ML20086L410

1990 Annual 10CFR50.59 Rept.Rept Received W/O Page 63

ML20086L410
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1990
From:	Miller D PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CCN-14209, NUDOCS 9112160220
Download: ML20086L410 (111)

Text

CCN- 14209 Plill.ADEl.PillA El.ECTHIC COMPANY PIACl!14WIOM ARB11C POV 1:R s1ATION

/ R D 1, llor 20H W. Dclu. lYnnsy lvania 17314 m,cm w ntw-t4:Nm ae ne e umit wi (717) 4 % 7034 D, B. Miller, Jr.

Vice rienutent December 9, 1991 Docket Nos. 50-277 50 278 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

SUBJECT:

Peach Ilottom Atomic Power Station (PilAPS)

Annual 10 CFR 50.59 Report For The Period 1/1/90 through 12/31/90

Dear Sir:

Enclosed is the 1990 Annual 10 CFR 50.59 Report as required by 10 CFR 50.59.

Should you have any questions, or require further information, please contact us.

Sincerely, f ,e

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DBM/A, JN/MJB:dit v i Attachment cc: R.A. Burricelli, Public Service Electric & Gas T.M. Gerusky, Commonwealth of Pennsylvania JJ. Lyash, USNRC Senior Resident inspector R.I. McLean, State of Maryland T.T. Martin, Administrator, Region I, USNRC H.C. Schwemm, Atlantic Electne J. Urban, Delmarva Power 1dtevr.ltr

$ N .?, .b ? b i 9112160220 901231 k PDR ADOCK 05000277 R PDR

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bec: J. A. Basilio - O.A.Ilunger J. A. Bernstein . - E. J. McDermott Commitment Coordinator - D. B. Miller T.J.Robb-Correspondence Control Desk E. J. Cullen D. M. Smith-A. A. Fulvio

. D.- R. Helwig m- n ,,,,,,,, ,

1990 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT Table of Contents nytt am Eane Units 2 & 3 1 Modifications 371 Structural 2 967 Offgas 3 1029A Fire Protection 4 1684 Feodwater 5 1993 Structural 6 5112 Radiation Monitoring 7 Temocrary Plant Alteration -

40-02 Ventilation 8 Ena. Work Reaug31 P 51614 Control Rod 9 1911 Emergency Service Water 10 Nonconformanco Rcppus P 891003 Reactor Protection 11 P-89236 Service Water 12 P-89380 Torus Water Cleanup 13 P-89985 Diesel 14 _

P 90001 13kV 15 P 90006 480 Volt 16 P-90010 480 VAC 17 P-90081 RWCU 18 P-90121 4kV 19 P-90142 Standby Gas 20 P-90143 Post Accident Sampling 21 P-90149 Circuit Breakers 22 P-90152 Condensate 23 P-90100 Containment Atmospheric Dilution 24 P-90191 Drywell 25 P-90224 Containment Atmospheric Control 26 P-90256 Diesel Generator 27 P-90305 Fire Protection 28 P-90351 Offgas 29 P-90364 Domineralizer 30 P-90408 125/250 Volt 31 P-90444 Instrument Nitrogen 32

Table of Contents

.SYllt!D EM12 Nonc_qnformance Reoorts P 90455 Radiation Monitoring 33 P-90510 Radiation 34 P-90512 Circulating Water 35 P-90514 Motor Control Center 36 -

P 90515 Emergency Service Water / Reactor Bldg. Cooling 37 P 90529 Instrume:11 Air 38 P-90587 Condensate Filter Domin 39 P-90588 Raw Water 40 P-90594 Offgas Recombiner 41 P 90619 Alternate Rod Insertion 42

- P-90638 Service Water 43 P-90644 - Auxillary Steam 44 P 90663 125/250 VDC ' 45 P-90672 Fire Protction 46

-: P 90687 Contalnment Atmospheric Dilution 47 P-90700 Reactor Water Cleanup 48 P 90743 Radiation Monitoring. 49 P-90759 - Control Rod Drive Hydraulics 50 P 90779 Diesel Generator 51 P 90786 Core Spray 52 P 90795 Standby Liquid Control 53 P-90796 Stanoby Liquid Control 54 P 90797 Standby Liquid Control 55 P 91007 - Condensate 56 Unit 2 57 Modificatlom .

=1352A HPCI- . .

58 5199 Drywell/ Control Room 59 5224 Condensate 60 Temocra Plant Alteration 02-1? - Recirculation 61 Other TS-01 Emergency Service Water 62 Emergency Service Water 63

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Table of Contents System Enqt Nonconformance Reoort P-89997 125/250 VDC 64 P 90093 Emergency Ventilation 65 P-90109 Instrumentation 66 P-90250 Residual Heat Removal 67 P-90353 Residual Heat Remova! 08 P-90376 Radiation Monitoring 69 P 90383 SeMce Water 70 P-90407 Reactor Water Cicanup 71 P 90569 Turbine and Extraction Steam 72 Unit 3 73 Modifications 955J Plant Moni:oring 74 1353A HPCI 75 5143 Radwaste 76 86414 Reactor Feedwater 77 Temocrary Plant Alteration G2-4 Cranes, Holsts, Tools 78 62-10 Rod Protection , 79 New Operating Condition Ofigas 80 Nonconformance Reoort P-89782 Main Steam 81 P-89923 _ High Pressure Coolant injaction 82 P-90077 - Condensate 83 P 30082 - ' Condensate 84

- P-90083 Fuel Pool Cooling 85 P-90092 Reactor & Recirculation 86 P-90201 Main Steam 87 P-90208 Turbine & Reactor Building Cooling Water 88 P-90209 Ch!!!od Water 89 P 90211 Emergency Cooling 90 P-90212 Instrument N:trogen 91 P-90219 Reactor Core isolation Cooling 92 P-90223 High Pressure Cooiant injection 93 P 90356 Containment Attr.ospheric Dilution 94

. P 90391 High Pressure Coolant injection 95 l P-90398 Condensate 96 P-90399 ' Standby Liquid Control. 47 P-90524 Radiation Monitoring 98 P 90690 - Primary Centainment High Range Monitoring 99 P-90710 Recirculation, Residual Heat Removal, Reactor Water Cleanup 100

Table of Contents

.SyllCE EARt P-90762 Radiation Monitoring 1(,1 P 90792 Starxiby Uquid Control 102 P-90793 Standby Liquid Control 103 P-90794 Standby Liquid Control 104 e

PHILADELPHIA ELECTRIC COMPANY

- PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 DOCKET NOS. 50-277; 50-278 1990 ANNUAL 10 CFR 50.59 REPORT

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I Docket Nos. 50-277 50-278 1990 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT This Repoit is issued pursuant to the reporting requirements of 10 CFR 50.59 for Peach Bottom' Atomic Power Station Units 2 and 3 (Facility License Numbers DPR44 and DPR 58 respectively). This report addresses, but is not limited to, tests and changes to the facility as they are described in the Updated Final Safety Analysis Report. This rt nort consists of those tests and changes that were completed in 1990. A summary of the safety evaluation for each item, concluding that an unreviewed safety question, as defined in 10 CFR 50.59 (a) (2), was not involved is included.

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_ Units 2 & 3 g.- .

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4 PEACil DOTTOM ATOMIC POWER STATION UNITS 2 & 3

• '- DOCKET No, 50 277 & 278 199010 CFR 50.59 REPORT Jg. Administr,glign And _ Shoo Facilities Mo$iricaTion Noa. 371 .

A. Sy_ntu; Structural B. Der.cmetion: ,

This modification added new administration and shop facilitics and consolidated and revitatized some existing facilities.

C Reason Fon Chang The existing facilities were inadequate.

.D. Surry Evawanow Suuuany:

1). Does this modification increase the probability of occurrence or the consequences of an accident

or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer; No. This modificadon adds new construction. No safety systems are involved.

2) Does this modification create the possibility for an accident or malfuncilon of a different :ype than -

.any evaluated previously in the safety analysis report?

Answer:

No. .No systems related to safety are being modified,

3) Does this modification reduce the margin of safety as def.ned in,the basis for the Technical

. Specifications'? -

Answer:

No. There are no changes that are related 'to the margin of safety as defined in the Technical Specificat cne. 8 M

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PEACH BOTTOM ATOMIC POWER STATION-UNITS 2 & 3 DOCKET No. 50 277 & 50-278 199010 CFR $0.59 REPORT Ambient Charcoal Offaas AQsorotion Delay Sgtgm Moomemon Not - 967 A. Sysnut Ofigas B. Descamnow:

The existing compressed storage port;on of the offgas system was replaced with an ambient charcoal delay system. Other changes were made to improve system operability and reliability. The changes included installation of instrumentation and controls, reloca' ion of some existing controls, replacement of the unit V specific annunciators, repainting the control board and application cf Human Factors Engineering enhancements to the panels. Controls in the existing system were modified to simplify operation end improve system flexibility and reliability, Active drums in the offgas system were replaced v.ith loop seals, where possible, to eliminate active components and improve reliability.

C.Hpsow Fon Cwawcr:

The purpose of this modification is to improve system operability and reliabilrty.

D. SurTv Evnumon Suuuany:

1)

Does this modification increase the probability of occurrence or tne consequences of an accident or malfunction of equipment important to safety as previously evaluated in the saf ety analysis report?  ;

Answer:

No. The modified offgas system operates at essentially atmospheric temperature ano pressure, which reducas pipe stress significantly, thereby reducing the probability of any leakage t illures which were analyzed for the FSAR. The system is designed to resist the forces due to a hydrogen detonation, and because active components in the main process stream have been minimized. the

_ probability of a detonation and/or instantaneous release has also been minimized.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?__

Answer: +

No accidents or malfunctions different from those previously analyzed are created by tne new system. The total release of all delayed gases has been ranalyzed previously and a full scale performance test has been performed to assure consistency with the PBAPS license.

3) Does this modification reduce the mary.n of safety as defined in the basis for the Technical Specifications?

Answer:

No. Release and dose rates are expected to be lower than the rates for the previous system. The revised offgas system is designed, fabricated and constructed to meet t! o intent of Regulatcry Guide 1.143.

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PEACH BOTTOM ATOMIC POWER STATION

, , UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199010 CFR 50.59 REPORT Encaosulate Raceways Pull Bovet and Junction Boxeji JAonneaTion Ptconi Not 1029A A. Systru: Fire Protection s

B. Desen.!ntgw_;

This modification involved encapsulating raceways that contain cables connected to C-listed equipment.

C. Reason Fon CHANGE:

~

The purpose of Mod 1029A was to encapsulate existing raceways in 1-hour or 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire barriers in specified locations throughout the plant and bring Peach Bottom into compliance with Appendix R to NRC Regulation 100FR50.

D. SusTv EvawAvios Snuuaav:

1) Does this modification increase the probability of occurrence or the consequences cf an accident cr malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. Ampacity dert.tes and add;tionalloads on hangers were considered.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evclusted previously in the safety analysis report?

,Answem

, No. The raceways are being encapsulated to protect the cables within. The safety related function of the equipment is not changed..

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. The ractway are nct listed in the Technical Spec!!ications.

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PLACH BOTTOM ATOMIC POWER STATIOf1

• UNITS 2 & 3 D 'CKET llo. 50-277 & 50 278 199010 CFR $0.59 REPORT ,

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i In5.tgl Control Valyg )

jdgoinemen Rgfon? Not 1084 A. S stru: Feedwater B. j}ngw now:

This modification involved the Installation of control valves on the 3rd and 4th foodwater hecto' extraction steam drain lines.

i C. Reason Fen Comoc i

This modihcation will otevent turbine water induction.

l D. SArrry liyjuyft!ow Spuuny:

1) LDoes this modification increase the probability of occurrence or the consequences of an accident -

- or malf uncilon of equ!pmen' important to safety as previously evaluated in the saf tti analysis report?

.ABSwer  ;

No. The new valves will close automatically and prevent turbine water induction. Th6 ade: tion of ,

valves only aNects feedwater heater extraction steam drain lines.

2) Does this modification create tha possibility for an accident or malfunction of a different type than l any evaluated previously in the safety analysis report? -

3.01EtU No. These lines are not required for s3fo shutdown.

3) Does this modification reduce the margin of safety as deid in the basis for the Technical Specifications?

• S.i1*EIG No. - The feedwater extraction steam drain liids are not addressed in the Technical Specif2ations.

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t PEACH B01 TOM ATOMIC POWER STATION UNils 2 & 3

. . DOCKET No. 50 277 & 278 199010 CFR 00.d9 REPORT

.Efiicnment_of RE! Lng M?r!! Lee.109 No.1 1993 A. itin u; Structural B._D n em m eg

4. e % t This modification replaces roofing on various structures with sing!e ply roofing. Section 12 of the UFSA9 9[

will te revised to show these changes. h C. Rrascu F9a_.Quawng; The old roofing shows damago and ago. [I D Sarm Evuyinpu Suuusavt

1) Does this modification increa ,e the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the cafe:/ analysis report?

Ansvttrj No. The new roofing will meet ol existing requirements. This change does not aff ect any equtment important tc safety,

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety rinalysis report?

Ariggg No. No designs or functions were changed. Th!: modification involves roof covering only.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

A._namn ,

No. There are no changes th:.1 would affect any system reliability.

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PEACH BOTTOM ATOMIC POWER STAT 10f4 UtilTS 2 & 3 DOCKET No. 50 277 & 278 i 1990 to CFR 50.59 REPORT

.Qetetten Of Yhe Control Room VentJMon isolation FunctiorLE_ rom padiation Mon; tors fdoomeariewNo.t 5112 A. Smtut Radiation Monitoring B. Drsenietion:

This modification ueletes the control room ventilaticn isolation function from the control room radiation .

mon 1 tors' high.high radiation trip, as previously described in UFSAR Sections 712.5 and 10.13.

C. Rrason FoLQf_ay_g;  !

f Eliminating the high hlgh trip would ensure that the control rcom retrained in a post accident filtered air intake mode with posltive pressure control, minimiting radiation concentrations.

D. Sartry Evatvatiew Suuuaav:

1) Does this modification increase the probability of occurrence or the consequences of an accident '

or malfunction of equipment important to safety as previously evaluated in the saf ety analysis repon?

Answer:

f

, No. The UFSAR design basis accidents do not eva'uate the centrcl room in the Isolated mode (High-High Radiation)and the subsequent NUREG o737 controt room habjtability analysis cencluded .

that the operator radiation exposur0 would be less if the control room would remain in the fl!!ered r intake mode rather than In the isolated tmace. -In addition calculation. ME 245, which determined ,

the High-High radiation setpoint,-Indicated that for the design basis 1.OCA the control room fatered  ;

intake radiation concentrations wou!d not exceed the monitors' high high setpoint.

2) Does this modification create the possibility for an ace; dent or malfunct!cn o' a different type than +

any evaluated previously in the safety analysis report?

Answer: ,

i No. This modification does not involve an initiating. event of any design basis accident: The

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consequences of the design basis accidelits have been evaluated with the control rcom in the filtered intake mode and the NUREG 0737 licensing basis controf room habitability analysis indicated

- that the operator radiation exposure is reduced by remaining in the filtered intake mcde cf the control room ventilation system.

- 3) Does this modification reduce the margin of safety as defined in the basis for the Technic.at Specifications? r Answer:

t No. 'This modification more accurately reflects the system as described in the Technical Specifications, section 3.11.

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PEACH DOTTOM ATOMIC POWER STATIOt1

• UtilTS 2 & 3 DOCOET No. 50 277 & 50-278 1990 to CFR 50.59 REPORT flaleJ!ntto!'alinnJor Ventdallen_0JacMoe itxtszu22.LhuMh9% 40 U2 A. Sans Ventilation i

U.ht%CR%),Q%  ;

I This Temporary Plant A!!cration installs plates to blank off the Reactor Building supply venblation to the A and C Residual Heat Removal (RHR) Rooms C. Braign Fon_Cnawu This will minimlre the pniential releau ni Activity outeldo secondary containtnent during remnval of the RHR hatch.

D. lurry livjwmon Ssnaan

1) Does this modification increase the probability of occunence or the consequences'or an accident l or malfunction of equipment important to safety as previously evaluated in the saf ety analysis repon? ,

A.Il1LW.fG t'

No. Blanking off the normal ventilation supply ducts to the RHR Rooms has no effect on safety related systems. If equipment operation in the RHR Room were fequ%d, the Compartment Room '

• coolers could periorm the required cooling.  ;

2) ~ Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? ,

i Answff;

, No. This Temporary Alteration allows the Reactor Building Ventilation System to perform its intended function. q

3) Does this modification foduce the margin of safety as defined in the basis for the Technical ,

l-_ Specifications?  !

w t bI.WS.C No. There are no Reactor Building ventilation requirements in the Technical Specifications. This Temporary Alteration does not reduce the margin of safety for Standby Gas Treatment or Secondary i Containment.

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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 50 278 199010 CFR 50.59 REPORT

.Use Of Loncer Ufo _Qgntrol Bod E!adn Enmuerawa Won.drovrst Noa P 51614 A. Svsmi Control Rod B. Descamveut Existing control rod assemblies will be replaced by longer-tife assemblies, C. Biato.w_Ee.a Cuawnt; These assemblics have a longer lifetime expectancy, are made of improved materials and have revised design features. They are designed to be direct standard replacements fcr the current controi rod assembiies in use. Section 3.4 of the UFSAR will bo omended to reflect this change.

D. jarrty Evatuanou SuuuAav:  ;

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis repon? *

.8.01.EtG No. These new assemblics are direct replacements. Their mechanical and nuclear properties do not differ from those of the original assemblies in any way that is significant for normal or accident conditions. i

2) Does this modification create the possibility for an accident or malfunction of a different type than '

any evaluated previously in the safety analysis report?

Answert

- No. The new ascemblies are operationally the same as the original assemblies. They have basically the same mechanical and neutron absorbing properties as the original assemblies.

3) Does this modification reduce the margin of safety as defined in the basis for the Techni:al

. Specifications? ,

_.y Answer:

No. The new control rod assemblies comply with Technical Specification requirements. Sections .

3.3 and 4.3 have been reviewed to make this determination.

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PEACH DOTTOM ATOMIC POWER STATION tJNITS 2 & 3 DOCKET No. 50 277 & $0 278 199010 CFR 50.59 REPORT DDin!nQ oLQms Tie Valves 1mi A. Snnul Emergency Service Watcr (ESW)/ Core Spray (CS)

D. Entalt2931 -

This is a temporary change h' the operating configuration of the Emergency Service Water (ESW) and Core Spray (CS) systems. The chege isolates ESW to the CS motor oil cooler.

C. BLs1qu.fon Cusuqn This is a test to demon, ttxte thM t WW ..i in this manner is acceptable. The test is required because of ESW flow rates to the cpe(t 'AM Ma Jwer than those recommended by the vendcr.

D. Sarcty Evagariew Syu_ua2yv

1) Does this modification increase the probabihty of occurrence or the consequences of an accider1 or malf ur.ction of equipment important to safety as previously evaluated in the saf ety analysis report?

Answn;-

No. The probability of an accident is independent of the presence of mitigating equipment. This temporary change, therefore, does not increase the probability of occurrence of an accident previously evaluated in the SAR, The consequences cf an accident will not be increased cecause

' the CS pumps will remain in service and will be available in the event of an automatic initiation signal.

At that time, the cooling water to the motor oil coolers will be valved in and the system will function as 6 signed. The pumps will be tested individually.

2) Does this mcdification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis recort?

Answer:

No.' The performsnce of this test can in no way create an accident of a different type than previously evaluated._ The bearing temperature will be continuously monitored to assure that the pump will' 4 not be made inoperable as a result of exceeding the maximum temperature as statec by the manufacturer. This temporary change maintains the CS systems operability, thereby ma:ntaining the ability of the plant to operate wichout creating any accident initiator not considered int the SAR.

3) Does this modification reduce the rnargin of safety as defined in the basis for the Technical ,

Spec:fications?

Answen No. Technical Specification Section 3 5A was reviewed to make this determinatiott i

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PEACH !!OTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 278 l 1990 to Cf'R 50.59 REPORT l l

fatit.10MD!k11100 l I

Rqygowtoepagtlktgat Not P-891003 l A. fitntui Reactor Protection il R ugawt gj .

This NCR evaluates the suitability of the stainless steel sheathed siliccn cable and conduit Installed by Mod 2383. UFSAR sectlens 7.2.3.10 and 0 4.5 will bo changed to reflect the use of the stainless stcol silicon

. dioxido cable.

C Dsucn Fon Cnsugc ,

This NCR is disposttionod to use as is.

D, Suny Evauta.nqw,JauMim.21

1) Does this modification increase the probab2lty of occurrence or the consequences of an accident or malfunction of equipn.ent important to saf ety as previously evaluated in the safety analysit report? ,

Answer: .

No. The new cable is equivalent to the previously installed cable and condult. *

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

AJ11EtG ,

No. The stainicss steel sheathod, silicon dioxide cable has been determined to be superior to the ,

original cables. As such, the cable used will not atiect the function of the system.

3) Does this modification reduce the margin cf safety as defined in the basis for the Technical Specifications? -

_Answen f No. No changes were mado to the Techn cal Spemfications. The new cable performa the same function.

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PEACH DOTTOM ATOMIC POWER STAT 10f4

' ' UtilTS 2 & 3 DOCKET No. 50 277 & $0-278 199010 CFR 60.59 REPORT ,

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.Brsolution of Elst[tNncyf ttMfDlidiz!!Liind Decumtritajlpa f)2wqpagnuagr Rtront t{pj P 0923G A. S n n uj Cervice Water

• B. Atseniettowl This NCR ider," r . discrepancies between the as built configuration. USFAR Figure 10A1. and associated i documentation. The document revisions added and deleted vent sample and drain lines, changes in valvo positions. and the addition of valve mark and identification numbers. The documentation wl!I be isvised to ICflect as bullt conditions. j C. Rrason Fon CHawor:

This NCR is dispositioned to use as is.

D. Sarrty Evawation Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an acc! dent or malfunction of equipment important to safety as previously evaluatnd in the safety analysis report?

Answer:  ;

No. No change was made to the plant configuration, The changes made do not affect the system design basis as described in the SAR.

2) Does this modtfication create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

- No. The revision was inillated so that the drawing reflects the "as built' condition of the Service Water system; System function is not affected.

3) Does this modification reduce the nwgin of safety as defined in the basis for the Technical .

Specifications? .,

Answff; i No. There is no description of the Service Water system in the Technical Specification.

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l PEACH DOTTCM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 278 199010 CFR 50.59 REPORT Q.h_An1ged Valye Pclit!RD l

f(p.ttq.que onuawcdeqatJt; P.09380 rev.1 A. Sottu; Torus Water Cleanup D. Enqnmfiow!

This NCR evaluates the valve positions for valves 2(3).14A 73 and 14 2(3) 0038 on the Torus Water Cleanup System The valve position for 14A 2(3)D038 was changed based on the later.t revision of the System Operating procedure SO 14A.1.A.2 (3). UFSAR Figure 7.4.4 and associated docurnentation will be revised '

to reflect this chsnges.

C. Rrasow Fon Cnau.gn This NCR is dispositioned to use as is. i D. .fiarrty Evat.yf.nqduuaft

1) Does this modification increase the probability of occurrence or the consequences of an accident l or rnalfunction of equipment important to safety as previously evaluated in the safety analysis repon?  !

ADLw3D No. This actMty does not affect the ability of any system or component to perform its intended safsty function. UFSAR Section 14.0, Plant Safety Analysis, was reviewed. ,

2)- Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?  ;

AI15EtG No. This act!vity revises the applicable P&l0 and UFSAR Figures and does not cause an accident

. initiator not considered in the UFSAR.

5) Does this modification reduce the margin of safety as defined in the basis for the Technical i Specifications? e t hat.w. 9E No. The Torus water cleanup system is not specifically addressed by the Technical Spec:fications.

This actMty does not affect the limiting conditions for operation or bases of Technical Specification Section 3.7

• Containment Systems".

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I, PEACH DOTTOM ATOMIC POWER STAllDt1 UtilTS 2 6 3 DOCKI'T fio. 50 277 & 278 199010 CFR 50.59 REPORT DJMcLG1!1EDltr ,,AltjystrnLQ,tgng;ji 1{gyqpyLqauawgJ1teent _Ngj P fl9985 A Dnnu; Diesel

. D, QUcmovoyi This NCR Identifies changes required to the locked valve arrangement on the diesel generator starting air system to ensure air supply to each engine. UFSAR Figure O S.1 sheet 1, and associated documentation

- will be revised to reflect these changes.

C. Etait!LEstCeten; -

This NOR disposition approves the additica of locks to valves $2C 10008A, B, C, O and 52C 10009A. B O.

D to ensure the availability of air from the reservolt tanks to the diesels, Removing the locks from valves

- 52C 10072A, B, C D (non safety related) does not affect the intended safety function of any component or system.

D.Jarrty Evawavow Suuuan,tj .

t 1)

Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?- l i

A01WUJ

- No. The changes described do not affect the passive safety related function of the valves as ,

pneumatic pressure boundaries. Adding locks to the existing normally open valves enhances the availability of auxillary support features provided by the diesel generators. Removing the locks from valves 52C 10072A through D does not affect diesel start since these valves are upstream of the air ,

reservoirs and are not safety related, .;

2) Does this modification c cate the possibility for an accident or : Tlfunction of a different type than any evaluated previously in the tafety analysis report?

Jnswer:

No. Revising the control of the valve position does not increase the potential for initiation of any I

- accident scenario and the safety related function of the valves is not affected.

3) Does this modification reduce the margin of safety as defined in the basis for the Tghnical'-

Specifications? .

h!nnH1  ;

No.- The limiting conditions for operation of the aunitary electrical system (Technical Specification 3.9) and surveillance requirements (Tecbr*al Specification 4 9) were reviewed. Operabihty of tho ,

4 diesels is maintained by this revision,

'14-

6 PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 278 -

199010 CFR 50.59 REPORT i t

! Resobt;Qn gf Qinffnancy Between As-Is and Documentation i

Nowgownopuawcr Re.paa_!Lol P 90001 A. Srstauf 13 kV B..QsAgrenoyi This NCR identifies discrepancies between the as built configuration and UFSAR Figure 8.41. The figure will be revised to correctly indicate that the 13.2 KV feeder terminates at the No. 6 L & P transformer, ,

Rrasou eon Cf.a.wgr.1 This NCR was dispositioned to use as is i

D. Sarrty Evatuangs Suuuany:

1) . Does this modification increase the probabillty of occurrence of the consequences of an accident or malf unction of equipment important to safety as previously evaluated in the safety analysis report?  !

i l

. Answer; No. This is drawing correction only and will reflect the plant as built conditions, 2)_ Does this modification create the possibility for an accident or malfunction of a different type than '

any evaluated previously in the safety analysis report?

Answer: ,

No. Correcting the drawings to reflect the as built conditions of the plar 'I not increase the l consequences of an accident previously evaluated in the safety analysis re e .

3) Does this modification reduce the margin of safety as defined in the basis for the Technical-

. Specificatlons? t Answerj .

No. Section 3.9 of the Technical Specifications was reviewed and makes no reference to margin of safety for the substation light and power distribution panels' Therefore, the margin of safety as defined in the basis of any Technical Specification is not reduced.

T t

15 i,, -a. ., . - . . _ _.. a ,. _ _ _ ,_ _ .. _ ._..__.. ,,:- .,.;. .. ;,,,. ,s.____ m,..,. ..m4,-,.._-,, . , . _ , . .

_ . . . _ . _ _ . . _._ ~ _ _ - _ . - - _

PEACH DOTTOM ATOMIC POWER STATION

• ' UNITS 2 & 3 DOCKET No. 50-277 & 278 ,

199010 CFR 50.59 REPORT l

- ,4MLVAC Drawinn _Discremnqy JLqwcourpauanctBtroat Not P40006 A. jifunu; 400 Volt D. ,QL5CalPT1oM This chango involves revising the 440V single lino drawings to chango the r.afety clasOfication of the 24V battery chargers. The singlo lino drawings and UFSAR Figures 8 4 6A throug 1.4.6D will be revised to show the chargers to be non. safety related.

. C.Erasow Foa__Qmqn The change corrects discrepancies and brinCs drawings into conformance with current safety class:fication of the equipment.

D. Som Evatuat.injiyuuaav:

1) Does this modification increase the probability of occurrence or the consequences of an accident

- or malfunction of equipment important to safety as previously evaluated in the sa!ety analysis report?

AnsWen No. This change is required only to correct drawing discrepancies and bring the drawings into conformance with the current safety classification of the equipment.

2) Does this modification create the possibihty for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answen ,

No. Correcting the drawings to reflect the current safety classWication of the equipment will not Increase the consequences of an accident previously evaluated in the safety analysis report.

3) Does this modification reduce the margin cf safety as defined in the basis for the Technical Specifications?

y

- AnsWyg No. Technical Specification a 9 was reviewed and makes no reference to margin of safety for the-24V battery chargers.

16

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i

, i PEACH DOTTOM ATOMIC POWER STATIOfi

'

• UNITS 2 & 3 DOCKET No. 50 277 & 278 1990 to CFR 50.59 REPORT i

.6nglution Of DiscreconcifMwrgn As. Built And Dggmtntatiqg l Newcoutoauanet Preont f!g P 90010 A. Systru: 480 VAC B.ptsem* TION:

This NCR Identities spare MCC compartments 52 2682. 2083,3754,3314. 3881. 3895,3901 and 3993 for i Peach Bottom Atomic Power Station Unit 2 and 52 2622 for Unit 3, which were removed for preventive maintenance but were not reinstalled for economic reasons. These are shown on UFSAR Figures 8 4 CA and 8.4.6B and associated documentation. The figures will be revised to reflect the as built condition.

C. ,Brasow Fon Cwawom The NCR is dispositioned use as Is.

D. SareTv Evuvation Suuuanv:

1)' Does thl4 modification increase the probability of occurrence of the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?  ;

Answer:

No. This change does not affect any of the plant systems or equipmeni function as described in the safety analysis report. The change involves showing ' spaces

• instead of ' spares

• on single line, tabulation, and secondary and control connection drawings to reflect the as built conditions of the plant.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. Correcting the drawings to reflect the as built conditions of the plant will not increase the consequences of an accident previously evaluated in the safety accident- report since the nonconformance report does not make an4 functional changes to the plant systems or equipment.

3) - Does this modificatien reduce the margin of Afety as defined in the basis for the Technical Specifications?

Answe,rj No. There are no Technical Specifications applicable to the mentioned MCC's compartments.

4 l

1.

i l 17 a - , -,-. . .- - . , . . - . - . - - . - _ .

I f

PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 278 199010 CFR $0.59 REPORT Mgation Of The Reactor Water Clean Un (RWCU) Valves JJoncouromu.awer Rrpont Not P.90081 A. Smrut RWCU B. Drscrupups -

Figure 4.9.1 of the UFSAR and associated documentation were changed to reflect the l

elimination of RWCU valves on:

1) Cooling lines outlet piping off the RWCU pump seal (3 valves.1 per pump),

2) Bearing Housing cooling lines outlet piping off the RWCU pump (3 vanes.1 per pump).

3) 1* HH header to the RWCU pumps drain lines (2 valves).

4) Outlet piping of the RWCU pump 'C' bearing cooler 3CE27, downstream of relief valve 8%

3705C (1 valve).

C Erasow Fon Csanot: j These changes are in accordance with the NCR disposition.

O. Sartry Evatvavon Suuuaav:

1) . Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the saf ety analysis repon?

_ Answer:

No. The non-existence of the valves has no effect on the ability of the RWCU system to perform -

its intended function. UFSAR Section 14.0, Plant Safety Analysis, was reviewed to make this determination.  ;

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer; I

No. The RWCU system provides clean up processes to the reactor water. The consequences of the anticipated and abnormal transients and Design Basis accidents described in the UFSAR Section 14.0 are not affected by this activity.

3)- Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? -

Answen .

No. The margin of safety as defined in the bases of Technical Specifications 3.1,3.2, anc 3.86 is not reduced by this activity i

l 18

._. _ __ _ _._ - _ _ _ . . . _ _ ~ - _ . - . _ . . _ . . _ . . . _ . - . . ~ . . _ . . _. , _ . . - _ . , _ . . , _ . . . -

PEACH DOTTOM ATOMIC POWER STATIOtl

• UNl1S 2 & 3 -

DOCKET flo. 50 277 & E0 278 I 199010 CFR 50,59 REPORT Epsolution Of Discrfpansylsveen Asjui!1 And Qogumentation

_Nowgntgauanct Rrront Ng1 P-90121 A. Systru: 4kV B. Descawtioni This NCR identifies e discrepancy between single lino diagram E-0, Sheet 2 and Table 3.2 B of the Technical Specifications regarding the 4KV Emergency Bus undervoltago relays trip setting and UFSAR Figure 8.4.3 sheet 2. The single line diagram end UFSAR will be revised.

C, ELAlgMon CHawcII Tho disposition of the NCR will bring the UFSAR and single ilne diagram into conformance with Technical specifications.

D. Sarrry EvatvAtto8 Suuuanyt

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the saf ety analyMs repon?

A01Ef.U No. This change does not affect or modify any of the plant safety or non safety related systems er equipmbnt as described in the SAR. It only revises the UFSAR and a single lino diagram to bring

~

them in conformance with the Technical Specifications.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

$Dswert No.- No changes or modifications were mado to the relays or trip settings.

. 3) Does this modification reduce the margin of safety as defined in tho basis for the Technical Specifications? #

Answen No. This NCR revises single line diagram E-8, Sheet 2 UFSAR Section 8.4.0.2 and Figuro 8.4 3.

Sheet 2 to clarify and make the trip setting value for the 4KV Emergency Bus undervoltage relays consistent with the Technical Specification value shown in Section 3.2 and Table 3.2.B.

4 19-

i PEACH 00VTOM ATOMIC POWER STATION

, , UNITS 2 & 3 DOCKET No. 50 277 & $0 278 199010 CFR 50.50 REPORT I

Eelolution Of Discrecancy Between As-QuilunclQngnjentation I

1 NewcocoauAuct Report Ngj P 90142 ,

A. Enns Standby Gas B. MisCRIPTtoN' This NCR identifies drawing inconsistencies between the instalkd configuration of the standby gas treatment [

system ductwork and the Unit 2 and 3 heating and ventilating physical layout drawings. The drawings included non-existent relief dampers, inspection doors, incorrect gauge of duct coristruction, and incorrect instrumentation nomenclature. The UFSAR Figures 5.3.1 and 5.3.2 and associated documentation will be revised io show the 8$ 15 condition.

C. Acasow Fon CHawo11

- This NCR is dispositioned to use as is. I D. jbrtTv Evatumow Suuuanyj  ;

1

1) Does this modification increase the probability of occurrence or the consequences of an accident I or malfunction of equipment important to safety as previously evaluated in the safety analysis report? j AMEffi No. Drawing changes reflecting the "as built

• condition of the plant have no impact on the accident analyses in UFSAR Section 14

2) Does this modification create the possibility for an accident or malfunction of a different type than

- any evaluated previously in the safety analysis report?

Aciwku No. The as built documentation will not impact the design or operation of any safety-related equipment.  ;

' 3) Does this modification reduce the margin of safety as defined in the basis for the Technical

• Specifications? +

- AM.WEi No. The as built condition of the plant does not alter the system design bases as addressed by the Technical Spec #ications.

k l

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i ,

PEACil DOTTOM ATOMIC POWER STAllON  !

. .- lJN11S 2 & 3 l DOCKET No. 60 277 & 50 278 f 1990 to CfD $0.69 REPORT l

[

fleinfulltu QLRiac.troancy OcMernAttitla!MendJk!st.t!nen!@p11 l

Rossmestattnat Nai N 9143 l A. Snng; Post Accident Sampling SyStern (PASS)

D.p.LSen+tiog l

This NCR Identitled valve arrangement discrepancies between various design documents and tho 'as- .

Installed condition for the Post Accident Sampling System (PASS). UFSAR Figure 7.20,1 and associated -

documentation will be revised to show the actual system configuration, i C. Hrasow Fom C t93011-This NCR is dispositioned to use as Is.

D. Engly.,1gaLu.a.nn9_Ama2yl i

1) Does this modification increase the probability of occurrence or the consequences of an accident '

or rnalfunction of equipment important to safety as previously evaluated in the nafety analysis report?

An1M11 No. This actMty makes drawings changes only, I i

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

An%E No. This actMty maintains the design function of the PASS and does not cause an accident initiator not considered in the SAR. ,

f

. 3) Ooes this modification reduce the margin of safety as defined in the basis for the Technical .

Specifications?

Answert

r. -

No. Technical Specification Section 019 was reviewed, i

l e

21 u . _ u - . _ ___ ., , _ - . , _ _ _ _ . . ~ . . _ . , _ . - , _ . . _ . . _ , , _ . . . . _ . . . - _ . _ _ , - , ~ _ . -

s 1

4 PEACH UOTTOM ATOMIC POWER STATION

.- , UNITS 2 & 3 DOCKET No. 60 277 & $0 278 199010 CFR $0.59 REPORT

. Br$Q!UdODALDRILRNDr1.EtMOULh1EUU AndERulH!CU!MRU NoNCoktoMMANCE N(Pont No.: P 90149 A. Airau,; Circuit Breakers B,png_nmuow:

• ThIs NCR identifies severai lose and circuit breaker ratings shown in figures 8 0.1 A. 8 0.10, and 8 6 2 of trio UFSAR that are not in agreement with the as built conditions of dl hilon panels 00YO3. 00YC$1. 00Y$3, 00Y54,00Y57,20Y34, and 30Y37. Some fuses and breakers will be replaced. while some are acceptable-as is, and will be left in place. UFSAR Figuros 8 6.1 A,0 6.10, LO 2 and associated drawings will be revised to reflect the fatings of the fuses and circuit breakers found to be acceptable as is, C.Brnen Fon_Cunon The physical work and drawing changes are being made in accordance with the disposition of tr'e NCR..

D. Sarm Evawation SpujtaaJ -

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis repon?

Armwer

- No, An evaluation of the as built condition of each fuso and circuit breaker was penctmed to ent!Jte that fuses or circuit breakers were adequately sized to protect the cable and equicment in the circuit, in some cases the fuse or circuit breaker must be replaced to adequately protect cable and equipment. In those instances the breaker will be tcfaced to conform to the rating indicated on the singlo lino diagrams.

2) Does this modification create the possibility for an accident or malfunction of a different type thar-any evaluated previously in the safety analysis repon?

A.QDERG No. The function of the circuits led from these breakers is not being Changed.

3) Does this modification reduce the margin of safety as defined in the basis fcr the Technical Specifications?

Answer:

No. Technical Specification Section 3.9 was reviewed to make this decision..

l:

22 l

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PEACH UOTTOM ATOMIC POWER STATION UNITS 2 E. 3 DOCKET No. 60 277 & $0 278 199010 CfH 50.59 REPORT -

fiesolution Of Distriparlq1DttAc!;n As-llullLAnd Qpcymtnjallag

.ll93GelGSMANSI-HtPoni f{pj P 90152 A. Silluyl Condensate

n. .h1S?It11951 This NCR Identified discrepancies between the as-built piping and existing design drawings at the condensate t,hort path recire control valves (CV2110 and CV3110). UFSAR Figuro 11.0.1 sheet 4 and associated documentation will be revised to reflect the as. built piping, location of the low point drain and '

valve installation detalls.

C. Hrasow EsLGna.snt:

This NCR is dispositioned to use as Is.

D. .Sant_v1vatuation Svuuanyj

1) Does this modification increase the probability of occurrence or the consequences of an accid .t or malfunction of equioment important to safety as previously evaluated ln the saf ety enalysis tepon?

Answef; No. This update reflects 'ar&dat' conditions. It does not make changes to system function er tests.

It does not affect the ability of the Conndensato System to perform.

2) .Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis repc,rt?

AURHl No. This activity has no impact on any malfunction of equipment.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifichtlons?

Answer: P-No. This 'as installed" configuration change does not affect the ability of the Condensate System to function as intended, 1

t 23

PEACH DOTTOM ATOMIC POWER STATIDH UNITS 2 & 3 DOCKET No. 50 277 & 50 278  ;

199010 CFR 50.59 REPORT Eciolution Of Discrep3.4C.ylthyeen As-Instatjed AndJpeumentati.gn figysyy.rppuanct Rurons No: PD0190 A.,5vstru; Containment Atmospheric Dilution D. .(2tlematioH_1 This NCR Identifies a discropancy between UFSAR section 5.2.3 9.4 and the actual configuration of 1 equipment in the control room. The UFSAR indicates a level alarm in the control room when one does not actually exist. The UFSAR will be revised to delete the toforence to a hydrogen alarm in the control roorrt C. Emon Fon CHawor:

This NCR is dispositioned to use as is.

D. Samy Evawation Stuua2y1

1) Does this modificatiollncrease the probability of occurrence or the consequences of an accident or malfunction of equipment impettant to safety as previously evaluated in the safety analysis report?

Answen.

No. Hydrogen and oxy;)en te e generated following a LOCA as a result of metal-water reaction and radiolytic deceir.pos41on of water, lack of an alarm feature in the control room for post LOCA hydrogen concentration level in the primary containment does not increase the probability of a LOCA, nor does it increase the consequences of an accident, Post LOCA oxygen and hydrogen concentration levels in containment atmosphere are indicated, recorded, and the oxygen level is alarmed in the control room. Primary containment oxygen concentration is maintained at less than 5% volumo by the addition of nitrogen through the Containment Atmosphere Dilution system.

2) . Does this modification create the possibility for an accident or malfunction of a differert type than

- any evaluated previously in the safety analysis report?

Answen No. The maintenance of a low oxygen concentration level (less than 5% volume)in the post LOCA containment atmosphere prevents buming"or explosion regardless of the hydrogen available.

3) Does this modification reduce the tmtgin of safety as defined in the basis for the Technical Specifications?

j  ; Answert ,

g' _ Nc. The lack of an alarm feature in the control room for the post LOCA primary contairenent

..3~

atmosphere hydrogen containment level does not reduce the margin of safety as defined in the basis of any Technical Specification. Technical Specifications 3.7.A,41A,3.7,D, and 4.7.D and associated bases were reviewed in making this determination.

24

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~ . - . . . . ... . _ . . - - - . . . - . - . . . . - - . _ . . _ _ - - - - . _ - . - . _ - -

PEACH BOTTOM ATOMIC POWER STATIOf1 UNITS 2 & 3

• DOCKET No. 50-277 & 50 278 199010 CFR 50.59 REPORT Pe*,olution Of Discrenancv Eetween As.l L And Qoggfr{11tMigg No co aopuwet Rrrent No: P 90191 A. _SY$ttMl IryWOlj ,

t B. Drsq!sSW This NC8 sg' s 4 itvs removal of the reference to the drywell temperature alarm in the Control Room from UFSAP, tahc. A N.12.

C Russ[h!Meun Thir, a!r.r' rne1, ret exist.

D.I EI E EP3.u M G1!Ji1Luu m :

1) Oce!< tt, s modification increase the probability of occurrence or the consequences of an accident

,or rnaltretion of equipment important to safety as previously evaluated in the safety analysis report?

b1tMC No. Increased drywell temperature would be detected by a sump level alarm in the case of steam lea'ks, a pressure alarm and/or recirculation seal alarm in case of a recirculation pump seal failure or valve packing teak, and by a fan discharge high temperature alarm, which detects Ngh temperature on the outlet side of the fan coil units.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

AnW!B Nolhis disposition does not delete or modify equipment required for safe shutdown or downgrade suppor, system performance.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical

' Spacifications? . , . ,

Antw3D No. Technical Specification Section 3.2 was revir'wed in making this decisiort ,

r i

i 1

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'b 25

,. . _ .an . -. _ _ ...._ __. _. _ ,_ _ , . . . . _ _ _ . . _ . _ _ . _ _ _ , . _ , _ _ _ _ . .

. - . _ . _ _ _ _ _ _ . _ _ _ _ - _.__.__._._.m._-- _. .___ ,_.,

PEACH 00TTOM ATOMIC POWER STATION UNITS 2 & 3  !

.c ,

DOCKET No. 50-277 & 50 278 199010 CFR 50.59 REPORT .

Resolution Of Discrepancy Between As trnt,Mkd And DqalrntnDtjen  ;

5 jducowoauaNCLBpont E Ng; P 90224 A. jirugu_j Containment Atmospheric Control System (CACS)

D. ,QJSca?flo$

This NCR 1dentifies discrepancies identified during a plant walkdown between the Unit 2 and 3 as installed

- CACS, UFSAR Figures S.2.7 S 3.2,1.3.1 and other documentation. The docurrentation will be revised, instrument valve numbers and valve positions will be changed and taferences to flow Indicators which do --

not exist will be doloted. ,

C. Brason Fon C8 anon This NCR is dispositioned to use as is.  ;

D. Sum Evatuavon Suuusay:

1) Does this modification increase the probabil!ty of occurrence or the consequences of an accident  !

- or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

AQ1WRG No. These are minor changes which reflect the as built condition of the plant and do not adversely affect the function or operation of the CACS or interfacing systems.  ;

5

2) Does this modification create the possibility for an accident or rnalfunction of a different type than any evaluated previously in the safety analysis report? (

Answer: '

No. The discrepancies do not create any new accident initiators or affect any existing accident  !

Initiators such that a differenct type of accident than previously evaluated would result.

3) Does this modification reduce the margin of safety as defincd in the basis for the Technical Specifications? v"

~

d!1Lw_tG 6

. No. The discrepanc8*s do not impact the Technical Specification bases of the Containment

. -1 Atmospheric Control System or any interfacing systems. This was confirmed by review of Technical Specification Sections 3.2,3,6,3.7,3.8, and 6.0, and associated bases.

l

-26

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PEACH DOTTOM A10MIC POWFR STATIOf4 UNITS 2 a 3 DOCKET No. 50 277 & $0 278 199010 CFR !!O.59 REPORT EC1019 tion Of Discrepanc.yfrlwrtrLUte A515 3 A d QCRnF1GfbD JJpassuronwawafteqat No: P to250 A. Ennu; Diesel Generator D. _Drsemeuout This NCR use as Is disposition involves correction of equipment and valvo discrepancies and revisions to the O list for the diesel generator auxillary systern. Corrections to UFSAR Fi0ure 8 5.1 and associated documentation are necessary. There are no physical changes associated with this NCR.

C.Brann Fon Cusuon .

The NCR is dir, positioned use as Is.

D. _ Sum EmVanW Slu_u8 Era 1)- Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previousf revaluated in the safety analysis ter ort?

Answer:

No. The changes addressed in this NCR do not affect the reliability of the diesel generators nor do they contribute to the probability of a fire. The emcigency diesel generator systems can perform their design basis safety function. No physical changes were made.

2) Does this modification create the possibility for an ccident or malfunction of a different type than any evaluated previously in the safety analyc!s report?

Answer:

No. These revisions do not change the function or operation of any component. The failure of the non-Q components included in this NCR does not prevent the diesel generator system from performing its safety function.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? F^

kMW.til No, The diesel generator safety design basis as described in Section 3.5 F and 3 9 the Technical Specification bases is maintained.

27

. ~ . . . - .- --. - . - - . . ~ . . - _ - . - - - . . . - - - . . - . - - - -

PEACH 00TTOM ATOMIC POWER STATlDN ,

UNITS 2 & 3 i i DOCKET No. 50-277 & 50 278 199010 CFR 50.59 REPORT Qigdenancies DstgrcfrL.1Epilt A and Dericn QtlL* jags Fire Protitq1onJntsm Fjagg poneewronuance Arront No: P@305 '

A. Annuj Fire Protection D. Descamnow; This NCR identifies discrepancies between the as built configuration of the Cardox Refridgeration Unit.

UFSAR Section 10,12 and associated documentation. Documentation will be revised to reflect the as bunt condition.

C. _ Reason Fon CHanorj This NCR was written specifically to intiate these documentation changes.

D. Sarrry Eynyanos Suuusay:

1) Does this modification increase the probability of occurrence or the consequencas of an accident or malfunction of equipment important to r.afety as previously evaluated in the safety analysis report?

Afl1W.1G No. The changea included in the subject NCR dio not affect equipment in the Fire Prctection System or interfacing systems in a inanner which would increase the probability of component failure The system is non safety related and the changes did not affect equipment important to  !

safety.

2) Does this modification create the possibility for an accident er malfunction of a different type than any evaluated previously in the safety analysis report?.

Answys.rj. 4 No, This NCR only changed the Fire Protection Program Figure B 2 Sheet 4 and the Instrument

> 4 Index and adds an Instrument Data Sheet for pressure switches PS 7593 and PS 7593A in the Fire' Protection System. .c

3) Does this modtfication reduce the margin of safety as defined in the basis for the Technical Specifications?

Answeg i

No. The subject NCR changed did not affect equipment in the Fire Protection System or interfacing systems in a manner that would create a different type of component malfunction. than previously evaluated in the SAR.

f 28

-t y r mr . ,cv r- y -----mse--,.,r*.--'.-,-- r. gy, , . ,.-eow-s w s...r--.s,. e +v.wwwr--n-.-++.w.e.,+~.re.m,,...%,1-r w e-,,.-r .-w.m-w.r,.--&-w w -t m mw- r se ey w e w .w s t e f + ea re-- + w . e

t PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3

. . DOCKET No. 50-277 & 50 278 1990 to CFR 50.59 REPORT pamaced Presnttttirnq.atQts figucowcauangrJeront No; P 90351 A.11111ul Ottgas B. Engstat,tqu.j The suttion and discharge pressure indicators Pt 8(9) 380 and 308 A and B for units 2 and 3 Ottgas System jet compressors have been damaged as a result of line pressure in excess of their capabillties. Damaged '

pressure indicators were replaced. To protect the new pressure indicators, the root valves for the B-Indicators will be shown normally closed. UFSAP. Figure 9.4.1 sheet 1 and 3 and associated documentation will be updated to reflect these changes to normal valve position,

~

C. hason Fon Csanot;

- This change is per the NCR disposition.

D. Sarny Evatyattow Suuusavt

1) Does tNs modification increase the probability of occurrence or the consequences of an accident or malfunction cf equipment important to safety as previously evaluated in the saf ety analytis report?

Alswert No Showing the root valves to Pls 0(9)389B and 8(9)3988 as normally closed does nct increase the consequences of an accident previously evaluated in the SAR. It will not change, degrade, or prevent the response of active /passivt systems described or assumed in UFSAR Chacter 14.0 of Appendix G. This activity has no affect on onsite/ottstte radiological effects previously approved in UFSAR Chapter 14.0 of Appendix G.

2) Does this modification create the pussit.lity for an accident or malfunction of a different type than any evalu2ted previously in the safety analysis report?

d AD1EeJl No. Isolation of Pts on the off gas loop not in service during plant operation will not cirectly or indirectly affect the performance of any equipment or systems its.

3) Does this modification reduce the margin of safety as defined in tne basis for the T?chnical Specifications?

ADswer:

No. There are no applicable Technical Specifications Sections that specifically address ths portion of the offgas system. *-

-29

l PEACH DOT 70M ATOMIC POWER STATIOf4 UfJITS 2 & 3

, _. DOCKET Ho. 50 277 & 50-278 199010 CFR 50.59 REPORT fliph,rtlQn Qffiserenancy Betgeen Ags And DnlurnentAtien JLoumgnuancs.Dsvont Ng1 P4C364 A. jl, tug u; Demineralizer B. pycnietion.j The NCR addresses an improper restorat;on of a temporary change to a demirieralized water supply connection. The demineralized water connection dov.nstream of check valve 3814170 was restered to a threaded cap instead of a hose connection UFSAR Figure 10.16.2 sheet 2 will be revised to show the threaded connection.

C. BLA ON Fon CHawon This NCR ls dispositioned to t,se as-is.

o..aucaum me - :

1) Does this modification increase the probability of occurrence or the consequences of ali ecciduti ,

or malf unction of equipment important to saf ety as previously evaluated in the safety unalys:s report?

Answer:

No. Installation of a threaded cap in lieu of a hose connection downstream of domineralized water system check valve 38D 14170 does not increa*e the probabihty of occurrence of an anticipated operational transient, an abnormal operational transient, a design basis accident or a "special event" .

previously evaluated in UFSAR Chapter 14.0 or Appendix G. ,

2) Does this modification create the possibility for an accident or ma! function of a different type than any evaluated previously in the safety analysis report?

Answer:

No. This actMty does not directly or indirectly degrade the performance of equipment important to safety as it is assumed to function in the accident analysis, to below the design basis.

3) Does this modification reduce the margirr of safety as defined in the basis for the Technical-Specifications /

Answen No. There are no applicabte Technical Specifications that address the makeup domineralizer system, l.

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1 30

PEACH DOTTOt.1 ATOf.ilC POWER ST ATIOf4 UfJITS 2 & 3 DOCKET No. 50-277 & 50 278 199010

• ~

CFR 50.69 REPORT SCIChiipn Of DmenticileMepn A@dLAMQ!a3in(1;.

tiowcouronwanet Ri pent Noa P 904C8 A. Entys; 125/250 Volt B. ,Duc enevowj This NCR does not affect plant safety relattd systems and equinmert its purpose is to update crawings to reflect previous deletion of non.Lafety loads from 125V DC safety-related distnbution panels. Trrte are no accidents or abnortnal occurrences listed in Chapter 14 of the UFSAR that are affected by this load's removal. UFSAR Figures 8.7.1 A. 8 7;A and ascociated documentation will be revised.

C. Erason Fon _Qt1NC{}

The NCR was wntten to initiate changes to the documents.

D. jannJvatuanow Supuapv;

1) Does this modification increase the probability of occurrence or the consequences of an au..Jent or malfunction of equipment important to safety as previoutly e /Liuated in the saf ety analysis report?

Answer; No. These changes are editoriai only. They remove non existent loads.

2) Does this modification create the possibihty for an accident or rnalfunction of a different type than any evaluated previously in the r,afety analysis report?

.$01Wtti No. No equipment was modified.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Arlm0!J 3-No. Technical specification 3 9 and 4.0 were reviewed to make this determination.

31

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PEACH DOTTOM ATOMIC POWER STATIO!J UfJITS 2 & 3 DOCKET No. 50-277 & $0 278 199010 CFR 50.59 REPORT

,Bgsdution Of Dit_creppneylffser' As-Uni!L.Artd_R.Qgymentatsit flpNcoN'opuanqtBreonT th; P 90444 A. Systru: Instrunient Nitrogen D. D.U.ctitiL "i -

As a result of the Valvo Labeling Project, various valves were found to be identified incorrectly. UFSAR ,

Figure 51.9 and associated documentation will be tevised to reflect as built conditions.

C.Busou Foa Canngsi This NCR is dipositioned to uco as is.

D. Sarriv Eyouanow Suf.mav:

1) Does this modification increase the probability of occurrence or the consequences of an accident ct malfunction of equipment important to safety as previously evaluated hnhe safety analysis report?

.A, rig,wtu No. This only corrects the reversal of tho *a* and "b' designators on the valves.

2) Does this modification create the possibility for an accident or malfuncuon of a different type than any 04aluated previously in the safety analysis report?

AD_fWfE No. This is an editoriat change only.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answen No. This does not affect the Technice' Spec'fications.

y 32

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PEACH BOTTOM ATOMIC POWER STATION

• - ' UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199010 CFR 50.59 REPORT Hesolution Of Diseracancy Between Asyuilt And Docum,gngligg Nosqsuromuanet Rr.tgatfigi P-90455 -

A. Sysnu: Radiation Monitoring 3, B, Dn; sevow!; ,

This NCR is written to request drawing changes on P&lD M 310 Sheet 2, to reflect a change from a 3/4" .

130 gate valve to a 1' 131 globe valve on the sine from the Unit 3 Offgas Recombiner and Hok'up to the Vial Sampler 20C105; this change was made as a result of ERR P 5619 for Mod 967. Also, the Unit 2 isometric M 1375 does not show a section of piping and a valve from the Ur'it 2 Ofigas Holdup to Vial Sampler 20C105, which is shown correctly on PalD M-3:0, Sheet 2 UFSAR 11.4.12 sheet will also be revised, to reflect the as-installed conditions.

C, Reawn Fon Csangrj This chance resolves discrepancies between drawings and as-installed conditions.

D, Sarm Evatuanow Symn,y; 4

1) Does this modification increase the probability.cf occurrence or the consequences of an accident or malfunction of equiptnent important to safety as previously evaluated in the safety analysis report?

Answer:

No. The drawing changes /correcticos, described in NCR P 90455, will make the drawings agree with existing plan

• conditions that were previously evaluated and approved under Mod 967,

2) Does this mod 4ication create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

, - Answer:

, Nc, The changes ri.ade by NCR P 90455 did not affect safety related equipment or equipment 7 important to safety, These changes do not directly or indirectly degrade the performance of a safety _

system, assumed to function in the accident analysis, below the design basis.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical.

Specifications?

Answer:

" , The margin of safety as defined in the basis of the Technical Specifications, was not reduced.

Technical Specificat)6n Sections 3.8C and 3.8E were reviewed.

1 1

9 33

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PEACH DOTTOM ATOMIC POWER STATION

. UNITS 2 and 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT Radiation Zong Clarification llenm1 Rate!!sLBPon? N3a P-90510 A. Jytnpi; Radiation B. pligenpyj

,, Addition of a clarifying note to figure 12.312 in the UFSAR to address radiation zones during refueling activities.

C.Byson Fon Quay _gst To clarify the appropriate radiologica. controls for this area during various refuel activities.

D. Sanry Evacuation Suuuany:

'1) Does this modification increase the probability of occurrence of the corcequences of an accident or malfunction of equipment Important to safety as pieviously evaluated in the safety analysis report?

3.01 ERG No. This adds a notation tot clarification only.

2) Does this modification croato the possibility for an accident or malfunction of a different type than P

- ' any evaluated previously in the estety analysis report?

O- SalERG No ' No physical changes have been mado.

3) Does this modification reduce the margin of sWN as defined in the bacis for the Technical Specifications? - .

a --.

AE1 ERG No. No physical changes have boon mado b

34

1 PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 50-278 199010 CFR 50.59 HEPORT

. Resolution Of Discrecancy Between As-Built and Documentation Newcouronuawer Rtpont Noj P 90512 A. _Systru: Circulating Water

9. Drsenianow:

-.On the circulating water discharge tunnel, there are two v3nts (standpipes), one inside and one cult.ide the

- Turbine Building. Documentation shows only one vent line located on this tunnel. UFSAR Figure 11.6.1 sheets 1 and 2 and associated documentation will be revised to reflect these conditions.

C. ,ErAsow Fon ('aN,gG

' To provide ccirect documentation for as.is plant conditions.

D. Sarrry Evavanow Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident ,

or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

- Answerj No. This is a drawing change only to reflect the as is plant condition. No physical changes are -

being made.

2) . Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

-No. This system is not safety related. No unanalyzed accident possibilities are being created 3)- Does this modification reduce the margin of safety as defined in the basis for-the Technical Specifications?

Answerj 7 No. No changes have been made. The Circulating Water System is not addressed in tt.e Technical Specifications. '

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PEACH UOTTOM ATOMIC POWER STATION .

• "' UNITS 2 & 3 DOCKET No. 50 277 & 50-278 100010 CFR 50.59 REPORT flesch> tion Of DJscrecangy Bhtweco.fijs And Ogumentallco Newconsonwancr_ Rupar Ng; P/J0514 A. Sysnu; Motor Control Center B. Desenietig31 Cable 0861311s shown on singte lino drawing E-1621 as three #4/0 conductors However, during a plant walkdown the actual Installation was found to be a 3 conductor #2 cable. Figure 8,4.78 of the UFSAR and associated documentatich will be revised to reflect the as. installed condition.

C. Brasow Foa Cw1=or:

This NCR is dispositioned to use as.ls.

D,larew Evatuation Suuuaav:

1) Does this modification increase the probability of occurrence or the consequences of an ace! dent or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

~

' Answer:

No, This is a drawing change orny to reflect the as.is plant condition. No physical changes are

being made. The cable is adequately sized to feed the 75kVA transformer 00X503.

2) Does this modification create the possibility for an accident or malfunction of a different typ9 than any evaluated previously in the safety analysis report? -

9 Answer:

No. This a drawing correction only.

3) Does this' modification reduce the margin of safety as defined in the basis for the Technical LSpecifications?

i Answer:

-7 No, This cable is not addressed in the Technical Specifications.

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L

PEACH BOTTOM ATOMIC POWER STATION *

,- UNITS 2 a 3 DOCKET No. 50-277 and 50-278 199010 CFR 50.59 REPORT Resolution Of Discrenqnev Between Aguilt and Documentation

,t(gucpufoauance Rracar No; P43515 A.ShsTru: Emergency Service Water (ESW)/ Reactor Building Closed Cooling Water (RSCCW)

B. Dycamnow1 This NCR involves the closure of the Emergency Service Water (ESW) system and Reactor Building Closed Cooling Water (RBCCW) system cross tie valves,33 520 A & B, on both Units 2 and 3. The ESW/RBCCW cross tie valves were locked closed in 1979 while concerns regarding the seismic design of the RSCCW heat exchangers and normal service water system piping were reviewed. The closure of these valves isolated ,

the non-seismic RBCCW system from the seismic ESW system te aoure the integrity of the ESW system pressure boundary during and following a design basis seismic eve Reportable Occurrence Report No.

2-7216/IP documented this change and check off list S.9.4.2.A was revised to reflect the valve closure.

There is no evidence that a 10CFR50.59 review was performed at that time nor was the UFSAR ever modified to reflect this change. Subsequent efforts (MOD 556) to qualify the RBCCW system as Seismic Class f were unsuccessful C, .Elusou Fon _CHwor:

The NCR was dispos:tioned to change the UFSAR to reflect the line-up discussed above and required a safety evaluation to be made, D. Suerv EvuvaneN Suuusav:

1)- Does this modification increase the probability of occurrence or ihe consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The accidents which form the plant design basis, including the loss of dam and flooding events, were reviewed to make this determination. Loss of Coolant Accident (LOCA), Loss of Offsite

- Power (LOOP), and LOCA in conjunction with LOOP were considered.

2) Does this modification create the pos$ibility for an accident or malfunction of a dif'erent type than

- any evaluated previously in the safety analysis report?

Answer;_

' No. The impact of the change on the following system and components was considered in making this evaluation: Reactor Water Clean-up. Reactor Recirculation, Drywell Chilled Water, Containment Isolation Valves, Control Red Drives, Fuel Pool Cooling, Air Compressors, Post Accident Sampling, Emergency Service Water, and Reac'ar Building Closed Cooling Water.

3)- Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. Sections 3.3, 3.5, 3.6. 3.7, 3.9, 3.10, 3.11, 6.9 c.nd 6.19 were reviewed to make this determination.

37

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PEACH BOTTOM ATOMIC POWER STATION

'L

• UNITS 2 & 3 DOCKET No. 50-277 & S0-278 199010 CFR 50.59 REPORT higlon Of Discrenancy Between As Built AnJ_Qgcum_entation JgwcowronuawceRuomrNpj P 90529 A.RyJnu; instrument Air B. Jh3 cniptioH_1 During a walkdown, drawing discrepancies were identified on valves and piping which provide distribution of air, These discrepancies included: differing line class representation, piping end connection, valve type or missing valve locked status. UFSAR Figure 10.17,1 and associated documentation will be revised to ,

reflect as installed conditions.

. C. RrasoH Fon Chance!

This NCR is dispositioned to use as-is.

D. Sarrty Evatuanow Suuuaav:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No.' This is a drawing change only to reflect the "as built

• plant condition. System function is not affected by ciiffering line class representation, piping end connection, valve type or missing valve locked status.

2) Does this modification create the possibility for an a'ccident or malfunction of a different type than i any evaluated previously in the safety analysis report?.

Answe't; No. This a drawing correction only.

3) Does this modification reduce the margin,of safety as defined in' the basis for the Technical Specifica!!ans? 7 s Answen No, The disposition of tNs NCR does not reduce the margin of safety as defined in the basis of any Technical Specification. Technical Specification Sections 3/4.7 were reviewed in making this-

' determination.

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a t

_ PEACH BOTTOM ATOMIC POWER STATIDti .

UNITS 2 & 3 DOCKET No. 50 277 & 50 278 199010 CFR 50.59 REPORT

_ R1 solution Of DiscrecancyEetween A,3;Dyilt And Opqymenta[!gg

,thwcowsoayawcr Rreoar No.: P 90587 A. Systrut - Condensate Filter Dernineralization System

.B. Descovow: ,

This NCR initiated changes to figure 11.7.1 of the UFSAR and other documentation to add speed control valves that were initially installed on all the air operated valves in the Condensate Filter Demineralizer System. These revisions reflect original plant condition.

C. Heason Poa Csawgn These drawing revisions are per the NCR disposition.

D. jaFETY Evatuarios Suuunan

1) - Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis repon?

Answert No. This is a drawing change only to retlect thc original plant design.

2) Does Inis modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

_No. No unanalyzed accident possibilities have been created.

3) ,Does this modification reduce the margin of safety as defined in the basis for the Technical

' Specifications?

Answen No. There are no Technical Specifications 3at are applicable 39

d PEACH DOrf0M ATOMIC POWER STATION

. . UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199010 CFR 50.59 REPDAT Resolution Of Qh.crenancy Between As.tnst a lled And Doc 1Lrnentation Nowcouronuawer Repeat No.: P-90588

. A. Systru; Raw Water B. DrscwTiow; This NCR initiates changes to UFSAR Figures 10.16.1 and 10.16.2 to account for the following as-installed conditions:

A recirculation line with a manual valve routed from the discharge of each Raw Water Sump Pump (OAP129 and OBP129) back into the surhp.

A trip of the Raw Water Sump Pumps on receipt of a high or low pH signal from existing pH transmitter (pHIT-0193).

C. Brason Fon CHawot:

This NCR is dispositioned to use as-is.

D. SartTv Evatuation Suuuaav:

1)- Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?-

Answer:

No. The Raw Water Sump Pump recirculation line and trip do not apply to any accidents evaluated in the UFSARi 2)~ Does this modification create the possibility for an accident or malfunction of a different type than

. any evaluated previously in the safety ana.ysis report?

Answer:

' No. This system is not safety re!ated. The design basis function of the Makeup Water System

. (which includes the Raw Water System) is unaffected by this activity.

3) Does this modification reduce the margin of safety as defined in the. basis for the Tecnnical-Specifications?

Answer:

No. There are no Technical Specifications applicable to this activity.

l 1

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1 -

PEACH DOTTOM ATOMIC POWER STATIOf4 UNITS 2 & 3 DOCKET No. 50-277 and 50 278 199010 CFR 50.59 REPORT Re12hJt.ipn Of Qb:c_remr2gyjgtween AMyllt And Qggimentatiqn HQHegggM A%gjfroMT No; P-90594 A.Snitu; Offgas Recombiner D. Jhicnietion_;

This NCR review documents the existence of two in-line 1/2' snubbers in the sensing lines to pressure controllers PC-8386A(B) In the Unit 2 Offgas RecomL'ner System, ano the absence of the snubbers in the same location in Unit 3. The present P&lD shows the sensing lines incorrectly and shows only 1 pressere controller on Unit 3. Snubbers will be added and FSAR Figure 9.41 will be revised to reflect the as-built condrtion.

C. Jie199.loe._QttAmon These changes are in accordance with the NCR disposMon.

D j aft TY (V_ALV ATiON $UMM ARY'

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the saf ety analysis repnrt?

.A_ngw_ct; No. The existing configuration of the air and sensing lines (including the snubbers) and pressure controllers has no connection to any accident scenarios previously evaluated in the SAR.

2) Does this modrfication create the poulbility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

AnsweJ1 No The design basis function of the Offgas Recombiner System is unaffected.

3) Does this modification reduce he margin of safety as defined in the basis for the Tecnnical Specifications? .

Answen No. Technical Specification Section 3.8 was reviewed to make this determination a

41

7,.

- {.

e PEACH DOTTOM ATOMIC POWER STATION

. .- UNITS 2 & 3 DOCKET No. 50 277 & 50-278 199010 CFR 50.55 REPORT Resolution Of Discrecan.syjgtwgg.n, As-Installed And Documentation Nowcowronuawer Remont No_; P-90610

' A. Sysnu: Alternate Rod Insertion (ARl)

B. Descaiaview: -

The NCR resolves an inconsistency between the 70 A fuse rating for circuits 29-2303 and 29-2406 shown on Drawing 8i 27.(UFSAR figure 8.7.2.a) and the field installed rating of 30 amps. The circuits affected provide DC power for ARI system solencids. Documentation will be revised.

C, Reasow Foa CuanagJ These changes are being made in accordance with the NCR disposition.

D. SurTv Evawariew Suuuany:

1) Does this modification increase the p obability of occurrence or the consequences of an accHent or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No; This is an editorial revision reflecting the correct fuse size as-installed. Based on the cable.

load size, and protection coordination, a 30A fuse was determined to be the correct size,

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

AnLw.2G No. The fuses are the correct size. No hardware changes have been mado.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: , , ,

No. No changes are being made to the Technical specifications. ~ Sections 3.9 and 4.9 were-reviewed.

l 42

+-

PEACH BOTTOM ATOMIC POWER STATION UNilS 2 & 3 DOCKET No. 50-277 & 50 278 199010 CFR 50.59 REPORT R,f,1glution Of Disgr3pangly.ctween As.13 And Other QggjmMatign NoncowropuAucr Rrpoar No; P-90638 A. 3.tsnuJ Service Water B. Orsemptiow: '

This NCR identifies inc- mong Operating Procedure SO 30.1.A.2(3) and other documentation' twJing Ur5AR figuh C. RcAsow Fon Chance; To revise documentation f-- nge normal valve position to 7HROTTLED OPEN' for vent valves HV-2 30-21727A,i,. WJ N1736 A, B. C and D, and addition of vent valves HV 2-30 217278 and HV 3 30-3u. s n3rmal position as " THROTTLED OPEN*.

D. S ArrTv EVALU ATioH

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis repon?

Answer:

No. This is an editorial revision to show the correct normal valvo position for these vent vaives and to add valves not previously shown.

2) Does this modification create the possibility for an accident or maltunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. This is in conformance with the design requirements of the system and is consistent with vendor requirements. .

3) Does this modification reduce the margin of safety as de"93d in the basis for the Technical Specifications?

Answer; No. The Service Water System is not addressed in the Technica! Specifications.

h 43

t

i PEACH BOTTOM ATOMIC POWER STAllOf4 Uf11TS 2 & 3 DOCKET No. 50-277 & 50-278 199010 CFR 50.59 REPORT

,Bgsolution Of Disqrrnancv Between As.BgMQgqqmJn,tggn RogqCN'oeuaNec Ryont Nni P 90644 A. Synut: Auxiliarf Steam B. QpenwrioN,,j This NCR identfies a discrepancy between the as installed condition of the plant auxiliary steam system, UFSAR Figure 10.23.1 and related documents. A field walkdown identified that the output of LIS4715 initiates an alarm only, not an alarm and pump motor trip, as currently shown on UFSAR Figure 10.23.1 sheet 2. The UFSAR Figure will be changed to reflect the as installed condition.

?

C. Reasow Fon Csawar:

The NCR was dispositioned use as is.

D,14srty Evatuation Suuuanyt

1) Does this modification increase the probabil!ty uf occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. Changes to the documents do not have any impact on the functions or operation of the plant auxdiary steam system.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. The changes in the documents, which reflect the as-built conditions of the plant, do not create any new accident initiators nor affect any existing accident initiators such that a different type of accident than previously evaluated would result.

3) . Does this modification reduce the margirLof safety as defined in the basis for the Technical Spe+fications?

Answer:

No. These changes do not reduce the margin of safety as defined in the bases of any Technical Specification. The changes to the Auxiliary Steam System addressed here do not im;act the Technical Specrfication bases of any interfacing systems. This was confirmed by review of all Technical Specifications.

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4 PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 and 50 278 1 199010 CFR 50.59 REPORT i

- Lack _ of Adeowlta.Dass 1E isolation

_Nowcouronuaner ReconT No; P-90063 A. Systru: 125/250 VDC System D, DrsenipTiow:

This NCR identifies the lack of adequate class 1E isolation between Class 1E and non-1E electrical circuits that are connected on Unit 2 and 3.

C. Reason Fon Cuawon The NCR is dispositioned to rnodify the equipment to return it to the condition described in the UFS AR. _ A safety evaluation was made which justifies continued use of the as-built configuration until tne modifications can be completed.

D. SarrTv Evatuariow Suuuany:

1)- Does this modification increase the probability of occurrence or the consequences of an accident or malf unction of eqdpment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The potential failure modes could only affect the 125/250 Vdc System. These potential failure modes havc been determined not to affect operation of the de System and are bounded by existing analyses.

2)- Does this modification create the possibility for an accident or malfunction of a different type then any evaluated previously in the safety analysis report?

- Answer:

No. The possibility of an accident of a different type than any previously evaluated in the UFSAR is not created, if the cables were to fail, they could only potentially aff:ct operation of the dc -

- system. . This type failure would not be an initiator of a different type of accident.

. 3) Dces this modification reduce the margir[Af safety' as defined in the basis for the Technical

- Specifications?

Answer:

No. Technical Specification Section 3.9 was reviewed in making this decision.

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.. . _ _ . _ ~ _ . - _ _. _ . _ . _ _ _ _ _ _ . _ . _ _ _ _ . _ . . . _ ._

PEACH DOTTOM ATOMIC POWER STATION ,

UNITS 2 & 3 DOCKET No. 50-277 and 50 278 199010 CFR 50.59 REPORT Discrengncy.fptween instded Plan (fgulpment And_ Dgqurnentgip3 floigoponuancf_BLP,oni No.: P M 672 A. _Systru: Fire Protection B. DJscairnowl The position of three solenold valves on the Fire Protection System (CO, system ) P&lD 6200 M 318, Rev.-

44, needs to be changed because they are shown in the operating condition Instead of the "shell condition?

P&lD M-300, Rev. 26. Note 1 states "all solenoid valves are shown in their shelf condition.' This actMty veill make changes to the SAR Fire Protection Program Figure B 2, sheet 4 of 4.

- C, Bfta_ tow Fon CManofj This NCR is dispositioned to use as-is.

D.- Surry Evatvavon Suuuany:

1) Does this' modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The "as built

• configuration of these solenoid valves is correct The document revis!cns are intended to eliminate any potential confusion.

' 2) Does this modification create the possibility for an accident or malfunction of a different type than -

any evaluated previously in the safety analysis report?

Answgn

- No. There are no physical changes being made, Revision to documentation is to show correct -

~'

position only.

3) Does this modificatloa reduce the margin of safety as defined in the basis for the Technical Specifications? - 7 Answm

-No. Technical Specification Section 3.148 reflects the correct position of the valves.

t 46

- . . . . . - . _ . - , _ . _ _ _ , , _ _ - -,- _.-. _~ _ . . _ . _ . . , _ , _ . . . .

PEACH DOTTOM ATOMIC POWER STATION

. . UNITS 2 & 3 DOCKET No. 50-277 & 50 278 199010 CFR 50.59 REPORT R%@tlon Of DistrSpMqv Between As-Is And DogyrnentJtjng Nowcowronuawct Arpont No; P 90087 A. Systruj Containment Atmospheric Dilution B, Drseniption:

The purpose of this NCR is to add four diesel generator daytank overflow vent valves to UFSAR figure 8.51 and associated documentation. These valves are shown on isometric drawings and appropriate system operation procedures and checkoff lists, but are not shown on the UFSAR figure or the associated P&lD and QAD.

C. Etasow Fon Cwawar: .

This NCR is dispositioned to revise the appropriate drawings.

D. Sarrry Evatuarios Suuusan

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?  ;

Answer:

No. This is an editorial change. It does not make a change, degrade, or prevent the respcnses of active or passive safety systems.

2) Does this modification create the possibility for an accident or malfunction of a ditterent type than sny evaluated previously in the safety analysis report?

Answer:

No. The valves meet the system requirements and are non safety related.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical "

Specifications?

Answer: T No. These valves are not addressed in the Technical Specifications.

47

PEACH DOTTOM ATOMIC POWER STAT!ON

, , UNITS 2 & 3 DOCKET No. 50-277 & 50-278 109010 CFR Sn.59 REPORT E210Migntp1DigtepiLnty_DatweenliK An@_ggngniggg,9 Newcouronuanct Rtreat No_; P-90700 A. Systru; Reactor Water Cleanup D. Desemanout This NCR ldentifies inconsistencies between Drawings M 354, sht.1, and M-354, sht. 2, which apply to the reactor water cleanup (RWCU) system for Units 2 and 3. Drawing M-354, sht.1, shows tne RWCU Pump Casing Vent Valves (2-12 21400A(B, C)) as normally open and Drawing M-354, sht. 2. Shows the corresponding Unit 3 Valves (3-12 31400A(B, C)) as normally closed. In adGition, the NCR also evaluates the normal position of tne RWCU Pump Casing Drain Valves (212-21401 A(B. C) and 312 31401 A(B CD The NCR also addresses the mislabeling of RWCU Pump Casing Vent Valve 312 31400A whicn is shov,n on Drawing M-354, sht. 2, as 3-35-31400A. The drawings will be corrected. The corresponding sheets cf UFSAR Figure 4.9.1 will also be revised C. _Rrasow Fon Csawort This NCR ls dispositioned to use as is.

D. SutTv Evatuattew Suuuanet

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction c f equipment important to safety as previously evaluated in the salety analysis repor:?

Answert No. NCR P 90700 revises the valve position of the RWCU Pump Casing Vent and Drain valves from open to closed and corrects a valve identification number. These changes are editorial and mace to reflect the normal configuration of the RWCU System. These changes do not affect the function or operation of the RWCU of interfacing systems.

2) - Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answert No. The change in valve osition from open to closed and the correction to the valve identicaticn number are in agreemua alth existing plant procedures and system operation

3) Does this modification reduce the margin of safety as defined in the basis for the Tec",nical Specifications?

.Artsmg No. A review of the Technical Speedication (Sections 3.6 and 3 8) indicates there is no section cf this document applicable specifically to the RWCU system.

48

l

C PEACH BOTTOM ATOMIC POWER STATION l UNITS 2 & 3 DOCKET No. 50-277 & 50 278 199010 CFR 50.59 REPORT

' Resolution Of Discrenancy Between As-Is and Documentation Nodcouronpawer Rrpont No; P 90743 -!

A SisTrut Radiation Monitoring B' Erseniption:

This NCR identifies the fact that, during normal station operation, two dilution air fans should be in operation even though UFSAR Section 9.4.4.2 states that only one of the three fans will provide the required dilution air flow for two-und operation.

C.' Reason Foa Chance:

- This NCR is .fispositioned to use as-is.

D, SarrTv Evatuation Suuurnv:

- 1) Does this modification increase the probabsrty of occurrence or the consequences of an acciden;

- or malfunction of equipment important te safety as previously evaluated in the safety analysis report?

Answer: +

No. A review of the UFSAR Plant Safety Analysis (Section 14) indicates that the dilution air fans are not involved in any existing postulated accident snenarlos.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

- No. The actrvity described does not involve any equipment important to safety previously evaluated

- in the SAR. -

3) - Does-this modification reduce the margin of safety as defined in' the basis for the Technical

- Specifications? '

.. y Answer:

No. Technical Specification Section 3.8.C for both units requires a minimum of 10,000 cfm cilution air flow. This specification will be satisfied by operating two dilution air fans in accordance with Station Operating Procedure SO 8.7.A, Rev. O.

49

.)

'?

PEACH 00TTOM ATOMIC POWER STATION o.

UNITS 2 & 3 DOCKET No. 50 277 & 50-278 199010 CFR 50.59 REPORT ECiQldlon Of Discrppancyj@een As.ls an1Qgtympngigg Spyggyloauaner Reoat No; P 90759 A. Systrul Control Rod Drive Hydraulics (CRDH)

8. ,,Q[sCRIPTIojj This NCR identifies discrepancles between the as built configuration of the plant and the associated drawings. This change involves those items dispositioned use as-is. FSAR Figures 3A.7. 3A.8 and 1.3.1 will be revised to add water header vent valves 2 3 21597 and 3 3 31697, and to change the designation of various valves from " globe

• to

• manifold block."

C. Egason Fon CHaworj .

This NCR is dispositioned to use as is.

D. Jamy Evatuation huuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The addition of two vent valves and the changing of other miscellaneous vent valves from

" globe

• to " manifold block' valves on the P&lD reflects the plant as-built condition. The function of these components is to provide a manual venting capability and manifold block valves adequately perform this function.

2) Does this modification create the possibility for un accident or malfunction of a different type than any evaluated previously in the safety analysis report?

. Answer: ,

No. These changes do not affect the function or operation of the CROH system or interfacing

- systems in a manner that would create the possibility of a accident or malfunction of equipment important to safety .not previously evaluated,.

- 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? .

' Answer:

No. . The addition of two vent va!ves (Dragon manifold block valves) and the changing of other miscellaneous vent valves from " globe

• to

• manifold block" valves on drawings and documentation reflects the plant as built condition _

j 50

. . - - . _ - ~ - - . . - . - . - . - . . . _

L

,_ , PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199010 CFR 50,59 REPORT Rework Irt Samole Une Succorts ancj, Associated Document Revisions Noncouronuance Repoar No: P 90779 A. Systru: Diesel Gerierator i

B. Descamriow:

The NCR addresses a number of discrepancies related to sample line supports on the standby Diesel

' Generator, Some of these discrepancies were dispositioned use as is, These require revision of tJFSAR Figure 8.5.1i and involve:

a. The removal of sample connection globe valves 52E-10043A,B,C.D.

b. - The correction of the sample connection identification to SX-7711 A,B,C D.

c. The repositioning of the sampie connection including RTV-52E-7711 A,P.C.D near the three way valve TCV-7239A.B.C,0.

C. Reascu Fon Chancy;

- These revisions are per the NCR disposition.

D. SartTv Evawarios Suuuaav:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

- No. =These drawing revisions reflect the as installed configuration.

2)- Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

y No,. These are drawing changes, and as such they do not affect the function or operability of the system.

3)- Does this rnodification reduce the margin of safety as defined in the basis for the Technical Specifications?

AnitG No - These are drawing changes and as such do not introduce any new accident initiatcrs.

e i

i 51 i

, , - - - , , , - - - , . - - - . - - , ,n. , -

PEACH DOTTOM ATOMIC POWER STATION UNITS 2 and 3 DOCKET fJo. 50-278 199010 CFR 50.59 REPORT Akilly.RLCort6nray P_utne1Ip P rform f Their B!n. sling 14oweowrpauggtJJnoatf{pj P-90786 A. _Systru; Core Spray D. _Dngnmuow:

During the monthly surveillance test, calculated motor bearing all temperature for the 2B Core Spray P was found to be higher than design as a result of degraded performance cf the Emergency Service Water System (ESW). The NCR and associated Safety Evaluallon justify continued use of the 2B Core Sprag Pump, and all remaining Core Spray Pumps, at river water temperatures below 60 degrees fahrenheit.

C. Iltasow Fon CHANOf.}

This is considered a change to the plant as described in the UFSAR because the analysis uses lower-than- '

design ESW Inlet temperatures.

D. jiurry EvawanoN Suuuaav:

1)

. Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the saf ety analysis report?

AntwG No.

The probability of occurrence of an accident previously evaluated in the SAR will not be increased because Core Spray is used li mitigating the consequences of an accident and is not involved in any accident initiation sequence. The consequences of an accident previously evaluated in the SAR will not be increased because operability of the Core Spray Pump will be maintained at the reduced river water temperature of 60 degrees fahrenheit. This lower than design coolant temperature assures that design heat transfer is maintained.

2)

Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? -

A!1PLE No. The existing condition of the motor 6f5ooler will not create the possibility of an accident of a different typo than any previously evaluated in the SAR, since there will be adequate cooling to the pump motor bearing oil during the period covered by the analysis for this NCR. Therefore. an accident initiator different than those described in the SAR is not created by this condition. UFSAR Chapter 14 has been reviewed in making this determination.

3)

Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

A9am No. Technical specifications 3.5A and 312 were reviewed to make this decision.

52

J; ';

4 i PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50 278 199010 CFR 50.59 REPORT Discrecancy Between "As-installed And Documentation Nowegwronuance Remont No.: P 90795 A. Systru: Standby Liquid Control (SLC) w B. D_tsemiption:

The 3* diameter Standby Lic,uld Control fanks 20(30) T18 outlets have 1" locked closed and capped intet valves for demineralized water connections, Valves HV 2(3)11-024 are not shown as locked closed and capped on P&lD 6280-M-358 or UFSAR Figure 3.8.1 sheets 1 and 2. The documents w;ll be revised to reflect the as-Intailed condition.

C. Reason Fon CHANGE:

This NCR is dispositioned to use as-is.

D. JimTv Evaluation Suuuany:

1)1 - Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis repon?

_ Answer:

No.- This is an editorial change only to revise documentation to reflect the "as installed" ccndition.

and does not change the ability of the SLC system to perform its intended function.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. This activity maintains the design basis function of this portion of the SLC system.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? -

. ern, Answer:

-No. Technical specifications 3A and 4.4 were reviewed to make this decision.

53

. - ~. . - . . - - . _- -- - - -

l PEACH BOTTOM ATOMIC POWER STATION UNITS 2 and 3 DOCKET No.50-27B 1990 to CFR 50.59 REPORT Resolutlen Of DisIreoancy Between As ineLalled And Docyrggntation Nowgowromuanet Rrmont No; P 90796 A. Systru: Standby Liquid Control (SLC)

B._Quenietiow:

The 3' diameter Standby Liquid Control Header to Pumps 2(3)A(B)P40 has 1" locked closed and capped drain valves HV 2(3)11-025. These valves are not shown as locked closed and capped on P&lD 6200 M-358 or UFSAR Figure 3.8.1 sheets 1 and 2. The Figures will be revised to reflect the as installed condition. ,

C. _ Reason Foa Cuawor:

This NCR is dispositioned to use as-is. ,

D. Sarrry Evatuatiow Suuuany:

'1) Does this modification increase the probability of occurrence or the consequences of an accide'11 or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

- Answer:

No. This is an ed'rtorial change only to revise documentation to reflect the "ac installed

• ccccition, and does not change the ability of the SLC system to perform its intended function.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. This activity maintains the design basis function of this portion of the SLC system.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?-

Answer: ~7-No. Technical specifications 3.4 and 4.4 were reviewed to make this decision.

54

I PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT E01Qhaion_Qipiscrggylttpphyggn "AsinstalPd Andpoqijm_entitjon t lloscowronuaner Rreont No.; P 90797 A. Sysuu; Standby liquid Control (SLC)

B. Dcsemenowj The 11/2' diameter Standby Liquid Control Pumps outlet piping has a dual isolation locked c!csed and capped drain connection assembly. Outer valves HV-2(3)11-033 are not shown as locked c!csed and capped on P&lO 6280-M 358 or UFSAR figure 3 81 sheets 1 and 2. The documents will be revised to reflect the as-installed condition.

C. Rf A50H Fon CHanorj This NCR is dispositioned to uso as is.

D. Sarrry Ewuanou Suuuan;

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated h the safety analysis report?

Answert No. This is an editorial change only to revise documentation to reflect the 'as-installed" ccr'dition, and does not change the ability of the SLC system to perform its intended functiert

2) Does this modification create the possibility for an accident or malfunction of a different type than any avaluated previously in the safety analysis report?

Answer:

No. This activity maintains the design basis function of this portion of the SLC system.

3) Does this modification reduce the margin of safety as defined in the basis for the Tecnnini Specifications?

Answer:

No Technical specifications 3.4 and 4 4 were reviewed to make this decision 55

. -- - ~. .~ .. -

i N *- PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199010 CFR 50.59 REPORT Resniution Of Discrepancv Between As-is And Documentation floheewronuawee Rcront Not P-91007 A. SysTru.1- Cordensate B. 9esenien9y; Condensate pump seal water drain valves HV-2(3) 5 2(3)6187ABC are now kept normally closed instead of open or throttled. This limits seal water leakoff and prevents unnecessary inputs to the radwaste system.

UFSAR Figure 11.8.1 sheets 1,2,4,5 and associated documentation have been c.nanged to reflect this condition.

C. Ruson Fon CHawac This NCR is dispositioned to use as is.

D. larrry Evawanow Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety a3 previously evaluated in the saf ey analysis repon?

Answer:

No. This NCR was initiated to revise drawings and documentation only. The valves have no affect on the ability of the Condensate System, or any p' ant systems to perform their intended functions,

2) Does this modification create the possibility for an accident or malfunction of a different type than -

any evaluated previously in the safety analysis report? -

Answer:

- No. This is a documentation change only.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

_ Answer:

No, There are no applicable sections of the Technical Specifications that specifically adcress this portion of the Condensate System.

56

_.,----v--- - - . _ ___-.

I 9

]Jnit 2

. t 57

. m._ . __ . . . _ _ _ _ __ ___ __ _ _ . _ _

• PEACH DOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 199010 CFR 50.59 REPORT High Pressure Coolant Injection (HPCI) Alternative Control Station (ACS) hheincanon No.: 1352A.

A. Svsnu: HPCI O. M$31PfloNJ This modification reroutes certain safety-related circuits to an ACS, where a safety related transfer / isolation

. switch will transfer the control location of safety-related equipment from its normal control panel to an ACS and isolate safety related system circuits that could adversely affect safe shutdown in the case of an Appendix R Fire. The controls and indications at the ACS are arranged to allow remote manual startup, operation and shutdown of HPCI. Automatic operations of the HPCI system, including Primary Containment isolation System functions are not required for attemative shutdown, and cons squently are nct re-established

- at the HPCI ACS.

- C. Reason Fon' Ceanou This modification was necessary to meet 10CFR50 Appendix R requirements.

D. Surry Evatuanon Suuuaav:

'1)- Does this modification increase the probability of occurrence or the consequences of FD 2C0! dent

. or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

,A, nswer:

No. In the event of an Appendix R fire, and the use of the ACS, the time available for manus!

. operation of HPCl is sufficient to keep the core covered.

2) 'Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. The possibility of an accident or malfunction of a different type than evaluated previously in the safety analysis report is nct created. This modification does not change the operation of HFCI and RCIC systems as described in the UFSAR when all of the transfer / isolation switches are in the

'normar or " test

• position. Test and emergency switch positions are annunciated in the Main Control

, Room to alert the operator of these abnormal conditions so that the system can ce restored to normal if there is no fire._

- 3) : Does this modification reduce the margin of safety as defined in the basis' for the Tec"q! cal Specifications?

Answer:

= No. The safety function of the HPCI and RCIC systems are not affected by the rerouting cf circuits and the addition of a safety-related transfer / isolation switches, and the automatic initiation and trip features of the system are not affected when the transfer switches are in the "normar or " test' positions. This modification does not reduce safety or raise an unreviewed safety question.

The use of this HPCI ACS panel (i e. When the transfer / isolation switches are in the alternative mode) to re+ pond to an Appendix R. fire is required to ensure safo shutdown. The Appendix H fire wil!

create a need for temporary departure from the Technical Specifications in order to mitigate the consequences of the fire.10CFR50.54 allows departure from the Technical Specifications in an emergency such as an Appendix R fire.

58

PEACH DOTTOM ATOMIC POWER STATION

• • . *= UNIT 2 DOCKET No. 50 277 199010 CFR 50.59 REPORT

'e

- p_ryw_gil Bulk _A,grqoe Temogratgr1 1nq!ig.ation

- JAopncaTio9 No.: 5199 <

A. SyJJuu; Drywell/ Control Room

, D.' Descamnout This modification adds a drywell bulk average tempirature indication in the control room.

C. RcAseu Fon CHawor:

This vcill allow automatic calculation of the average drywell bulk temperature, display tha temperature in the control room, and provido analog signals to the Process Monitoring System (PMS) computer.

D. Rutry Evatuanow

SUMMARY

1) -Does this modificatien increase the probability of occurrence or the consequences of ea ac;ided or malfunction of equipment important to safety as previously evaluated in the sticty analysis tsport?

. Answer:

No. This modification will enhance the operator response under accident conditions since the drywell temperature will be automatically computed by the new indicating unit.

. 2) Does this modification create the poscibility for an accident or malfunction af a different type than any evaluated previously in the safety analysis report?

Answer; No The new indicating unit added bij this modification has no control function, and coes nct offect

- the operation of any safety system.

3)  ; Does this modification reduce the margin of safety as defined in tho' basis fcr the Technica!

-Specifications?

. Answer:

.. No. Technical Specification Section 3.2 was reviewed to make this determination.

59-

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 199010 CFR 50.59 REPORT

,Fesolution of DiscrenJ!ncyJetween As-ts and Dqqumentation f[onncATiow No.: 5224 A.Eultul Conderuate D. DESCRIPTION:

This modification replaces the obsolete Leeds and Northrup model 1911 flow transmitter FT-2110 with a new Leeds and Northrup modei 2610 transmitter. The power feed to the new transmitter will be changed from 24 VDC to 120 VAC and the input resistor to the square root converter will be changed to accommodate the new transmitter current output.

C. Rrason Fon CHANGE]

The existing transmitter failed.

D. _Samy EvatuaTion Suuuans

1) - Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to sfety as previously evaluated in the safety analysis report?

. Answer:

No. The design function was not changed.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? .

Answgn No. Since the operation and separation of plant safety-related systems will not be affected by this rnodification, the possibility for an accident or malfunction of a different type than any evaluated previously in the SAR is not created.

3) Does this modificailon reduce the margin of safety as defined in the basis for the Technical Specifications?

-v.

Answer: * - '

No. There are no Technical Specifications applicable to this modification.

60

p_..-._._-._._..-__.___ _ . . _ _ _ . _ . _ _ . . _ -- - _ _ _ . - _ . _ _ _ _ . _ _ _

i PEACH DOTTOM ATOMIC POWER stall 0N

• . UldlT 2 i DOCKET No. 60 277  ;

199010 CFR $0.59 REPORT i i

I ulumoeted Out Point On Temocrature RtterAg;  ;

1 Temoorarv Planr 611trJil!0.rj 02 13 '

A. Sysnu; Recirculation i D. ,Qtsentiow;  !

This change temprarily jumpered out po. t 15 on the temperature recorder for the 'B' recirculat:on pump motor upper guide bearing.

3<

. C. ,Brasow Fon Cuawor;  ;

An open thermocouple was causing erratic readings.

D. ;Sarm Evatuapow Suuusay: ,

1)= Does this modification increase the probability of occurrence or the consequences of an accirient or maitunction of equipment important to safety as previously eve'i ated in the safety analy!,is rcport? i A!1t M :  ;

No. High bearing temperature on recirculation pumps is nvt an initiator or <:entributing factor to any accident evaluated in the Safety Analysis Report.

[

Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

4 At.1&fli i No. The worst case of failure would be a selzure of the recirculation pump which is analyttd in SAR sectiort 14.5.5.4. {

3) Does this modification reduce the margin of safety as defined in the basis for the Technical  ;

Specifications? l Answer: v" No. Technical Specifications do not address the motor upper guide bearing' temperature.

1 f .

. t' s

0 b

61 )

v- ,-w ,,---er., v,..<n--,-~,nn+ .. ,-- ~nn.r--,,r-an.,;.-,,>n,n,.,,,,,,,w-----n,-,--a,.--n,-.,,ww--- n,+,r. -..,.r-, --m-*

6 PEACH BOTTOM ATOMIC POWER STATION

, , UNIT 2 DOCKET No. so277 199010 CFR 60.59 REPORT Temportyy_ljdron 61 The StangylCfAS ACLQ Unit Cr@r Auecasay Stanen Rtreat No; TS 01 A, ,$3nul - Emergency Service Water B. _Drsemiscui This charige involves a realignment of the ESW system to isolate the standby ECCS unit cooler in each of the Unit 2 ECCS and the RCIC pump compartments, and the positioning of the control switch for these unit coolers to the off position, it also n!!ows the unit cooler air operated valves (AOVs) to be *lai!ed open to enhance coder reliability, C. Reason Fon Csawan Corrosion of the Emergency Service Water (ESW) System's carbon steel pip!ng has led to fouling of components and a reduction in flow capacity. The corrosion of system piping has required the cleaning of equipment, the replacement cf affected ESW piping i Unit 3, and a schmjuled replacement of the affected

. piping during the next refueling outage on Unit 2. The temporary realignment cf the system will make ESW operable on Unit 2 until the sene:!uled pipe replacernent. .

D .SusTv Evatuatigy.Juuuaayt Does this mod'rfication increase the probabihty of occurrenc9 or the consequences of an accident 1) or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

AD1E!EG No. The isolation of the standb/ unit coolers will not impact the safety function of the emergency ventilation equipment. 'lhls change will ensure that sufficient cooling water flow is provided to the unit coolers consistent with maintaining pump- compartment ambient temperature within the Environmental Qual!fication profiles. Flow to the RHR seal coolers and Core Spray motor til cocers will not be degraded by the proposed alignment.

• Falling open* the primary unit cooter AOVs will.

climinate the possibility of the valve falling to actuate. _

2)

Does thl; modification create the possibility for an accident or malfunction of a different tyos than any evaluated previously in the safety analysis report?

e Answer; No. This change in operating configuration of tte emergency ventilation unit coolers maintains the operability of the Emergency Core Cooling and Reactor Core Isolation CooHng system;

~

3)

Does the modrfication reduce the margin of itfety as defined in the basis for the Technical Specificatione?

Answen No - Tne margin of safety as defined in the basts of Technical Specifications 3.5A 3.5 B,3.5.C.

15.F.3.S.H,3.9.A,3.9.C and 6,16 has not been reduced, since only one cooler as defined in Section 3.5 H is requirei 62 l

._-j 1

PEACH DOTTDM ATOMIC POWER STATIDf1 U NIT 2 DOCKET No. 50 277 199010 CFR 50.59 REPORT Pesolutbn of OlstregantyJetween AsJullt anJ_ Documentation NoncourommNet Rteeat No; P 89997 A. Sy5itu: .125/250 VDC Station Batteries D. Mj1ca1Pflo4: .

This NCR Identifies discrepaneles between the as built configuration. UFSAR Figure 8.7.1 A. and associated documentation. The 2$0 VDC MCC 20008 has 12 breaker cubicles. However Drawing E 26. sheet 1, orly shows 11. The breaker cubicle not shown on the print is a spare with a 30 Amp disconnect switch and 12

- Amp fuses. This chavige will update the drawings to ?cflect the as. built condition, and install the correct nameplate for the spare breaker.

C. Reason Fea Cwauot:

This NCR was dispositioned to use as4s-D. Satm Evatuatio4 Suuusay:

.1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. - Correcting the drawings to reflect the as built condition will not increase the consequences cf an accident previously evaluated in the safety analysis report.

2) Does this mocification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analys!s report?

. Answer:

No. This nonconformance report does not change any of the plant equipment function or uperation -

important to safaty. It reflects the as-built condition.

)

3) _ Does this inodification reduce the margin,of safety as defined in the basis for the Technical Specifications?

Answer.;

No. Technical Specifications do not lis'. individual breakers in an MCC. Section 3.5 of the Technical Specification was reviewed to make the above determination.

64

PEACH DOTTOM ATOMIC POWER STATION UNIT 2 i DOCKET No,50 277 l 199010 CFR 50.59 REPORT I

f.cs2MkLrLQf E!Eclgranev Betwqef.dthdt_.aM_D,ogun_watation

.tJ_gagg.n.tqauwtg tR eeat No: P 90093 I

A. Smm Emergency ventilation D. ,QMCatPTIOM

• NCR P 90093 identflies inconsistencies between design documentation for the Emergency Ventilation '

System, the Project 0 Ust. Prclect Quality Assurance Diagrams, System Diagrams and UFSAR Figure 5.3 t.

Two space heaters in the Diesel Generator Building are shown as safety related. They are not safety-telated.

C. Buton von Cuaner:

The NCR disposition specifies that documentation be revised to reflect the downgrading of the two space heaters located in the Diesel Generator Building from safety related to non safety related, D. SmTv EvuuAtmg Sy.wuAq

1) Does this r70dification '.ncrease the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as prev'ausly evaluated in the safety analysis repon?

Answer:

No, Removal of the space heaters fr3m the O.Ust and the redesignation to a non safety related qualification does not compromise the safety of the plant. The enange reflecting the as built condition of the plant has no impact on accident analyses in UFSAR Section 14. ,

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Id12WEl No. The as-built documentation does not impact the design or operation of any safety related equipment.

3)' Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? -i Answert No. The Diesel Generator Building space heaters are not addressed in the Technical Spec 1 cations.

Sections 3 9. 4.9A. 3.5F. and 4.5F were reviewed to make this determination.

P 65

- . .-, , - , + , . - - _ . . , _ . , - . , . . - , , . ~ , . . - . - . - , . - - - - _ - , - . . . - - . . . - . _ . - - , - - - -- - --

L i

PEACH BOTTOf.i ATOMIC POWER STATION UNIT 2 l DOCKET No. 50 277 199010 CFR 50.59 REPORT <

Egsolut!cn of O'screcancyJetween As Bjl!!t An@gqumg0. tit 020 JLqttq,quronuanet Rtront Ngj P 90109 [

- A. SJtutul instrumentation D. peseniptiow:

l NCR P90109 was written to addtcss a discrepancy between UFSAR Figure 0.4.1. associated documencition, and the physical plant condition. Valve 843047A is shown open, however it is closed in the field. Thb is '

a vent valve on the steam supply line to the 'A' jet compressor 2AK35. The drawing will be revised to show the valve position closed.

C. HrAsow Fon CpAN0rt Thls resisjon is per the disposition of the NCR.

D. SartTv EvauLATION Suuuanyt 3

1) Does this modification increase the probabihty of occurrence or the consequences of an ace! dent or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

$nSwert No. The only change is a correction to the drawings.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? l

- Answer:

No. The change in position of Vent Valve 843047A does not affect safety related equipment or equipment important to safety nor will it cause an accident inillator not considered in the SAR. >

Answer:

3) Does this modification reduce the marginf ol safety as defined in the basis for the Technical Specifications?  ;

t No. Technical Specifications 3.6,3.8C and 6.18 were reviewed.

I l

60

. PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277  ;

199010 CFR 50.59 REPOHT Eciglytlon Of DAc.rppancy Getween As.ls and Documentatipa Houcoureau_aws.!LBsecar Nej P 90250 '

i A. Systrut Residual Heat Removal (PHR)  !

B. Rusano_Hj The deminerallad water isolation valve 38D-47073C to RHR heat exchanger 2N0?4 rXJd 1:e M con as locked closed on UFSAR figure 10.7.1 sht 1 of 4 and other associated docu'aentation. Currently inis valve i is shown only as closed, Documentation wn revtsad te t&pw this cMn;;e.

C. Egason Fon CHawogJ i

The NCR is dispositioned to use a vt.

D. Sarm Evawanow Souuany:

1) Does 'his modifbation increase the probability of occurrence of the consequences of an accident or malfunct ion of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. Locking closed this valve maintains its safety related function and provides positive isolation of the domineralized water system from the High Pressure Service Water (HPSW) system. Locking ,

closed valve 380 47073C is consistent with the condition of the identical valves for the remaining three Unit 2 RHR heat exchangers and for all four Unit 3 RHR heat exchangers. .

2) Does this modification create the possibility for an accident or malfunction of a different type than ,

any evaluated previously in the safety analysis report?

Answer; No. This actMty does not directly or indirectly degrade the performance of tt;c HPSW or RHR Systems, or etner equipment important to safety, as it is assumed to function.

3) Does this modtfication reduce the margin of safety as defined la the basis for the Technical i Specifications?  ? '

, A0swer:

No. Technical Specifications 3 5 and 6.8 were reviewed to make this determination-4 67 s-

, . _ . , . _ ..-. , , . , . . . - , , - _ _ _ . - . . - . . - - - . _ _ . , - . _ , - _ _ , . . - , - . . , _ . , , . . - , . . - . - - - . , , _ . , - . _ _ _ _ , , , , - . ~ _ . _ . . - - -

PEACH DOTTOM ATOMIC POWER STATION UNIT ?

DOCKET No. 50-277 199010 CFR 50.59 REPORT EeduticafL0nneratc1 Bet ^een Asjug And QpgumentWm JJoncontgouanet R(po=T L4oj P-90353 A. SysTIuj Residual Heat Removal B. QLscn'etignj The domineialized water bolation valve HV 2 38D 47074A to Ri;R iiesi excin%ei ;AEG24 s, hajj not be shown as locked closed on UFSAR ligure 10.7.1, sht,1 of 4 and associated documentation. Documentation will be revised to reflect these changes, C. Jhason Fon CHaworj This is por the NCR disposition.

D. ,$attTY Evatuation Suuuaav:

1) Does this mcdification increase the probability of occurrence or the consequences of an accW nt or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. Domineralizer water valve HV 2 38D47073A is upstream of valve 47074A and is locked closed to provido positNe isolation of the domin water system from HPSW. Locking closed vane 47074A would be redundant and therefore unnecessary.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

ArL4Wf!G 4'

No. Not having valve HV 2 38D-47074A locked closed maintains the safety related function of that valve and does not prevent other safety related systems or eq'ulpment from fulfilling the;r safety -

related function.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specificat;ons?

Answer:

No Technical Specifications 3.5 and 6.8 were reviewed to make this determination.

68

PEACH DOTTOM ATOMIC POWER STATION UNIT 2 DCCKET No. 50 277 199010 CFH f.0.59 REPORT EMQh1Mn Of DisegnantyleMetrL1QinitAd.20ND!$t.gjjpg A RoK9H'o*EELBRonJ{pj P 90370 A. ly.ntsj Radiation Monitoring System j B. Quenwvou; 1

This NCR involves a drawing discrepancy and is taspoClioned to change the UFSAR Figure 712 3 description cf GL194278 from a globo ulve to its as.is configuration of a gate valve  ;

C. Epsow Fon__Cuawgil This NCR is dispositionec' to use as-is.

D Som Evowmum Suuuiim

1) Does this rnodification increase the probability of occurrence of the conscauences of an accident or malfunction of e tulpment important to safety as previously evaluated in the safety anJysis repon?

Answeg  ;

No. This is a description change only. Since the probability of failure of a globe valve is the same as that of a gate valve, the failure of 63L 19427B will not cause an accident of a diflerent type than previously evaluated in the SAR.

t

2) Does this modification create the possibility for an accident or malfunction of a different type than -

any evaluated previously in the safety analysis report?

Answer:

No. Since the probability of failure of a globe valve is the same as that of a gate valve the failure of 63L 19427B will not cause a malfunction of equipment important to safety different than any previously evaluated in the SAR,

3) Does this modification reduce the margin of safety as deflncd in the basis for the Technical Specifications?

.7, AES.YitG No. Technical Spectiication bases 1.0,3/4.2,3/414 were reviewed for making this determination.

e 69 iisierr su-Wm*- --ee wui,w4>+r---++-ee ew-.us-N-~ w --=--+495mT**'"'-- Y"t^-T**W5MM*"TUM**=*Me"MfT-4'"'V"t--"WWV't9*4 - "T' T e vyv'r7-~'T-"TWN'W**TW

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1 PEACH BOTTOM ATOMIC POWER STAT 10fJ tlfJIT 2 DOCKET No. 50-277 199010 CFR 50.59 REPORT Sn0Ntgn Of DisnenantLaet_*een As Built And Other OptVmepnta.tjgtn

.dPECON'SaMMELBff91LUp; P @ 383 A. Systtu: Service Water B.Ancaintiost This NCR identifies a 1* vent valve located on the 6' service water supply header upstream of valve 2 33-514. This valve does not appest on the UFSAR Figure 10.G.1 or associated drawings. The Figure and associated drawings are being revised to reflect the as built condition.

C. B.Latow Foa Cua9au This NCR is dispositioned to use as is, D. SAf TTY kVA,[QATION Souuany;

1) Does th!s modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety ann / :.imntt?

Answen No. The existence of the vent valve has no a'fect on the abiDy of the service water system to perform its design function.

2) Does this modtfication create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answert No. The vent valvo provides a passive system pressure boundary and venting function. The vent valve has no adverse affect on the system d4 sign or power generation objective.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: y No. Sections 3.5 and 3.9c were reviewed.

t 4

e t

70

- = - - _

I PEACH DOTTOPA ATOPAIC POWER STATlotJ )

UtJIT 2 -

DOCKET No,50 277 l 199010 CFR 50.59 REPORT jligrepancies Between As. Built aniAgg1@rgMMg! ,

f.{pywp.guepawawqr Rtreat Nod PM107 A, Ennuj Reactor Water Cleanup (RWC)

B. prsemption:

This NCR identifies a condition where the as built system does not agree wnh the analytical stress model.

Reactor Water Cleanup System check Valve 212 02 was abandoned (with internals remosed) and a replacement valve was installed upstream of support 12DE H33 The replacement valve is 90 lb heavier than the original valve. The stress calculation dd not include not evaluate the increased weight of the replacement valve and the additional weight of the abandoned valve in the analysis Cf the piping system.

C. Rrasow Fon Cwawor: -

' Analysis shows that, while the new check valve and the abandoned check valve have no adverse eMect on piping stresses and pipe supports, two supports exceed code allowables and design enteria. To resche this, a rigid support will be installed in place of one of them. ,

f D. Sarm Evawa.nqLSuuusay:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malf unction of equipment important to saf ety as previously evaluated in the saf ety analysis report?

Answer:

No, System operability and the structural integnty of the piping and supports will be mair' rained.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?  ;

ADiw.tn - -

No. Support will remain adequate during the installation of the new support. The abandoned check valve and new check valve have no adverse effects.

3) Does this modification reduce the margin of safety as defined in the basis fcr the Technical Specifications? ,

Answer:

No, Technical specification 3.5C,3 5 0 and 3.68 were reviewed to make this determination, f

i I

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- - _. _ . . - - - - - _ . . . -_. -- - - - ~_.-.. - - - -. - .-.- -.~ -. ..-. -

PEACH BOTTOM ATOMIC POWER STAllON UNIT 2 DOCKET No, 50-277 199010 CFR 50.59 REPORT Resolution Of Discrecancy Between Ash And DocumentalirAn i

Newecutoapaner Rracat No.: P 90569 A. SyntuJ Turbine and Extraction Steam D. Descaintiqw_;

This NCR Identrfies discrepancies between the as is condition of the Turbine Extraction Steam System, UFSAR Figure 11.8.3, and related documents. A field walkdown identified tnat three dump valves, HO 406-1 1 A B & C, shown on FSAR Figure 11.8.3 sheet 1 Rev. 8, do not exist. The UFSAR Figure will be revised to -!

reflect the as-installed condition the single existing dump valve.

C. Reasow Foa Csance; l The NCR is dispositioned use as is.

D. Sarm Evatuation Suuuaarl

1) Does this modification increase the probability of occurrence or the consequences of an accident or malf unction of equipment important to safety as previously evaluated in the saf ety analysis report?

Answer:

No. This is a drawing change only to reflect the as-is plant condition. The existing single duma valve adequately interrupts air to the extraction steam line drain valve. It causes the drain valves to open, allowing any collected moisture to drain to the main condenser. This ensures that no water remains in the extraction line to flash and re-enter the turbine to cause damage or an overspeed condition,

2) - Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the sa.ety analysis report?

_An}wer:

No. The changes are minor system changes which reflect the as is condition of the plant. The

- single air dump valve will function to cause the extraction steam drain valves to go to the open position on turbine trip, draining any water or steam to the main condenser. It does not create any new accident initiators, nor does it affect any existing accident initiators such that a different type of accident than previously evaluated would result.

- 3) Does this modification reduce the margin of safety as defined in the basis for the Technical-Specifications?

Answer:

No. There are no Techrdeal Specifications that are applicable.

72 l

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Linit 3 73

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PEACH DOTTOM ATOMIC POWER STATIOf1

. . UNIT 3 DOCKET No. 50 277 & 278 199010 CFR 50.59 REPORT Phnt Monitorino_Sntfm (PMS) Cpngtgr I,rg13 MpmuqatioH No.: 955J A. _Sy.nuj PMS B. Dmcniption:

This modification involves installation of Unit 3 conduit. raceway cables, cable terminations, loop taps. and IE multiplexers for the new PMS. Some existing equipment was modified to make circuits compatiole with the new PMS. Appropriate sections of the UFSAR will be revised to show these changes C. Reason Foa Csance:

This modification was necessary to implement the new PMS system. It allows input of new signa!s.

D. SartTv_fvatuaTion Suuumav:

1) Does this modlfication increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

_ Answer:

No. The additionalloads have been evaluated and will have no adverse impact on bus loading This modification does not affect the capability to shut down the plant in the event of a fire. Iso!ators have been properiy separated and qualified such that common mode failures have been eliminated.

The input taps do not change the way instrument loops function.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. All applicable design criteria have been met for safety related circuits. Installed input taps do not change the way in which instrument loops function or introduce any new failure modes.

3) Does this modification reduce the margirwof safety as defined in the basis for the Technical Specifications?

Answen No. Technical Specifications 3.5C. 4.5C,3.7D. 4.7D were reviewed in making this decision.

74 e

0 PEACH DOTTOfA ATOMIC POWER S1 ATIOf1 UfllT 3 DOCKET fl0. 50 273 1990 '0CFR 50.59 Repcrt Hich Pressgrdgpjjnllrgion (HPCI) Alternative CQntrqLSMtiqo fri.DI!Diall2! TEN 1353A A. S2mu; HPCI B. pl}cmirtioN:

This modification provides and alternative control station fc? the HPCI system to supply water to maintain reactor vessel water level and to provide a mechanism for removing energy from the reactor vessel fcr attemative shutdown.

C. ,Brason Fem CwaNot:

This modification was necessary to meet 10CFR Appendt< R fire reqirements.

D. Sartry Eynvation Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

.A_n,swer:

No. This modification does not change the operation of the HPCI system as described in the UFSAR when all the transfer /lsolation switches are in the ' normal' or " test" position.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer; No. This modification does not change the HPCI system a defined in the UFSAR.

3) Does this modtfication reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. The safety function of the HPCI system is not affected by rerouting of circu.ts and the add tion of safety related transfer / isolation switches.

75

'

• PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT P2 pige Thg3ajgute Di'fuser To Discharcggrg]

Mooiricatiou No; 5113 A. BATLui Radwaste D. Sn.qairtion:

This modification replaced the treated liquid radwaste distribution lino with a single dischargo orifice located at the east edge of the canal and climinated the concreto block and cable supporting system.

C. Reaspw Fon CHauct:

One of the concrete blocks supporting the discharge pipe for discharging treated liquid radwaste into tno circulating water discharge canal has moved allowing the pipo to swing downstream. This condition overstressed the pipe and broke the flange bolts.

D. Sartry Evatuation Suuuany:

1)

Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. Failure of the orifice would not create a release point difierent that that of the original distribution line.

2)

Does this modification create the possibility 1or an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. The possibility of failure of the orifice is less than that of the distribution line.

~

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specccations?

Answer:

No. Technical Specification 3.9 was reviewed to make tnis decision.

76

PEACH DOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No,50 278 -

199010 CFR 50.59 REPORT  !

EfiCMl0!LO_D!it!fDitncy.fetweeri l AMAncLaocun2Matist) fdppmcalingEAi 06-014 A, Ettigj Reactor Feodwater D. DLL9f)f.IL2911 The pressure taps on the reactor foodpump minimum flow recirculation line were removed and replaced with wold caps Figure 11.0.2A of the FSAR will be updated to reflect this change-C.fitnow Fom_CHawotl The pressure taps were not used and woro frequently damaged.

D. Sarrty Evatvanow Suuuaar:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malf unction of equipment important io safety as previously evaluated in the safety analysis report?

$n#Effl No This will not effect the min flow operation.

2) Does this modification create the possibility for an accident of malfunc'lon of a difieront type than any ovaluated previously in the safety analysis report?

Answer:

No. The modification permits the system to operate within the bounds of the previous feodwater {

system evaluation,

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? -

_ Answer:

No. The Technical Specification does not address these pressure taps.

l I

I l

l r

l 1

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77

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PEACH DOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT

p. cleat Interlocks Assotiated With Rc'ueLElootStry.i ttLJat!prm P Cranelicht liR2eany plant Antaation Npj 02-4 A. Signuj Cranes, Holsts, Tools D.ptscwitent This TPA delcats tho interiocks associated with the refuel floor service platform jib crano " hoist loaded

• as described in the FSAR section 7.6.3. This IPA also removes the 120 volt AC logic power to the service platform jib crane.

C. [1 tag _qw Fon Csaggij The llb crano has been scrapped and is no longer available for use.

D. SarfTY Evatvation Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report? -

Answer:

No. Since the crano no longer exists, defeat of its refueling intertocks does not increase the probability of occurrence or the consequences of an accident or malfunction.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?  !

Answer:

No. All other refuel Interiocks are proven to be operational in accordance with Technical Specification 4.10.8.1 via performanco of refuelinterlock STs 12.1,12.1.A. and 12.1.B as required.

Therefore all applicable refuoi interiocks will be in effect, serving to prevent inadvertent errlicality during refueling operations.

3) Does this modification reduce the marglrrof safety as defined in the basis for the Technical Specifications?

• Answer:

l No. The margin of safety as definod in the basis for the technical specifications is not reduced by l the defeat of the interiocks since the jib hoist has been removed, l i l

i A 78 ,

_ . _ ~ . . . - _ _ _ _ _ _ _ _. . _ _ . _ . _ _ _ - . . _ ., _. _ _ _ _., _ - ... _ _ ... _ - _-. _._

PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 ,

199010 CFR 50.59 REPORT Control Rod Position Indicatica

,Iturgaany Ptawt Attraws 6210 l A. Systru: Rod Protection ,

D. DescamtioN:

This Temporary Plant Alteration wdl clear the

• Rod Drift Annunciator

• by removing the extra indications in the ones digit and leave only the actual position in the ones digit.

C. Reason Fon CHANOt:

Control Rod 22 27 has an attA position in the ones digit which causes the

• Rod Drift Annuncialor* to remain alarmed.

D. Sarrty Evatuattow Syger I

1) Does this moBhcation increase the probability of occurrence or the consequences of an accident or malfunction of equipment important io safety as previously evaluated in the safety analysis report?

Annwer:

No. The rod switches on the position indicating probe and the rod drrft alarm ara not important to I safety, so the probability of occurrence of a ma: function of equipment impurtar. to safety previously -

evaluated in the SAR is not increased.

2) Does this modification create the possibility fcr an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. The reed switches on the position Indicating probe and the rod drift ciruitry cannot effect equipment important to safety. Therefore the possibility of a different type of malfunction is not

' Created.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answen No. The control rod position indications are not referred to in the Technical Specification Dasis.

79

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t i

PEACH DOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 199010 CFR 50.59 REPORT Egggmbina'lon Procn3 in the Steam Jet Alt Elector (SdAf Condenser i Nrw Ortnatiwo CONDITION ,

A. Arntvl Offgas i

B. Drseniettow:

i This evaluation was written to determine the operability of the Off-Gas system during the recombining process.

C. Reason Fow CHawor:

During Offgas System trouble-shooting, a system porturbation occurred which resulted in a sustained hydrogen burn in the SJAE aftercondenser. The unit was operated for 3 days in this condition.

D. Sarryv Evaluattow Suuuany:

1) Does this modificatlon increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

A lawer:

No. Hydrogen concentrations downstream of the recombiner were always below flammability limits and gaseous radwaste and releases at the main stack did not increase during the new operating conditions.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

_ Answer:

N o. . No accidents of a different type than previously evaluated in the FSAR are created in the new operating condition of the non-catalytic recombination process occuring in the SJAE aftercondenser since the recombination process was occurring in the offgas system.

3) Does this modification reduce the margin *of safety as defined in the basi. for the Technical Specifications?

Answer:

No. Technical Specifications section 3.8.C were reviewed and continuous monitoring assured that dose rates of radioactive materials and noble gases did not exceed limits specified.

• PEACH DOTTOM ATOMIC POWER STATION j UNIT 3 DOCKET No. 50 278 199010 CFR 50.59 REPORT De-eneralzation Andfetirement Of MO-3ts JJ.pncouronuanet Rr*ont No: P 89782 l

A. Systru: Main Steam D. Dr$CRIPT1o$

This NCR evaluates the do-onergization and retirement of MO 3469. UFSAR Figuro 11.2.1 Shoot 3 and associated documentation will be revised to show these changes.

C.BIAson Fon CHanor:

The existing valve stem is undersized and could fall if operated repeatodly. PBAPS Operations personnel no longer open th!s valvo and have no intention of using this valve in the future.

D. MAFETY EvAtuation Syuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

AnJLsn No. MO 3469 is 'O' to maintain system pressure boundary (passive operational mode). Do-energizing the MO with the valvo in the closed position does not affect the ability of this valve to perform its safety related function of pressure boundary,

2) Does this modification croato the possibility for en accident or malfunction of a differeat type than any evaluated previously in the safety arialysis report?

AntWRG No, Valve MO-3469 is not involved in the determination of consequence for any accident described in Section 14 of the UFSAR. De energizing this valve still allows the valve to maintain its 'O' function and does not affect any other systems or components.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Autatu No. Converting the valve from an active to a passive component maintains tho 'Q* function of the valve and reduces the possibility for creating different types of accidents. De-energizing the valve eliminates any active functionhig cf the valve and thus reduces the probability of a malfunction.

81

5 PEACH BOTTOf.i ATOf.itC POWER STATIOf1 UNIT 3 DOCKET fJo. 50 278 199010 CFR 50.59 REPORT Resolution Of Discrecancy Between As Built And Deslan Drawki.gs NoncouronwNcr Rrront No; P-89923 A. Sysnut High Pressure Coolant Inloction (HPCI)

B. Orscamtiowt This NCR identifies a discrepancy between UFSAR Figure 7.4.1B sheet 4 of 4 rev. 7, associat and the actual piping configuration. Specifically, the Figure shows a test connection consisting o of 1/2" pipe, a valve and a cap between valves VRV-5998A and cot.dition.

C. Reason Fon Csaugfl This NCR is disposttioned to use as-is.

D. SastTv Evawation Suuuanyt Does this modification increase the probability of occurrence or the consequences of an accident 1) or malfunction of equipment important to safety as previously evaluated in the safety analysis report Answer:

No. The test connection would facilitate testing or troubleshooting of vacuum relief valve S998A.

Technical Functional testing of the VRV is not required by the Technical Specifications.

Specifications 3.5.C, and 3.7.C, were reviewed to make this determination. Although not a Te Specification requirement functional testing of VRV 5998Ais an NRC commitment. Surveillance

- procedure, ST 12.14, HPCI Vacuum Relief Valve Functional, satisfies this commitment. This procedure does not utilize the subject test connection.

. ~

Does this modification create the possibility for an accident or malfunction of a different type than 2) any evaluated previously in the safety analysis report?

Answer:

No. This activity maintains the operability of VRV 5998A.

Does tais modification reduce the margin of safety as defined in the basis for the Technical 3)-

Specifications?

ADMAG No. Technical Specifications 3.5.C and 3.7.C were reviewed in making this determination.

)

82

PEACH BOTTOM ATOMIC POWER STATION

, , UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT Resolution Of Discrecancy Between As-Buitt and Documerjugn flouconsonuAsiqe Rrnont Ng1 P 90077 A. Systru_1 Condensate B. Desewtion; A 3/4" vent valve (3/4* 130) located on the 6" condensato filter demineralizer precoat tank fillline shown on UFSAR Figure 11.7.1 does not exist in the plant. The as-built plant configuration is a 1 1/2' nipple and cap installed on line 18 HG-6". The as built piping configuration also has an 8' x 6" reducer installed between lines 18 HG 8' and 18 HG 6'. The figure will be revised to reflect the as built conditions.

C. Eplow Fon C,tasot:

This NCCt is dispositioned use as is.

D. Sarm dvaluanoN Suuuany:

1) Does this modificat;on increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The Condensate Filter Domineralizer System provides a power generation support function.

As this actMty does not affect the operation or design basis of the systtm. there is no effect on Accident Scenarios evaluated in UFSAR Section 14. ,

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. Deletion of this vent valve and addition of the reducer will not affect the function of the Condensato Filter Demineralizer System.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? ,

Answer:

No. This portion of the system (condensate filter demineralizer) is not addressed in the Technical Specification.

83

.. _. _ _ _ _ . .~ _- - . _ __ _ _ _ _ _._ ._ .._._ _ . _ _ _ __ __

4.

PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT l

l Resolution Of Discrecancy Between As-Insta!!ed And Documentation 11QNCoNFogANes Rrront Not P 90082 A. SmLM.] Condensatn j D. Drsemiption,J .

This NCR identifies a discrepancy between the as built condition of the plant and UFSAR Figure 11.4.1 to ,

eliminato a 3/4*:130 valve and line from one end of the condensate service off gas line. The Figure will be revised.

C. Reason Fon CHANor:

This NCR is dispositioned to uso as-is.

D. J AFETY Evatuam

SUMMARY

l

1) Does this rnodification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answtrj No. The existence of the 3/4*:130 valve and lino has no effect on the ability of the condensate service off gas line to perform its intended function. UFSAR Section 14.0.

• Plant Safety Analys!v i and Appendix G. Plant Nuclear Safety Operational Analysis were reviewed to make this determination.

2) . Does this modification create the possibility for an accident or malfunction of a different type than l any evaluated previously in the safety analysis report?

Answer:

No. The function of the cap on the end of the condensate service off-gas line is to have no flow on the line with or without the 3/4*;130 valve in between and is not safety related.

3) Does this modification reduce the margin' of safety as defined in the basis for the Technical Specifications?

Answer:

No. Technica! Specification 3.78. 3.7C,3.88,3.8C were reviewed to make this decision e

a 04 L

. . _ . , . , . _ , . . _ - , . _ . . . . - . .._,,,,_._.m . . _ . , _ _ _ . . . _ _ _ _ . . . , . . - _ _

PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPOR7

&SD ution l Of D,i.sgicoancy Between As-Instalkd And Documentati.QQ P

.tlpmouronuanet Rrront No; P-90083 P

A. Synnu; Fuel Pool Cooling B. DrsemenoNI This NCR evaluates the removal of the Condensate Refuel Water vent line from UFSAR Figure 10.5.1 and '

other related documentation to conform with existing conditions.

C. Rrasow Fon CHa.yoi; The NCR was dispositioned to use as is.

D. Sarrty Evatuanow Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment importi,nt to safety as previously evaluated in the safety analysis report?

Answer:

No. Another vent line exists on the same elevation. This vent lino performs the regulred system venting function.

2) Does this mod'.?lcation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. The existing vent is perf6rming adequate venting functions.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Speelfications?

Answer:

No. Sections 3.5 and 3.10C were reviewed in making this determination.

85 t

- -~ - . __. - - ._ . - - _ - . . _ . - - - - . _ _

• PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 199010 CFR 50.59 REPORT Resolution Of Discrecancy Between As-Built Andjgumenta([gg Newcouronuawcr Rtcent Ng; P.90092 A. _Systrut Reactor & Recirculation D ,QDcniptiog The NCR identifies discrepancies between UFSAR Figure 8.7.2A, associattd documentat built conditions During the implementation of Mod 1536, compartment 30D1120 of MCC After the mod was installed. Maintenance a spare by removing motor operated valve MO 310-33.

Engineering removed the bucket from compartment 30D1120 and used it during th Program. The bucket was not re installcd for economic reasons.

C. Reasow Fon Cwawot This NCR is dispositioned to use as.ls.

D. Sarm Evatuanow Suuuanz Does this modlfication increase the probability of occurrence of the consequences of a 9 accid 1) or malfunction of equipment important 1o saf ety as previously evaluated in the saf ety a Answer; r No. Single line, tabulation and secondary and control drawings are being revised 13 reflec built

• condition to show

• space

• instead of ' spare" for MCC 30011 comp. 30D1120 2)

Does this modification create the possi' i y for an accident or malfunction of a different type th any evaluated previously in the safety L..atysis report?

Answer; No. This change does not make any functional changes to any plant systems or equipment.

Does this modification reduce the margin of safety as defined in the basis for the Technical 3)

Spec?ications? i e

Answer; No. There are no technical specifications applicable 10 the mentioned MCC Compartment. S 3.9 of the Technical Specifications were reviewed.

I I-I i

i 66

~ - - ~ . - _ - . _ _ , , _

. - . -_~ .. - ...-. -- - - -_-. - . . - - - - - _ - _- - - -. . -

PEACH DOTTOM ATOMIC POWER STATIOld l UNIT 3 l < .

DOCKET No. 50 277 & Su-278 199010 CFR 50,59 REPORT f

pesolution Of Discrepancy Between As Btst.Ani_Qecumentation l

Nosgsw!an.uawce Rega.130J P 90201 l

A. .f nguj Main Steam >

D. pngatavowt The NCR identifies discrepancies between as built conditions. UFSAR Figure 1121 sheets 3 and 4, and UFSAH Figurc 9.4.1 sheet 3. The Figures wdl be revised to reflect the as. built conditjons. These are min revisions such as the addition of pipe caps, the addition of vent lines, the relocation of drain vah,es, and the addition of line sizes.

ClEgnpu Fon Pti'.!!9ti This NCR is dispeirioned to use as is.

D. lamy Evatuanow Suuusay!

Does this modtfication increase the probability of occurrence or the consequer es of an accident 1) or malf unction of equipment important to safety as previously evaluated in the saft " analysis report? ,

ADDREG No. These char,ges did not affect the Main Steam System or interiacing systems in a manner which would have impact any accident evaluation. These were minor system changes which reflected the "as built" condrtion of the plant and did not adversely affect the function or operation cf ine Main Steam System or interfacing systems.

Does this modlfication create the possloility for an accident or malfunction of a different type than 2) any evaluated previously in the safety analysis report?

Aoiw._rU No. The changes do not create any new accident initiators nor do they affect any existing accident initiators such tht! a different type of accident than previously evaluated in the SAR would result.

Does this modification reduce the margin of safety as defined in the basis fur the Technical 3)

Specifications?

A01F.1G No, There were no Technical Specification applicable specifically to the Main Steam System and the changes to the Main Steam System addressed here did not mpact the Technical Specification bases of any InteMacing systems. Technical Specifications 3.5,3.6. and 3.7 were reviewr.d b

87

PEA.?H ROTTOM ATOMIC PONEH STATION UNIT 3 DOCKET Ho. 50 278 199010 CFR 50.59 REPORT Eesolutio1LcLDActcoancyJhtnetL A ljy!MMpagnegiall.oD R9HQ2nLQaMaaGLBirMLRo_i P-90208 A. Snnuj Reactor Building Closed Cooling Water (RBCCW)/

Turbine Building Closed Cooling Water (TBCCW)

O. Dngawtiow:

This NCR Identifies discrepancies between UFSAR sections 10.8 ard 10.i0 und Figures 4 9.1 and 7 20.1 and with the as built plant condition for the RBCCW ard TBCCW systemt,_ These systems are described in UFSAR sections 10 0 and 10.10. The OUAR and the documentation were revised C.Egalen Foa Cnawou The NCR was dispositioned to use as Is.

D. Syntf;.yAwattow Sepany:

1) .Does this modification increase the probability of occurrence or the consequences of an accident or ma!! unction of equipment important to safety as previously evaluated in the saie ty analysis repon?

ADD M G No. There are no physical changes made to the plant by this NCR. The NCR only documents the as-buXt configuration of the system and its components. As a result, all the identified changes were to drawings only.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the cafety analysis report?

A01ECG No. Since this NCR reflected the as-built condition, eil non safety related piping, tubing and components are considered to be installed in accordance with specification M 301 (tof 2). Therefore, the physical arrangement of the piping, tubing and components were considered acceptable.

3) Does this modification reduce the rmrgin of safety as defined in the basis for the Technical -

- Specifications?

A09MG No. The margin of safety as defined in the bases of any Technical Specification was not seduced since the RBCCW nnd TBCCW systems were not addressed in any Technical Specification.

Additionally, the etwiges specified to the RBCCW and TBCCW systems did not affect the Techntral Specification bases of any interfacing system.

B8

i PEACH DOTTOM ATOMIC POWER STATION UNIT 3 DOCKET Ho. 50-278 199010 CFR 50.59 REPORT Resolution Of Discrenancy Betwem As Byilt And QQcumentation

-140Ncouromuawer Rtront No: P.90209 A. 31Hiul Chuled Water

9. Desc w tion:

UFSAR 10.11.1 sheets 3 and 4 and associated documentation will be changed to reflect the as built piping configuration as identified during a field walkdown. These are minor changes such as:

the addition of normal and failed condition designations to air operated valves.

addition of pipe nipples and cups.

. the addition of vent at.d drain lines.

C. Bsssow Fon Curytggi This NCR is dispositioned to use as is.

D. Sutw EvnuatioH Spuum:

1) Does this modification increase the probability of occurrence or the consequences of an accident or nelfunction of equipment important to safety as previously evaluated in the safety analysis teport?

AD1Effi No. These were minor system changes which reflected the as built condition of the plant and did not adversely affect the function or operation of the Chilled Water System or interfacing systems.

These changes de not afioct the system in a manner that woutd adversely affect the function or operation of the O-Listed sections of the Chilled Water System or interfacing systems.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. The changes did not affect component _s in the Chilled Water System or interfacing systems in a manner that would create a different type of component malfunction or accident than previously evaluated.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Ananu No. There are ca Technical Specifications appihble specifically to the Chilled Water System and the changes to the Chilled Water System addressed here did not impact the Technical Specification bases of any interfacing systems.

89

4 PEACH BOTTOM MOMIC POWER STATION UNIT 3 DOCKET No. 50 278 199010 CFR 50.59 REPORT 1

.Besolution of Discrenansies Between DocumentationAnd_As-Built Conditions f[quso,wf,qepange Orege,y '{g; P-90211 A. Systruj Emergency Cooling B. Qtsemer.tgyl

. This NCR identifies discrepancies between the as built configuration, UFSAR Figure 10.24.1, and associated documentation. The figure will be revised to show Valve 48-11207 as a globe valve in the closed position.

Four pressure points will be labeled on the Figure and removed from the O-List.

C. Reason Fon Cuawor: .

Th:; NCR was dispositioned use as-is.

D. Sarrry EvAtuation SouuAny:

1) Does this modif' cation increase the probcbility of occurrence or the consequences o,' ea accident or malfunction of equipment important to safety as previously evaluated !n the safety analyses report?

Answe,ff No. The changes made are either minor or for documentation only and reflect the "as-built

• condition of the Nnt. They have no affect on the function or operation of the Emergency Cooung System.

2) Does this modificat on create the possibility for an accident or malfunction of a different type tfun any evaluated previously in the safety analysis report?

Answer:

No. Cooling water will still be provided to the Emergeng Sewice Water system in accordance with design requirements. These changes do not affect the function or operation of the associated vent lines or process lines. . Removing ' hem from the O-List is acceptable since they are isolcted from the safety-related process line in which they are attached by normally closed manual root valves.

and they have no safety related active function.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. The changes do not reduce the margin of safety as defined in the bases of any Technical Specification. Section 3.9 and 4.9 of the Technical Specification discuss the Emergency Cocling System; The changes do not alter the function or operation of the system since the changes are g for documentation purposes or for vent and test connections which have no active function during system operation.

90

PEACH 00TTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT BIsolution of Discrenancies Between Documentation And As-Built Conditiong Noucouronunner Repont No: P.90212 A. Systru: Instrument Nitrogen System

8. Desenistion:

This NCR identifies discrepancies between the as built configuration and documentation. UFSAR Figure 5.2.9 and associated documentation will be revised to show the as is condition of the traversing incore probe ourge take-off location and insulating fittings which exist on lines to instruments on nitrogen receivers A &

B. Alsc, valve number 36B 55222B will be labeled, a tubing symbol will be added to TS 5234, and a reducer e id 1/2" line designation will be added. _

C. Erasow Fon Cuawor:

This NCR was dispositioned use as is.

- D. Sartry Evatuation SUMMaRv:

1) Does this modification increase the p obability of occurrence or the consequences of an accident -

or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answe_r:

No. These are minor drawing changes which reflect the as-built condition. -

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

AnswE No.' The changes included in this NCR do not affect components in the instrument Nitrogen System or interfacing systems in a manner which would increase the probability of a component failure. -

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? .

Answer:

No There are no Technical Specifications applicable specifically to the Instrument Nitrogen System.

Changes addressed here do not impact the Techolal Specifications of any interfacing systems 91 e-. . . . . . . . .

PEACH BOTTOM ATOMIC POWER SI ATION UNIT 3 DOCKET No. 50-278 1990 to CFR 50.59 REPORT ECjiDgion i Of Dhcustncyletween As MIL &KLD9cume_ntation n RoNQOyonMANC[,,8(PonT yp.; P-90219 A. Sysnu: Reactot Coro isolation Cooling (RCIC)

B. Dnc!En9M P&lD's M-359 and M-300 did not agree with'the as-built condition of the plant. The turbine exhaust drain pot was improperly indicated. The dcstination locator of the RCIC barometric condenser vacuum pump discharge line required revision, and a turbine exhaust drain line was deleted from two P&lD's.

C, Brason Fon Cnawon Revisions were made to reftect the as built condition of the plant D. 33fts;Tv Evatuanon Suuuany: ~

1) Does this mod #ication increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The RCIC turbine exhaust line, from which the drain line is being deleted, is not related to the initiation of any accidents. The affected exhaust piping is downstream of the exhaust line containme'it isolation valves and is not connected to the reactor coolant pressure boundary.

2) Does this triodification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answen 5

No. Since the RCIC System functions are unaffected, no new failure modes for RCIC equipment are created.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answen No. the bases for Technical Specifications 3/4.5C&D were reviewed in making this determination.

92

n . . . .. - .

PEACH DOTTOM ATOMIC POWER STATION

, , UNIT 3 i DOCKET No. 50-278 199010 CFR 50,59 REPORT RegglMllon Of Discrepancy Be. tween As Built _ And Dagurnentaligt)

.Nowcowronuawge Repontfigj P 90223 A. Kyluu; High Pressure Coolant injection B, Ptsemieiion:

This NCR identifies several inconsistencies between the as-built condition of the Unit 3 HPCl System and P&lD's M 365 sheet 2 and M-366 sheets 3 and 4 These consist of numbering errors conection of drafting errors, clartfications or corrections to agree with the as-built configuration of the plant.

C. Erasow Foa C.tewor:

This NCR is dispositioned to revise documentation to agree with as built configuration.

D. lartty EvamaTion Suuuany: _.,

1) Does this modtlication increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

~

Answen No. The changes identified had no effect on accident initiators, HPCI System function or operability.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answen No. Most of the identified changes are drawing clarifications or corrections of drafting errors.

Location changes and added vents, drains and root valves are consistent with design standard instrument installation,

3) Does this modification reduce the margin of safety as defined in the bas ls for the Technical Specifications?

Answer:

No. The margin of safety for the HPCI System is unchanged. The bases for Technical Specificatioc 3/4.5.C were reviewed in making this determination.

l 93

PEACH BOTTOM ATOMIC POWER STATION

'- 4 UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT Rfolacement Of Resistor and Multiolier Divider Ca_rd NONCoNFORMANge Rt, ont No: P 90356 A. Ststru: Containment Atmosphere Dilution (CAD) l B. Descwriow: ,

This NCR evaluates the replacernent of precision resistor and a multiplier olvider card in the mass flow computer, for the CAD system.

C. B5ason Fqn _CuaNor:

To proyide more accurate indication of the nitrogen flow to the containment. This will enhance the responso of the control room operator to the oxygen concentration limita.  !

D. SAFETY Evatuation Suuuany:

1)- Does this mcdification increase the probability of occurrence or the consequences of a.1 accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The multiplier divider card is a one-tomne replacement. Replacement of the precision resistor in the mass flow computer will increase the accuracy of indication and recording.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis repott?

~ Answer:

No. The new components are similar to existing components.

3) Does this modtfication reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. Technical Specifications section 3.7A and 4.7.A were evaluated to make this determination.

94

PEACH BOTTOM ATOMIC POWER STATION

. . UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT Discrenan:v Between P&lD and Schematig NoncoNFoRMaNee Repont No: P-90391 A. Sysitu: High Pressure Cooiant injection (HPCI)

B. DesemiprioN:

This NCR identifies discrepancies between P&lD M 365 and schematic M-1-S-36, and UFSAR Figura 7.4.1 A sheets 1 and 2. The P&lD incorrectly identifies the power supply for HPCI turbine control instrumentation.

C. Reasow Fon CHawo1 The NCR is dispositioned to revise PalD M 365 and UFSAR Figure 7.4.1 A sheets 1 and 2 to agree with schematic M 1-S 36 sheets 12,13,27, and 28 which show the correct power supply.

D. Sartry Evatuation Suuuamv:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. This is an editorial change only. The correct power supply performs the same function as the one being changed was thought to have performed.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. No physical changes are being performed. No new failure effects are created by the change.

3) Does this modrfication reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. Technical Specifications Section 3.5 was reviewed to make this determination.

95

• a PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 1990 to CFR 50.59 REPORT d

Reolacement Of ('!DI.tLEcedlq_Candensate.Eumn Transferrner flowcouronuawcr Repon_ trol P-90398 A. Systtu_; Cordensate B. Drseniption:

This NCR addresses the replacement of the failed 500 MCM Cable that feeds the Transformer 3AX07 for the Condensate Pump 3AP03 with a 15KV, 750 MCM Triplex Aluminum Cable. UFSAR Figure 8.4.28 and associated documentation will be revised to show this replacement.

C. Reason Fon CHANG 1 This NCR is dispositioned to replace the 15KV,31/C-500 MCM aluminum cable between the 13.8KV,30A01 bus and the 3AX07 transformer that feeds the 3AP03 condensato pump with a 15KV, 750 MCM triplex aluminum cable.

D. Sarriv Evaluation Suuuans

1) Does this modification increase the probabikty of occurrence or the consequences of an accident or malf unction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The operability of the safety related systems or equipment will not be compromised.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

.A_njinpfri No. The replacement of the power cable associated with the Condensate System will not make any -

functional changes to the plant safety-related systems.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. This NCR does not make any changes to the operation or function of the plant safety related systems or equipment.

96

PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT Englution of Discreoancy Between As-in AnglJocumentallag NowcouroRuawce Repont Npj P-90399 A. Systru: Standby Liquid Control (SLC)

B. DescamTio.un This NCR addresses discrepancies in design information with regard to the number and identification of temperature switches used it' the heat tracing and alarm circuits of the Standby Liquid Control System.

UFSAR Figure 3.8.3 and associated documentation will be revised to show the correct r; umber and identification of these switches.

C. Reasow Fon Cwawar The NCR initiated changes to electrical drawings. Figure 3.8.3 of the UFSAR corresponds to one of these drawings.

D. Sarriv EVALUATION

SUMMARY

I

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment irr.portant to safety as previously evaluated in the satety analysis report?

Answer:

No. This is an editorial change only to revise documentation to reflect the current plant condition.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

Nc. This activity maintains the design basis function of this portion of the SLC system.

3) _ .Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

- Answer:

No. Technical specifications 3.4A, B, & C were reviewed to make this decision _

97

, - - - ._ = - - - -- - . -

., , PEACH DOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 1991 10 CFR 50.59 REPORT Discrecancies Between UFSAR and Actual Plant Installation 14oncouronpance Rrpont No.: P.90524 A, SYSTEM: Radiation Monitoring B. Drseniengs This is a change to table 7.13.2 of the UFSAR to show the actual location of radiation detector, RE 318-30AG.

C. Rrason Fgn Csauct:

To provide correct documentation for "as is' plant conditions, D. SarcTv Evatnapon Suuuany:

1) ~ Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. This is a drawing change only to reflect the "as is' plant condition. No physical changes are being made.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. No unanalyzed accident possibilities are being created

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. No changes have been made. Technical Specifications sections 1.0. 4.2. 3.5, and 3.8 were reviewed to make this decision.

I L

98 i

(

m_.._m _ . - . _ _ _ . _ - _ .. _- .m.____. - _ m __ __

E .

PEACH BOTTOM ATOMIC' POWER STATION

-)

UNIT 3 - ,

DOCKET No. 50-278 199010 CFR 50.59 REPORT Resolution Of Discrecancy Between As-Installed And Documentation NowcowropuAwer Re=onT No: P-90690 -

A' Systru;- Primary Containment High Range Monitoring

, i B, DESCRIPt10Nl The NCR identifies a discrepancy between the installed coaxial cable used in primary containment for the Primary Containment High Range Radiation Monitoring System and the system design documentation; The installed cable was identified during a walkdown as Brand Rex. The system design documentation identifies -

the cable as Rockbestos. 2 During extreme high temperature, low radiation level conditions, the radiation -

monitoring system may not operate within the factor of two accuracy requirements of Regulatory Guide 1.97 -

- with Brand-Rex cable installed because of the cable's insulation resistance.

C. RcAsow Fon CwAwart

-The NCR was dispositioned use as-is.

D. SArrry EvAtuArion SuuuAny:

1)' Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report? -

AnswJ No. The Primary Containment High Range Radiation Monitoring-System provides an indication

'only and does not serve as an initiator or contributor to any LOCAs evaluated in the SAR. Brand.

Rex is a qualified cable for use in Primary Containment,

2) -

Does this modification create the possibility for an accident or malfunction of a different type'than

- any evaluated previously in the safety analysis report?

- Answen-No. The Brand-Rex cable satisfies original design requirements. Possible system inaccuracies-would only invalidate radiation readings that are well below the levels requiring any operator action.

per Peach Bottom Emergency Procedures /

3) Does this modification reduce the margin of safety as defined in the basis for the Technical

- Specifications?- "

Answer:

z No.- The Primary Containment High Range Monitoring System provides only indication of radiation =

- levels in containment to the plant operator. The operator wil! use this indication along with other .

data in making an evaluation for implementing emergency action plans.

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PEACH DOTTOM ATOMIC POWER STATION I UNIT 3 )

DOCKET No. 50-278 1990 to CFR 50.59 REPORT Resolution OfSiscrecancy Between As-Is and DocumentatiGD Nowcouronuawce Rrpont No: P 90710 A. Systru: Recirculation. Residual Heat P.emoval, Reactor Water Cleanup B, pesemenoH:

Temporary cables between the Distribution Panet 30Y50 and the Vibration Monitoring Computer Panels 30C850 and 30C851 that were temporarily installed during Mod 1536, have been removed but the Single Line Diagram E-29, UFSAR Figure 8.6.2 and other as-built drawings were not revised to delete the feeds.

C.Frason Fon CMawos:

To revise drawings and documentation to agree with current plant configuration after removal of temporary items.

D. SAFETY Evawancu Su_uuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. The safety evaluation performed for Mod 1536 has determined that the consequences of an

. accident or malfunction previously evaluated in the SAR will not be increased as a result of this change, This NCR only revises the electrical drawings and the UFSAR figure to bring the drawings in conformance with the scope of Mod 1536.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. The safety evaluation performed for the Mod has determined that the possibility of a different type of accident or malfunction of equipment important to safety is not increased as a result of the Mod. Revision of the drawings to show the a+ built conditions does not affect the operation or function of any plant systems or equipment important to safety.

3) Does this modification reduce the margin of safety as defined in the oasis for the Technical' Specifications?

Answer:

No, Technical Specifications 3.9,4.9. and the associated bases were rev!ewed.

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l 100 i

.-- , PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT Correct Discrecancies NonconroRMANCE Rrpont No: P-90762 A. Systru: Radiation Monitoring B. Desenieriow:

This NCR identifies a non conformance between the UFSAR Table 7.13.2 and actual plant installation. The non-conformance involved wording in UFSAR Table 7.13.2 describing the actual location of Radiation '

detectors RE 018-30J and RE 0-18-30K. The Table will be revised to reflect as-installed conditions.

C. REASON Fon CHANGE:

This NCR is dispositioned to use as-is.

D. larETv Evatuation Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

Answer:

No. Changing the nameplate legend description of UFSAR Section 7.13. Table 7.13.0 to agree with actual plant installation and system design documentation did not impact any accidents evaluated in the SAR. UFSAR Sections 7.13, 9.0, 14.0 and Appendix J were reviewed in making this determination.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated prev!ously in the safety F.1alysis report?

Answer:

No. Changing the location description in UFSAR Table 7.13.2 does not increase the probability of '

occurrence of a malfunction of any equipment important to safety as evaluated in the SAR.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer:

No. The Technical Specifications bases were reviewed and none were fans.. to be applicable for the Radwaste Radiation Monitors.

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PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET Mo. 50-278 199010 CFR 50.59 REPORT Resolution Of Olscrecancy Between As-Installed And Documentation NoncouronuAwen ReponT No.: P-90792 A. Systrut Standby Liquid Control (SLC)

B. DrsemipTion:

The 1 1/2' dia. Standy Liquid Control header to Reactor Pressure Vessel has a dual isolation test connection assembly in which both valves are locked closed and capped. The outer valve, HV-3-11-037 Is not shown as locked closed on the P&lD 6280-M-351 sheet 3 and UFSAR Figure 4.3.2 sheet 3. NCR P-90792 was initiated to show this valve locked closed and capped (LCC) on P&lD 6280-M-351 sheet 3. The UFSAR Figure will also be revised to reflect the as-installed condition.

C. REAsow Fon Cuanoc:

This NCR is dispositiened to use as-is.

D. SartTv Evatuanon Suumany:

1) Does this modification increase the probabil*y of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

.enswer:

No. This is an editorial change only to revise documentation to reflect the as installed condition, and does not change the ability of the SLC system to perform its intended function.

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer:

No. This activity maintains the design basis function of this portion of the SLC system.

3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

An1wm No. Technical specifications 3.4 and 4.4 were reviewed to make this decision.

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102

' PEACH BOTTOM ATOMIC POWER STATION

1. UNIT 3 DOCKET No. 50-278 199010 CFR 50.59 REPORT Resolution Of Discrepancy Between "As instaHe i And Documentation Noncouronuawce ReponT Nog P 90793 s A. S<sTru: Standby Liquid Control (SLC)

B. Descamnow:

The 1 1/2" diameter test valve, HV-311-031B Standby Liquid Control Header to the Reactor Pressure Vessel has a dad isolation locked closed and capped vent connection assembly. The outer valve is not shown as locked closed and capped, on UFSAR Figure 3.8.1 sheet 2. The UFSAR Figure will be revised to reflect the as-installed condition.

C, Reason Foa Cuawcr:

This revision is per the NCR disposition.

D. Sartry EvatuaTion Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?

n newr.

• No. This is an editorial change only to revise documentation to reflect the "as-installed

• condition, and does not change the ability of the SLC system to perform its intended function.

~

2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

, Answer:

No. This activity maintains the design basis function of this portion of the SLC system.

3) Does this modification reduce the margin of safety as defined in the casis for the Technical Specifications?

Answer:

No Technical Specifications 3.4 and 4.4 were reviewed to make this decision.

103 s

PEACH BOTTOM ATOMIC POWER STATION

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UNIT 3

[ DOCKET No. 50-278 199010 CFR 50.59 REPORT Resolution OLpiscrepancy Between As Installed And Documentation NONCoNFoRuaHer REPoMT. No; P-90794 A, jyts.nul Standby Liquid Control (SLC)

B, Dgsemir.is The Standby Liquid Cor.irol Tank 30T18 has a single 1 1/2" locked closed and capped drain valvo connection HV-3-11-023 which is not shown as locked closed and capped on UFSAR Figure 3.8.1, Pl&D 6280-M-358, shoot 2 and OAD 6280-M 858 sheet 2. Documentation will be updated to reflect the as installed condition.

C, fiteson Fom CHaHor:

This NCR ls dispositioned to use as !s.

D. SarrTV EvatuarioH Suuuany:

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important ta safety as previously evaluated in the safety analysis report?

Answer:

No. This is an editorial change only to revise documentation to reflect the "as-installed" condition, and does not change the ability of the SLC system to perform its intended function.

2) Does this modification create the possibility for an accident or malfunction of a different 'lpe than any evaluated previously in the safety analysis report?

Answer; No. This activity maintains the design basis function of this portion of the SLC system.

3) Does this modification reduce the margin of safety as oefined in the basis for the Technical Specificatic m?

Anmer:

No. Technical Specifications 3.4 and 4.4 were reviewed to mat e this decision.

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