ML19324C433

Forwards Listing of Changes,Tests & Experiments Completed During Month of Oct 1989

ML19324C433
Person / Time
Site:	Quad Cities
Issue date:	11/01/1989
From:	Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-89-75, NUDOCS 8911170006
Download: ML19324C433 (20)

Text

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N .k Commonwealth Edison 1

-[%d ,b= Ouad Cities Nuclear Power Station j i

s 22710 206 Avenue North

' Cordova, liltnois 61242 9740 I

' Telephone 309/654 2241 m <

4 RAR-89-75

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November 1, 1989 i

Director of. Nuclear Reactor Regulations .

U. S. Nuclear Regulatory ~ Commission .

Hall Station P1-137 -l

> Hashington, D. C. 20555 Enclosed please find a listing of those changes, tests, and experiments completed during the month of October, 1989, for Quad-Cities Station. .

Units 1 and 2, DPR-29 and DPR-30. A summary of the safety evaluations ,

are being reported.in compliance with 10CFR50.59 and 10CFR50.71(e). ,

e Thirty-nine copies are provided for your use.

Respectfully. .

COMMONNEALTH EDIS0N COMPANY ,

QUAO-CITIES NUCLEAR POWER STATION 1,

1 R.hth R. A. Robey Technical Superintendent

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RAR/LFD/djb 1

Enclosure ,

1 cc: R. Stols T. Hatts/J. Galligan

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L !9911170006 891101 PDR ADOCK 05000254

'R' PDC 0027H/0061 Z s, -. - - . . . - ,. . _ , - . . . , - . . , . - , , . - , , - , - .

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s Safety Evaluation #89-540 Increasing the CRD Cooling k'ater Flow Special Test 2-92 Sp6cial Test 2-92 van completed on October 20, 1989. The purpose of this test is to identify if Increasing the cooling water flow will increase the efficiency of the control rod drive pump. At steady-state plant conditions, the flow control valves X-302-6A(B) will be adjusted to allow 60 gpm of cooling water flow rather than 40 gpm at 5 gpm increments.

1. The probability of an occurrence or the consequence of an accident,

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or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because Section 10.5 the cooling water extends the lift of the graphitar seals and elastomer o-rings. Increasing the cooling water flow would provide note prctection from the reactor temperatures. The 60 gpm in within the design bases of the CRD cooling water system of 40 - 60 gpm.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the increase of cooling water flow to 60 gpm is within the design criteria of the system. In addition. increasing flow would establish a pump operating point within the manufacturers secommendations.

3. The margin of safety, as defined in the basis for any Technical Spect-fication, is not reduced because the Technical Specifications do not address the issue of cooling water flow, but increasing flow will not jeopardize the function or availability of the Control Rod Drive System.

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procedura Change QAp 300-2 i

this revision changes some paragr=phs around for easier understanding of steps. Title, change.due.to re-organization have been implemented. The location i,

of keys have been corrected. A corporate directive providing guidance on operation

-above 100% power hes also been implemented.

' 1 '. -the probability of an occurrence or the consequence of an accident, e

or malfunction of equipment important to safety as previously evaluated

q. in the Final Safety Analysis Report is not increased because the majority 6' of changes were editorial in nature. The change-implementing guidance F '

on operation above 100% only provides written instru:tions on the actions to be taken when power drifts above 100%. No FSAR evaluations are affected by these changes.

2. - The possibility for an accident or malfunction of a different type than r? any previously evaluated in the Final Safety Analysis Report is not' created because no unevaluated FSAR conditions are created by this procedure change.

. 3. The margin of safety, as defined in the basis for any Technical Speci-

', fication, is not reduced because the changes made to this procedure do not affect the margin of safety. The changes are basically administrative.

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This procedure has been rearranged to provide easier completion and reduce

., .the possibility of mistakes. '

1. The probability of an occurrence or the consequence of an' accident, i or malfunction of equipment important to safety as previous 3y evaluated L in the Final Saftsty Analysis Report is not increased because this procedure

!' is a surveillance procedure that has been changed in format only. This procedure will not change the probability of an accident. i

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not  ;

L cr.ated becaurae the testing. required by this procedure is not different -

than before the change. The change doesn't create a new unanalyzed j

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3.

The margin of safety, as deflned in the basis for any Technical Speci- ,

o fication e is not reduced because this terting is done to assure Technical 5 i ..

Specification compliance. This has not changed with procedure revision.

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Safety Evaluation $89-550

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Drywell Personnel Airlock Local Leak Rate Test i-l' This is to change the Local Leak Rate Test. procedure for the drywell airlock.

to require a test pressure of 48 psig and to reorder the steps in the procedure.

1. The probability of an occurrence or the consequence of an accident, r or malfunction of equipment important to safety as previously evaluated

[ in the Final Safety Analysis Report is not increased because testing 3- the airlock at pa (48 peig) yields results which are representative h

of the postulated' accident conditions. The order of the steps as specified vill allow the test to be conducted more smoothly and will not increase the possibility or consequences of an accident.

2. The possibility for an acciderit or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the door.is capable of being tested at pa (48 psig) l with'the otrongbacks installed on the inner door. The test ensures proper containment capability of the airlock, p 3. The margin of safety, as defined in the basis for any Technical Speci-

j. fication, is not reduced because testing the airlocks at pa yields more l conservative results than reduced pressure testing.

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Safety Evaluation #89-561 Service Air Crosatie to Turbine Coupling g s- Removal Propane Heater  ;

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i. This is to crosstie service air to turbine coupling propane heater.

[ 1. The probability of an occurrence or'the consequence of an accident,.

p or malfunction of equipment.important to safety as previously evaluated >

in the-Final Safety Analysia Report is not increased because service air is a non safety-related system and does not support any safety- ,

related systems or equipment. (

2. The possibility for an. accident or malfunction of a different type than f any previously evaluated in the Final Safety Analysis Report is not i

! created because propane can not enter the service air system because [

it is maintained at lower pressure than service air, a check valve is installed at.the service air supply, and the burner will be monitored.  ;

n addition, the burner is open to atmosphere and does not provide a i

solfd link betwcen station air and propane.  !

L i 3.- The margin of safety, as defined in the basis for any Technical Speci-  ;

fications, is not reduced because service air is not included in any ,

L basis for Technical Specifications.  ;

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safety Evaluation #89-569 J, RCIC Turbine Area High Temperature Calibration 7

This reduces'the number of temperature switches from 16 to 4 and changes the trip setpoint from 185'F to 155'F on the Unit One RCIC Turbine Area High Temperature Isolation system.

-1. The probability of an occurrenr4 or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because reductag the number of temperature elements will not degrade the integrity of the leak detection system. Decreasing the trip level setting will reduce response time and maintain radiation releases within acceptable limits.

Therefore, the probability of an occurrence or consequence of an accident u is not increased.

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2. The possibility'for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not P created because the modified system will still maintain one-out-of-two j taken twice trip logic and separation criteria for electrical power i

supplies. The system will still ensure isolation in the event of an '

L actual steam line break but should preclude spurious isolations due .

.to small localized steam leaks. Therefore, there is no possibility {

for an accident or malfunction created. l t

3. The margin of safety, as defined in the basis for any Technical Speci- l fication, is not reduced because this change requires a revision to  !

Technical Specifications. However, the change will not reduce the -!

effectiveness of the steam leak detection system. The modified system  !

should increase the reliability of RCIC by reducing the probability of sporadic isolations. Therefore, the margin of safety has not been  !

reduced.

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Safety Evaluation $89-570 L HPCI Turbine Area High Temperature Calibration  !

L This reduces the number of temperature switches from 16 to 4 and changes-the trip setpoint from 185'F to 155'F on the Unit One HPCI Turbine Area High Temperature Isolation system.  !

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F x 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated

, in the Final Safety Analysis Report is not increased because reducing ,

L- the number of temperature elementa will not degrade the integrity of i

. the leak detection system. Decreasing the trip level setting vill reduce  ;

F response time and maintain radiation releases within acceptable limits.

, Therefore the probability of an occurrence or an accident is not increased.

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, 'M 2. The possibility for an acc'ident or malfunction of a different type than f l

>T any previously evaluated in the Final Safety Analysis Report is not ,

l- created because the modified system will still maintain one-out-of-two {

taken twice trip logic and separation criteria for electrical power L supplies. The system will still ensure isolation in the event of an .[

! actual steam line break but should preclude spurious isolationo due l l to small localized steam leaks.- Therefore, there is no possibility l h for an accident or malfunction created. -

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3. The margin of safety, as defined in the basis for any Technical Speci-  ;

fication, is not reduced because this change requires a revision to i Technical Specification. However, the change will not reduce the i effectiveness of the steam leak detection syotem. The modified system should increase the reliability of HPCI by reducing the probability

, of sporadic isolations. Therefore, the margin of safety has not been l

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! Modification M-4-1-81-21 5

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Description l The part of the modification being done at this time is the installation  :

of test jacks and relays to facilitate testing of the ITT Barton level instruments.

7 The modification itself involved a variety of work to the Scram Discharge Volume -

as a result of the ' failure to scram' event that occurred on June 28. 1980 at I TVA's Brown's Ferry Unit 3. The specific design change requirements were delineated L in NRC Bulletin 80-17 and Supplements 1 through 5. The following changes were f completed by the end of 1986.

i' l. The existing instrument voluaea were removed and separate instrument volumes were provided for each bank of Scram Discharge Headers.

2. A latger drain header and larger drain lines were installed between the f. cram Discharge Headers and the new Instrument Volume.

3. New drain lines were installed between the new instrument volumes and L

the drain header of the reactor building equipment drain tank.

4. Diverse and redundant water level instrumentation were instal'ed on b each instrument volume.

5. All Scram Discharge Volume vent line piping was rerouted to ensure positive venting.

6. ' Double isolation valves were installed on all vent and 42ain lines.

7. The 3" Scram Discharge Header cross-tie connection was replaced with a 4" cross-tic connection in order to conform with the seismic analysis and the design drawings.

Evaluation L

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment importatet to safety as previously evaluated in the Final Safety Analysis Report is not increased because the additional redundancies and inprovements provided by this modification will reduce the probability of accidents or malf;nctions as stated in the FSAR.

2. The possibility for an accident or mal' unction of a different type than any previously evaluated in the Final Safety Analysis Repart is not created because this modification is the result of an event that was not evaluated in the FSAR. The intent of this modification is to reduce the possibility of the accident or malfunction modes.

3. The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because the additional improvements provided by this modification will increase the margin of safety as defined in the Technical Specifications.

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Modification M-4-0-83-010 and M-4-1(2)-83-029 i

i l' Description  !

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j These modifications were initiated to replace check valves 1-3999-86,  !

l 2-3999-86.and 1-3999-85 with new wafer type check valves. In addition to  :

L' this replacement..the check valve and its adjacent isolation valves vill be .

l_ relocated downstream several feet where less turbulent flow is expected. i

[ 'The current location of these check valves subjects them to direct pump discharge  ;

flow er.d thus excessive wear. As a result of relocating the new valves they

, will :.o longer be accessible from the floor, therefore, a chainf all or valve i platform will be built.

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Evaluation

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I 1. The probability of an occurrence or the coneequence of an accident.

- or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the integrity f of the Diesel Generator Cooling Water Pump Discharge System has been 1 increased by relocating and installing more reliable valves away from  ;

the present turbulent location. Relocating these valves will help f

climinate the probability of any future valve deterioration (internal) from reoccurring. These mods will not change any of the design criteria of the Diesel Generator Cooling Water System as designated in the FSAR.  ;

2. Tha possibility for an accident or malfunction of a different type  ;

than any previously evaluated in the Final Safety Analysis Report is l i not created because the relocation of the new valves just further l upstream from the present location, does not change the design or i operability characteristics of the Diesel Generator Cocling Water  ;

j System.' Therefore, t.he possibility of an accident or malfunction  !

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not evaluated in the. FSAR is not created. j h 3. The margin of safety, as defined in the basis for any Technical Sptei- l fication. is not reduced because the function of the Diesel Generator '

Cooling Water System nmains unchanged, the original design basis as defined in the Tech. Spec. remains unchanged. Therefore, the margin of safety is not reduced.

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L Modification'N-4-1-86-010 Description Modification M-4-1-86-10 involves changes to the wiring of the Quad Cities  ;

Unit One Diesel Generator HACR Synchronizing Relay. The modification is being  ;

implemented for two main reasons: 1) To replace the existing HACR-1 relay [

with a HACR-1V' relay and 2) To improve potentially poor electrical connections  ;

from the terminal block to the RACR relay.

The modification involves replacing the existing wiring from the associated l terminal block to the HACR relay with hinge wire lugged with a calibrated  !

crimper.' The. vires will be tied down around the cabinet hinge with screw '

down anchors. In addition, the battery negative from the relay will be removed and a tsrminal of the HACR relay will be grounded.

y Evaluation

[ 1. The~ probability of an occurrence or the consequence of an accident, i l or malfunction of equipment important to safety as previously evaluated  :

in the final Safety Analysis Report is not increased because this l L aspect of Diesel Gn*nerator control is not discussed in the FSAR. Under '

accident conditions, this relay is not in the Diesel Generator start circuit. . (Diesel Generator Auto Starts.) '}

L 2. The possibility for an accident or malfunction of a different type l than any previously evaluated in the Final Safety Analysis Report ,

is not created because a portion of wiring is deleted which reduces [

the possibility of a malfunction.  ;

I L 3. The margin of safety, as defined in the basis for any Technical Speci-

[ fication, is not reduced because this portion of the Diesel Generator  ;

, Control does not enter into the considerations for margin of safety. t b

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\ Modification M-4-1-87-017A [

I l -J 9 Descriition L Thii modification revised the standby condensate booster pump auto start

logic. Before this modification was made the standby Condensate 3ooster Pump j p (CBP) atito started e , low ' suction presrure (160 psi) to the reactor f eed pumps. >

(This fasture was retu ned as a backup to the new auto start logic.) The standby  :

c CBP now .; tarts on the loss of any running CEP. The modification was initiated l

! to elimir.-ute the delay in starting the standby CBP while waiting for the Reactor Feed Pump (RFP) suction pressure to drop to 160 psi. This delay and the time  ;

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required for the standby CBP to come up to speed caused a further reduction j in RFP suction pressure which often tripped the RFP's which in turn lead to a reactor scram. j q

Evaluation  !

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1. The probability'of an occurrence or the consequence of an accident, i or malfunction of equipment important to safety as previously evaluated  !

in the Final Safety Analysis Report is not increased because this l modification does not alter any equipment or systems important to +

L safety as previously evaluated in the FSAR. In fact, the CBP system l reliability will be enhanced. However, this would have no bearing .

on the probability or consequence of an accident or malfunctioni since i analyses take no credit for this system.  !

2. The possibility for an accident or malfunction of a different type f than any previourly evaluated in the Final Safety Analysis Report i

! is not created because.the potentir.1 failure modes of the CBP's (such as failure to auto start when needed or inadvertent start when not l needed) are not different than the ones that presently exist. Failure  :

effects are bounded by existing failure analysis of the feedwater

.l control system, as provided in Section 11.3.3 of FSAR/UFSAR. j t 3. The margin M safety, as defined in the basis for any Technical Speci- [

fication, is not reduced because this modification does not alter  :

or affect any equipment described in the Technical.Specificationi..  !

Therefore, the margin of safety will not be reduced.

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Modification M-4-1/2-87-22A Description ,

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.This modification consists of installing conduit and cable routing for  !

the 1/2 Diesel Generator betveen panels 901-8 and 902-8. This mod will provide  ;

the Unit 2 operator with swing dierel generator controls and indication on Ji the 902-8 panel identical to those on the 901-8 panel. This mod was initiated by Human Engineering Discrepancy (IIED) Nos. 0418 and 0271 related to swing l

diesel controls. Commitments were made to the NRC by Ceco to resolve tFr

, HED's and bring them within the guidelines specified in NUREG 0700. No cermina- .

tions are required for this mod.

, Evaluation

1. The probability of an occurrence or the consequence of an accident.  :

or malfunction of equipment important to safety as previously evaluated l in the Final Safety Analysis Report is not increased because the  !

structural loads added by this partial are within the capabilities e of the existing structures. This partial mod does not include cable terminations. Therefore, no single failure event nor design basis accident, as evaluated in the FSAR, is a4fected by this partial mod.

2. The possibility for an accident or malfunction of a different type I than any previously evaluated in the Final Safety Analysis Report  ;

is not created because no interfaces with existing systems are  :

created as this partial mod does not include cable terminations. The  ;

only interactions are the conduit runs which are located _ncar safety-y_ related components. These interactions are mitigated by the use of j seismically designed conduit supports.

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3. The margin of safety, as defined in the basis for any Technical Speci- '

fication, is not reduced because the cable being added by this partial  :

mod has been verified to be within the capabilities of the existing fire detection and cuppression systems. Those fire stops being disturbed,

are of a tested configuration for the amount of cable involved. Therefore, j~ the margin of safety defined in the basis for Technical Specifications (Section 3.12) are not reduced. No other systems discussed in the s, . -

Quad Cities Technical Specifications are affected by this partial mod.

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Modific4 tion H-4-1/2-87-22B, C, D Descrf3 tion

,, These moda provide additional meters for the 1/2 DG parameters. Indications of 1/2 DG operating parameters are available on both the 901-8 and 902-8. Also the Unit 1 and Unit 2 DG meters have been rearranged providinp identical meter locations.

Evaluation

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipmes.t important to safety as previously evaluated in the Final Safety Analysis Report in not increased because meters added to 902-8 will not significantly change the structional capabilities of 902-8. All new meters are Class IE components. These partial mods include cable tarminations. Relocation of existing meters on the 901-8 and 902-8 does not change the probability of an accident. There-fore, no Single Failure I. vent (SFE) nor Design Basis Accident (DBA),

.as evaluated in the FSAR, is affected by these partial mods.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because interfaces with the existing swing DG voltage regu-lator and governor control circuits are created by these partial mods..

Interactions with the swing DG differential overcurrent relay art created by the addition of the new meters to the relay 'pT' circuit.

The new meters and switches are mounted near existing safety-related equipment. - These interactions are mitigated through the procurement of Class 1E equipment and the seismic mounting of this equipment.

3. The margin of safety, as defined in the basis for any Technical Speci-fication is not reduced because the nee meters on 902-8 are procurred safety-related and are of the same type as the existing swing DG meters on-901-8. The metering for DG is not needed for the DG to be operable.

Therefore,'the margin of safety defined in tne basis for Technical

. Specifications Section 3.9 is not reduced. No other systems discussed in the Quad Cities Technical Specifications are affected by these partial modificaticus.

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w ED Modification M-4-1/2-87-22E

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' Description This modification consists of installing red and green indicating lights on 901-8 to provide indication that the DG1 cooling water pump is running. This l mod will also terminate spare conductors on existing cables and instal! new ,

! fuseblocks at switchgear 19 and panel 901-0, "ew cable will be routed between L local panel 2251-98 and 901-8. This mod was .uitiated by Human Engineering .

Deficiency (HED) Nos. 0418 and 0271 related to swing diesel controls. Commit-l ments were made to the NRC by CECO to resolve the HED's and bring them within -

the guidelines specified in NUREG-0700.

k Evaluation  ;

1. The probability cf an occurrence or the consequence of an accident, i or malfunction of equipment important to safety as previously evaluated ,

in the Final Safety Analysis Report is not increased because the new -

lights have na adverse effect on the existing panel. This partial i mod does not affect the control of the DG1 cooling water pump and hence the ability of the Unit I diesel generator to operate and provide ,

emergency power Therefore, no Single Failurc Event (SFE) or Dcsign Basis Accident (DBA), as evaluated in the F3AR, is affected by this i partial mod.

2. The possibility for an accident or malfunction of a different type ."

than any previously evaluated in the Final Safety Analysis Report is not created because an existing divisional cable with spare conductors is used and is isolated from the non-safety relared indicating lights

circuit. The new lights, fuseblocks and conduits will be seismically mounted to mitigate failure. The new lichts circuit does not interact with the DG1 cooling water pump controla. The new cable will be procurred  ;

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3. The margin of safety, as defined in the basis for any Technical Epeci-fication, is not reduced because the existing and new cabies have been ..'

L verified to be within the capabilities of the existing fire detection and suppression systems. Existing fire stops are being resealed with approved firestop material. The new lights circuit does not affect the operation of the DG1 cooling water ' ump controls and hence the i

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availability of the Unit I diesel gener ator. Therefore, the margins of safety defined in the basis for Technical Specifications 3.9 and 3.12 are not reduced. No other systems discussed in the Technical Specifications are sffected by this partial mod.

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r Modification M-4-1/2-87-2M i O

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I Descriptions This modification consists of inctalling red and greeu indicating lights

>l on 901-8 to provide indication that the DG1/2 cooling water pump is running.

This mod will also terminate spare conductors on existing cables and install l

'new fuseblocks at switchgear 18 and panel 901-8. New cable will be routed  !

l, between local panel 2251-100 and 901-8. This mod was initiated b;r Human D Engineering Deffeiency (HED) Nos. 0418 and 0271 related to swing ditsel controls.  !

, Commitments were made to the NRC by CECO to resolve the HED's and bring them .

within the guidelines specified in NUREG-0700.

,g Evaluation  !

L i The probability of an occurrence or the consequence of an accident.

1.

,' or mal (unction of equipment important to safety as previously evaluated  ;

l in the Final Safety Analysis Report is not increased because new lights j have no adverse effect on the existing panel. This partial mod does '

not affect the control of the DG1/2 cooling water pump and hence the i ability of the swing diesel generator to operate and provide emergency  ;

power. Therefore, no Single FaiTure E' cent (SFE) or Design Basis Accident i (DBA), as evaluated in the FSAR, is affected by this partial mod. ,

2. The posaibility for an accident or malfunction of a different type than l any previously evaluated in the Final Safety Analysis Report is not ,

created because an existing divisional cable with spare conductors is  !

. used and is isolated from the nan-safety related indicating lights circuit. l l

The new lis, hts, fuseblocks and conduits will be seismically mounted to r 1 mitigate failure. The new lights circuit does not inte-act with the DGl/2 cooling water pump controls. The new cable will be procurred  ;

safety-related. i t

3. The margin of safety, as defi.ed in the basis for any Technical Speci-  !

fication, is not reduced because the existing and new cables hav$ been  !

(- verified to be within the capabilities of the existing fire detection I l and suppression systems. Existing fire stops are being resealed and t

': tne naw core hole sealed with approved firestop material. The new lights

. circuit does not affect the operation of the DC1/2 cooling water pump l controls and hence the availability of the swing diesel generator. There-fore, the margins of st.fety defined in the basis for Technical f>pecification l Sections 3.9 and 3.12 are not reduced. No other systems discussed in the ,

Technical Specifications are affected by this partial mod. [

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Modification M-4-1-87-71 i n

Dgpeription.  :

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-o The purpose of this modification is to add vibration isolator assemblies ,

to the assemblies to the Main Steam Line low pressure switches 1-261-30A, B, C D located on instrument rack 2251-1. Installation of the isolators will l help to eliminate cpurious actuation of the switches due to excessive vibration  !

at the switch mounting. Tne switch will be mountea on a plate. Another prate is mounted to the instrument rack, and the vibration isolators are sandwiched

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, between the plate. The conduit and tubing attached to the switch are replaced with more flexible type.

- Evalu a t ic.) >

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1. The probability of an occurrence or the consequence of an accident,  !

or malfunction of equipment important to safety as pteviously evaluated in the Final Safety Analysis Rcport is not increased because FSAR does i not specifically address the main steam low pressure switches. However, '

this modification does not change either the switch function or per- ,

-formance requirements.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not  ;

created because the FSAR does not specifically address the main steam low pressure switches. . However, since this modification does not change elthee the switch function or performance requirements, the possibility  ;

for an accident or malfunction of a different type than any previously i evaluated is not created.

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3. The margin of safety, as defined in the basis for any Technical Speci-ficat.in, is not reduced because modification does not change switch ~

performance requirements and, therefore, doos not affect the margin of safety used for the basis of any Technical Speicifcation. The margin  ;

of safety, therefore, is not reduced. The limiting safety system settings  ;

are not changed as a result of this modification and the low pressure ,

isolations of the main steam lines at 850 psig minimum will still protect against rapid reactor depressurization (thermal hydraulic safety limit)  ;

and the resulting rapid cooldown of the vessel. l r

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Modification M-4-1-87-73C t j Description s

This modification relocated the position indication lights and lenses of

, the four (4) RHR service water valves essociated with each RHR heat exchanger.  :

(!: These indicating lights on the 901-3 panel in the Quad Cities Unit 1 control r00m were rearranged to satisfy an NRC commitment as well as Human Engineering Discrepaney (HED) concerns. Ihn rearrangement resulted in the valve's indicators t being grouped together, the valve " closed" indicating light located on the left and the valve " opened" indicating light located to the right. This presented  ;

an arrangement more consistent with normal control room panel arrangements.

i Evaluation '

i 1. The probability of an occurrence or the consequence of an accident, i or malfunction of equipment important to safety as previously evaluated ,

in the Final Safety Atalysis Report is not increased because this

partial consists only of rearranging the RHR service water heat exchanger i

! valve positions indicating lights. No new equipment is being added {

, by this nod. Therefore, no single failuro event nor design basis l

• accident, as evaluated in the FSAR, is affected by this modification.

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2. 'The possibility for an accident or malfunction of a different type

k. than'any previously evaluated in the Final Safety Analysis Report ,

f' is not created because no change has been made which affect. any of l l the bounding conditions in the FSAR accident analysis. All bounding i b conditions remcin the same, no new acef.ents are introduced by this  ;

( t modification.

3 .' The margin of safety, as defined in the basis for any Technical Speci- }

fication, is not reduced because the function of the residual heat removal system' remains unchanged by the rearrangement of these indicators.  ;

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Therefore, the specification 3.$.A. B. F. G and H are not reduced.

No other systems discunsed in the Quad Cities Technical Specifications i are'affected by the partial modification.  !

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Modification M-4-1-88-27A, B, and C d'

pescription

'These modifications will remove the CRD system return line from the reactor vessel nozzle to a flanged connection outside primary containment. 'The process pipe through the drywell penetration will remain. This modification is removing stainless-steel piping susceptible to Intergranular Stress Corrosion Crecking 4

(IGSCC). The balance of carbon steel is also being removed to reduce further L. .

amintenance costr and to supply a spare drywell penetration. Control room flow

. indication and local pressure indication is being removed.

!- geluation 1.s The prebability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated

in the Final Safety Analysis Report is not increased because its

. removal reduces the potential for IGSCC, thereby decreasing the prc,bability of a line break.

2. The passibility for an accident or malianction of a different type i than any previously evaluated in the Final Safety Analysis Report l is not created because no new interfaces with safety-related equipment, I systems, or structures or new systems subject to failure or malfunction have been introduced. The drywell capping detail is evaluated and qualified using the loading utilized at other areas of the containment where steel is not backed by concrete.

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3. The margin of safeti, as defined in the basic for any Technical Speci-fication, is not reduced because the CRD return line is not discussed in any technical specification and was not included in the original plant safe shutdown analyt.is. Removal does not reduce any aargins of safety or limiting condition of operation. Primary containment integrity will Se maintained by the installation of a new cap which will be qualified using the FSAR-specified design criteria.

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• Modification M-4-2-89-50 Description While running a routine surveillanco, it was discovered that the ground water level $ncreased when the Unit 2 RHRSW pumps were run. ' Investigation determined that the 2B RHRSW discharge line 2-1005-B-16"-D was leaking. This modification will reroute this pipe above ground and isolate the old pipe in

. place.

Evaluation.

1. The probability of an occurrence or the consequence of an accident,

'or malfunction of equipment important to safety as pravious)y evaluated in the Final Safety Analysis Report is not increased because the new line functions in exactly the same way as the old one. The routing above ground does not adversely affect the performance of this or any other piece of safety-related equipment, because the line is

' seismically supported.

2. The possibility for an accident or malfunction of n different type than any previously evaluated in the Final Fafety Analysis Report is not created because no new single failure mode is introduced bj this rerouting, either foe the pipe or any other piece of safety-related equipment.

3. . The margin of. safety, as defined in thu basis for any Technical Spdei-fication, is not reduced because the routing of the new line and installation of the plate do not impair the ability of plant personnel to meet the testing and surycillance requirements or the ability of RHRSW system tn perforn ac required.

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