ML19332D053

Proposed Tech Specs Re post-accident Monitoring Instrumentation Requirements,Primary Containment Isolation Valves & Primary Containment Leakage Test Penetrations

ML19332D053
Person / Time
Site:	Quad Cities
Issue date:	11/15/1989
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML19332D020	List: ML19332D019 ML19332D053
References
0371T, 371T, NUDOCS 8911290344
Download: ML19332D053 (10)

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. QUAI > CITIES DPR-29 TABLE 3.2-4 POSTACCIDENT MONITORING INSTRUMENTATION REQUIREMENTS [2]

Minimum Instrument Number of Readout Operable Chan- Location Number nels[1] [3] Parameter Unit 1 Provided Range 1 Reactor pressure 901-5 1 0-1500 psig 2 0-1200 psig 1 Reactor water level 901-3 2 -243 inches +57 inches 1 Torus water temperature 901-21 2 0-200'F 1 Torus air temperature 901-21 2 0-600'F  !

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Torus water level 901-3 1 -5 incta +5 1 incicator inches (narrow range) 2[6] Torus water level 901-3 2 0-30 feet indicator (wide range)

Torus water level 1 .18 inch rahge sight glass (narrow range) ,

1 Torut piessure 901-?- 1 -5 inches Hg to

• 5 psig 2 Drywell pressure 901-3 1 -5 inches Hg to 5 psig

-10 inches Hg to 70 psig 2 0 to 250 psig 2 Drywell temperature 901- E 6 0-600'F 2 Neutron monitoring 901-5 4 0.1-10 8CPS 2 E43' Torus to drywell 2 0-3 psid ,

differential pressure 1[8] Drywell Hydrogen 901-55, 56 2 0-4% '

, concentration 2 EI3 Drywell radiation 901-55, 56 2 1 to 108 R/hr monitor L

L 3.2/4.2-21 Amendment No. 114

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TABLE 4.2-2 e

POSTACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS hinimum Instrument '

Number of Readout Operable Location Instrument -

Channels

• Parameter Unit 1 Calibration Check ,

1 Reactor pressure 901-5 Once every 3 months Once per day }

I Reactor water level 901-3 Once every 3 months once per day 1 Torus water 901-21 Once every 3 months Once per day temperature .

1 Torus air 901-21 Once every 3 months Once per day  !

temperature Torus water level 901-3 Once every 3 months Once per day

, indicator (narrow range) 2 Tores water level 901-3 Onca every 18 months Once per 31 ,

itidicator (wide range) days Torus water lavel N/A None sight glass

• 1 Torus pressaire 901-3 Once every 3 months Once per day 2 Drywell pressure 901-3 Once every 3 months Once per day l 2 Drywell terperature 301'M Once every 3 months Once per day 2 Neutron monitoring 901-5 Once every 3 months Once per day ,

2 Torus to drywell Once every 6 months None differential pressure  :

1 Drywell Hydrogen 901-55, 56 Once every 3 months Once per 31 concentration days 2 Drywell radiation 901-55, 56 Once every *** Once per 31 monitor 18 months days Main Steam RV 901-21 ** Once per position, acoustic 31 days monitor 2/ valve Main Steam RV 901-21 Once every 18 months Once per position, 31 days temperature monitor 3.2/4.2-30 Amendment No. 114

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TABLE 3.7-1 (Cont'd)-

PRIMARY CONTAINMENT ISOLATION  !

Valve Number Number of Maximum for Power-Operated Operating Normal ' Action on; Isolation valve Units Valves Time Operating Group Initiatin!

Identification 1 and 2 Inboard Outboard (sec)' Position Signal 2 Reactor shutdown cooling supply MD-1001 1 <40 C SC 2 Reactor ~

shutdown rooling tupply MO-1001-50 1 <40 C

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Pressure Suppression 2 Drywell purge valve A0-1601-21 1

<10 C SC 2 Vent. valve A0-1601-22' 1 710 C SC 2 Orywell -

vent valve A0-1601-23 1 <10 C SC 2 Vent to- -

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exhaust system A0-1601-24 1 (10 C SC 2 ~ Nitrogen purge A0-1601-55 1 <10 0 GC 2 Torus purge valve A0-1601-56 1 <10 0 GC 2 Makeup valve MD-1601-57 1 0 GC 115 2 Torus makeup valve A0-1601-58 1 115 C SC 3.7/4.7-34 Amendment No. 114

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PRIMARY CONTAINMENT LEAKAGE TEST PENETRATIONS 1

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-Double-Gasketed Seals-

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X-1: Equipment-hatch '

i a- X-2 -Personnel air lock 1 X Control-rod' drive hatch

X-35A through G- TIP_ drives . _ . o

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, . Drywell head access hatch Suppression chamber access hatch-X-200A & 8- .

1 through 8l Drywell head shear lug (inspection hatch)

-Pipe Penetrations ,

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X-7A-through 0; Main staam ';

, .X-8 -Main steamline drain '

'X-9A & 8- Feedwater

-:X Reactor core isolation cooling '

LX High ;ressure coolant injection ,

JX Residual heat removal ~  ;

-X-13A & B Residual-heat removal-  :

X-14' - Reactor. coolant cleanup

-X-16A ,,

& B- C.___ore spray- .,.

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?X-23:&-24' Reactor building closed cooling water  !

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.X-25.4 26 -Drywell ventilation .

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X-47: Standby liquid control-  !

Electrical' Penetrations

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L ~X-101A, B & D': "

LX-1028 u

X-103. .

X-104A through 0 & F ['

X-105C.  :

X-106A & B: '

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s ATTACHMENT 3 SQHMARY OF CHANGES Page 3.2/4.2-21,-3.2/4.2-30 .

Change'drywell' temperature instrument location from "901-21" to "901-3".

Page 3.7/4.7-34 Delete: . Reactor head spray valves H0-1001-60 and H0-1001-63 Page 3.7/4.7-39 Elpe.,hoetrations Delete: )0-17 Reactor vessel . head spray-X-36 Control Rod Drive Return i

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o ALTACHENLA DESCRIEU OMDJAS ES_f_0P.RdENDM E NLREQU E SI Current Technical Specifications 3.7.D.2 requires that all Isolation Valves ' contained in Table 3.7-1 be operable during reactor power operation, Table 3.7-1 contains a description and associated requirements for operating position and operating time for reactor head spray valves M0-1001-60 and M0-1001-63. In addition, Table 3.7-2 describes the Primary Containment Leakage Test Penetrations which includes penetrations for the reactor vessel head spray and CRD return lines. Tables 3.2-4 and 4.2-2 describe the location for drywell temperature instrument as 901-21.

The proposed Technical Specification deletes the reference to the reactor vessel head spray and CRD return lines from Table 3.7-1 and 3.7-2 as a result of the removal of these lines. The drywell temperature instruments are now located on Panel 902-3.

The modifications to the reactor vessel head spray and CRD return l lines were performed as part of the Station's Intergranular Stress Corrosion Cracking mitigation program, as put forth in the response to Generic Letter 88-01. The piping was determined to be unnecessary and susceptible to IGSCC.

The head spray system provided a means to augment reactor cooldown and reduction of pressure following a shutdown. The head spray system is part af the Residual lleat Removal system and allows water to be diverted to a spray nozzle in the steam dome of the vessel. The operation of the head spray system during reactor shutdown is optional. The system is not used to mitigate accidents or during normal reactor. operation. j The CRD return line was previously removed from service to prevent thermal stresses on the reactor vessel due to temperature differential caused by the return of cooler CRD water. The CR0 return line provided a return flowpath to the reactor vessel following CRD movement. The current return flowpath is reverse flow through the exhaust header and return to the vessel through'the CRD seals. Since the return line was capped, the piping was not utilized during any mode of operation.

As a result of the removal of the llead Spray piping, valves MO-1001-60 and M0-1001-63 were eliminated. These valves provided for containment isola-tions and were normally closed. The valves received a close signal during a ,

Group II inttlation signal (low reactor water level, high drywell pressure, or high drywell radiation levels). The remaining piping (reactor side) is blind flanged which provides for isolation of the pressure boundary. The blind flange provides the same or greater level of reactor pressure boundary as the

, removed isolation valve.

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, The removal of the CRD return line and head spray piping resulted in

~the closure of two drywell penetrations which were listed on Table 3-7.2.

'This Table delineates the penetrations which require Type C local leak rate  ;

testing. The integrity of the penetration closure will continue to be tested during containment integrated leak rate-test. The closure of the penetrations ,

e eliminate possible leakage path from containment.

The relocation of the drywell temperature was performed to enhance human factors.. The temperature instrument readout was previously located on c

'the back panels'. Due to the need to monitor drywell temperature during.

. accident conditions, the instruments were relocated to the front panels to allow easy access for monitoring during reactor operations.

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t 611ACHMENI_5 BASLS FOR NO SLGN1f1 CANT HAZARDLDIIERtilBAIl0B As~ discussed in the Description of the Proposed Amendment Request, the proposed change revises the current Technical Specification listing for the primary containment isolation valves, primary containment leakage test ,

penetration, and location of drywell temperature instrument location.

The changes have been reviewed by Commonwealth Edison. Based on this review, Commonwealth Edison does not believe the changes present a Significant Hazards Consideration. The basis for the determination is documented below.

Commonwealth Edison has evaluated this proposed amendment and determined that the changes does not involve a significant hazards consideration. In accordance with 10 CFR 50.92(c):

.I. The proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.

The probability of a pipe leak and/or break due to intergranular stress corrosion cracking (IGSCC) is reduced by the removal of the Head Spray and Control Rod Drive return lines. The. removal of the Head Spray line reduces the consequences of a loss of coolant accident (LOCA) due to the elimina-

-tion of a vessel leakage path. The hydrostatic test of the vessel, performed _each cycle, will assure the integrity of the installed blind flange.

The probability of an accident is not affected by the closure of the penetrations. The consequences of an accident is reduced since the closed penetrations will not be available as a possible Primary Containment-leakage path. -The leak tightness of the penetrations will be verified by the integrated leak rate test (ILRT).

The relocation of the drywell temperature nonitor does not affect the probability or consequence of an accident previously analyzed. The relocation of the instrument was performed to enhance human factors.

Finally, since neither the CRD return line nor the Head Spray system are utilized to mitigate any accident scenario, the elimination of the Head Spray system and the CRD return line does not increase the probability or consequences of any accident previously evaluated.

l The proposed change will not create the possibility of a new or different 2.

kind of accident from any previously evaluated.

No new interfaces with safety-related equipment, systems or structures or i any new systems. subject to failure or malfunction have been introduced.

l The proposed change does not introduce any new operational modes. The 0371T:8 1

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Head-Spray system and.CR0 Return Line were not used for accident mitiga-tion, normal operation or shutdown (use of Head Spray during a shutdown is ,

optional). . Consequently, the removal will not result on the use of other systems in new or unanalyzed methods. No new or different kinds or accidents from any previously evaluated is therefore created.

-g The relocation of the drywall instrument does not introduce any new modes of-operation, therefore, no new or different kind of accident from any~

accident previously evaluated is created. The proposed Technical

Specification change is administrative in nature and does not involve accident mitigation.

3._ TI'e proposed change does not involve a significant reduction in the margin of safety.

The margin of safety is unchanged as a result of the removal of three possible pathways for leakage. The possibility of failure due to IGSCC in the removed piping systems is eliminated as is the possibility of leakage through the Head Spray Containment-Isolation Valves. The current configuration of th'e blind flange (reactor side) and pipe cap (RHR side) provides for an adequate isolation of the piping. Reactor vessel-hydrostatic testing will ensure integrity of the current. configuration..

Closing the. penetrations in the-drywell also reduces the possibility of Primary Containment leakage through these paths. . Penetration closure integrity is verified using ILRT.

The relocation of the drywell monitor does not affect the margin of safety. The monitor _was relocated to enhance factors in the control room. -;

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