ML20128D369

Annual 10CFR50.59 Rept

ML20128D369
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1991
From:	Miller D PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CCN-92-14133, NUDOCS 9212070244
Download: ML20128D369 (72)

Text

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Dear Sir:

Enclosed h the 1991 Annual 10 CFR 50.59 Report u rec,ulted by 10 CFR 50.59. Should you have any questions, or requile further information, please contact un. Sincerely, ./ rfd1Yp76 { Ok lk - 4 DBM/AAF/JhM/MJB/ LT l. / ~ ~ ". 4/ /,' Attachment ec: R.A. Burricelli, Public Service Electric & Gas T.M. Gerusky, Commonwealth of Pennsylvania LI. Lyash, USNRC Senior Resident inspector R.I. McLean, State of Maryland T.T. Martin, Administrator, Region I, USNRC ILC, Schwemm, Atlantic Electr c [_ O I C.D. Schaefer, Delmarva Power 4 ') 3 /.'M d 8I idtevr.itr .A / t 07 021 /, 92t2070244 921231 / jDR ADOCKOS00g7

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t a r-h PHILADELPHIA ELECTRIC 00MPANY Pti.ACH 00TPOM ATOl41C POWER STATION tlNITS 2 AND 3 DOCKET NOS. LO277; 50-278 9 ( 1991 ANNUAL 10 CFR 50.59 REPORT a ---_-2 --.w--

} 4 Docket Nos. 50-277 50 278 -1 1 i a: .j" 1991 PEACH BOvTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT I This Report is issued pursuant to the reporting requirements of 10 CFR 50.59 for i Peach Bottom Atomic Power Station Units 2 and 3 (Facility License Numbers ] DPR44 and DPR 56 respectively). This report addresses, but is not limited to, tests and changeu to the facility and procedures at they are described in the l Updated Final Safety Analysis Report. This report consists of those tests and ) changers that were completed in 1991. A summary' of the safety evaluation for O

each item, concluding that an unreviewed safety question, as defined in 10 CFR =-

50.59 (a) (2) was not involved is included. I i d 5 I t n J+ h 1 . = -_ 3-/ - ~ ..,..;_.~,_---_-.

4 1991 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.50 REPORT Table of Contents AY11 TID .P>as Un!to 2 & 3 1 Mod 11 cat!ms 1490 HPCI, RClO 2 1729 Miscollaneous 3 5041 Firo Protoction 4 5085 Turbino 5 5243 125/250 V DC 6 5346 ECC & RC10 7 Nonconfotr anco Reports P 9u288 Diesel .8 P 90529 Instrument Air 9 P-90580 Reactor Water Cloanup 10 P-90010 Altomate Rod insortion 11 P 90640 Residual Haat Removal. 12 P-90050 Reactor Core isolation Cooling 13 P-90725 Stoam Leak Detection 14 P 90762 Radiation Monitoring 15 P-90700 Fire Protection 16 P 90005 Residual Heat Removal 17 P-91317 Emergency Auxilliary Transformer 18 Review ESW, RBCCW 19 Emergency Service Water 20 Procedures AO 38D-1 and AO 380 2 Domineralizer 21 Clearance and Tugging. - Performance Monitoring 22 -

\\ ? Table of Contents Unit 2 23 l .!thMUCall003 955E Plant M>nitoring 24 1310 CAC, CAD 25 1542 Drywfdl Cooling 20 1832 Sprink!st 27 1801 Vartous 28 2075 Primary Containtnent 29 22G9 250 & 125 Volt 30 5040 Emergency Service Water 31 5110 ESW,ECW 32 5209 4kV 33 5240 Reactor Core Isolation Cooling 34 Ronconformance Reoorts P-90214 Nuclear Dolor Vessel instrumentation 35 P 90216 Reactor Water Clean-up System - 30-P-90353 Emergency Service Water, High Pressure Service Water 37 P E 376 Radiation Monitoring 30 P-90509 Recombiner 39 P-90789 Fire Protection - 39 P-91150 High Pressure Coolant injection 40 P-91193 Steam Separator '42 P-91283 Residual Heat Removal 43 Temocrary Plant Alterations TPA 213 07 Reactor Cora isolation Cooling 44 TPA 2-2347 High Pressure Coolant injection 45 TPA 240 07 Service Water 46 TPA 24018 Reactor Protection 47 UnR3 48 Honconformance Reoorts P 80232-3113 instrumont Nlttogon 49 - P-90538 - Residual Heat Removal 50 P 90678 Circulating Water 51 P-91116 Foodwater. 52 P 91107 440 V Emergency Auxilliary 53 P-91347 High Pressure Coolant injection 54 1

)4 m ~ \\ i Table of Contents Unit 3 hrlE2tAIYllarj Afterations TPA 34-7 Foodwater Hoater SS TPA 3184 Refuel Platform 56 TPA 340B-4 Ventillation 57 TPA 342 22 Control Rods 58 1 TPA 342-21 Control Rods 59 TPA 342 25 Control Rods 60 TPA 342 27 Control Rods 61 EIK9dul9 ST 13.18 3 Standby Uquid Control 62 Miscellaneous 63 II111al!all0D installation Flood c4 l l l

O e e 4 N UNITS 2 & 3

PEACH UOTTOM ATOMIC POWEFI STATION UNITS 2 & 3 dor;KET No. 50-277 & $0 278 199110 CFH 50.59 REPORT N0! Yid 3dEtllRDI i WD8fSAHOff Na; 149a A. .SiSIEMJ High Prossuto Coolant lejection (HPCI), Roactor Coro isolation Cooling (RCIC), Coro Spray D. .DMCBIPJ10ri; This modification replaces two tilting disc chock valvos in the coro spray injoction line, and one swlng chock valvo each in the HPCI and RCIC injection linen with swing chock valves. A minor rotouto of Core Spray loop *D' piping was toquired for this installation. A HPCI and a flCIC equalizing tAock valve woro rulocalod to outt,kle it.o outboard MSlV room. O. IEASON,NILQiM10K; The replacornont of the iltting dit,k check valvos was nocessary becauso valvo indication was nbt rollatic, the disc was sticking in the valvo seat, valvos would not opon fully, and Iwo valvos would not pass tostlng requlrornonts. The rorouto of the coro spray piping was dono to elimitato an intorforence whh a junction bot The coro spray piping chango was based on ISI W/ing results, and the block valvos were inovod to minimizo doue. D. SAfm.flAWAT10fLRlM!dM111 1) Does this modification increaso the probabl!!!y of occurrence or the consequertos of an accident or nmilunction of equipment imponant to safety as previously evalua.ed in the safety analysis report? Anwen No. The now valves do not impact any test procedures and are silil flow actuatod compononts. The replacomont of coro spray piping moots the original design requiremonto. 2) Doos this (nodifkation create the possibility for an acckJent or malfunction of a differeA type than any ovaluated previously in the safoty analysis report? At1 wig 14o. The scopo of tils actMty only involvos the replacement of oro stylo of tostable check l valvo with another, and a minor block valvo relocation. 3) Does this modification rodJce;ho margin of saf9ty as defined in tho basis for the Techn19 31 Specifications? g lc i ADtWSU No. Technical Specificadon Soctiors 3.5 wes svlowed and it was detorminod that thoro ara no j. timits or regulatiom specifically listod for thow chock valvos and their auxiliary components. l; l p

PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199110 CFR 50.59 REPORT .QQukidEQOIILUpcindo r I 1AQDiflQAllQlt.LQJ 1729 and NCRs 90 423,421,425 A. G1310E; Mlscollaneous B. IRS.QB11gjpK; This modification rearranged contro! room office spt,co, added 2 lunchrooms with kitchonottos, and an Instrumant lab. UFSAR Figure 10.13.1 was revisod f a show the new air flow panern. Figuros 12.1,4 and f~ 12.3.4 were tovlsod to show the shloiding for the arosn, and FPP figuro B 7 was revisod to show virr barriors aN COL hose tool locations. The associated NCHs address the changes to the PetDs made by this modification. C.IV,ASDN FOR CHANGFJ This modit' cation was completed to optimize the work!ag environment and !mplomont human factors recommendations. D. SArETY EVALUATION

SUMMARY

k 1) Doos this modification increase the probabl!!ty of occurrone - % consequences of an accident or malfunction of equipment important to safey to provk aluated in the safety analysis report? An1Etc No. The consequsnces of fire, earthquako, and LOCA have boon dotermined not to be any greator \\ as a result of this modification. j 2) Oces this modification croato the possibility for an accident or malfunction of a different type tron any evaluated oroviously in the sshty ann ysis report? AnfESU No. Appropriato provisions woco taken to assure phytical security of the control room complex. Additional electric loa !3 have been cons!dorod in the heating. ventilation, and air conditioning and electrical dotigns. 3) Does this modification reduce the margin of safety as defined in the basis for t*n Technicu Specifications? A01WRU k No. Technical Specification Soction 3.14 was reviewod for making this dotormination. Tablo 13.14.C.1 was changed to reflect the addition of one smoke detector. 3 l

[ PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50 278 1991 10 CFR 50.59 REPORT BemoynLQJ_Cardox Systpq) h%2DIFICATION IF2.J 5041 A. SYSTEM; Fire Protection D..DESCBIEI1Qti; This mc *M emoved the ardox hso reels and m.ated piping, hangers, and Oloctrical equipment from tL some twm. UFSAn cPF Sections P.U, 3.1,3.3, and associated Tables and Figuros will be changod. C IEASON FOR CHANGE.; This modification ollminated a hazard associated with control roorn habitability. D. SAFETY EVAt.LLATION

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equ:pmorit important to safety as previously evaluated in the safety analysis repor;? Answer: No. The probability of a fire is not Inc.aased by the removal of the hose tools. The control room operators and firo brigado are adoquately equipped to liandle a firo in the control room. 2) Does this modification croato the possibility for an accident or malfunction of a different typo than any evaluated previously in the safety analysis report? Answer: No. The ollmination of a redundant fire supprossion system doos not croato the possibuity for a now or differoi9 :po of accident. 3) Does this modification reduce the margin of safety as defined in the basis for the Toch.ilcal Specificaticns? .8D1WMJ No. Technlut Specification Section 3.14 was reviewed for making this determination. 4

m-.,. . - -. -. ~ -. - q PEACH COTTOM ATOMIC POWER STATION - UNITS 2 & 3 - DOCKET No. 50-277 & $0 j 1991 10 CFR 50.59 REPORTr M Turbine Valve Setnoint Chanags jeQQlDCATION 2.r 5085 A. Jy11ggj Turbine B. DESCRIPTION: This modification revised the turbine stop valve closure (TSVC) and turbine-control valve fast closure' (TCVFC) scram bypass setpoints. FSAR Sections 7.2,14.5, and associated documentation will reflect tfme - changes. C. FEASON.R)R OiANGE: This mcdification corrects the TSVC and TCVFC scram setpoint. These changes are based on GE SIL 423. D. SAFETY EVALUATION MIMMARY: 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 setpoint change conservatively decreases the minimum reactor power level for which the TSVC and TCVFC scram function is operab!3 and insures consistency with assumptions in the safety analysis report. 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 TSVC and TCVFC scram continues to function to insert negative reactivity into the rea'ctor in anticipation of the pressurization transient resulting from turbine trip, TSV closure and TCV fast abnormal operational transient, but will now nerform this function over a wider range of reactor power levels. ~ 3). Does this modification reduce the margin of safety as defined in the basis for' the Technical Specifications? j Answer: . No. This is a more conservative approach than the Technical Specifications recommend. Sections. 3.1,3.3.8,4.1, and 4.3 were reviewed in making this decision. l -1 4 I,. 4 5

- ~. F PEACH BOTTOM ATOMIC POWER STATION i UNITS 2 & 3 DOCKET No. 50 77 & 50 78-199110 CFR 50.69 REPOHT - fuse Roolacement. installation._And Class Seoaration ,MQDIFICATION NO.; 5243 A. SYSTEM: 125/250 V DC B DESCHlPTION: This modification replaced fuses that did not meet test criteria, removed redundant fuses, and provided separation between Class 1E and non 1 E electrical circuits associated with cables between 2(3) A(B)D18 and 2(3)A(B)D36, and 2(3)A(B)D18 and 2(3)A(B)38. FSAR Figures in Section 8.7 and FPP Section 6.1 were updated to reflect these changes.. C.JFASON FOR CHANGE: This modification was necessary bacause the fuses failed PECo tests, to create separation between Class - 1E and non 1E, and to ensure the functional requirements of the circults are utisfied. D. SAFETY EVAL.UATION

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 ; I report? InfE2G No. The fuse replacement and fuse block installations'do not affect any existing accident analysis. The operation of other systems required for accident investigation are not affected, and the radiological consequences of an accident are not increased. 2) Does this modification create the possiblilty for an accident or malfunction of a different type than : any evaluated previously in the safety analysis report? - Answer: No. The new fuses have the same failure modes as the_old fuses, but the new fuses have improved - design characteristics. No new failure modes are introduced.

3) -

Does th!s-modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Technical Specification Sections 3/4.9.A and associated Bases were reviewed to make this deterrrdnation. -t 6 1

PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50 278 199110 CFR 50.59 REPORT Establishment Of Sinalo Url!LCooler Onomtio9 fMpjf1C6 HON fC.: 5346 A. SYSTEB; Emergency Core Cooling (ECC) and Reactor Core isolation Cooling (RCIC)

8. pl$QBIPTION:

This modification implomontod a change in systom operational alignment to isolato cooling water and electric power to the standby coolor in each of the ECC system and RCIC pump compartments. Also the valve - positioning for the ernorgency service water (ESW) manual throttle valves on the cooling water outlet lines from all ECC and RCIC room unit coolers, residual heat removal pump seal coolors, and coro spray coolers has been revised from locked throttle to locked open. FSAR Sections 5 and 7, Table 8.5.2F and Figure 10.7.1 reflect the changes made. C. DCASON FOR f>iANGji.; This modification was necessary tc soturn the system to a configuration consistent with tho original design intont of a single cooler operation and will permit system flow balancing to provide the maximum cooling water flow to the operating cooler. The changos will also ensure that maximum ESW flow Is available for components. D. SAFETY EVALUATION

SUMMARY

') 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. None of the changes will have an adverse impact on the operation or maintenance of equipment. The probability of occurrence or consequeaces of previously evaluated accidents and - malfunctions romains unchanged. 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 new accident initiators or malfunctions aro created by these changes. 3) Does this modification reducc the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Technical Specification Sections 3,4. and 6 were reviewed to make this detarmination. 7

-. n -.. c.: 3-PEACH COTTOM ATOMIC POWER STATION UNIT 2 - DOCKET No. 50-277 199110 CFR 50.69 REPORT-NONCONFORMANCE IEPORT.NO,t P-90288 A. SYSTEM; Diesel B. DESCRIPTIONt This NCR addresses the consistency of the diesel enerators and plant installed equipment agreement with design drawings, and the replacement of a 1/2' globe valve with a plug valve (petcock) at the E4 diesel jacket coolant cooler vent connection. UFSAR Figure 8.5.1 has been changed to reflect these changes. C. IEASON FOR CHANGE; This NCR is dispositioned rework to restore consbtency to the diesel generators and to assure plant installed equipment agrees with design drawings. The change in valve type provides for consistency-between the four diesel engines. The other three cooler vents utillze plug valves. Replacement with a 'O' plug valve provides acceptable design, safety and functional characteristics. D. SAFETY EVALtJATION SJMMARY: 1) Does this modification increase the probabi!!!y of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis _ report? Answat; No..This activity did not affect the Intend 9d function of the Jacket cooling or standing AC power systems. ' UFSAR Section 14.0 " Plant Sa'ety Analyt!/' was reviewed. The replacement valve. performs an identical safety function (maintains pressure boundary) as the original valve. 2) Does this modification create the poss!bi'ite for an accident or malfunction of a different type than

any evaluated previously in the safety analysis report?

Answer: No. This actMty maintains the design function of the affected systems ar'd components. 3) Does this modification reduce the margin of safety as defined in the bas,. for the Technical-- Specifications? Answer: No. Technical Specifications Section 3.9 A and C were reviewed in making this decision. l. 8 l g

PEACH COTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 1991 10 CFR 50.59 REPORT & solution Of Diserocancy Between As -Built And Documentation HONCONFORMMICE FEPORT fM.: P-90529 A. ESIEM; instrument AJr D. pCSCRIPTjQff; This NCR addresses drawing discropanclos, identified during a walkdown, on valvos and piping which provide distribution of air. These discropancies included: differing line class representation, piping end connection, and valve type or missing valve locked status. UFSAR Figure 10.17.1 and associated documentation woro revisod to reflect as Installod conditions. C..TASQ14 FOR CHANGE: This NCR is dispositioned to use-as-Is. D. SAFETY EVALUATION

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safoty as previously evaluated in the safety analysis report? Answer; No. This is a drawing change to reflect tho 'as built' plant condition. System function is not affected by differing lino class representation, piping end connection, valve type or missing valvo lockod status. 2) Does this modification create the possibility for an accident or malfunction of a differont type than any evaluated previously in the safety analysis report? Answgg No. This a drawing correction only. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? SIllMU No. The disposition of this 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. 9

.m._ 4 4 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 50-275-199110 CFR 50.59 REPORT: Resolution Of Discreoancy Between As-Built And Documentation. NONCONFORMANCE fEPORT No.: P-90586 A.JiSTEM: Reactor Water Clean-up System B. DESCRIPTION: This NCR is dispositioned to change the SAR Figure 4.9.1 and associated documentation to reflect better. detaR and to add a symbol for a Thermowell. C. FEASON FOR. CHANGE: This NCR is dispositioned to revise documentation. D. SAFETY EVALUATION SJMMARY: 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? Aonwas No. These are drawing changes only and do not reflect any changes to the physical 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: No. The changes make the drawing agree with the as-builiconditionc These changes do not affect the integrity, operabRity, or function of the system in any way, 3) Does this modification reduce the margin of safety as defined in the basis for the Technical: w Specifications? l. -- Answer: L L No. Technical Specification Sections 3/4.6.A were reviewed for making this determination. e s I 10

PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199110 CFR 50.59 REPORT Resolution Of Diserocancy Betwoon As-Installed And Documentation IDRC.QNFORMANCE FEPORT No.: P-90619 A.jLSILMJ Attorrate Rod insortion (ARI) B. DESCRIPTION: This NCR resolved an inconsistency between the 70 amp fuse rating for circuits 29-2303 and 29-2406 shown on Drawing E 27 (UFSAR Figure 8.7.2.a) and the field Installed rating of 30 amps. The circults affected provide DC power for ARI system solenoids. Documentation was revised to show the 30 amp fuses. C. fEASON FOR CHANGEj This NCR is dispositioned to uso-as-Is. D. SAFETY EVAL.UATION

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of an accident ur malfunction of equipmont important to safoty as previously evaluated in the safety analysis report? Answer; No. This is an editorial revision roflocting the correct fuso size as-installed, based on the cable, load size, and protection coordination. A 30A fuse was dotormined to be the correct size. 2) Does this modification create the possibilrty for an accident or malfunction of a different type than any ovaluated previously in the safety analysis report? Answer: No. The fuses are the correct size. No hardware changes have been made. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? AnPE.tG No. The Technical Specifie.ation Bases applicable to this NCR are not affected. Technical Specification Sections 3.9,4.5 and 4.9 were reviewed. ? l 11

l PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199110 CFR 50.59 REPORT Extemal Deslan Pressure For Sucoression Chamber Discrecancy fj[LNCONFORMANCE FEPORT No.; P-90640 A. SYSTEM : Residual Heat Removal (RHR) B. DESCRIPTION: This NCR identified inconsistencies the UFSAR and actual suppr9ssion chamber extemal pressure which is identified as 2 psig but is actually 2 psid. UFSAR Section M.3.2.4.9 and Table M.3.4 will be changed to reflect 2 psid. C. FEASON FOR 01ANGE: This NCR is dispositioned to change the FSAR. D. SAFETY EVAL.UATION

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 report? Answer: No. This is an editorial correction only and makes no physical changes to 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. This is a text change only and makes no physical change to the plant. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Technical Specification Sections 3.7.A ard 4.7.A were reviewed for making this determination. 12

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

1991 10 CFR 50.59 REPORT-

- External Deslan Pressure For Sunoression Diserocancy MLMCONFORMANCE FEPORT No.: P-90640 A. SYSTEM: Residual Heat Removal (RHR) B. DESCRIPTION: This NCR identified inconsistencies the UFSAR and actual suppression chamber extemal pressure which is identified as 2 psig but is actually 2 psid. UFSAR Section M.3.2.4.9 and Table M.3.4 will_ be changed to - reflect 2 psid. C. FEASON FOR CHANGE: This NCR is dispositioned to change the FSAR. D. SAFETY EVALUATION SJMMARY: 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 correction only and makas no physical changes to the ' plant. 2) Does tis 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 text change only'and makes no physical change to the plant. l l.- 3) Does this modification reduce the margin of safety as defined in the basis for the Technical E Specifications? Answer:- No. Technical Specification Sections 3.7.A and 4.7.A were roviewed for making this determination. - ^ i-l- L l i. 12' i e r -u J. -

PEACH BOTTOM' ATOMIC POWER STATION ' UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199110 CFR 50.59 REPORT Resolution Of Discreoancy Between As-Built Plant Condition Nu Existino Documentation BQRQQHEQflM&MQg_gf0RT No.: P-90650 A. SYL* TEM: ' Reactor Core Isolatica Cooling (RCIC) B..QEECRIPTION: This NCR Ident) fled inconsistencies between Functiord Control Diagram M-1-CC-43 sheets 4 and 10 and the as built configuration of the RCIC turbine speed control system. UFSAR Figure 4.7.2D and associated. documentation were revised to reflect the as built condition. C. FEASON IDR CHANGE: This NCR is dispositioned to use-as-is, and revise diagrams. D. SAFETY EVALUATION SJMMARY: 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 3 report? Answer: No. This is a drawing change only and ' oes not alter any equipment or hardware in the' plant. d 2) Does this modification create the possibility for an accident or malfunction of a different typo than any evaluated previously in the safety analysis report? Answer; No. There has been no change to the plant configuration. This is a drawing change only.- 3) Does this modification reduce the margin of safety as defined in the basis for the Technical' Specifications? Answer: No. Technical Specification Sections 3.5D and 4.5D were reviewed for making this determinatlort 13

PEACH BOTTOM ATOMIC POWER STATION - UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199110 CFR 50.59 REPORT Resolution Of Discrecancy Between As-Built Plant Condition And Existina Documentation if0NCONFORMANCE FEPORT No.: P-90725 A. jh'gIgM; Steam Leak Detection B. JE$RRIPTION: This NCR clarifles documentation to identify that no area temperature monitoring exists in the reactor water cleanup pump room. UFSAR Sections 4.10 and 7.3 were revised. C FEASON ibR CHANGE: This NCR is dispositioned to use-as-Is. D. SAFETY EVAL.UATION SUMMAR1; 1) Does this modification increase the probability of occurrence or the cornquences of an accident or malfunction of equipment important to safety as previously evaluated in the safety enalysis report? Answer; No. Plant functional capabilities, as previously evaluated, are unaffected. 2) Does this modification create the possibility for an accident or malfunction of a different type than - any evaluated previously in tha safety analysis report? Answer: No. There are no hardware changes or changes to the operation of installed equipment as a result of these changes. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Technical Specification Sections 3/4.2,6,7, and 8 were reviewed for making this determination.- 14

l PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199110 CFR 50.59 REPORT Correct Discrecancies LONCONFORMANCE FEPORT NO.: P-90762 A. SYSTEMJ Radiation Monitoring B. RSCRIPTION: 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 radlation detectors RE 0-18-30J and RE 0-18@K The Table was revised to reflect as-installed conditions. C. FEASON FOR CHANGX; This NCR is dispositioned to use-as is. D. SAFETY EVAL.UATION

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of en 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 7.13.2 to agree with actual plant installation and system design documentation did not impact any accidents evaluated i 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 previously in the safety analysis report? Answer: No. Changing the location description in UFSAR Table 7.13.2 does not create the possibility of an accident of a different type than previously 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 found to be applicable for the Radwaste Radlation Monitors. 15

~ PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199110 CFR 60.69 REPORT - Resolution Of Discreoancy Botwoon As-Built Plant Condition And Existina Documentation BQRQQNFORMARGE FEPORT PC.: P-90766 A. SYSTEM: Fire Protection B. DESCRIPTION: UFSAR Figure B-2 of the Firo Protection Program was revised to show the 4' fire water piping to the Vendor Building sprinkfor system as a 12' diamotor Main Firo Water Line. The drawing and the associated P&lDs were changed to reflect the as-built condition. C.RASON FOR CHANGE: This NCR is dispositioned to uso-as-Is. D. SAFETY EVALUATION

SUMMARY

1) Does this modification increase the probab.'lity 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 design base function of the fire protection system is unaffected by this drawing change. 2) Does this modification croate the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? Answer: No. This change maintains the design baso function of the fire line. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Technical Specification Sections 3.14 and 6.12 were reviewed for making this determination. 16

.1 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3

^ DOCKET No. 50-277 & 50-278-199110 CFR 50.59 REPORT Resolution Of D12Cf.0Daugyjggeen As-Built Plant Condition And Existino Documentation - NONCONFOBMANCE fEPORT No.: P 90005 A. SY.gIIMJ Residual Heat Removal (RHR) B. IESCRIPTIONJ This NCR Identified the following valves that should be shown as normally closed to reflect the actual:- configuration of the RHR system. HV-210-21572A,B HV-3-10-31572A,B HV-210 21573A,B HV-310-31573A,B HV 210-21018A,8 HV 310 316188 HV-210-21619A,B ' HV 31041619B in addition, the associated nitrogen bottle regulating valve must be shown as normally closed. UFSAR - Figure 4.8.2 was changed to reflect these changes, t C. FEASON FOR_ CHANGEj This NCR is dispositioned to use-as-is and show these valves to be normally closed. D. SAFETY EVALUATION RJMMARY: 1) Does this modification increase the probability of occurrence or the consequences of an accident : ~ or malfunction of equipment important to safety as previoudy_ evaluated in the safety analysis report? .80tWet; No. These valves are used for surveillance testing. This change does not affect the performance of the RHR system.- 2) Doas 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 performance of the RHR system is'not degraded by the normally closed position. 3) - Does this modification reduce the margin of safety as defined in the basis for the Technical: Specifications?

A0swer;

'No. There are no applicable Technical Specifications that specifically address this portion of the RHR system. 17

PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 278 199110 CFR 50.59 REPORT Power Cable Failurg LCfLQONFORMANCE FEPORT Ngh; P-91317 A.SGTEM.; Emergency Auxiliary Transformer B. ESCRIPTION: This modification is an interim repalt which involved abandoning cable 20AX04R(A, B, C phase) in place, and removing a section of Cable OAX04T (A, B, C phase) between MH (manhole) 401 and MH 09 and replacing it with 15kV cable and splicing in MH49. The total number of conductors between the transformer and the bus duct has been changed from 27 to 24. The installation may be reworked in the future pending additional evaluations. C.LEASON FOR CHANGEj Two of the 27 cables connecting the 2 emergency auxillary transformer OAX04 to the non.seg phase bus duct OOA19 failed. D.1AFETY EVALUATIOy_

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of an accident l or malfunction of equipment important to safety as previously Pvaluated in the safaty analysis report? Answer: No. The circuit has the current carrying capacity to supply the actual worst case loading. The original design requirements are satisfied with respect to current carrying capacity and an acceptable voltage drop is maintained. l l-2) Does this modi!lcation create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? A01 ERG No. The use of 8-1/C 1000 MCM cables and added splices has been evaluated and found to meet design requirements of the Auxillary Power system. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Technical Specification Sections 3/4.9 and the associated Bases were reviewed for making this - determination. f 16 .g

I PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 54277 & 547h 1991 10 CFR 50.59 HEPORT - Soolina Water Cross Tie Vrivny KYlGY A. SYSIEM; Emergency Service Water (ESW), Roactor Building Closed Cooling Water (RBCCW) B. DESCRIPTIOR; This revlow addressed the closure of the 33 570 'A" and ?B* ESW te Reactor Buitrung Clomi Cooling Wator (ROCCW) cross tio valvos. This cicsure was performod in 1979 to isolate the non selsmic RBCCW sy6 tem from the seismic ESW to assure the integrity of the ESW pressute boundary during and following a design. basis solsmic event. Reportable Occurrence Report No. 2 79-16/IP documented this change. UFSAR Sections 5,7, and 10 were changed to reflect this closure. C. RA ON FOB _CHAN91; J To document the adoquacy of the design configuration change in accordance w}th the cnhanced 10CFR50.59 Review mquirements. D.1AFETY EVALUfgIJQMQMM&B2t 1) Does this modt!! cation increase the probabDity of occurrenco or th6 consequences of ari accident or malfunctJon of equipment important to safsty as previous!y ovoh.iated in tho safety analysis y report? At19xec No. These accidents which are potentially negatively impacted by the chango are accWnts or events that involvo a loss of Offsito Power (LOOP) or Loss of Coolant Accident (LOCA). Ali systems that could be impacted by this chango waro indMdually evuluated it was determined that no increase in the probability of an accident or maltunction was created. 2) Coos this modification ciaato the possibility for an accident or malfunct!an of a different typo than any estuand previously in the safety analysis report? An1EtrJ No. Systems that could be impacted by thh modif; cation were indMdually evafnatd it wan-determined that the modification does not creato the possibi!ily of a diffemn! type of accident or maifunctsn. 3) Does It'Js moMication reduce the margin of safety as defined in the basta for the Technical Specifications? An3Ett No. Technical SpecificaDon See-3 and 6 wwo teWuwed for rnaldng this determination. v

PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 A 50-78 1991 to CFR 50.59 REPORT MD!Yftl.'DfAD..ChT1LW fulnY ? A. j;MIEM.; Emergotry Servico Water (USW) D. pg.S.Q,R!EUptt; This change closed the bypasa valve arourd the alt operated valves upstream of each insoMee operating Emergoney Coro Cochng Systorn (ECCS) and P,oactor Coro lejection Cooling (RCIC) System room cooler. C. JEASDB.fplLCIWiGC This temporary change is invokal when the ESW Chemical In>ction System is looperablo. D. fiAfEIL.tNe1.UAl10N,SJRtMRY.; 1) Does this rnodification Irv; tease the probability of occurrence or the consequences of an accident - or malfunction of equipment important to safety es previovoly ovaluated in the safety analysis report? AlttMG No. The posillon of those valvos has to hanring on the probabtity of an accident. The temporary linoop will not afIoct thre ESW oystom's capabiitty to wpport ECCS ar4 RCIC equipment if it is calltd on to mP.1 ato the consequonias of an accident. 0 2) dom this modification create the possibMty for an accidant or malfattetton of a citforunt typa than any ovaluated pmvlously in the ufoty analyals report? Atisma No. Thb changs does not at;'oct the ESW systom's tapability to mpport ECCS and fiCIC squipment !f it is necessary for mitigation of the consequencos of an accident. 3) Doca th!s inoclifice:bn reduca the margin of safety as defined in the basis for the Tochaical SpecNications? AG1MU No. Technical Specification Sections 3.5 and 3.9 were reviewed for making tNs dctormination. l 20 j -. l

t-i- PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 ?77 & 50-278 1991 10 CFR 50.59 REPORT E0!1aldtDEtsMIAUZiG fl1QCIQuatg A048D.! and AO.380 2 A.firHJJM; Domineralizer B. LX1CBJg[(ph; This rodston providos instruct'.ons for Zhu use of a traitored online domineralizor. C. LW.A301(JQfLQfM19E; A tempomry poru& dentrera lzac and teller were insta!!od. D..WE!1.RYL4Mh1\\DHD)M$hSY1 1) Does th!s mod!flcation Mr.roaso the pobability of occurrenco or the consequences of an accVent or malfunction of equipment important to safety as previously evaluated in the safety analysis report? b.DDMD No. Water uuality is monitored and maintained to the samo specificifications as the existing system. 2) Does thts modification create the posalb!!Ity for an accident or malfunction of a different type than i any evaluated previously in the safety analysis report? Autun No. Equipment important to safety la not impacted by the use of a portable domineralizer. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? AG.PMG No. The Makeup Domin system is not mentioned in the Technical Specifications. 21 L

c; - PEACH COTTOM ATOMIC POWER STATION-- UN'TS 2 & 3 - DOCKET No. 50-277 & 50-278-199110 CFR 50.59 REPORT-Manual For Clearance And Taaalna - PRODEEJ185 - Clearance and Tagging Manual A. SYSTEM: Performance Monitoring B. ESCRIPTIOR; implementation of a Clearance and Tagging Manual to provide methods of protection for personnel and plant equipment during operations, maintenance, and modification activities. C. FEASON FOR CHANGE: This change is necessary because of the changes instituted with the new computer. --D.' SAFETY EVALUATION RIMMARY: 1) Does this modification increase the probability of occurrence or the conscquencec of an accident : or malfunction of equipment important to safety as previously evaluated in the safety analysis report? Answer: No. This procedure will only be used for Load Dispatcher's Permits, No accidents are described - in the FSAR which involve this procedure. 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 change is administrative only. No changes were made to the levels of control of ~ equipment; 3) Does this modification reduce the margin of safety as defined in the' basis forthe. Technical.- Specifications?- ADAntti No. : Technical Specification Section 6.8 was reviewed to make this determination,' ) \\ O =. ha

' 4 a a muses UNIT 2 l b 23

[ PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50-277 1991 10 CFR 50.59 REPORT Elant Monitorina System fPMS) Installations LAODIFICATION NO,: 955E A. SYSTEMJ Plant Monitoring B. DESCRIPTION: This modification Installed raceway, cable, cable terminations, loop-taps, and class 1E multiplexers for the new Plant rAonitoring System (PMS) computer, in addition, this modification removes some temperature elements for monitoring secondary containment area tomperatures from the Q List. UFSAR Sections 4.8, 9,10; 7.1, 2,3,4,5,8,12,13,16,19,20; 8.5; 10.9; 17; and 11.3,15,6 and associated documentation were changed to reflect the changes. C.EMON FOR CHANGE: This modification was necessary to make the circuits compatible with the new PMS. D. SAFETY EVAL.UATION

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 report? A_nswer: No. For equipment that is important to safety, qualified isolators will be used to isolate the non-safety related PMS from the safety related circuits. The signal isolators being added are more reliable than other components in the loops and do not change the way the 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? l AnsweG No. The PMS input taps do not change the way the instrument loops function or Introduce any new failure modes. Temperature elements deleted from the O List do not affect the capability to safely shutdown the plant in case of fire. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? AGRERG No. Technical Specifications Sections 3.5C,4.5C,3,7D,4.7D and associated bases were reviewed and are not affected. 24

.._ ~ .] - C i j - PEACH COTTOM ATOMIC POWER STAT 10N

q UNIT 2 -

N DOCKET li>. fl0-277 - - ~ 1991 10 CFR 50.53 REPORTL installation Of Safety GradgE23uppl.y MODIFICATION No.t-1310 s A. SYSTEM t Primary Containmeru, Containment Atmosphere Control (CAC), Containment Atmosphere Dilution (CAD)' B. DESCRIPTION: Installation of a cafety grade N2 supply to the containment isoWlors wives' and an upgrade of CAC and CAD - components, including work on mechanical piping and tubbg. w ves, fittings and switches. e C %$3ON FOR CHANGEJ This system will rernove the need for the bottled gas supply, and ins associated requirement to estabilsh ! and maintain air supply systen leak rate cilterla for indMdual useic, arU t#11 allow verification of adequate ' ' system performance by functiona4 testingc The upgrade was also done to assure compliance with 1.Ec Bulledn 7941B and Regulatory Guide 1.97. D. SVETY_Ey4!,yAllQtL1UMARY,; 1) Ooes tNs modification increase the probability of occurrence or the consequences of an accid 4nt i or malfunction of equipment important to safety as previously evaluated in the. safety analysis-a report? _ Answer: - -No. - Desl n requirements applied to tNa modification include l but are not limited to,-seismic 0 yaalification, L quality assurance, testaulty,i maintainability,, origbal equipmentgnerformance ; i' specification, erwironmental _ qualificatie, separation critoria - safeguard ' power Lsources,1and protection from jet impingement. t a 2): Does tNs modificatiottereate the possibility for an accident or malfunctioniof a different type thani 3sj; any ovahlated prevk usly in the safety analysis report? f Aapsac I .s No. iThe criteria used as a basis for the protection and evaluation of one plant component tsm; adverse interactions of another plant compormnt during various operating condi' ions comply with' the UFSAH arid ANSI standards.- ~ ~ ~

3) ~

. Does this modification reduce the margin of safety as defined in the basla'for the Technical Spedications? Answen! - No,' Technical Specifications Sections 3.7 and 4.7 have been reviewedc Allload stress acceptance. criteria for all components evaluated and/or repaired by this work comply with the original design L code requirements specified in the UFSAR or as a!! owed by ASME Soction Xl; 25 .,J'l[-l 5 t-4

e . _ _. _ ~ q-m_ PEACH BOTTOM ATOMIC POWER STATION ~ UNIT 2 ' DOCKET Ho. 50 277-1991 10 CFR 50.59 REPORT: .. Drvwoll Cooler Fan Loalg MODIFICATION No.t 1542 A. SYSTEM: Drywell Coding B. DESCRIPTION: This modification changes fhe control logic of drywoll coolci fans 2AV26 through 2GV26 to bypass the high drywell pressure or low te;.ctor water level trip signal. C. FEASON R)R CHANGE: This modification implements the NRC recommendation to improve the plant recovery from high drywell-pressure (above 2 psig) per 1.E. Information Notice 84-35. D.JAFETY E%UATION

SUMMARY

1) Does this mWication increase the probabiltty of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report? m Answer: Ths probabulty of occurrence or the consoquences of an accident or malfunction of equipment - -{ important to safety previously evalmisd in the safety analysis report has not been increasedf Fagure of the new bypass ekcult may result in increased loading on the emergem',y buses, however,y the additional load will not exceed the capacity of the d!osel genemtors. The equipment added by 1 this modification has been located.in safety related panels and has been mounted in a simRar. manner to that of safety related components to prevent @rnsge to the safety related equipmerttj inside the pa'nel. q 2) Does this modification create the possibatty for en acektent orynalfunction of a different type than j any evalmted previously.in the safety analysis report? :

i AalMIG The possibility for an accident or malfunction of a d& rent type than any evaluated previously in the :

safety analysis report has nolt been created. Fafuro of the bypass circuit does not result in the : j placernent of unacceptable loads on the emergency bus or its associated diesel generator. This; a change allows the operator to restart the dryws!! cooler fans from the control room to improve the : plant recovery frorn high drywell pressure? 3); Does this modification reduce the margin of safety as defined In the basis for the Technical; l .y Specificatharis? 1 Ante!rG' The margin of safety as defined in the basis for the Technical Specifications has not been reducedc Technical Specification Sections 3/4,9 were reviewed in making thFa detem11nationJ ~ M li ~ 1 ) /.. i 1-N y' v m

._.m __.m_. c ~ PEACH BOTTOM ATOMIC POWER STATION" - UNIT 2 - DOCKET No. 50-277 - 1991 10 CFR 50.59 REPORT . Automatic Sorinkfor Extension for Turbino/ Generator Bearinas And Front Standard MODIFICATION No.: 1832 A _ SYSTEM: Sprinkler B DESCRIPTION: This modification provided wet pipe automatic sprinkler protection for the turbine bearings, generator bearings, and front standard, and removed the dry chemical piping and rh.zzles. UFSAR Section-10.2.3.4 and FPP Sections 2.2, 2.9, 3.3.1, 5.3.40, and associated tabics and figures were revised to show these: modifice*%ns. C, HM1CW TAM 91.; This modification was an enhancement to comply with Insurance (ANI) recommendations. D. SAFETY EVAL.UATION RIMMARY,;- 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.' There is no equipment important to safety affected by this modification. The modification does j not require a revision to the analysis of safe shutdown in the event of a fire, thereforn,- the-l. - modification has no effect on the probability of occurrence of any postulated accident. r 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? AMEtG No. This modification provides automatic sprinkler protection. It does not affect any safety related or safe shutdown equipment.

3).

~ Does this modification reduce the margin of safety as defined _in the basis-for the Technical-l Specifications? Answer: I _No. - Technical Specification Sections 3/4.14 were reviewed for making this determination. t' !y 27 h - -m

il ~ PEACH COTTOM ATOMIC PCWER STATION : w UNii 2, DOCKET No. 50-E77 1991 to CFR 50.59 REPORT q Roolace Existina Flow Transmitters MODIFICATION NO.: 1891 A. SYSTEM: Various B. IESCRIPTIOtu This modification replaced existing flow transmitters with environmentally qualified flow transmitters in various systems. It also replaced existing signal conditioning Instruments associated with these flow transmitters to accommodate the new transmitter with 4 20 ma output. Appropriate UFSAR Figures in Sections 4 and : a 7 were updated to reflect changes made by this modification. C.MSON FDR CHANGE: This modification was necessary to meet a Reg. Guide 1.97 commitment. D, SAFETY EVALUATION ElMMARY: 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 analysts report? Answer: No. This modification replaced existing flow transmitters with environmentally quallfled flow : transmitters in various safety systems. This modification did not change the operation of the safety. H systems involved. 2) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysts report? Answer: No. This modification does not change the operation of the affected system and has a negligible . effect on loop accuracy and reliability. It does not introduce any new accident initiators. 3) Does this modification reduce the margin of safety as defined _in the basis for.the Technical-Specifications? Answer: -: No.. Technical Specifications Sections 3.4.2 and 3.5.A. 8, C, and D were reviewed in making this determination. 28 t

~ O l PEACH BOTTOM ATOMIC POWER STATION l; UNIT 2 DOCKET No. 50-277 199110 CFR 50.59 REPO?tT ApM J Modification fd?OlFICATIO]LflQJ 2075 A. JiyjilfjL Primary Containment D. QtlAQRIEIqt TNs modificatton added block valve an6/or test connections to ensure local leak rato lrsts (WP) of containment isdation u.lves are performed in full compliance with 10CFR50, Appendtx J. This modification also reverses manua) block valves between the primary containment and the inboard containment isolation vabo to include valve stem packing in the boundary and modifles twenty-three penetrations. C. FEASQN [WLOMHGE: This mcdification was dono in response to inspectbn Report 50-277/85-21 and 50-278/85-23. D. jp2TY EVALUATIQHJUMMA9X; 1) Does thb modification increase the probability of occutTence or the consequences of an accident or matfunction of equipment important to safety as previously evaluated in the safety analysis report? Actwar; No McdificaOon of the containtnent isolat!on boundary will allow a more conservative Local Leak Rate Tmt for the pawtrations sinco previously untested valve stem packing will now be included, it is now more likely 2.at packing leaks w13 be detected by routine testing and ropalred, lowering the probability of radiation rel ease outside cortainment. 2) Does this modificatbn create the possibility for an &ccident or malfunction of a different type than any evaluated previously in the safety analysis report? All1ECG No. The design of the systams affected by this modification are essentially unchanged with the execption of an enhancement of the leak rate testing of the containment boundary. No new acekjent initiators were introduced. 3) Does Ws modification reduce the margin of safety as derbed in the basis for the Technical Specificatirms? AnswtG No. Techn! cal Specification Sections 3.7 A and D were revlowed for making tNs determination. i i 29

e-PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50-277 199110 CFR 50.59 REPORT Intia!!al!Qn2f.ustsAttd.f.rafl!ach ACDIE1CKi10tt1DJ 2200 A.LUIIMJ 250 & 125 Volt B, DESCBJEIl010 This modification installed fu o and fuse tdocks in the bus undervcitage rolay circuits contained in the 250 VDC motor control contV 10D08 and the 12S VDC distribution panols 20D21,20D22,20D23,20024, a*d g s 20073. UFSAR Figuros 8.7.1a,0.71b, ard 0.7.2b wi,i be revised to show thoso changos. C. B21Q11RILCtL'dtGEJ This ri.odification was made so that the undervoltago rolays can be testod and calibrated without a safeguard bus outago. D. !MFETY EVALUA110fLSUMMAtl1; 1) Does this modification increase the probebi!!!y of occurrence or the consequences of an accident or malfunction of equipment Important to safoty as previously evaluated in the safoty analysis report? AntWec No. The function of the DC system is unchanged. It will make testing and calibration easier, 2) Does this modWication croato the pos lbility for an accident or malfunction of a utfferent type than any ovaluat6d previously in the safot, ahalysis report? AntERG No. Those fuses and fuse blocks are quallflod for the w~st-caso environmental and Solsmic condition for tholt locations. This modification did not introduce any new failurn modos. 3) Does this modification reduco the it'argin of safoty as definod in the basis for the Technical Specifications? AnlHfD No. Technical Specification Sections 3/4.9.A were reviewod for making this dotormination. 30 .=__

o PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 54277 199110 CFR 60,59 REPORT gmergeogy.Sanic.g Water (ESVoflalA23c!peation Ard Chanoes WJDif1CM101LBQ4 5040 A. SYSIEMJ Emergoney Service Weier B. DESCn!PT10B; This modification and its associatod Engincoring Review Roquest Form roplaced and rotocatod the ESW ring ^ headore end associated pipbig in the Torus room and provided ting heador segmontation valves and plug valves with position indicators for maintenance purposos. To completo the modificatlan temporary benacning of steam / flood barriors was required. Testir.g was per ormed to determino flowrates and the plug d valvos may be left in the throttled position during optrations. Appendix A of the FSAR ks boon changed to Indicato codo updatos, valve typo

• and to update valvo positions. Associated Chock L., Usts have boon g

changod. C. [EASQH FOR OWJQE; This modification was necessary becauso u problems with piping corrosion and flow balancing of the ESW system. D. }AFETY EYALUAT'Q![jUMMARY: 1) Does tha modification increase the probability of occtNonce ol the consequences of an accident or malfunction of equipment important to safety as previously evaluuM in the safety analysis report? Anuer: No. This modtfica:lon makes no functional change, nor introduces any now hazards, 2) Does This modification croato the possibutty for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? AD.lERG t'o, Tla modification makes no functional changes to the system and Introduces no now hazards. The tredlfication does not create the possibility of an accident of a differont type than any evaluated in the Safc./ Analysis Roport. 3) Does this modification reduce the margin of safety as defined in the basle for the Technical Speciftcations? A01 ERG No. Technical Specification Sections 3 5 and 3/9 were reviewod for making this detentbation. 31 77

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 54277 199110 CFR 50.59 REPORT Emcigency Sorvico Water (ESW)/Emeroency Coolina Water (ECW) Flow Instrumentation Changel IdQDif1 CATION PC) 5110 A. Sf.111M; ESW/ECW D. DESCRIPTION: This xxilfication replaced a soction of the carbon stool plpo with a flangod stainless stool spool ploco and - provklod flow measuring devices. FSAR Sections 10.9 and 10.24 woro rev! sod to show these changes. C. IEAS9fdD!LQJAt(GRJ This modification was necessary to comply with ASME Xl requirements. D. SAFETY EVAL.UATION

SUMMARY

1) Does this modification incroano the probability of occurronce or the consequences of an accident or malfunction of equipment important to safety as provlously evalested in the safety analysis report? Answe.c No. This modification was installod in accordance with the original design basis. 2) Does this modification croato the possibility for an accident or malfunction of a diflorent typo than any evaluated previously in the safety analysis report? Ariswic No. This modification does not affect any current event initiators or introduce any now event initlators. 3) Does this modification reduce the rnargin of safety as defined in the basis for the Technical Spocifications? Answer: No. Technical Specification Sectiont 3.9.C and 3.11.B were reviewed in making this determination. 1 32 J

.n.-- PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET !!o, 50 277 199110 CFR 50.59 REPORT binnaio Food To Batterv_Chmg9rs Arv) 120 VAC Distribution Panell i 1MOH.lCARQEDQA 5209 A. SfSI m ; 4kV B. Ill1CHIE110HJ This modification installod attornato power foods for battory chargors 2AD03,2BD03 and 120 VAC distribution panois 20Y33, 20Y34, 20Y35, OOYO3. Changes were inado to FSAR Sections 8 and 9 as well as Appendicos F arKI J, and associated documentation. C.IEMQN fDR QiMQfj This rnodification was necessary to resolve two outstanding problems associated with maintalning qualified control power to the common diosol genorators, their associated 4kV omor00ncy busos, and to Unit 2 instrument AC panels when performing tests and maintenanco on DMslon I or 11 equipment during a Unit 2 cold shutdown or refueling mode. D. SAFETY ENALLIAHQH,j)MtiMy,; 1) Does this modification inercaso the probabl!!!y of occurronco or the consequences of an accident or malfunc*lon of equipment important to safety as provlously evaluated in the safety analysis topct17 An1 ERG No. Thoro are no specific UFOAR Chaptor 14 accident analyses that discuss taking crodit for the 125/250V DC system and the 120V AC instrument power supply systems. The modification does not alter any assumption mado in ovaluating the radiological consequences of an accident. 2) Doot this modification create the possibility for an accident or malfunction of a different typo than any ovaluated previously in tho safety an4ysis report? Antwea Nc. This modification does not chango the function of the battery chargors or the instrument AC r.cnols 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Anants No. T0chnical Sponification Soctions 3.9 arvi 4.9 wort reviewod for making this determination. 33 $P mm -ym + g

==-se

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50-277 199110 CFR 50.59 REPORT Leyff Switch Chanoo fdQDIDCATIOfLth; 5240 A.St.SIIR; Roactor Coto it,dation Codirig B.DgGCBIEI1Qtt; This modification removed levol switch LS 2-13 74 and replaced it with a similar level switch. A minor change to the drain lino piping was nocessary to accommodate the now switch. FSAR Figure 4.7.1 A reflocts those changos. C.1EASQB FOR CHANGE.; This modification was nocessary because the existing model level swMch was obsolete. D. SAFETY EVA.LtLATION SUMMAR1; 1) Does this modification increase the probability of occurrence of the consequences of an accident or malfunction of equy> ment Irnportant to safety as previously evaluated in the safety analysis report? ALr1Lugg No. The now switch performs the same function as the original switch. 2) Does this modification croato the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? ADIERG No. The now levol switch moots the camo design critoria as the original switch. 3) Does this modification toduce the margin of safety as definod in the basis for the Technical Specifications? ARILW2G No. Technical Specificat'an Sections 3.5 and 4.5 woro reviewed in making this dotormination. 34 )

e FEACil COTTOM ATOMIC POWER STATION UNfT 2 DOCKET No. 50-277 199110 CFR 60.59 REPORT Epsoludon Of Discronancy Between As Bullt And Documentation fMfgQNfDRMANCE FEPOffi NQ; P-90214 A. jitgIjE Nucioar Bollor Vessd Instrumentation D. DiljiCRIP110N: This NCR addressed valvo numboring changos, correction of drafting errors, and makes clarifications to existing documentation. P&lD M 52 shoots 3 & 4 and SAR Figure 7.3.1 will be changed. C.NASDN FOR CliffGEj 1 This NCR is dispositionod to ur.9-as-Is. D..SArETY EVALV.A. TION

SUMMARY

1) Does this modification increase the probability of occurrence or the consequentos of an accident or rmlfunction of equipment important to safety as prs /lously evaluated in the safety analysls report? hSEEtti No. Thoso are drawing changes only and do not reflect any changes to the physical plant condition. 2) Does this modification croato the possibUty for an accident or malfunction of a different typo than any evaluated previously in the safety analysis report? Answer: No. The changos provido a greator levol of detall and clarity si the PalD representation of the plant's actual as built condition. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Technical Spocification Section 3/4.2 was reviewed for making this determination. l l l 35

e PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 60-277 199110 CFR 50.59 REPORT Epsolution Of Discronm1cyJetween As Built Arv.f Documentation MRCQH[ORMANCE FEPORT PC.: P-90210 A. SYSIER; Roactor Water Cloan-up Systom B. (EscalimQR; This NCR addrossos inconsistenclos betwoon the as-built condition of the roactor wator clean-up systom and documentation. PalD M 354 Shoot 2 and the SAR Figure 4.9.1 shoot 2 were changed. C.JEASON FDR CHANGIJ This NCR ls dispositionod to uso as-is. D.RE.ETY EVALUATION SijMMARY: 1) Does this modification increase the probabCity of occurrence or the consequences of an accident or malfunction of equipmont important to safety as previously eva!u: od in the safety analysis report? AillESC No Those are drawing changos only and do not reflect any changes to the physical plant condition. i 2) Does this modification create the possibl!!ty for an accident or malfure..:'on of a different type than any evaluated previously in the mfoty analysis report? An1WE No. Those changes do not affect the integrity, operability, or function of the system in any way. The consequences if a SAR ovaluated occident are unchanged, 3) Does this modification reduce the margin of safety as definod in the basis for the Technical Specifications? AD.anto No. Technical Specification Section 3/4.6.A werc reviewed for making this determirmtion. 36 l

[ i PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 199110 CFR 50.59 REPORT ResoMion Of Discreoancy Between As Bullt And Documentation RQR.qQNfDBMANCE FEPORT to.t P-90353 A.EfAIEM; Emergency Service Water and High Pressure Service Water B. DESCRIPTION: This NCR identiflod inconsistenclos betwoon docuinentation and actual valve position. The dominerallzod water isolation valvo HV 2 380-47074A to RHR heat exchanger 2AE024 should not be shown as locked closed on UFSAP figure 10.7.1, sht.1 of 4 and associated documentation. Documentation was revisod to 1 rofloct those changos. This change is consistent with that of the romalning Unit 2 RHR hoat exchangers and is in accordance with the Unit 2 locked valve list. C. FEASON FOR CHANGE; The NCR is dispositioned to uso-as-is. D. SAFETY EVALUATION SUM _ MARY: 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 topon? ARREtG t No. Doloting the drawing notation that the valve is locked closed has no effect on the consequences of an accident or malfunction of equipment as previously evaluated in the SAR. 2) Does this modification create the possibi; tty for an accident or malfunction of a differont type than any evaluated previously in the safety analysis report? Answer: No. This drawing change does not offect the safety (ola'od function of the valve and does not prevert, other safety rolaiod systems or equipmer't from fulfilling their safoty related function. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No Technical Specification Sections 3.5 and 6.8 were reviewed to make this determination. I 37 1

-i-PEACH COTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 199110 CFR 50.59 REPORT Resolution Of Descrocancy Between As-Built And Documentation NONCONFORMANCE FEPORT ND. P90376 A. SYSTEM; Radiation Monitoring System B. DESCRIPf10N: This NCR involves a drawing discrepancy and is dispositioned to change the UFSAR Figure 7.12.3 description of G3L 194278 from a globe valve to its as-Is configuration of a gate valve. C.HMSON FOR CHANGE: This NCR la dispositioned to use-as.ls. D, SAFETY EVAltJATION RJMMARY: 1) Does this modification increase the probabRity of occurrence or the consequences of an accident - or malfunction of equipment important to safety as previoudy evaluated in the safety analysis - report? Answer: 3 No. 63L.194278 provides a system pressure boundary and does not impact an accident or malfunction important to safety as previously evaluated in the SAR. 2) Does this modification create the possiblity for an accident or malfunction of a different type than any evaluated previously in the safety analysis report? Answer: No. This is a description change only. Since the probability of failure of a gbbe valve is the same - as that of a gate valve, tha failure of 63L 19427B will not cause an accident of a different type than previously evaluated in the SAR. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?. Answer: - No. Technical Specification Sections 1.0,3/4.P. 2/4.14 were reviewed for making this determination. i-l' l- ' 38 1 b. J

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50-277 1991 10 CFR 50.59 REPORT Resolution Of Discrecancy Between As-Dullt Plant Condition And Existino Documentation p HQRQQNFORMANCE FEPORT NO.: P 90599 l A. EYJlfMJ Rocombiner D. DESCRIPTIOt#: This NCR Identiflos inconsistenclos betwoon documentation and actual configurotion of the off-gas recombiner $1oam supply drain piping installed correctly, but not shown correctly on UFSAR Figures 9.5.1, 11.2.1, and associated documentation. C. IEASON FOR CHANGE: This NCR is dispositioned to rovico documentation to show correct configuration. D. SAFEiY EVALUATION GUMMARY: 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 safoty analysis report? Antwsc l No. This NCR makes no physical changos to the system. 2) Does this modification croato the possibility for an accident or malfunction of a different type than any evaluated precusly 10 the safety analysis report? l Aointu No. This NCR changos documentation only. 3) Does this modification reduce the rnargin of safety as defined in the basis for the Technical Specifications? Aritwec No. This portion of the gas recombiner system is not addressod in the Technical Specifications. l 39 N

h PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 54278 1991 to CFR 50.59 REPORT Reiglution Of Dhcreonnev Betwoon As Bulit Plant Conaltig And ExistinaIlocumentatim R2tlCONFORMAtiCLIEPORT WL; P-90789 A. SYSTEM: Firo Protection D. St1CBIEIt0R; This NCR indicated that a wot standpipo in the turblno building is connected to the 10 Inch fire main upstream of HR 116-19 only. The FSAR and associated P&lD showed the standpipo to be connected to the main in two places, UFSAR Figure B 2 of the Fire Protection Program and associated documentation were revised to reflect the as Installod condition. C.RMON R>n QlM9g.) This NCR is dispositioned to uso as-is. D.faAFETY BfAt.tlATION

SUMMARY

1) Does this modification increase the probabliny of occurrence or the consequences of an acckjont or malfunction of equipment important to safoty as previously evaluated in the safety analysis report? Answen No. The possibility of a fire or lino break in the turbino building is not increasod. 2) Does this modification create the possiblitty for an accident or malfunction of a different type than any evaluated previously in the safety analynts report? hG5EtG No. The drawing chango does not croato the possiblilty of an accident or malfunction of a different typo than previously evaluated in the SAR. 3) Does this modification reduce the margin of safety as definod in the basis for the Technical Specifications? _ Answer: No. Technical Specification Sections 3/4.14 woro reviewod for inaking this dotormiration. 40 9- * - --*

m

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~ e PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50-277 199110 CFR 50.59 REPORT Resolu'lon Of DistwLqogyjotween As Bulti Plant Condition And Estina Documentation 1MLCD1flQEMMCE FEPORT NO.: P 91159 A. SGIr&' High Prossure Coolant injection D.L U CRign0R; This NCR involved drawing changos to accurately reflect circuit breaker sizo and replacement of a 25 Amp circuit broakor with a 10 Amp circuit breaker, UFSAR Figure 8.4.0, and associated documents wero rovised. C.IEAlatf FDR OiANGI; This NCR is dispositionod to reviso drawings to show this as insta!!od,10 Amp circuit breaker for MCC 20B30 compartmont 52 3614 and to replace the Installed 35 Amp circuit breaker in MCC 20B37 compartment 52-3752 with a 10 Amp breaker. D. $AfgIYJ/AWAHQN

SUMMARY

1) Does this modification increase the probability of occurronco or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report? Answer: No. The intended design function and operation of the plant Electrical Distribution system and the and the High Pressure Coolant Injection system are not affected by this modification. 2) Does this modification croate the possibility for an accident or malfunction of a different type than any evaluated provlously in the safety analysis report? Antwin No. The failuro modos of the 25 Amp and 10 Amp circuit breakers are the same. No now accident initiators were introducod. 3) Does this modification reduce the rnargin of safety as defined in the basis for the Technical Spocifications? Ant.W90 No. Technical Specification Sections 3./4.7,3/4.9 and associated bases were reviewod for making this dotormination. l i 41 t l^

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 199110 CFR 50.59 REPORT J,.mt Parts in Reactor Vestel .t*2flGONf0RMAllGE_BEPORT NQj P-91193 A. SfSIER; Steam Separator D.[ncmEnow Whi!o removing the lockpin to manipulato the steam separator latching tool, the lanyard cllp became dotachod from the lock pin and foil into the reactor vossol. The lost part analysis ovaluated the offect on safo roactor operations of the lost part and the cumulativo offect of the clip and previously lost parts. C. IEASON FDR CHANGE: This NCR was dispositionod to havo GE perform a full lost parts analycis to dotormino a use-as Is assessmont because tho SAR assumos that lost parts do not oxlst. D. SAFETY EVALUATION

SUMMARY

t 1) Does this modification increaso the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as provlously evaluated in the safety analysis report? AntMD No. Tho lost parts do not intorforo with control rod motion or fuol bundle coolant flow. The lost objects do not creato a chemical or corrosion concom. 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? Annus No. The lost objects will not block control rod operation, cause fuol bundio flow blockage or, cause damage to reactor Internal components. The ability to safely shutdown the plant is maintained. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? Answer: No. Reactor water chemistry limits and roactor performance will be unaffected.- The Bases for Technical Specification Sections 1.1, 3.3A, 3.6A and 3.6B were reviewed in making this determination. 42 L i

s PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 60-277 199110 CFR 50.59 REPORT Resolution QLQhCgoancy Between As-Built Plant Condalon And Existina DocumentatioD }MfCQHf0RMANCE FEPORT 70.; P-91283 A. SYSTEM: Residual Hoat Rornoval (RHR) B. L M QBLEil014i This NCR addressed the replacement of a 480 VAC Westinghouse typo HFB circuit breaker fooding the RHR recirc loop A roturn valvo MO-210-25A The breaker trippod due to high inrush current experienced during valve operation. Testing of the broakor found the breaker tripped at a much lower current on the A phase than on the B and C phases. In addition, the instantaneous trip setting of the breaker was not set high onough to provent the broakor from trippin0 as a result of high in-rush current. C.1EASON FOR QiAllQ1; lhls NCR is dispositionod to replace the breaker to provent tilp on in-rush current, and increase the Instantaneous trip sotting. D. SAFETY EVAt1AllQN _

SUMMARY

J 1) Dcos this modification increase the probabl!!ty of occurronce or tho consequences of an accident or malfunction of oquipment important to safety as previously evaluated in the safety analysis report? Answon No. The breaker was replacod with a functionally equivalent one. The chango on the instantanoous trip sotting of the now breaker to provent inadvertent tripping will have no offect on any accident identWlod in the SAR. The RHR system functions have not been changed. 2) Does this modification create the possibility for an accident or malfunction of a different typo than any evaluated previously in the safety analysis report? Ant)MD No. The now breaker will perform the same function as the original breaker and the ability to open and closo valvo MO-210 25A is maintained.

3).

Does this modification roduce the margin of safety as defined in the basis for the Technical Specifications? P An1ESU No. Technical Specification Sections 1/2.2 and 3/4.5 woro reviewod for tr aking this determination. 43 y .m um. .me--ie e g ->.-.s 2.~-.-..-3i,. e w w s ,m., s,a +-p

PEACH OOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 199110 CFR 50.59 REPORT iC' Channel Trio JUdEQRARY ALTERATIOPe PC.: 21347 A. 9(STEM; Roactor Core Isolation Cooling (RCIC) D. IESCRIPIlONJ installat!on of a jumpor to insert 'C' channel trip. C. RA'3ON FOR CHANGE: The Technical Specifications state that when the number of operable channels is less than required the channel should be placed in the tripped condition or RCIC should be dociarod inoperable. D. SAFETY EVAWATION RLh1MAfD'; 1) Does tias modification inctcase the probabi!!!y of occurrence or the consequences of an accidont or malfunction of equipment important to safoty as previously evaluated in the safety analysis repott? A!11HfG No. This configuration is safe because it is more conservative, 2) Does this modification croato the possibility for an accident or malfunction of a difforent type than any evaluated previously in the safety analysis report? AlltEtG No. This is only a jumper to initlato a trip. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? All1H9D No. Technical Specification Soction 3.2 was reviewed for making this determination. r 5 ? U

PEACH BOTTOM ATOMIC POWER STATION-UN!T 2 DOCKET No. 54277 199110 CFR 50.59 REPORT .Qefeat Of (ms OfPower Alarm .IEMEDRARY PLAffT ALTERATIOR; TPA 2 2347 A. SYSTEM: High Prossure Coolant injection (HPCI) D. DESCRIPTION: This TPA was insta!!od to defeat the HPCl gland soal blower DC motor power loss alarm. C. EEASON FOR CHAEGEj ] The undervcitage relay for the gland seal blower motor had failed and was bringing up an alarm. masking any problem with critical equipment. D. SFETY Ev6LLIATION SUMMAB1; 1) Does this modification increaso the probability of occurronce or the consequences of an accident or malfunction of equlpment important to safety as previous.ly evaluated in the safety analysis report? AD1WPG No. Undervoltage. conditlans on the HPCI gland soal blower are not detected tiy alarm, and have no impact on HPCI operablUty. 2) Does this modification croato the por.sibility for an acciderit or malfunction of a dhfarent type Ifun I any evaluated previously in the safety analpla repert? - hMWtG No. Operators woro able to monitor for degradution of tha hPCI syWom. The gland seal blower is not requfied for HPCI operabilky and the loss of power nlarm for it proddes no safety function. - l 0) Doos this modification reduco the margin of safoty as de'inod in the basta for the Technical Specificatiorm? l hDWtG No. Technical Specification soction 3.5.C was reviewed for making this determination. I L i d E 45z s

PEACH 130TTOM ATOMIC POWEH STATION UNIT 2 f' DOCKET No. 50 277 1991 10 CFR 60.59 REPORT ScDdtR,WW2L8dVD1kmEW!ha.CMRLhtltROAD99 JXftTQH&I1B&fLALTIMI1OR; TPA 2 3047 A. b1SJUd; Sotvico Water D. DCMEEUQlO This TPA involved opening valves HV 2 30-217!,1 un;; 21792 to provido cooling ve. tor to a refrigerant vapor reccuory system during the repalt of a drywc4l chiller.

c. fiUAS9ftE0fLCliat1GE This modification was required to prevent loss of the refrigerant to the atmosphoro during repair of the cinilor D. EMIDLWA1ARTMLGIMMMtD t)

Doos this modifk.ation increase the protvMity of occurrence or the consequences of an accident or malfun tlon of equipment important to safety as previously evaluated in the safety analysitJ report? EDRMD No. Servico Water is not a safety related system. 2) Does this modificatbn cicate the possibility for an accktont or malfunction of a ddforent typo than any ova!uat9d previour.ly in the safety analysis apott? Aaima No, Sorvico Water is not a safety related system. 3) Does this modification reduce the margin of safety as defined in the bags for 1 Technical Spocifications? ADhYtG No. Service Water is not mentioned in the Tecni.lcal Specif%tions. [ ] 'I 46 m. ~

4 e PEACH is0TTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 199110 CFR 60,09 REPORT E0El0LE(Q!Elb!1.MQinLGi!MtfilDLMdd.l.fiii.ldW1.10Ckt1 i .lfMEQBARY RANLALIISADQX; TPA 2 0010 A.#fGIEMJ Roactor Protaction D. 0gSQLIE.HQff; This TPA a ided a voltage recordor to the 0 'B' M/G wit. C.1EMMLIDII.Q!AttGB) i This n' lowed bottor mordtoring c' M/G voltage output and voltage rogufator input and output. i D. TaArETY C%1LUAltQfLDJMMalir. 1) Does Ith modification increase the probability of ocenitence or the consequences of an accident or malfuncilon of oquipment important to safoty as previously evaluated in the aaloty analysis report? AGutt0 No.l The M/G at is protected from a la#ure in the monitoring equipment by fusos, arxi the equipment is %1smically rnountrd 2) Does ibis inod!fication creaty rho pos4bility for an accident or malfunction of a different type than any ovuluated pluN.,usly in the safot, analysis report? .8118 WAG No. This TPA only connects monitoring equipment.u the M/G sot, j 3) Does this modification reduce the margin of safety as definod in the basis for tho Technical-SpectiMations? AME9t; Nn. Technical Specification Soction 3.1 was rovlowod for making this dotermination. 47

if PEACH BOTTOM ATOMIC POWER STATION Ul41T2 s

DOCKET No. 50-277 1991 10 CFR 50.59 HEPORT UNIT 3 .a 3 , - - = +

=-.~.. -_ ] PEACH BOTTOM ATOMIC POWER STATION ) + UNIT 3 DOCKET No. 50-278 199110 CFR 50.59 REPORT .8tgfution Of Discrenancy Botween As-Dul:t And Ey;btina P81Ds MRCONLDAMatiCLRP_0RI.BL; P89232 313 A. S131M; instrument Nitrogon D. ERCRif'11QlfJ This NCR Identiflos changos to PalD 0280-M 333 and rotated Pal 0 0200 M 327 as a result of discropanelos with the as-built plant condition for the Instrument Nitrogen system for Unit 3. NCR P09232-313 Identifies nino individual items requiring evaluation, including nuinerous changos to valvos, draln laps, piping, and pressure switches. Thoso discropanelos woro generated as a result of a field walkdown of the system. C.f]CA$0N FOR QiANQEj This NCR ls dispositionod to uso as>ls. D.,$AFETY EVALVAll0N

SUMMARY

4 1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safoty as previously ovaluated in the safety analysis repon? .Answec l No. Those changor do not affect the function or operation of the instrument Nitrogon system or inledacing systems in a manner that would increase the probabulty of occurrence of a translont i accident, or evont pros.ausly evaluated in the UFSAR Chaptor 14.0 Appendix G. 2) Does this modification create the possibility for an accident or malfunction of a different t)po than any evaluated proviously in the safoty analysis report? l ArltMC No. Those changos do not affect the function or operation of the instrument Nitrogon system or intortacing systems in a mannor that would affect the saloty related function of any componont j required to mitigate the consequences of a trans!ont, accident, or ovent previously evaluated in the i - UFSAR, including Chapter 14.0 and Appondtx G. l 3) Does this modification roduce the margin of safoty as defined in the basis for the Technical Specifications? ' AlltMU l No. The Instrument Nkrogon system is not addressed in any Tochnical Specification. Additionally those changos specillod to the instrument Nitrogen Systom do not affoct the Tochnical Specification' bases of any interfacing sys!am. 49-t; l l

l PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET Ho 50-277 1991 to CFR 50.59 REPORT flf;$olution Of Disc 1ppaDCy_Detwoon As-Qllitt F" ant 0017M20 And Exist!O2.Qocumentation ljQfiqQNFORMANCLIEPORf 61; P 90538 A. StSIER; Residual Heat Rornoval (RHR) D. I E S M EIl01fJ This NCR ident!flod Inconsistenclos between documentation and actual valve position of HV 31045. UFSAR Figuros 4 0.2,7A6,13 3 and assoclatod documentation wore revisod to show this valve in a normal!y closed posdion. C. IEASQ11.IDILQlAllGE.; This NCR is dispositionod to show this valve as normally closod. D. SAFETY EVAU)ATION

SUMMARY

1) Does this modification inc ;ase the probability of occurronce or the consequencos of an accident or malfunction of equipment important to safety as previously evaluated in the safety analyslb report? Answer: No, This change has no uffect on the existing accident analysts or oporational analysis as praviously evalualod in te SAR. 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? AQ1W1T) No. Closing the Innot block valve doos not croats an accident initlator not provlously evaluated in the UFSAR. 3) Doos this rnodification reduce the margin of safety as defined in tho basis for the Tochnical Specifications? AntWE; No. Technical Specification Section 3.5 was reviewod for making this determination. 50 b i lu um iii. Su n d m nu mu u ng u u g u -u.....

. - -.. -. - -. ~. -.. _.. -. o PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 1991 10 CFR 50.69 REPORT l RQ20!Mt.km Of Discrenancy Between As BMt Plant Condition And Existina Documenta'dQD .tDflCQREQHff,&[lCE REPORT %t P 90678 i A. Sy1IIM; Circulating Water D DESCRIMlONt This NCR idontiflod inconsistenclos between PalD M 312 Shoo! 2 and as-Is condition of the plant. UFSAR Figure 11.0.1 was revisod to show that valves HV 3 28A-38011 A, B, C and HV 3 28A4012A.0,C are carbon stool instead of brass valves. C. BLAJON FOR CHANGEt This NCR is dispositioned to use as-is. D.JAFETY EVA11IAI1QtLSI,l}AMARy; 1) Does this modification increase the probability of occurrence or the consoquences of an accident or matfunction of equipment important to safety'as previously evaluated in the safoty analys!s report? An1ERC No. These valvos rneet system design requirements and have no impact on the plant accident scenario described in the UFSAR. 2) Does this modification create the possibility for an accident or malfunction of a different type than any ovaluated previously In the safety analysis report? AtllEfC No. This is a drawing change only and doos not change the current plant configuration. 3) Does this modification reduce the margin of safety as defin9d m the basis for the Technical Specifications? l Antrec l No. Technical Specifications do not address those valves. i l l 51

PEACH DOTTOM ATOMIC POWER STATION UNIT 3 ) DOCKET No. 50-278 199110 CFR 50.59 REPORT hohltion Of Discrecancy Betwren.As-Bultt Plant Condition AniExistina Documentation

1. 2flGQllEQfLMANCE FEP0FiIhQ4 P-91110 A.EyJIgMJ Foodwater B. 001C111EI1011J j

This NCR identiflod inconsistenclos between documentation arx1 actual plant configilration. Modification x 2301 replaced conical orificos with flat plato orificos. UFSAR Figure 4.3.2A and associated documents did not reflect thoso changoc. 1 C. J]riASQfLIDILQiANGE; j This NCR is dispositionod av reviso documentn:lon. O. SAFETY EVALVATION SJMMARY: 1 1) Does this modification inctnaso the prot > ability of occurrence or the consogences d an accident or malfunction d equipment important to sofoty as previously waluated in the safety entdysis report? An1 ERG Nn. Tids is a document updato only. 2) Does this modification creato the possibility for an accident or rT'P function of a diffotont type than l j. any evaluated previously in the safety analyslo report? AntWOU No. This is a document change only. 3) Does this modification reduce the margin of safety as defined in the basis for the Technical i Specihcations? AD1W2f1 No, Technical Specification Sections 3.5,3.6, and 5.0 woro reviewod for making this determination. l t l l l l L 52

PEACH DOrTOfA ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 199110 CFR 50.59 REPORT Resolution Of Discremncy Detween As Butjt Plant Condition And Existino Documentation IMREQllf_QHt&AJ[CE FEPORLtD.J P-91167 A. SYjiInt; 440 V Ernorgency AuxRiary D. 0gspniPTioru This NCR Identiflod inconsistenclos betwoon documentation and octual configuration. The UFSAR Sin 0* Lino drawings E 1017 and E 1717 in section 8.4. woro be revised to show broskor 3081 and 3803 as 100 Amp breakers. C.11tAjiON FOR Q1Al[QE; This NCR is dispositionod to reviso drawings. D. SAFETY EVALUAllQR311@jABY,; 1) Does this modification increaso the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis toport? Antwtc No. The breaker slzot were s120d for modification 5045. When the rnodification was complete those drawings woro ovoriooked and never changod. T his is a drawing change only and doos not chango currant plant ecnfiguration. 2) Does this modification croate the possibility for an accident or realfunction of a difforont type than any evaluated previously in the safoty analysis report? Antwec No. This is an editorial change only. 3) Does this rixxlification reduce the margin of safety as defined in the basis for the Technical Specification:;? Antb2G No Technical Specification Sections 3/4.9 woro reviewod for making this determination. l 53 --msLa- - _ _ _ - - - - - - _ _. - - - - - _ - - - - - - - _- - - - _ _ - -. - - - - - _ - - - - - - - -

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L PEACH DOTTOM ATOMIC POWER STATION 4 UNIT 3 DOCKET No. 50 27b 199110 CFR 60.59 REPORT Emai.9fli!glLPIt3MO C91/auun[cctb1LOLCogler

1. 2tLCD11LQILWCLfEPQllLtD; P-91347 A. Gy.SIIM; High Prossuto Coolant Injection (HPCI)

8. DESCMEHQll; This NCR waluated plugging one tubo of the HPCI lubo oil cooler, C.IEA1QN FDFI CHANGE:

This NCR is dispos 4loned to ropalt the HPCI lubo ou coolot D. SAFETY EVAlyAIjQtjj)MPLAJJT] 1) Does this modification inctcaso the probability of occunonco of the consequences of an accident tA muttunction of equipment important to safety as previously evaluated in the safety.analysh report? AD1WRC No. The p!ugging of the tube wid increaso rothbility of the optem by proventing water irom loaking. Into the HPCI lubo oil. 8 2) Does this modification crnate the possibilny for an occident or nu!! unction o a tjilferent type than any evaluated previously in the safety analysis report? Ainec No. Thoto la no risk that a dL4odged f ug wedd travel beyond the naat exchanger. The plug was J installed in the second pass of a four pass exchanger. 3) Dos th4 modMeetion roduto the nwgin of cafety as defined in the basis for the Technical Spocl%tions? AG G M1 No Technical Spectication Sections 1/4c5A woro revlowai for making this deintmitustkirt 1 u

PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 199110 CFR 50.59 REPORT ftMit/al2LldWitLYd19 .IE!!EQfStlbeLIEILATJD!ill!L; TPA 34-7 A. ElIEld; Foodwater Hoator B.Dr&GBleILQti; This TPA installed a temporary hand valve and plpo cap downstream of HV-345317PA C. !EASQ!ilDfLOM! int.1 This vs9e and cap wote added to isoloto leaktgo. D.. SAFETY I;MUARQQMl{ASy.; 1) Does this modification increase the probabWty of occurrence or the consequences of an accident or malfunction of equipment important to safety as prov'ously evaluated in the safety analysis report? AMERG No. No accidenta are mentioned in the FSAR only a drawing of vont linos is shown. 2) Does this modification croato the possibhty for an accident or malfunct!on of a diffotont type than i any evaluated previously in the safety analysis report? A7Lw.tG No. No accidents are montioned in the FSAR. 3) Does this modification reduce the margin of satoty as defined in the basis for the Technical Speci!! catkins? At11 ERG No. The Foodwater Heater system is not shcrwn in the Tocholcal SpecFications. 55

PEACH COTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199110 CFR 50.59 REPORT LImlLSE11Ghra.Qn.[bfnel Platiortp JEMEQllaflY_/AIIIIAINILR;b; TPA 318-4 A. SY31[ld; Refuol Plattorm B. IEECIllEll010 This TPA liftod limit switch loads on the refuel platform. C.,[FA1Q11 fon..Q1Afsgj To f acilitato major maintenance without causing rod withdrawal block when over reactor cavity shield plugs. ? D. S&[ETY EVAt.UAT101L

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to saloty as provlously ovaluated in the safety analysis report? AillME No. No procoduro exists in the SAR which govems bridge position over the reactor vesso! with the vossol completoly assombled and the coro inaccessible. 2) Does this modification croato the possibility for an accident or malfunction of a different typo than any evaluated provlously in the safety analysis report? Antwen No. No those interlocks are intended to provent two independent means to chango core reactivity when the coro is exposed, The core was not exposed during this TPA. 3) Does this modification reduce the margin of safoty as defined in the basis for the Technical Spocifications? A!1sur.; No. The plant was not in the refuel modo during this TPA. SG l u

PEl.CH DOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199110 CFR 50.59 REPORT jmialla.tlp!LQLDlattLElanaes i JLMEQRAHL/LIIRA110N 1)L; TPA 3-4004 A. SYSTEM: Vontillation D. IK1CHIEDQM1 This TPA installed blank flanges on two reactor building ventillation supply ducts. C. IKA1QtLipJ1.DiARMj To help rnalntain the residual hoat removal (RHR) rooms at nogative pressure whon the hatch is removod. D. EffrLINALUAIl0R_9)EMABY; 1) Doos this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously ovaluated in tho safety analysis report? Imwsc No. There are no procedures in the SAR which portain to the teactor building ventillation. 2) Does this modification create the possibility for an accident or malfunction of a differont typo than any evaluated previously in the safety analysis report? Answer: No. Both RHR pumps had clearancos applied during this TPA. 3) Doos this mod!!ication reduce the margin of safety as defined in the basis for the Technical Specifications? Ar11 wen No. Tochakul SpocXicatk>ns Section 3.5 was reviewod in making this determination. 4 57

6 g PEACH BOTTOM ATOMIC POWER STATION ' '+ UNIT 3 DOCKET No. 50-278 1>';i 10 CFR 50.69 REPORT - 4 ^ QQD.ted Rod Circqy ? '3.6RY PLANT Al.TERAUQN.; TPA 342 22 fc. JIM; Control Rods a &f ix 9 0RIPr;0'.; j - e TPA wp. a.oroptoted to jumper in the full-In Indication for control rod 50-39. C.1PAWB_P?tLQi&Nara This control rod has a full in position Indicating probo (PIP) problem. D. SAFETY EVAL.UATION

SUMMARY

1) Does this modWication incrosso the probability of occaen.., or the consequences of an acciderK 6' or malfunction of equipment important to - '// as puesly evaluatud in tho safety anai/ sis report? U No. Rod 50-39 will be blocked full-in while this TPA is appiled so that rod 02-43 and any other rod can not be withdrawn simultaneously. 2) Does th!s morilhcation create the possibility for an accident or malfunction of a ridforent type Na any Aaluated previously in the safoty analysis report? An1W2C No. Thic TPA cnly allows the refuel mode one rod prmissive interlock to te operational. 3) Does this mrxilfication reduce the margin of safety as defineo in the basis for the Technica S, ecifications? AnswgG No. Technical Spnfication Soctions 3.1LA and 4.10.A woro reviewad for making this determination. 58 a-m.

PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 199110 CFR 50.59 REPORT Control Rod Circuitry JLMf_QH6RY PLANT 4.TERATION: TPA 3432-24 A. SYSTEM: Control Rods B. jf iCRIPTION: This TPA jumpered in a green back!;ght fun-in Ind% tion for rod 30-55. C. FEASON FDR OiANGE: The rood switch is defective. D. SAFETY EVALUATIQt[3)MMARY: 1) Does this mocification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in tha safety analysis report? Answer: No. Rod 30 55 will be blocked full-in while this TPA is applied so that rod 02-43 and any other rod can not be withdrawn simultaneously. 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? Answer: No. This TPA only allows the refuel mode one rod permissivo Irsterlock to be operational. 7 3) Does th!s modification reduce the margin of safety as defined in the basis for the Technical Specifications's AGEWRC No. Technical Specification Sections 3.10.A and 4.10.A were reviewed for making this determinatk t 59

PEACH COTTOM ATOMIC POWER STATIOM ' UNIT 3 DOCKET No. 50-278 - 1991 10 CFFI 50.59 REPORT 1:sninflod Cfreultry TEMPORARY PLANT /tTERATIQB; TPA 3424: A. SYSTEM; Control Rods D. EFR:PTION: This TPA was completed to jumpor in the full in indication for control rod 10-19. C. FEASON FOR OiAPiQEj This control rod has a full-in PIP problom. D. SAFETY Q%WAT10# SUMMAnY: 1) Does this modification increase the probability of occurronce or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safoty analysis renc.17 Antw.cc No. Rod 10-10 will be blocked full Ir. while this TPA la appiled so that for' 02-43 and any other rod can not be withdrawn almidtaneously. 2) Does this trM ration create th* oossibility for an acciden.t or malfunction of a different type than any waluatoo previously in the difety analysis report? Antwrc No. This TPA only allows the refuel mode one rod permissive interlock *) be operational. 3) Does this modification roduce the margin of afety as defined in the basis for the Technical Specifications? Answer; No. Technical Specification Sections 3.10.A and 4.10.A were reviewod for making this determination. 60

- ' e_ ;... 31 PE ACH COTTOM ATOMIC POWER STATION UNIT 3 - DOCKET No. 60 274 1991 10 CFR 50.59 REPORT - Contrd Rod Circuitry TEMPORARY PLAKI ALTERATION: TPA 342 27 A. SYSTEM; Control Rods - B. DESCB1EDQNJ This TPA jumpered in a clasure signal for rod 34 27. C..TASQN PDR OiANGE: The reed switch on the PIP probe is defective. D. SAFETY EVALUATIOfL

SUMMARY

{ 1) Does this modification increase the probablity of occurrence or the consequences of an accident - or malfunction of equiprnent important to safety as previously evaluated in the eafety analysis'-- y report? - Artswec No. Rod 34 27 will be blocked full-in whue this TPA is applied so that rod 34 27 and any other rod 1 l} can not be withdrawn simultaneously. 2) .Does this inodification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analys!s repor? Answer:.- t No. This TPA only allows the refuet mode one rod permisske Intertock to be operational. 3) Does this modification reduce the margin of safety as defined in the basis for the: Technical- ~ Specifications?- Antiwna No. ; Technical Specification Sections 3.10.A and 4.10.A were reviewed. for making this determination. l: f ) 61 g, i:{[jip: , L:~ ~,

6 o PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 1991 10 CFR 50.59 REPORT Bal]ef Valvo. Infection. And Recirculatina Testing filqQIRUBij ST 13.18-3

n. fy11gM.; Standby Liquio Control (SBLC)

B. QESCRIPTION: This procedure makes clarnications, adds unit designatK>ns for valving, updates 13T criteria to reticct the new velocity readings, and inakes minor changes to the chemical enalysis after flushing. O. IEASON FDA CHANGE; These changes make the procedura easier to use and provides more accurato data. D. RAFETY EVA1UATION

SUMMARY

1) Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipmont irrportant to safety as previously evaluated in the safety analysis report? Answer: No. The SBLC will not be required dunng the period w-hen testing is being done. 2) Does this modification create the possibility for an accident or malfunc' ion of a different type than aay evaluated previously in the safety analysis report? ADLW2D No. The equipment continues to operate in the same manner as previous to the procedure revision. 3) Does thia modification reduce the margin of safety as dafined in the basis for the Technical Specifications? AE1 LEG No. Technical Specification Sections 3.3/4 and 4.4 were reviewed for making this determiration. fs2 I

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PEACH BOTTOM ATOMIC f OWER STATION UNIT N/A DOCKET No. 50-277 & 50-278 199110 CFR 50.59 REPORT instalbtlon Of A Fish Lndder At Conowinag_Qam .tiSIALMT10ff A. jiyJIgMj Flood D.1%SCRIPTION; t This review discusses the fish ladder installed at Conowingo Dam and its impact on Peach Bottom Atomic Power Station during a major flood event. C.IEASON FOR CHANGE: Installation,i the Ilsh ladder reduced the regulating gates at Conowingo from three to two and made cMnges to operational procedures for Conowingo Dam as described in UFSAR Section 2.4.3.5,4. D. SAFETY EVAL.UATIQRjil)16MARYJ 1) Does this rnah!.c~ tion 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 changes will not affect flood levels at Peach Bottom as described in the UFSAR. 3 9) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?- AI12W2f3 No. Changes at Conowingo do not affect flood levels at Peach Bottom. 3) Does this modification reduce the margin of r y as defined in thc basis for the Technical - Specifications? A!19 M U No. Technical Specification Section 3.12 was reviewed for making this determination, i 64}}