(a)

Letter from R.C. Mecredy, RG&E, to A.R. Johnson, NRC,

Subject:

Inservice Pump and Valve Testing Program, dated May 23, 1989.

(b)

Letter from A.R. Johnson, NRC, to R.C. Mecredy, RG&E,

Subject:

R.E.

Ginna Nuclear Power P1ant IST Program for Pumps and

Valves, 1990-1999 Third 10-Year Interval, dated April 15, 1991.

(c)

Letter from D. DiIanni, NRC, to R.

Kober, RG&E,

Subject:

Resol uti on of USI'-2, "Asymmetric LOCA Loads ",

dated September 9,

1986.

(d)

NUREG-0821, Integrated Plant Safety Assessment Systematic Evaluation Program, R.E.

Ginna Nuclear Power Plant, December 1982.

I Changes Attachment C

Technical Specification Changes Effect 2.

3.

Revise Table 3.6-1 Penetration 112 to delete valves 204A and 820 and add valves 200A, 200B,

202, and 203.

Revise Table 3.6-1 Penetration

206, 207, 321I and 322 to replace manual isolation valves with S/G closed system.

Also delete Note 7 from these penetrations.

Delete current Table 3.6-1 Note 17 and replace with new Note 17.

Penetration now satisfies explicit requirements of GDC 56.

Penetration is now consistent with the main steam and feedwater penetrations.

S/G closed system provides an acceptable containment barrier.

Current Note 17 is no longer used.

New note provides information with respect to S/G closed system.

No technical change.

TABLE 3.6-1 CONTAINMENT ISOLATION VALVES PENT.

NO.

29 IDENTIFICATION/DESCRIPTION Fuel Transfer tube 100 Charging line to "B" loop 101 SI Pump 1B discharge 113 SZ Pump 1A discharge 120 Nitrogen to Accumulators 120 Pressurizer Relief Tank (PRT) to Gas Analyzer (GA) 102 Alternate charging to "A" cold leg 103 Construction Fire Service Water 105 Containment Spray Pump 1A 106 "A" Reactor Coolant Pump (RCP) seal water inlet 107 Sump A discharge to Waste Holdup Tank 108 RCP seal water out and excess letdown to VCT 109 Containment Spray Pump 1B 110 "B" RCP seal water inlet 110 SI test line 111 RHR to "B" cold leg 112 Letdown to Non-regen.

Heat Exchanger ISOLATION BOUNDARY flange CV 370B CV 889B CV 870B CV 383B welded flange CV 862A CV 304A AOV 1728 MOV 313 CV 862B CV 304B MV 879 MOV 720 (20)

AOV 371 CV 889A CV 870A CV 8623 AOV 539 MAXIMUM ISOLATION TIME *(SEC)

NA NA NA NA NA NA NA NA 60 60 NA NA NA NA 60 NA NA NA 60 ISOLATION BOUNDARY (1)

(2)

(5)

(5)

(2)

MV 5129 (3)

(2)

AOV 1723 (4)

(3)

(2)

(5)

(6)

~%4A XOV:'::200A m Sue AOi~~20,0E,

+14~+

AOV,".".:":202:.

RV,'.::.:2,03 (5)

(5)

AOV 846 MV 546 (7)

MAXIMUM/

ISOLATION TIME

• (SEC)

NA NA NA NA NA NA NA 60 NA NA NA NA NA RA 60. ~

i@

NA NA 60 NA

4l Ch I

Ch PENT.

NO.

IDENTIFICATION/DESCRIPTION 141 RHR-¹1 pump suction from Sump B

142 RHR-¹2 pump suction from Sump B

143 RCDT pump suction 201 Reactor Compart. Cooling Unit A & B 202 "B" Hydrogen recombiner (pilot & main) 203 Contain.

Press.

Transmitter PT-947

& 948 203 Post accident air sample to "B" fan 210 Oxygen makeup to A & B recombiners 204 Shutdown Purge Supply Duct [Purge Supply Duct]

205 Hot leg loop sample 206 Przr. liquid space sample 206 "A" S/G sample 207 Przr.

Steam space sample 207 "B" S/G sample 209 Reactor Compartment.

Cooling Units A

& B ISOLATION BOUNDARY MOV 850A (13)

MOV 850B (13)

AOV 1721 MV 4757 (16)

MV 4636 (16)

MV 1076B MV 1084B PT 947 PT 948 MV 1563 MV 1566 flange (22)

[AOV 5870]

AOV 966C AOV 966B AOV 5735 AOV 966A AOV 5736 MV 4758 (16)

MV 4635 (16)

MV 1080A MAXIMUM ISOLATION TIME *(SEC)

NA NA 60 NA NA NA NA NA NA NA NA NA [5]

60 60 60 60 60 NA NA NA ZSOLATZON BOUNDARY MOV 851A (13)

MOV 851B (13)

AOV 1003A AOV 1003B SOV-ZV-3B SOV-IV-5B MV 1819C MV 1819D MV 1565 MV 1568 AOV 5869 (22)

MV 956D (14)

MV 956E (14)

MV 956F SOV IV-2A SOV ZV-2B MAXIMUM/

ISOLATION'TIME

• (SEC)

NA NA 60 60 NA NA NA Normally Closed NA Normally~

Closed NA NA NA NA NA NA NA NA NA m ~

NA Normally Closed NA Normally Closed 8

0

a.

PENT.

NO.

IDENTIFICATION/DESCRIPTION ISOLATION BOUNDARY MAXIMUM ISOLATION TIME *(SEC)

ZSOLATION BOUNDARY MAXIMUM/

ISOLATION'TZ~

• (SEC) 316 Service Water to "B" fan cooler 317 Leakage test supply 318 Dead weight tester (decomissioned)-

319 320 Service Water from "A" fan cooler Service water to "C" fan cooler 321 A S/G Blowdown 322 B S/G Blowdown 323 Service Water from "D" fan cooler 332 Leakage test and hydrogen monitor instrumentation lines 324 Demineralized water to Containment 332 Cont. Press.

Trans.

PT-944, 949

& 950 MV 4628 (16) flange welded shut MV 4629 (16)

MV 4647 (16)

AOV 5738 AOV 5737 MV 4644 (16)

CV 8419 PT 944 PT 949 PT 950 MV 7448 MV 7452 MV 7456 SOV 921 SOV 922 SOV 923 SOV 924 NA NA NA NA NA 60 60 NA NA NA NA NA NA NA NA NA NA NA NA (11)

MOV 7443 welded shut (ll)

(11)

(ll)

AOV 8418 MV 1819G MV 1819F MV 1819E cap cap cap (21)

(21)

(21)

(21)

NA NA Normally Closed NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA

~

P (10)

(11)

(12)

(13)

(14)

(15)

(16)

The pressure transmitter provides a boundary.

Normally operating incoming and outgoing lines which are connected to closed systems inside containment and protected against missiles throughout their length, are provided with at least one manual isolation valve outside containment (FSAR 5.2.2 pg. 5.2.2-2).

The single remotely controlled containment isolation valve is normally open and motor operated.

The cooling water return line is not directly connected to the reactor coolant system and, should remain open while the coolant pump is running.

A second automatic isolation barrier is provided by the component cooling water loop, a closed system.

(FSAR 5.2.2 pg. 5.2.2-la)

See FSAR Table 5.2.2-1 and Figure 5.2.2-2.

Sump lines are in operation and filled with fluid following an accident.

Containment leakage testing is not required.

The valves are subjected to RHR system hydrostatic test.

Normally operating outgoing lines connected to the Reactor Coolant System are provided with at least one automatically operated trip valve and one manual isolation valve in series located outside the containment.

In addition to the isolation

valves, each line connected to the Reactor Coolant System is provided with a remote operated root valve located near its connection to the Reactor Coolant System.

(FSAR 5.2.2 pg.

5.2.2-1)

See FSAR Table 5.2.2-1 and Figure 5.2.2-17.

The Service Water system operates at a pressure higher than the containment accident pressure and is missile protected inside containment.

Therefore, these valves are used for flow control only and need not be leak tested.

(17)

(18)

MNg+P">UP 'P~.. 'S%%..g?N~"y+8%CP

. ~ MgWC YW:, Ã>Y>g'e QP4CAVPehSWPV Y V~'P~4Q Tne::,:Sy'G:;;::tubes',:".:::,::and:::;::s'e'conairj:.:."s'iai

.':jr'ov'icte ji.:.:,:::,:glosid~ iyjjtem!rxniiide::,'.:':b'oi"i,0'a!i'n'ieri'iij Fire Service Water will be used only to fight fires inside containment.

AOV 9227 is closed during power operation.

A containment isolation signal to automatically close this valve is not required because a spurious signal during a fire may be hazardous to personnel and may impede fire suppression activities.

(19)

See FSAR Table 5.2.2-1 and Figure 5.2.2-16.

(20) Containment leakage testing is not required per L. D. White, Jr. letter to Dennis L. Ziemann, USNRC dated September 21, 1978.

3.6-10

r, 0,

I t