ML19325D498

10CFR50.59 Annual Rept. W/ 891012 Ltr

ML19325D498
Person / Time
Site:	Braidwood
Issue date:	10/12/1989
From:	Hunsader S COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
NUDOCS 8910240264
Download: ML19325D498 (7)

Text

-- e f ') Commonwoelth Edison l 72 West Adams Deot. Chcapo, libros s . \ l' KhiiGEpy to. P6si'O%be b6DW L

,, \v' Cheago, Ilhros 60690 0767 October 12, 1989 )

I L Dr. Thomas E. Murley, Director l Office of Nuclear Reactor Regulation  ;

U.S. Nuclear Regulatory Commission l Washington, DC 20555 -

i r Attn: Document Control Dest I

' Subject + Braidwood Station Units 1 and 2 i 10 CFR 50.59 Annual Report i NRC Docket Nos. 50-45fi and 50-452  ;

Dear Dr. Hurley:

I Pursuant to 10 CFR 50.59(b)(2), Commonwealth Edison is providing the i required annual repori for Braidwood Station (facility Operrting License Nos. j NPF-72 and NPF-77). The annual requirement is based on the Unit I fuel load t license (NPF-59) issuance date of October 17, 1986.

l The report consists of descriptions and safety evaluations for ,

changes to the facility as described in the safety analysis report, The ,

l report only contains changes made to the facility; no tests or experiments i governed by paragraph (a) of 10 CFR 50.59 were executed. l

Please direct any questions regarding this matter to this office.

Very truly yours,  !

l{l' , fr- __- _la f

S.C. Huntader l

Nnlaar Licensing Administrator l l

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Attachment ,

l cc: S. Sands-NRR J. Hinds-RIII [F'l7 Resident Inspector-Braidwood I L 8910240264 891231 FDR ADOCK 03000456 R PDC ,

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Braidwood Nuclear Power Station i 10CFR50.59 Annual Report 1989 i NRC Docket No. 50-456 and 50-457 '

License No. NPF-72 and NPF-77 I

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l Modification N20-0-88-031 l

i Descriptions  !

This modification provides an automatic makeup feature for the Component  ;

Cooling System Surge Tank from the Primary Water system and Domineralised  ;

Makeup Water system, via control valves 2CC182 to the surge tank Train "A" side i and 2CC183 to the Train "B" side, respectively. Also, a provision is made for I Main Control Room operation of the subject control valves, as well as  !

associated alarms to indicate that makeup water is being supplied to the surge  !

tank and Component Cooling system discharge header pressure indication at Main  !

Control Board 2PN06J. This mod!!! cation will increase the reliability and ava!! ability of the Component Cooling system la the event of a loss of cooling {>

water capacity.

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Safety Evaluation Summary ,

1. i The probability of an ocenavence or the consequence of an accident, or i ma*. function of equipment important to saf6ty as previously evaluated in tov '

FirAl Safety Analysis Report is not increased bocause the osiatir.g Cnmponent  :

Cooling system malfunction analyals of TSAR Section 9.2.2.4 le not affected by [

this modification, rad the sedkability of the Couponent Coollag f,ystem for '

operation as described in f6*.k Section V.2.2.3 is enhanced by its 1.tata11ation. ,

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

the function of the Component Cooling System, as outlined in FSAR Section (

9.2.2, is not affected. Further, modification failure modes and effects do not j impact any system safety-related functions and can be mitigated by design. and  !

the ability of the Component Cooling and associated systems to perform their  !

intended functions is not impacted by the installation of this modification.

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

Specification, is not reduced because estating bases for Technical [

Specification 3/4.7.3 regarding Compement Cooling system operability, to ensure that sufficient cooling capacity is available for continvec operation of t safety-related equipment during normal and accident conditions, are not

• Impacted. Accordingly, the margin of safety is improved by the installation of -

this modification to incr6ase the reliability and availability of the Component Cooling System, in the event of a loss of coollag water capacity.  !

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, l Modification: N20-2-88-024 l

Descriptions f.

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This partial modification improves the reliability and operation of the unit Fire  !

Protection system through replacement of the esisting upper and lower cable spreading rooms and electrical cable tunnel lonisation (POC) detectors with one circuit each of rate compensated heat (thermal) detectors and a more reliable model of ionisation detector. Additionally, a cross-sono detection scheme is 1 introduced in the fire protection circuitry to require a detection signal from l both detector circuits in a specified some in order to initiate the associated .

gaseous fire suppression system. This partial modification also provides an  !

Interval time delay in the fire detector reset circuit for Fire Detection Control i Cabinets 2PA39J and 2PA49J and an independent battery supply to each cabinet, to l limit the duration of circuit power interruption and preclude falso indication ,

upon detector power restoration and allow for normal system operation during l periods of AC power failure. i

. Safety E*seluation Summary: '

3. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final i Gafsty Analysis Roport is not increased because the lastallation of this modification will improve the reliability and operation of the unit fire Protection system. 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 created because the ability of the unit fire protection system to perform its latended functions is enhanced by the lastallation of this modification.

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3. The margin of safety, as defined in the basis for any Technical Specification, is not reduced because this modification is non-safety related and '

system function is not impacted by its installation. r

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Modification M20 1-88 023

Description:

This partial modification improves the reliability and operation of the unit Fire Protection system through replacement of the esisting upper and lower cable spreading rooms and electrical cable tunnel ionisation (POC) detectors with one circuit each of rate compensated heat (thermal) detectors and a more reliable model of ionisation detector. Additionally, a cross-mone detection scheme is letroduced in the fire protection circultry to require e detection signal from both detector circuits in a specified sono in order to initiate the associated gaseous fire suppression system. This partial modification also provides an interval time delay in the fire detector reset circuit for Fire Detection Control Cabinets 2PA39J and 2PA49J and an in$ependent battery supply to each cabinet, to >

limit the duretion of circuit pswer interruption and preclude falso indication {

upon detector power restoration sod allow for normal system operatica during t periods of AC power failure. l t

i Safety Evaluation Summary:

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

malfunction of equipment important to esfoty as previously evaluated in the Final l Safety Analysis Report is not increased because the installation of this i modification will improve the reliability and operation of the unit fire '

protection system.

2. The possibility for an accident or malfunction of a different type than any f previously evaluated in the Final Safety Analysis Report is not created because the ability of the unit fire protection system to perform its intended functions  !

la enhanced by the' installation of this modification.  ;

3. The margin of safety, as defined in the basis for any Technical Specification, is not reduced because this modification is non-safety related and  !

system function is not impacted by its installation. I l

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Modification N20 2-88-029 ,

i Descriptions i This modification will remove the feedwater (rW) bypass line check valves and replace them with piping spool pieces. Additionally, the LOW TAVE interlock with ,

reactor trip for rW isolation will be eliminated. This modification will i

, eliminate flow anomalies associated with erratic operation of the TW bypass line i check valves. The unit will then be able to operate at 100% power without steam  ;

generator pre-heater section flow limitations.

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Safety Evaluation Summary: "

1. The probability of an occurrence or the consequen:o of an accident. or j malfunction of egulpment important to safety as previously evaluated in the Final i St foty Analysis Report is not increased becacee the ability of the Main reedwater [

(FW) system to perscrm its design function as described in Chapter 10.4.7 of the i TSLR is not impaired by deleting the controlled closure check valves from the FW l bypass lines. Deleting the check valves will allow the TW bypass line to perform i its design function of delivering a minimum of 10% of tots! FW flow to the upper

• nossle of the steam generator. The ability of the Abrillary reedwater (AF) i system to perform its late 9ded safety function as described in Chepter 10.4.9 of  ;

the FSAR le not Ispaired. Deletion of the Lov TAV* Interlock coccurrent wit h '

reactor trip fenn the 7W leolation initiation is consistent with the design intent of protecting the Reactor Coolant (RC) system from an escessive coollowa ,

following a reactor trip. The risk of unacceptably high stresses in the tw i bypass line due to water hammer following a feedline break upstream of tt'e check j valve is reduced since the closure time of the rW bypass isolation valve is j bounded by the closure time of the controlled closure check valve. The following  ;

I has also been provided: 1) A detailed discussion of the modification functional '

changes and their effects on plant operation. 2) An integrated safety i evaluation which analyses the effects of the modification on the Non-Loca, t. ora. l and Loca-Related accident scenarios as analysed in Chapter 15 of the FSAR. 11 A [

description giving the redundancy of the instrumentation and controls associated i with the modification (rW isolation interlock to reactor trip) and the l modification's ef fect on Ar system performance. 4) A safety evaluation which l demonstrates that the modification does not compromise the plant's abilit y 'n l 1solate a main steam line break outside of containment due to environment al '

effects. 5) Additional information regarding prevention of Ar Flow to ete e ' e sa i generator preheater. ,

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Modifications H20-2-08-029 (cont'd) l t

a Safety Evaluation Summaryl 1 j

2. i The possibility for an accident or malfunct44u of a dif ferent type than any ,

pre?!ously evaluated in the Final Safety Analy.;s r eport is not created because I the TW and Ar systems will still perform their respective latended safety i functions as described in FSAR Chapters 10.4.7 and 10.4.9. The use of the main .

FW isolation valve (rWOO9) and TW bypass line isolation valve (rWO39) will j prevent backflow of aus!!!ary feedwater to the steam generator preheater  !

section. Therefore, the performance of the A7 system is not impaired and the i formation of bubble collapse water hammer in tho steam generator preheater '

section is precluded. The risk of rapid valve closure induced water hammer in '

the bypass line is reduced since the FW bypass isolation valve closes slower t5an  ;

the controlled closure check valve which is being deleted. Sufficient redundancy '

esists in the instrumentation and controls abocciated with FM leolation surN that '

a single active failure will n>t rasvat'in tu u7eerlyJed accident. Siero brth  !

the pain rw isolatica valve (rt!0P9) and the rh b/ pass isolation valve wir oo -

used to leo' late the IT f aow f rms the Pese gea*.tetor p oheat er sectJoe, A s!Ugle !

f ailure of eithr4 valvo to close will nct res>1t in an v.nr.nalysed At flowpoth.

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3. the mergin of safety, as definwd it the basin for Eay Technical Specificetion, is not reduced because Lt.e Ar Svetem will still be espable of '

removing dacsy heat and reducing the EC syatsm temperature to less than 350 r from norma) opuratlag conditions (basis for Tech Spec 3/4.6.1.2). The IN I isolation valves will continue to function as designed to meet the intent of the bScis for Tech Spec 3/4 6.3. The Reactor Trip systen and Engineered Safety '

reatures Actuation System Instrumentation will continue to function as designed -

to meet the Intent of the bases for Tech Spec's 3/4.3.1 and 3/4.3.2. These bases I requiro revision to clarify the newly defined function of p-4 on page 83/4 3-3. l In accordance with 10CTR50.36, the bases are not considered part of the Technical I Specifications. Therefore, NRC acceptance of this revision is not required prior i to implementation of this revis2on is not required prior to implementation of  ;

e this modification. The margin of safety as defined in the bases for Technical 5 l Specifications 3/4.7.1.2, 3/4.6.3, 3/4.3.1, and 3/4.3.2 la not reduced by this I modification.  !

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