ML20058J958
| ML20058J958 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/31/1989 |
| From: | Miller D PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9012060159 | |
| Download: ML20058J958 (75) | |
Text
.
gCCN 90-14220 4
- 9 I%
PIIILADELPIII A ELECTRIC COMPANY
':(%5 PEACil llOllDM KlDMIC POWER STATION R. D.1. Ikix 20N 8% @.#
- Delta. Pennsyhanta 17.414 j
nxn norrow-me row an os tu ntenu (17)456 7014
~ D. B. Miller, Jr.'
- Vice President i
November 29, 1990 u
l Docket Nos. 50 277 50 278 Liscense Nos. DPR-44 DPR 56 i
U.S. Nuclear Regulatory Commission i
Docu.. ea' Control Desk Washington, DC 20555
.l
SUBJECT:
Peach Bottom Atomic Power Station (PBAPS) l Annual 10 CFR 50.59 Report l
For The Period 1/1/89 through 12/31/89 J
Dear Sir:
Enclosed is the 1989 Annual 10 CFR 50.59 Report as required by 10 CFR 50.59. Should you have any questions, or require further information, please contact us.
Sincerely, l
w
}
-P/p' Men 1
/
i r
DBM/AAF/w /MJB:dit
]
RAK
- dttachment cc:, R.A. Burricelli, Public Service Electric & Gas
' T.M. Gerusky, Commonwealth of Pennsylvania JJ. Lyash, USNRC Senior Resident Inspector
. R.I. McLean, State of Maryland T.T. Martin, Administrator, Region I, USNRC H.C. Schwemm, Atlantic Electne J. Urban, Delmarva Power i
i cvrltr i
9012060159 691231
.J A
j PDR ADOCK 05000277
/f R
Pac
/
I
\\
c
- -- J
y
-f..
e, i
' "t -
bec:
Commitment Coordinator Correspondence Control Desk J.F. Franz i
A.A. Fulvio D.R.' Helwig a
G.A. Hunger J.M. Madara D.B. MiUer J.T. Robb
'D.M. Smith PBNR k
,\\ ;
ts 1'
- e 1
'. I i t'
g 4
a 1
'i!;
)
I r.iU i..
lo li 1-
?!
n'i h
i I
l 1
?-
., { ".
lf 1-
' I. :
A E
'i PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 DOCKET NOS. 50 277; 50 278 1989-ANNUAL 10 CFR 50.59 REPORT J
l
)
Docket Nos. 50 277.
50-278 t
1989 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT l'
This Report is issued pursuant to the reporting requirements of 10 CFR 50.59 for Peach Bottom Atomic Power Station Units 2 and 3 (Facility License Numbers L
DPR-44 and DPR 56 respectively). This report addresses, but is not limited to, any changes to the facility or procedures as described in the Updated Final Safety Analysis Report, and any tests or experiments performed which were not.
described in the FSAR. A summary of the safety evaluation for each item is 1.
included Each safety evaluation concludes that an unreviewed safety question, as defined in 10 CFR 50.59 (a) (2), was not involved.
~
I:
L l
l l
b b
1
1 IABLE OF CONTENTS 4
1989 PEACH BOTTOM ATOMIC POWER STATION ANNUAL MODIFICATION REPORT
. Modifications System Eacn Units 2 & 3
-865 Control Rod Drive 1
)
1029B
_ Main Steam Heating and Ventilation Air Cond.
2-t 1029L 480 Volt Motor Control Center 3
1505-Standby Gas Treatment _
4
{
1660 Automatic Depressurization/ Main Steam 5
1890 P Imary Containment
-6 2081 Emergency lighting 7
4 2132 Miscellaneous 8
2371/
Emergency Service Water-9 2390-Fire -
j 10 2544 Pipe Supports 11 2564-Miscellaneous 12
-2579-Diesel Generator 2580 Cooling Tower.
13 i
14 5002
'4KV' 15 5017 Emergcacy Serrice Water 16 i
5083 13K/ '
17 5119'
' M'ator Control CenterControlCenter--
18 5125 H.gh Press. Coolant inject./ Residual Heat Rem.
19 i
'88-088 FeedwePr 20 4
Unit 2 -
i
'1160' Condenser
'21
-1352A.
High Pressure Coolant injection'.
22 -
1418-
. Containment Atmospheric Control / Dilution 23 5006 Containment Atmospheric Dilution L '
24 1
5084 High Pressure Coolant injection 25 5157-Reactor Building Coo!!ng Water 26 88 096 Condensate 27 '.88-097 Reactor Water Cleanup _
28
.i 6
9 I
Modification System Egge Unit 3 603B' Suppression Pooi 29 664 Primary Containment isolation 30 867
. Standby Liquid Control 31 916 Reactor Feed Pump 32 936 Core Spray 33 1243 High Pressure Service Water 34 1316 Primary Containment 35 l
1353E 4KV Circuit Breakers 36 1353G Automatic Depressurization System 37 1353H Miscellaneous 38 1359 Reactor Protection 39 1404 Reactor Protection 40 L
1536 Reactor and Recirculation 41 L
~_1542_
Drywell Cooling 43 h
1636 Fire.
44 1678 Reactor and Recirculation 45
- t L
1684-Feedwater 46 1713 Feedwater 47 1891 Various 48 l'
1892 Ventilation 49 l
1950A.
Automatic Depressurization 50 2079 -
High Press, Cool, injec./ Reactor Core isolation 51 I.
2080:
High Pressure Coolant InJcction 52 12083 480 Volt 53 p
"2085 Process and Diagnostic Instrumentation 54 2106 Emergency Service Water 55-2224 Feedwater Heater 56 1
2275-Instrument Alternating Current 57 i
2285-Residual Heat Removal 58 L
2354 Reactor Core Isolation 59 2389 Feedwater - -
60 2489A Secondary Containment -
61 l:
25178 480 Volt Motor Control Center 62 2532 Reactor Vessel Internals 63 2578.
Residual Heat Removal 64 5001 Reactor Core isolation Cooling 65 -
3 5156 Turbine Generator 66 j
79416 Instrument Nitrogen '
67
'j 83-037
. Miscellaneous instrumentation 68 L
86-095 Recirculation Pump and Valves 69 l
c e
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 198910 CFR 50.59 REPORT Afternate Rod Insertion i
MooiricAtion No.: 865 A. SysTru: Control Rod Drive B. Desenimou:
Addition of an attemate rod insertion (ARl) system and revision of the recirculation pump trip system This modification ensures that RPT and ARI system actuations occur simultaneously.
C. Reason Fon CHawor:
The ARI system was installed to satisfy the ATWS rule 10 CFR 50.62 and the OA guidance provided in generic letter 8546. The RPT was modified to reduce the number of spurious recirculation pump trips.
, D. SArETY EVALUATION Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?j
' Answer:
No. This modification has been installed as a backup system to trip the reactor in the event the Reactor Protection System falls to trip the reactor during an anticipated transient condition. This modification also revised the ATWS recirculation pump trip logic so that the spurious recirculation I
pump trips will be reduced, thereby reducing the number of challcnges to plant safety systems. The most serious malfunction this modification can cause is inadvertent simultaneous tripping of both recirculation pumps concurrent with a reactor trip. The recirculation pump trip evaluation in chapter 14 of the UFSAR envelops this malfunction.
2)
.Does this modification create the possibility for an accident or malfunction of a different type than -
E~
any evaluated previously in the safety analysis report?.
L
' Answer:
' No. The ARI and RPT systems are designed to be independent of the Reactor Protection System.
All non safety related circuits are Isoland from safety related circuits. Therefore, this modification cannet prevent any safety system from performing its intended function. The most serious malfunction this modification can cause is inadvertent simultaneous tripping of both recirculation pumps concurrent with a reactor trip. - The recirculation pump trip evaluation in chapter 14 of the -
S UFSAR envelopes this malfunction.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specificetions?
Answea m
Nc. This modification will actually reduce the probability of occurence of an anticipated transient without scram.
l 1
W: _.
~
=
- PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 198910 CFR 50.59 REPORT q
Roroute And Encansulate Cables in A Fire Barrier Areg Moomication No.: 1029B A. System; Main Steam Heating Ventilation Air Conditioning B. DESCRIPTION!
Rercute SRV cables, encapsulate 0. ables in a three hour fire barrier in specific areas, and designate cables as safeguards.
C. RsAson Fom CHANQJJ To meet requirements of Append'x R to 10 CFR 50.
l i
D. HanTY Evatuarlow
SUMMARY
l 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?
i Answer:
l The modifications are within UFSAR and Tech Spec guidelines; Rerouting was dono in accordance with approved installation specifications. There are no adverse effects on plant load requirements.
Rerouting of cables does not affect the function of related panels or oquipment.
l 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?
8at.w._.tfl '
I No. Installation was done implementing UFSAR physical arrangements and separation requirements.
3)
Does this modification' reduce the margin of safety as defined in the basis for the Technical f
Specifications?
Answer:
No. This mod follows the separation requirements presented in sections 7.1.6 and 8.4 of the UFSAR.
and compiles with criteria outfined in Appendix R to 10 CRF 50.
s.
h,'
T 2
-L
<g t
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 -
DOCKET No. 50-277 & 50 278 198910 CFR 50.59 REPORT Installation of Thermal Maanetic Leakers MooncAnow No.: 1029 L A. Sysieu: 480 Volt MCC B. Desenmnon:
Installation of thermal magnetic breakers or power fuses.
C. RsAsow Fon CHANGE:
An i ppendix R associated circuits analysis revealed that certain MCC circuits need to be modified to provide electrical isolation at the MCC without effecting the power to safe-shutdown equipment powered from the same MCC.
D. Sarriv Evatuanow
SUMMARY
l 1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
L Answer:
No. This modification maintains the fault protection that the magnetic trip breakers presently provide, and in addition provides circuit overload protection.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis repo_rt?
AntwtG No. This modification improves the overcurrent protection for the motor control centers.
.t 3)
Does this modification reduce the margin of safety as defined in the bas!s for the Technical Specifications?
Answer:
No.. This modification improves the electrical coordination of the emergency a.c. power system.
i
('.o L
3 3-
e PEACH BOTTOM ATOMIC POWER STATION 3'
UNITS 2 & 3 DOCKET No. 50 277 & 60 278 198910 CFR 50.59 REPORT Remove Standbv Gas Treatment System fSGTS) Differential Pressure Switches Mooineation No.: 1505 A. Systau: SGTS B. Diseawtion:
The purpose of this modification was to remove differential pressure switches from the fan discharge duct delete the ' Loss of Emergency Gas Treatment Fan' alarm; and change the operation of the SGTS fans to simultaneously start the primary and standby fans upon receipt of a system initiation algnal. Manual operation of the far.s was not affected by this modification.
C. Raasow Fom Cuawan Differential pressure switches failed to start the standby fan when the primary fan discharge dampers we closed. Other flow measurernent techniques were luund is M lnadequate because of the configuration of the SGTS ductwork, To ensure that one SGTS fan is available, the electrical design of the primary and standt>y fans was changed to cause both fans to start upon receipt of a system initiation signal.
i' O. SArrrY EvAwarion Suuuany:
1)
Oces this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The design of this modifica!!on is consistent with the operation of other safeguard systems in that redundant systems are started simultaneously when required. This modification inltlates operation of the SGTS standby tan simultaneously with the uni primary fan whenever system initiation 9 regolted. Further, the availability of the SGTS is imprcved by eliminating the possibility that the backup fan would not start if a ortrnary fan damper failure occurred.
l ~
2)
Does this modification create the posslbility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
l No. The modification will start the unit's standby tan and its assigned fan simultaneously upon receipt of a system initiation signal. Such operation does not lead to an accident or malfunction scenarlo. The modification ensures that the standby fan will be av&llable when required.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technica' Specifications?
Answer:
No. This modification does not affect the margin of safety as defined in the Technlcal Specifications.
its Implementation is to correct a design deficiency.
i i
4-
7 e
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 50 278 198910 CFR 50.59 REPORT Automatic Droressurization System FADS) & Main Steam Isofation Valve iMSM Nitrocen Acc kkD!lli Mooricatiow No.: 1660 A. Systant ADS and MSIV B. Desenwriox Install new ADS and MSIV air supply accumulators.
C. McAsow Fon Can_g3J To increase the reflability of the pneumatic systems, improve maintainability and reduce cos repair and maintenance of the ADS arid MSly air supply accumulators.
D. Surty Evuuatiow SuuuAny:
1)
Does this modification increase the probability of occurrence or the consequences of an acciden or malfunction of equlpment important to safety as previously evaluated in the safety analysi Animen No. The accumulators are similar in design and identical li, fu% tion.
2)
Does this modification create the possibility for an accident or.idunction of a different typ any evaluated previously in the safety analysts report?
ADAEtu No. The accumulators are simllar in design and Identical in function.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answen No; The new accumulators meet the same design requirements and perform the same functions as the original accumulators.
i 1
l' L
l i
i '
i 5
- b
PEACH BOTTOM ATOMIC POWER STATION E
UNITS 2 & 3 DOCKET No. 60 277 & $0 278 198910 CFR 60.59 REPORT Primarv Qg-rrent iso'ation Valvo Position Indication Imorovement MooincATio9 No.: 1893 A. SYSTEM: Primary Containment B. DESCRIPTION:
Replace 43 limit switches on Units 2 & 3 primary containment isolation valves with selsmicelly and environmentally qualified equivalents. Provide Indirect position Indication for the Unit 2 and 3 containment atmosphere dilution system valves.
C. BLASON FOR CHAJ.Qtg; These changes are required to comply with NRC Order dated June 13,1984.
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 safety as previously evaluated in the safety analysis report?
801EtII No. The design features to monitor primary containment isolation valve position are maintained.
2)
Does this modification create the possibility for an accident or malfunctio i of a different type than any evaluated previously in the safety analysis report?
Answer:
No. Requirements applicable to the primary containment Isolatlon system were applied to this modification.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. Technical Specification Sections 3.2.A,3.7,0. 4.2.A. 4.7.0 and the associated bases were reviewed and are not affected by this modification.
6
i PEACH BOTTOM ATOMIC POWER STATION UNITS 2 a 3 DOCKET No. 64277 & 64274 198910 CFR 60.69 REPORT Install Emeroency Llahtnina Per hre Protection Procram Moomication No.: 2081 A. Srsitu: Emergency Lighting B. Df SCR* Tion!
Installation of emergency lighting to support operation of safe shutdown equipment in the event of a postulated Appendix R fire.
C. BgASON Fon CMawor:
This modification is necessary to fulfill the requirements of Section lli Appendix R to 10 CFR Part 50.
D. Harrry Evaluation SuuuAmy:
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?
AD.tERG No. This modification only adds emergency lighting to illuminate certain plant areas In the event of an Appendlx R fire.
2)
Doca 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. Lighting fixtures were seismically mounted as required to preclude Interaction with nearby safety related equipment.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. Emergency lighting is not covered in the Technical Specifications.
l 1
l l
l 7
8 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 60 277 4 50 278 198910 CFR 60.59 REPORT Human Factors Enhancement To The Control Room PanNs Meow! cation No.: 2132 A. Systau: Miscellaneous -
B. Drscamnow:
This modification makes various improvements to the control room such as: Rearrangement of certain instruments and controls, addition of annunciator matrix identification, repainting and relabeling of panels, addition of manual initiation switches for certain water systems, installation of a new multipo temperature recorder, certain changes in procedures, access to the Public Address System. addition of ne temperature Indicator, and enhancements of the remote shutdown panels.
C. Raasow Fon CHAwoa:
i This modification was completed to implement the recommendations of the Control Room Design Re and to fulfill the lequlrements of NUREG 0737 Supplement 1.
D. Syrty Evatuanow SuuuAev:
i 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.
By improving the operator / machine interface with human factors enhancements, operator response to high stress translent situations will be improved. The onlyimprovement provided by this modification that has control action is the addition of switches wblch provide manual Initiation of systems for whleh automatic Initiation is already provided. A malfunction of one of these switches could only affect one dMslon of one system. Such a single failure has the same effect as other single failures previously evaluated in the UFSAR. The temperature recorder replaces one that is already installed.
2)
Does this modification create the possibility for an accident or malfunction of a different type than l
i any evaluated previously In the safety analysis report?
n Answer:
No. The improvements made by this modification will not create any different type of accident or malfunction.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
{
Answer:
i i
Noi The margin of safety as defined in the basis of the Technical Specifications is not reduced.
i The Technical Specifications were reviewed and there are no applicable sections for this i
modification.
l i
i I
8
i e
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 60 277 & 60 274 198910 CFR 60.59 REPORT Emeraenev Servlee Water Smal! Pinino Repjactment Moe'neatiow No.: 2371 A. Systrut Emergency Service Water (ESW)
B. Drscamtiow:
This modification replaces two inch diameter and smaller piping in the Core Standby Cooling System R and replaces 11 with the same class of pipe. Valves that were purely for redundancy were eliminated and other valves were replaced with new valves. Some globe valves were replaced with lower resistan valves and the valve immediately downstream of each cooler was replaced with a plug type balancin Low point drains and lateral clean out fittings were added for improved maintainability.
C. RaasoH Fon Cnawor:
This modification was necessary because of blockage caused by pipe wall corrosion and fouilng due to of untreated river water.
D. Sarrty EvatuatioH Suuuamv:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safetyanalysis report?
Answer:
No. This modification ensures the system can function as designed and resolves the problems of flow blockage.
2)
' Does this modification create the ;tssibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
At1LWAG No, The replacement pipe and valves were procured, installed and tested in accordance with the same, or more stringent codes and specifications than those governing the former components.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. The Technical Specifications are not affected by this modification. Section 3/4.9.C were reviewed to make this determination.
i l
l l
l 9
i
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 198910 CFR 50.59 REPORT Diesel Generator Room Fire Protection System Flooineation No.: 2390 A. Systru: Fire B. DrsCa*No,N,}
Replacement of control cabinets, master selector valves, electromanual pilot cabinets and push button stations with seismically quallfied safety related components.
C. Epsow Fon CHAnot:
A design deficiency was identified which could result in the common mode failure of all four diesel generators during a loss of off site power event.
D. S.agny_Evatuanow Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This modificative aplaces non seismic, non safety related Cardox system control components with seismically qualified, safety related devices to prevent a common mode failure of the diesels while meeting all applicable fire suppression system requirements.
2)
Does this modification create the possibility for an accioent or malfunction of a different t/pe than any evaluated previously in the safety analysis report?
Answer: -
No. The sole effect of this modification will be an increase in the reliability of the diesel generator Cardox fire protection system.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. Technical Specifications 3.9 and 3.14 were reviewed to assure an adequate source of electrical power to operato the engineered safeguards following an accident and to provide fire suppression -
capability for the Diesel Generator Rooms.
1-l-
10 1
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 60 278 198910 CFR 50.69 REPORT RHR Drain Line Retoutino And Exoandon Loco Installation Moemeation No.: 2544 A. Sysitui Pipe Supports B.pneneriow:
This modification assessed the adequacy of 0 listed, small bore pipe,3' nominal size and smaller. Where the assessment indicated a need, the piping or supports were reworked or repaired. Most of the repair and i
i rework was minor, related to pipe support hardware. In addition to this, two Residual Heat Removal draln lines were rerouted. The reroute involved the addition of expansion loops in the non-Q portion to brin the Q portions of the lines to within original design basis stress limitations.
C. Raason Fon CHAnor:
1 The walkdown was performed to provide a level confidence in the installed conditions of the small pip systems. The reroute was performed to bring the Q portlon of the lines to within original design basis stress limitations.
l D. SAnry Evaluation EguuAny:
s 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 evaluated in the safety analysis report?
1 Answer:
No. The piping systems involved in this modification meet the requirements set forth in the UFSAR and/or other NRC approved st&iscards for such piping systems previously approved for Peach Bottom.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
L Answer:
No. All Mark l load effected piping involved in this modification comolles with UFSAR requirements as modified by NUREG 0661, ' Safety Evaluation Report, Mark l ContaMment Long Term Program,
' Appendix A, July 1980. All non Mark i load effected piping compiles witi; the original design code -
l requirements specified in the in the UFSAR or with latti editions of industry Codes and Standards.
j 3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. The piping systems involved in this modification meet the requirements set forth in the UFSAR and/or other NRC approved standards for such piping systems previously approved for Peach '
Bottom.
All Mark i load effected piping involved in this modification complies with UFSAR requirements as modified by NUREG 0661, ' Safety Evaluation Report, Mark l Containment Long Term Program,
' Appendix A, July 1980. All non-Mark i load effected piping complies with the original design code requirements specified in the in the UFSAR or with later editions of industry Codes and Standards.
11
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50 278 198910 CFR 50.59 REPORT Chance Time Settina On Residual Heat Removal (RHR). Core Sorav fCS). Emeroency Service Water (ESW).
Emeroonev Coolina Water (ECW) Pump.s Moomicariou No.: 2564 A. Systru: Miscellar.cous B. QtsemptioNl Change the time settings to start the RHR, ESW, ECW, and CS pumps for Units 2 and 3 upon recelpt of a Loss Of Coolant Accident (LOCA) signal.
C. Reaso,4 Fon Csawor:
This modification is necessary to improve voltage regulation in the event of a LOCA.
D. Earriv Evatuation Suuuaav:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The new pump settings will have no effect on final coolant availability to the vessel in the event of a LOCA. The control relays added and replaced are the same as those presently in service except for their available time delay, 1
2)
Does this modification create the possibility for en accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No. No nevt failure effects are introduced into the systems by this modification. The replacement relays are identical to the relays in service except for the time delay range. The CS and RHR system response times fell within the limiting response times.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Spec;!ications?
Answer:
No. The new initiation times will minimize delay in the ECCS initiation sequence yet provide improved starting voltage for each pump motor. The new Initiation times fall within the limiting response times of the 10 CFR 50 Appendix K analysis of the UFSAR, t..
Y
)
12
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No.60-277 198910 CFR 50.59 REPORT Start Diesel Generator Fans Uoon LOCA With Offsite Power Available Maomcatiow No.a 2579 A. Systru; Diesel Generator B. Desenwriow:
This modification changes the start time of the four Diesel Generator room vont supply fans from approximately 8 seconds after diesel generator start to time t = 0 on receipt of a LOCA signal with offsite power available. The Diesel Generator supplemental supply fans also were modified to receive a permissive to start at time t = 0 on LOCA, but the supplemental fans remain interlocked with outside alt temperature.
C, Erasow Fon Cwas.:=:
The purpose of the modification is to assure adequate control voltage during the start of Diesel Generator vent supply fans. There was a concern that delayed start might make the supply fan motor contactor unable to pick up because of the voltage across the coil bcIng below its pick up vatuo. This condition was due to a control circuit voltage drop along with the usual degradation of voltage, i
D. Earrry EvAtuation SuuuAny:
1)
Does this modification increase the probability of occurrence or the consequences of an accident l
1 or malfunction of equipment important to safety as proviously evalualod in the safety analysis report? l Answer:
No. On the contrary, this modification improves the voltage regulation of the electrical distribution system of the plant and improves the temperature profile of the Diesel Generator rooms.
2)
Does this modification create the possibility for an accident or malfunction of a different type thanj 1
any evaluated previously in the safety analysis report?
Answer:
No. Since this modification improves the voltage regulation and does not introduce any new hazards, no accident or malfunction could occur that could not have occurred in the design previously evaluated.
3)
Does this modification reduco the margin of safety as defined in the basis for the Technia.:.1
?
Specifications?
Answer:
No. No technical specification addresses the starting times of the Diesel Generator Room vent supply fans.
13 l-
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT Trio Coofina Tower Loads on Loss Of Coolant Accident ROCA)
Mooineation Nod 2580 A. Systru: Cooling Tower B.DitcamTiow:
The cooling tower loads (pumps and fans) will be automatically shed upon recelpt of a LOCA signal from 1
Unit 2 or 3.
C. EsasoN Fon Cnawot:
To improve the voltage levels on the 13.5kv unit auxillary buses and the 4.16kv omorgency auxillary buses.
D. SAFETY EVALUATION Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the saf ety analysis report?
SatERG No. This modification improves the voltage regulation of the electrical distribution system of the plant. UFSAR Sections 8.0 and 11.6 have been reviewed to make this determination.
2)
Does this modification create the possibility for an accident or malfunction of a different type tren any evaluated previously in the safety analysis report?
Answer:
No. This modification improves the voltage regulation, maintains adequate channel separation, and does not introduce any new hazards.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
AnsweG No. Technical Specifications for Peach Botton'.do not address the cooling tower load shedding.
1 14
i f
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 & 50 278 198910 CFR 50.59 REPORT imorove Protection Provided BY Undervoltace Relava Mgpineaview No.: 5002 A. SysTru: 4KV B. Desen*tiow:
This modification improves protection provided by the undervolta level of off site power sources to the safety related 4 KV buses. ge relays (127Y) which sense t C. RtaSoN Fon CHAwot:
Improvement is needed to prevent unnecessary operation of the 127Y protective relays within the scenario due to large motor starts when only one off site source is available.
D. Santy EvawavioN Summany:
l 1)
Does this modification increase the probability of occurrence or the consequences of an acciden or malfunction of equipment important to safety as previously evaluated in the safety analysi hPLEtti rio. This modification provides greater assurance of the operation of the AC powei supplies 2)
Does this modification create the possibl!!ty for an accident or malfunction of a different typ any evaluated previously in the safety analysis teport?
Ananer; No. This modification improves safety related power supplies by using higher accuracy u l
relays, adding a level of protection, and by increasing the time delay to trip the off site po l
feed to the 4 KV buses due to motor starting voltage transient.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
An6wer:
No. The margin of safety as defined in the Technical Specifications for Peach Bottom has not be reduced, but a change to the Technical Specifications has been required and a license amendment -
and prior NRC approval was granted.
l 15
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 60 277 198910 CFR 50.59 REPORT i
Start Loalc of Emeraency Servlee Water (ESW) Pumo Modification Mooincariow No.: 5017 A. SY$TEMt ESW B. Desemmnow:
This modification changes the start logic for the OAP57 and OBP57 ES'V pumps so that whenever one of the pumps is running, the other pump will receive a start signal upon oss of discharge pressure on the running pump.
C. Raasow FoR CHawor:
This will eliminate operator action requirement and improve the reliability of the ESW system.
D. Sarrry EvatuanoN
SUMMARY
i 1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment Irnportant to safety as previously evaluated in the safety analysis report?
Answer:
No. This modification maintains the sollity of the ESW system to provide a re!!able supply of cooling water to the diesel engines and selected equipment coolers during a Loss of Offsite Power, 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
i Answer:
No. This modification maintains the capability to safely shut down the plant, adds no new electrical i
loads, does not affect any radwaste system, and equipment is seismically qualified to operate in the worst case envlornment.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical-Specifications?
s Answer:
No. Technical specification sections 3.9.C and 4.9.C have been reviewed. Operation of the ESW system remains unchanged.
1 i
l 1-l L
l l
1 r
1'6
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50 277 1989 ANNUAL MODIFICATION REPORT ReDlate 10805 Leeds & Northruo Transducer And VAR Transducer Mooincarrow No.: 5083 A. Sysisu: 13KV B. Desen*tiow:
This modification replaces the obsolete L&N 10805 watt transducer for start up and emergency auxillary regulating transformer 00X05 and var transducer for emergency auxillary transformer OBXO4 for the 13 KV power system.
C. Reasow Fon CHawar:
These obsolete transducers were out of tolerance and could not be calibrated.
D. Sorry Evatuatiow Sumuany:
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 No. This modification does not functionally change the operability of the 13 KV unit auxillary switchgear buses.
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 replacement transducers have the same specifications as the obsolete 1(pe.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical.
Specifications?
Answer:
No. Technical SpeFutions have been reviewed and there are no sections applicable to the 13 KV power system.
i l
1 17
I I
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 5 4 277 & 5& 278 198910 CFR 50.59 REPORT Motor Control Center (MQC) 120 Volt Control Circuit Modification Mooincarrow No.: 6119 i
A. SysTru: Motor Control Center (MCC)
B. Descamriow:
This modification revised the control circuits of certain safety related Motor Operated Valves and loads to assure operation of the MCC contactors and control components under design basis conditions.
C. BLaspN Fon CHawor:
This will ensure the plant's ability to mitigate a loss of Coolant Accident (LOCA).
D. Sorry EvatuatioN SUMManf!
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_nawtr; No. This improves the calculated voltage to the selected MCC contactors for the limiting condition LOCA event to a value which ensures contactor pickup.
1 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated provlously in the safety analysis report?
Answerl No. All changes made by this modification are in accordance with the existing design criteria for electrical equipment.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. The operating requirements of the associated MCC's are not impacted.
18
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 60 277 & 60 278 198910 CFR 60.69 REPORT Resistors in DC Motor Ooerated Valve (MOV) Strinns Moomcation No.: 5125 A. Sysnu: High Pressure Coolant injection (HPCI) And Residual Heat Removal. (RHR)
B. Desemieriow:
This modification removes motor step starting resistors frorn electrical circuits of active safety related DC MOV's in the HPCI and the RHR systems.
C. Reasow Fon CwAnos:
These changes will ensure the DC motor operatou will accomplish the required value safety function under the worst DC power distribution system conditions combined with elevated temperatures resulting from postulated design basis events.
D. Sarny Evatuariow Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This modification does not functionally change the operability of esiher HPCI or RHR valves.
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 end result of this modification is the improved operabilty and reilability of the motor operated valves.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answeti No. The function of the valves withln the scope of this modification will not be Impalred.
e l
19 i
i PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 60 277 & 60 278 198910 CFR 60.59 REPORT Rerrovat Of Lona Path Recirculation Vent Valve Downstream of RO 2663A i
Mppnc.ation No.: 88488 A. Ststru: Feedwater B. DrsemiptioN!
This modification removes the long path recirculation vont valve downstream of RO 2663A (RO 3663A) and plugs the piping connection.
C. Reason Fon CHANor:
This modification is necessary to eliminate the possibi!Ity of further piping cracks and valve packing leakage.
D. Sartry Evatuatiow Suuuany; 1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equiprnent important to safety as previously evaluated in the safety analysis report?
Answer:
No. The material and installation of the plug was in compilance with pipe class HF Specification M 300.
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:
1 No. The vent valve that was removed was not used for system operation. It was installed for testing purposes only.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer; No. Technical Specifications Section 3.6 was reviewed. This modification does not effect the margin of safety as dehnod in the Technical Specifications.
20
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No.60-277 1989 ANNUAL M00iFICATION REPORT Installation Of Suonorts To The Condenser Fairina Envelogg Mooineatiou _No.: 1160 A. SystrM: Condensor B. QtscaienoN:
Install additional supports to the condenser fairing envelope and addition of rollof cutouts in the envelope bottom.
C. Reason Fon Cwawor:
j Reduce vibration and essociated plate cracking and to provide pressure rellof in the event of future expansion joint failures.
D. SaPETY Evatuanow
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 modification effectively reduces the size of unsupported panels and reduces the magnitude.
of vlbration and associated plate cracking.
2) dmo 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. TNs modification effectively reduces the size of unsupported panels and reduces the magnitude of vibration and associated plato cracking.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. None of the equipment is 0-Ilsted or safety related, No additional electricalloads are imposed by this modification.
l 21
j PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50-277 1989 ANNUAL MODIFICATION REPORT i
High Pressure Coolant injection (HPCI) Alternative Control Station (ACS)
Mooineation No.: 1352A A. SYsnu: HPCI B. Drsen*Teow:
This modification reroutes certain safety.retated circuits to an ACS, where a safety-related transfer / isolation switch will transfer the controllocation of safety.rolated equipment from its normal control panel to an ACS and Isolate safety related system circuits that could adversely affect safe shutdown in the case of an Appendix R Fire. The controls and Indications at the ACS are arranged to allow remoto manual startup, oporttlon and shutdown of HPCI. Automatic operations of the HPCI system, including Prirnary Containment isolation System functions are not roquired for attemative shutdown, and consequently, are not re-established at IFo HPCI ACS.
C. Big.ow Fon CHawor:
This modification was necessary to meet 10 CFR 50 Appendix R requirements.
D. SMETY EvatuatroN
SUMMARY
1)
Does this modification increase the probability of occu rence or the consequences of an accident or malfunction of equipment important to safety as previausly evaluated In the safety analysis report?
Answer:
No, in the event of an Appendix R fire, and the use tI the ACS, the time available for manual operation of HPCI is sufficient to keep the core covered.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated prevlously in the safety analysis report?
Answer:
No. The possibility of an accident or malfunction of a different type than evaluated previously in the l
safety analysis report is not created. This modification does not change the operation of HPCI and RCIC systems as described in the UFSAR when all of the transfer / isolation switches are in the
' normal' or " test' position. Test and emergency switch positions are annunciated in the Main Control Room to alert the operator of these abnormal conditions so that the system can be restored to normal if there is no fire.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. The safety function of the HPCI and RCIC systems are not affected by the rerouting of circuits and the addition of a safety related transfer / Isolation switches, and the automatic initiation and trip features of the system are not affected when the transfer switches are in the ' normal' or ' test' l.
positions. The use of this HPCI ACS panel (le when the transfer / isolation switches are in the altomative modo) to respond to an Appendix R fire is required to ensure safo shutdown. The Appendix R fire will create a need for temporary depanure from the Technical Specifications in order to mitigate the consequences of the fire. 10 CFR 50.54 allows departure from the Technical Specifications in an emergency such as an Appendix R fire.
l 22
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 198910 CFR 50.59 REPORT i
CAC And CAD Llauld N2 StoraceNaporization Modification MoomcatioN No.: 1418 A. Systru: Containment Atmosphere Control (CAC) and Containment Atmosphere Dilution (CAD)
'l B. Drsemption:
This modification provides automatic closure of control valves on sensing cold nitrogen & 40 degrees F) in the injection lines. New trip units, which take signals from temperature elements in the injection lines provide a signal to the control room annunciators to indicate closure of the valves.
C. ReasoH Fon CMawor:
This modification Improve's the system's capability to detect and mitigate the consequences of cold k40 degrees F) N2 delivery.
D. Sarrty EvatuatioH Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This modification improves the system's capability to detect and mitigate the consequences of cold (140 degrees F) nitrogen, and provide isolation signals to nitrogen injection line control valves. The design basis of the CAD system includes the capability to prevent the occurrence of e
flammable mixture of gases in the primary containment. Although this modification provides automatic Isolation of the CAD Injection lines, the flammable gas generation rates are low enough
- so that a temporary Inability to inject nitrogen into containment will not adversely affect plant safety
. prior to the operator taking manual corrective actions.
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. Although this modification will provide automatic isolation of the nitrogen injection lines, the flammable gas generation rates are low enough so that a temporary inability to inject nitrogen into contalnment will not adversely affect plant safety prior to the operator taking manual corrective i
actions.
- 3) -
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. Technical Specifications 3/4.7.A.5, 3/4.7.A.6, 3/4.7,0, and Table 3.7.4 along with all associated bases which were applicable were reviewed. All design requirements app!! cable to the original equipment were app!Ied.
t 23
PEACH BOTTOM ATOMIC PLWER STATION UNIT 2 DOCKET No. 50-277 198910 CFR 50.59 REPORT Lnstallation Of A 3-Way Diverter Ball on Valve Unstream Of The Relief Vatves On Containment Atmosoherie Dilution (CAD) System Moowication No.: 5000 A Smsut CAD B. Desemmtion:
This modification installs a 3-way diverter ball valve upstream of the nitrogen storage tank pressure relief valve and provides flanged rollel valve inlet piping connections.
C. Reason Fon CHANoE!
To upgrade testing and maintenance provisions for CAD system Ilquid nitrogen storage tank pressure rellef valves. This modification allows isolation of one relief valve for testing or repair while the second tellef valve continues to provide system protection. Flanged relief valve Inlet piping connections facilitate removal of the relief valves.
D. Sartty Evatuatiow Suuuaav:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The diverter valve is a full ported ball valve and is locked in the correct position to ensure there -
is no restriction in the flow path to the relief valve.
1 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?
Answiti No. The only safety function of the diverter valve and new flangee is a passive one (maintaining pressure integrity under all loading conditions).
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Speelfic?tions?
Answer:
No. The addition of a diverter valve will result in only one of the two rollel valves being available i
at any one time. However, she margin of safety is not reduced, since the design intention was for l
the rupture disk to provide the necessary redundancy for single failure protection.
I
\\;
i I
24
PEACH BOTTOM ATOMIC PEWER STATION UNIT 2 DOCKET No. 50-277 198910 CFR 50.59 REPORT HELB Vent Path Modification Mooineatiow No.: 5084 A. SysTru: High Pressure Coolant injection B Drscuriow:
This modification secures the blowout panel in the ceiling of the HPCI toom (room #6) of Unit 2 and strengthens the door from stalt tower 24 to room 105. Other openings were evaluated during a walkdown and were left as found.
C. Reason Fon CHAwor:
To assure rollef during design basis events such as a high energy line break (HELB), tomado, or Internal flood.
D. Sarrty Evaluation Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. All equipment in those rooms which experience an increase in environmental conditions following a HELB have been reviewed and found acceptable for those conditions. Also, the environmental conditions for five other rooms were found to be less severe than their original design.
Neither the effect of a tornado nor internal flooding is affected by this modification. Structural integrity is maintalned for HELB, tornado, and internal flooding.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No. This modification does not affect normal environmental conditions; it only effects environmental conditions following a HELB or depressurization following a tornado. Additional accidents following a HELB need not be postulated and the modification does not affect the tornado analysis.
l l
0)
Does this modification reduce the margin of safety as defined in the basis for the Techr.lcal l:
Specifications?
AQl. WAG No, The increase in temperatures and pressures does not affect the qualification of equipment and structures. The environmental conditions in five rooms are less severe than their original design and neither the effect of a tornado nor intemal flooding is affected by this modification.
25
i PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 60 277 198910 CFR 50.59 REPORT Edocate 2A (Reactor Buildina Coofina Water) RBCW Pumo Motor Fee;l MoovicaTion No.: 5157 A. Systau: Reactor Building Closed Cooling Water B. Dnga,ignes This modification re routes the power feed and associated control wiring for the 2A "CW pump from Motor Control Center (MCC) 20B36 to MCC 20B27.
C. Bg ng1 on CHAwoa:
F This will tqualize loading between the diesel generators and their associated power distribution equipment followir's a postulated Loss of Coolant Accident coincident with loss of offsite power.
D. Rarrty EvatuatioN
SUMMARY
1)
Does this modification increase the probability of occurrence or the consequences of an acciden' or malfunction of equipment Important to safety as previously evaluated in the safety analysis report?
Answer:
No. This modification oniv relocaterithe power feed. The pump is not being physically disturbed or relocated. The pump s functional operation is 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?
Answe.n As the RBCW system has not been physically changed, no new equipment malfunctions outside those previously evaluated in the FSAR are deemed credible.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No margin of safety addressed in the Technical Specifications is affected.
(
l 26
1
]
e PEACH BOTTOM ATOMIC POWER STATION UNIT 2 DOCKET No. 50 277 1989 to CFR 50.59 REPORT Condensate Transfer Jockey Pumo Discharae Line Ploina Chance Out i
Mooineatiow No.: 88-096 A. SysitM: Condensate 1
B. Drsemmtion:
This modification changes the 3' pipe from schedule 10S to schedule 40S.
C, Reason Fon Cwawor:
The change to 40S is more advantageous since it will provide a greater wall thickness to weld the two 3/4' hall couplings required and withstand the vlbration that occurs in the cond6nsate transfer piping.
D. Sarrry Evatuatiow SUMMany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The new pipe will exceed present pipe class pressure and temperature ratings.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answert No. This modification will not effect pump capacity or pump characteristics.
3)
Does this modification reduce the margin of safety as defined in the basis for ihe Technical Specifications?
Answer:
No. This modification makes no changes to the pump capacity or characteristics. The rnaximum allowable worxing pressure of the new piping exceeds the system pipe class design rating.
E 27 L
i
PEACH BOTTOM ATOMIC POWER STATION UNIT i DOCKET No. 50 277 198910 CFR 50.59 REPORT Reactor Water Clean Uo (RWCU) Backwash Transfer Check Vwve MoonientioN No.: 88497 A.' Systru: 'RWCU
'~
B. Drsemmtiow:
j This modification adds a 3/4* check valve to the seal water lines on the RWCU backwash transfer p;mp.
C. Reasow Fon Cwawoc This modification will prevent backflow from the RWCU into the condensato system.
D. Surry Evatuattow Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This will reduce the possibility of cross contamination to the condensato system from the RWCU.
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. FSAR sections 9.2 and 9.3 were reviewed for effects of this modification and none were found.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. The Technical Specifications are not affected by this modification. Sections 3.8 and 4.8 were reviewed to make this determination.
28
1 e
PEACH BOTTOM ATOMIC POWER STATION i
UNIT 3 DOCKET No. 50-277 & 50-278 198910 CFR 50.59 REPORT Redunditnt Temnerature Monitorina Sucoression Pool fAceineattow No.: 6038 A. Systrut Suppression Pool
,l B. Drsewtiow:
This modification upgrades and replaces the existing temperature monitoring system with a redundant suppression pool temperature monitorint, system. The redundant system consists of thirteen resistance temperature detectors mounted in previously installed thermowells in the torus, a processor / indicator / printer located in the control room, and recorder located in the control room.
C. Reason Fon Cuawor:
The modification was required to meet the requirements of Appendix A of NUREG-0601,
- Mark l Containment Long Term Program.' It also satisfies the requirements of Regulatory Guide 1.97 Revislen 2, ' instrumentation for Light Water Cooled Nuclear Power Plants to Access Plant and Environs Conditions During and Following an Accident.'
D. Sarrry Evawatiow Supuany:
1)
Does this modification increase the probabi'lty of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
i No. This modification provides quality assured systems that are environmentally and seismically quallfled and installed as an engineered safeguard system.
l 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?
8.D.9EtG No. This modification provides quality assured systems that are environmentally and seismically quallfled and installed as an engineered safeguard system.
3)
Does this modificatlon reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. Thls change is necessary to meet the requirements of NUREG 0661 and to satisfy the requirements of Regulatory Guide 1.9).
29
1 PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 54278 i
198910 CFR 50.59 REPORT Hiah Radiation Trio of Containment Vent and Purae Unos
.hiooincatiow Ngg 064 A. Systru: Primary Containment Isolation System (PCIS)
B. Drsemmnow:
Thls mcdifies the primary containment isolation system on Peach Bottom Unit 3 by adding a trip signal from the offgas stack radiation monitors to the control circuit for the containment vent and purge isolation valves.
C.Bpsow FoM CHawgt; This change is required by item ll E.4.2(7) of NUREG 0737
- Clarification of TMl action Plan Requirements *,
D. Sarrry EvatumQN SUMMARJ 1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The modification adds an additional containment isolation signal not previously part of the plant design. The non-safety related trip signalis isolated by qual!fied relays, so that failure of this signal leaves the plant in the same configuration as evaluated in the UFSAR.
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?
Ariswe.n No. The modification only adds an additional contalnment isolation signal. The fa!!ure of this signal would leave the plant in the same condition as evaluated in the UFSAR.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
.Answen No. The addition of the new isolation signal has no effect on existing trlp settings as evaluated using Technical Specifications 3.2D,4.2D,3.70,4.70,3.8.C. 8 A and their associated bases.
P L
I' l
l L
30
PEACH BOTT@M ATOMIC POWER STATION UNIT 3 t
DOCKET No. 50 277 & 50 278 198910 CFR 50.59 REPORT Boron Solution Enrichment Mooincation No.: 867 A. Systru: Standby Liquid Control B. Drsenetton:
This modification changed the Standby Uguld Control (SLC) solution from the existing solution to a solution enriched with Boron-10. The tank low and high level alarms, low level heater permissive, and low temperature alarm setpoints were changed to meet the new solution properties.
C. Reasow Fon CHawor:
The modification was made for compliance with 10CFR50.62, which requires the SLC system to have a minimum flow capacity, boron concentration and enrichment equivalent in control capacity to 86 gpm of 13 weight percent natural sodlem pontoborate solution.
D. Sorry Evatuation Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
AILIMG This modification will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety as previously evaluated in the UFSAR because the use of the SLCS is not speelfically required for any translent or accident evaluated in the UFSAR. Also,
.the new solution will not change the probability of a malfunction of the SLCS and since this modification will cause the reactor to shut down faster, the effectiveness of the SLCS will be improved.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
A!1tEtG This mod!!! cation does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the UFSAR because the use of the SLCS is not speelfically required for any transient or accident evaluated in the UFSAR. Also, the Inst " %n of the enriched solution does not create any new types of accidents.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical 3
Specifications?
Answer:
This modification does not reduce the margin of safety as defined in the basis for any technical specification since this modification results in the SLCS shutting down the reactor faster than the current design during an A1WS event.
31
PEACH BOTTOM ATOMIC POWER STATION
-a UNIT 3 DOCKET No. 50-277 & 50-278 198910 CFR 50.59 REPORT Reactor Feed Pumo Sealinlection Ploinc Mooincanow Nod 916 A. S.yngu_1 Reactor Feed Pump B. Desenenow:
This modification involves the removal of existing 2 Inon and 1 inch Reactor Feed Pump (RFP) seal injection piping and replacing with 4 inch and 2 inch piping respectively.
C. RsAson Fon CHANGE:
I This modification was made to reduce pressure losses and allow RFP seal in,ection water to be supplied at a pressure to properly seal the pump shaft and prevent damage to the sea s and bearings.
D.- SanTY EvAtuanow Suuuany:
i, 1)
Does this modification increase the probability of occu.rence or the consequences of an accident l.
or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
AD1EEG No. None of the equipment involved is safety related. The new piping will reduce pressure losses y
and akw RFP seal injection water to be supplied at a pressure to properly seal the pump shaft and i'
prevent damage to the seals and bearings.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previous v in the safe'y analysis report?
Answer:
i
?,
No.' Notia of the piping or equipment is safety related. There is no electrical work involved.'
i I
N 3)
Does this modification reduce the margin of safety as defined in the basis for the Technical
- Specificatior:s?
l.
Answer:
No. This modification does not make any changes to the Technical Specifications. None of the work is O listed or safety related.
s y
i e
it
,I 3
'l'.
.i j
k
.y
l 1
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT Reolace Differentia! Pressure Switches MODWic4 TION No.: 936 e
A. Systru_i Core Spray B. Drsenwrios:
i Replace differential pressure switches DPIS 3-14 43A and B C. Reason Fon CHAnos:
Existing switches did not have a negative value scale and were calibrated to read 0 during cold shutdown, and positive if a core sprayline break occurred during normal operation. The lastruments were pegged at 0 during normal operation. This modification permits a negative pressure differential indication during normal operation.
D. Sarrry Evagation SUMMAnY:
- 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?
AfilXtG
.No. These switches are an Improvement over switches that already exist and the design criteria of
.this modification is in accordance with all criteria applicable to original equipmen'..
i 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated provlously in the safety analysis report?
Answer:
No. The design criterla of this modification is in accordance with all criterla applicable to original-equipment.'
- c. ;
i-3)
Does this modification reduce the margin of safety as defined in the basis for.the Technical Specifications?-
t Answer:
No. The switches will be surveillance tested as required in the Technical Specifications Table 4.28.
l l
r i
i 1
33
se PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 198910 CFR 50.53 REPORT Installation Of Flances on Hiah Pressure Service Water System Moovication No.: 1243 A.Aystru: High Pressure Service Water (HPSW)
- B. Assenetion:
installation of flanges on the HPSW System to allow for a temporary flow path of the Residual Heat Removal (RHR) heat exchanger cooling water in the event that the normal flow path becomes unavailable.
C. REASON Fon CHANOE*
This modification was completed to create an alternative for RHR flowpath if needed.
' D. Sarriv EvaluatioH Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The HPSW temporary flow path will be used only during periods of cold shutdown and when
. fuel is not being moved in the reactor building.
- 2)
Does this modification create the possibility for an accident or malfunellon of a different type than
- any evaluated previously la the safety analysis report?
Answer:
The only possible failure would be a room flood' caused by a hose rupture. Should a hose fallure occur, the consequences are bounded by a pipe failure of the RHR or HPSW piping in the same.
1
. room. The consequences of internal flooding of the RHR rooms have been previously analyzed in
. the UFSAR.
1 3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? '
Answer:
No. The function and reliability. of the system is unaffected using the bases in Technical Specifications 3.5B and 3.70.
t 34 e
.=
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DCCKET No. 50-278 198910 CFR 50.59 REPORT Installation Of Safety Grade N2 Suoolv jdooirecAtioN No : 1316 A. Systru: Primary Containment B. DesCRIPiloN:
Installation of a safety grade N2 supply to the containment isolation valves C. REASON Fon CHanor:
This system will remove the need for the bottled gas supply, and the associated requirement to establish o
and maintain air supply system leak rate criteria for individual users, and will allow verification of adequate system performance by functional testing.
D.,$.ArrrY Evatuatiow SUMMAnY:
1)
' Does this modification increase the probability of occurrence or t.
ansequences of an accident or malfunction of equipment important to safety as previously evalual. 'in the safety analysis report?
Answer:
No. Design requirements applied to this modification include, but are not limited toi seismic qualification, quality assurance, testability, maintainability, original equipment performance specification, environmental qualification, separation criteria, safeguard power sources, and-protection from jet Impingement.
- 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 criteria used as a basis for the protection and evaluatlon of one plant component from adverse interactions of another plant component during various operating conditions comply with
'the UFSAR.
s 1
3)
Does this modification reduce the margin of safety as defined in the basis for the. Technical t
R Specif: cations?
Answer:
No. Technical Specifications 3.7 and 4.7 have been reviewed. Allload stress acceptance criteria for all components evaluated and/or repaired by this work comply with the original design code.
requirements specified in thw UFSAR or as allowed by ASME Section XI.
f 1
35
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT l
AC Power Distribution for the Altemate Control Station MRpmeATiow No.:
1353E A. Systru:
4KV Circuit Breakers.
?
- ~
, B. Deseniptiom Provide alternative control stations for the 4KV circuit breakers that feed theUnit 3 *B' and 'D' safeguard channelload center transformers.
-l C. B.sAsow Fon Cuawor:
This modification assures safe shutdown in the event of a design basis fire, and is ne. assary to meet the requirements of Appendix R to 10 CFR 50.
D. SarrtY Evaluation SUMMAnVl
' 1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safetyas previously evaluated in the safety analysis report?
Answan No. The safety function of the circult breakers and their loads are not degraded by rerouting circuits through the transfer / Isolation switches at the Alternative Control Station.
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:
1 No. The modification does not change the AC power system as described in the UFSAR.
3)
.Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? -
Answer:
No. The use of the alternative shutdown mode is not described in the technical specifications. This
-modification was performed to conform with 10 CFR 50 Appendix R requirements.
i
)
36
i PEACH BOTTOM ATOMIC POWER STATION UNIT 3
{
DOCKET No. 50-278 198910 CFR 50.59 REPORT Deslan and Installation of Alternative Shutdown Caoability MoomicAviow No.: 1353G A. SYstru: Automatic Depressurization System (ADS)
B. DesenieTiow:
This modification reroutes safety related circuits to transfer / isolation switches and to alternative controls for three ADS valves and two Nitrogen supply isolation valves from the normal control panel to the HPCI Alternative Control Station (ACS), and Isolates safety related ADS circuits that could adversely affect safe shutdown in the event of fire.
C. RcAsow Fon CnAwar:
This modification was necessary to meet 10 CFR 50 Appendix R requirements.
D. Sorry EVALUATloN
SUMMARY
1)
Does this modification incrase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The safety function of the Autornatic Depressurization System is not affected by the rerouting of circults or the inclusion of a safety related transfer / isolation switch. There are no new unanalyzed operating conditions introduced.
j 2)
Does this modification create the possibility.for an accident or malfunction of a different type than
[
any evaluated previously in the safety analysis report?
I-Answer:
No. Since the ADS is: not: affected by rerouting circuits or inclusion-of a. safety related transfer / isolation switch, this modification does not pose a threat of an accident or malfunction of l
a different type than previously evaluated.
3)
Does this modification reduce the margin of ' safety as defined in the. basis for the Technical Specifications?
q Answer:
[
No. This modification is necessary to meet.10 CFR 50' Appendix R requirements. It does not
' degrade the Integrity of the circuits involved.
37
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 54278 198910 CFR 50.59 REPORT Process Monitorina Instrumentation MooiricatioN No.:- 1353H A. S m eu; Miscellaneous B. Desemption:
This modificadon provides process instrumentation necessary to assure safe alternative shutdown following a fire in chner the Main Control Room (fire area 29), the Cable Spreading Room (fire area 20), or the Emergency Shutdown Area (fire area 25). The alternative process instrumentation includes reactor vessel water level and pressure, suppression pool water level and temperature, drywell pressure and temperature, safety relief valve discharge temperature, and condensate storage tank water level. Indication for this instrumentation will be provided at the Unit 3 HPCI Alternative Control Station.
C. Reason FoR Cuanoe:
This modification is necessary to satisfy 10 CFR 50 Appendix R Alternative Shutdown Requirements.
D. SareTy Evatuation
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 modification will not affect the operation of the equipment monitored. No Chapter 14 analyses as previously evduated will be altered as a result to this modification.
t 2)
Does this'modillcation create the possibility for an accident or malfunction of a different type than -
any evaluated previously in the safety atWy=Is report? -
AnlW1r; No. The modification is being installed to meet the fire protection' requirements of 10 CFR 50 Appendix R.
y 3)
Does this modification reduce the margin of safety as defined in the basis'for the Technical -
i Specifications?
A:
Answer:
L No. Th!s modification will not affect the operation of the systems to be monitored. The monitoring instrumentation is not discussed in the Technical Specifications.
4
- i p
q L
38
~
PEACH BOTIOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 198910 CFR 50.59 REPORT MoomicAvion No.: 1359 A. SysTeu: Reactor Protection Systern (RPS)
B. Desenmtiow:
This modification replaces the existing static inveiter and adds an adjustable voltage transformer between the inverter and the RPS protection panel, and relocates the RPS alternate feed to the distribution panel supplied by the new inverter.
C. REA$oN Fon CHANOE:
This replacement is necessary because of repeated component failures and unavailability of replacement parts. The change in power supply will reduce challanges to the RPS.
D. SArrry EVALUATloN
SUMMARY
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malf unction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The existing isolation of the non. safety related inverter from its safety related power supplies is maintainedc 1
2)
. Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, This modification increases the rellability of the RPS attemate supply and the uninterruptible l
power supply system. The station batteries and chargers have sufficient capacity to supply an RPS -
channel load through the Inverters.
'. 3)
Does this modification reduce the margin of safety as defined in the basis for the Technical F'
Specifications?
- Answer:
l Based on a review of Technical Specification sections 3/4.1 and 3/4.9, this modification does not reduce the margin of safety. The invertor and its loads are not addressed in the Technical.
Specifications. RPS power supply trip points are unchanged.
s 4
, :i.
{
i 3 l~
39 l'
PEACH BOTTOM ATOMIC POWER STATION UNITS 3 -
DOCKET No. 50 278 198910 CFR 50.59 REPORT Remove 600 oslo Scram Slonal Mooincarios No.: 1404 A. SysTru: Reactor Protection B. Descamtion:
' ^ ~ '
Eliminate the 600 psig scram signal in startup, refuel and shutdown modes as it relates to low condenser vacuum or main steam isolation valve closure.
C. REASON Fon CHANGE!
This modification is necessary to facilitate warming of the main turbine.
D. Sarrry Evatuation Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment lmportant to safety as previously evaluated in the safety analysis report?
Answer:
No. This is not a safety limit setting. A test performed by General Electric on another BWR 4 plant demonstrated that the 600 psig setting is not required.
i 2)
Does this modification create the possibility for an accident or malfunctiori of a different type than any evaiuated previously in the safety analysis report?
Answer:
. No. Tests conducted by General Electric demonstrate that removing the setting does not create -
any safety concerns.
3)
- Does this modification reduce the margin of safety as defined in the basis for the Technical-Specifications?
Answer:
a No. This modification does not reduce the. margin of safety. as defined in the Technical q
Specifications. However, changes were made to the Specifications to reflect the change made by
.this modification. These changes were approved by the NRC on 3/14/86.
l u
li
~ ~.
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 198910 CFR 50.59 REPORT t
Unit 3 Reolacement of Recirc & RHR System Ploina MooincanoN No.: 1536 I
A. Sysicu:
Reactor and associated piping systems,
B. Descamnow:
This modification involves the removal and replacement of the following piping or components unless otherwise noted:
1.
The Recirculation System piping (Loops A and B).
2.
. The Residual Heat Removal (RHR) System Shetdown Cooling Suction and Return piping inskle containment and containment penetrations.
3.
RHR Head Spray piping insido containment.
4.
The Reactor Water Clean Up (RWCU) containment penetration including the piping through the first inside containment.
5.
The RWCU piping outside containment from the containment penetration to the RWCU pumps and
~ from the pumps to the re(,enerative heat exchanger.
6.
The Reactor Pressure Vessel (RPV) Drain piping.
7.
The 2 Jet Pump Instrument Seals, at the N-8A & N-88 RPV nozzles.
E 8.
The ten Recircu!ation inlet nozzle safe-end N 2.
i 9.
The two Recirculation Outlet safe ends N-1.
10.
The two core spray safe ends N 5 and closure spools.
11.
The existing Recirculation and RHR Shutdown Cooling Systems mirror insulation with soft, fiberglass,.
blanket type insulation.
i
' 12.
Removal of the recirculation pump flow splitter.
13.
Installation of new larger torus strainers for RHR & core spray.
i i
- 14. '
Permanent removal of cross tie piping and equalizer valves.
Permanent removal of pipe whip restraints associated with arbitrary Intermediate breaks (AIB) on 15.
4 the recirc system; 3
1
- 16. -
CRD-HSR nozzle cap (N-9).
'i C. Reasow Fon CHawor:
L This modification is being performed to replace existing IGSCC susceptible Type-304 stainless steel piping and safe ends with materials that do not sensitize during welding and have a low susceptibility to IGSCC.
u PEACH BOTTOM ATOMIC POWER STATION i
UNIT 3 j
DOCKET No. 50-278 198910 CFR 50.59 REPORT Mooincarion No.:.1536 (continued)
I D. SareTy Evatuation Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident 3
or malf unction of equipment important to safety as previously evaluated in the safety analysis report?
i
~ ~
~
Answer:
h No. The design and material standards specified moet or exceed the design and material standards 1
currently specified in the UFSAR. The overall system performance due to the redesign will not result j
in the systems being operated outside of thelt design or tested limits and will not lead to excessive
~
vibration or waterhammer, The design of the electrical portions of the modifications is in accordance l
with applicable safety design specifications.
+
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
l No. The design and material does not add any new fluid systems, add any new permanent electrical systems, add any combustible loading, change any fire detection or suppression capability, change j
any penetration location or type, change any safety related function, or change any safety related conditions.
i, 3)
Does this modification reduce the margin of safety as defined in the basis for the Technical
)
Specifications?
Answer:
No. The design and material will not adversely affect the operatlon of a system important to safety considering such items as flow, chemistry, power, instrument set points, level, pressure, temperature, or radiation levels. It wil not create a system configuration or operating condition such that the Technical Specification limiting conditions for operation or surveillance requirements are no longer -
1
. adequate. The design and materlats will not bypass or Invalidate automatic activation features ;
y required to be operable according to the Technical Specification.
p 1
i L
1 1,
b 1:
1 p.
42 l
PEACli BOTTOM ATOMIC POWER STATION UNIT 3 j
DOCKET No. 50-278 198910 CFR 50.59 REPORT i
Drvwell Cooler Fan Loalc Mooincarion No.: 1542 1
A. SysTru: Drywell Cooling
.- s.
B. DESCRWTION:
h
- This modification changes the control logic of drywell cooler fans 3AV26 through 3GV26 to bypass the high drywell pressure or low reactor water level trip signal. This change implements the NRC recommendations 3
to improve the plant recovery from high drywell pressure (above 2 psig).
C. Reasow FoR CHawor:
This modification Implements the NRC recommendation to improve the plant recovery from high drywell
.{
pressure (above 2 psig) per 1.E. Information Notice C4 35.
I L
D. SAreTv EVALUAftoN
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:
a h
The probability of occurrence or the consequences of an accident or malfunction of equipment Important to safety previously evaluated in the safety analysis report ' es not been increased, ' Failure of the new bypass circuit may result in increased loading on the emergency buses, however the additionalload will not exceed the capacity of the diesel generaters.. The equipment added by this j
' modification has been located in safety related panels and has t een mounted in a similar manner.
to that of safety related components to prevent damage to the s ifety related equipment inside the i
. panel.1 1
, 2)
Does this modification create the possibility for an accident or malfunction of a different type than 1
any evaluated previously in the safety analysis report? '
i
! Answer:
1 h
The possibility for an accident or malfunction of a different type than any evaluated previouslyin the
'l safety analysis report has not been created. Failure of the bypass circuit does not result in the placement of unacceptable loads on the oriergency bus or its associated diesel generator.- This i
change allows the operator to restart the dr) well cooler fans from control room to improve the plant recovery from high drywell pressure. Loss of con:rol power or loss of the trip signal (high drywell "N
pressure or low reactor water level trip signal) de energizes the bypass relay and automatically.
removes the trip bypass signal.-
3)
Does this modification reduce the margin of safety as defined in the basis for the Technicali Specifications?
Answer:
i L
The margin of safety as defined in the basis for the Technical Specifications has not been reduced since there are no applicable technical specifications basis sections.
43 b/
1 3,-
PEACH BOTTOM ATOMIC POW' R STATION d
UNIT 3 00CKET No. 50-278 -
198910 0FR 50.59 REPORT-Recirc M.G Set Room Sorinkler Exnansion l
h MoovicATiow NoJ 1636 A. SYSTEu: Fire s
B. DESCRIPTION:
The Reactor recirc pump's M-G set sprinkler system was expanded to provide total room coverage.
C. REAsow Fon Cwawor:
l This change was made to protect structural steel members which support 10 CFR 50 Appendix R safe shutdown fire barriers.
D. Sarriv EvatuaTiow SuuuaRv:
,1)
Does this modification ir:rease the probability of occurrence er the consequences of an accident -
or malfunction of equipm r nt important to safety as previously evaluated in the safety analysis report?
Answer:
No. The overall result of his modification is increased assurance of reactor safety by improving plant -
fire protection and saft shutdown capability in accordance with Appendix R without adversely affecting any safety features.
' 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
I.
Answer:
No. Safety equipment which could b:: ::Wemely affected in the unlikely occurrence of inadvertent i '.
water discharge was protected from direct water impingement by sealing the raceway penetrations into the top of the equlpment and by providing deflector hoods over the equipment.
u 3)
Does this modification reduce the margin of safet', as defined in the basis for the Technical':
-I L
Specifications?
L Answer; -
a 7
No. The potential damat e to electricai equipment that may be caused by actuation of the sprinkler -
system is enveloped by the Appendix t' Safe Shutdown Analysis described in the Fire Protection -
3 Program.
'i
!7 l
44
r l'-
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-277 & 50 278 198910 CFR 50.59 REPORT Gate Valves Sublect To Therraal Pressure Lockina MODWICATION No.: 1678 A. S m tut Reactor Recirculation
_ B. Desemimqm
- This modification involves drilling a small hole in the downstream disk of the reactor recirculation discharge valve to prevent thermal pressure locking.
t C. REASON Fon CHANGE:
Some gate valves have been found to bind during cool-down of the system. These valves will not reopen on heat up until the original temperatu o (at valve close down) is reached.
This modification will assure proper opiration of safety related valves.
D. SArrry EvAtuATioN SuuuAny:
1)
Does this modification increase the p.obability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
' Answer:
No. The small hole in the downstream disk of the reactor recirculation pump discharge isolation valve does not compromise the safety of the plant. The drilled portion of the disk is not a part of the required downstream pressure boundary. The hole is of_ insignificant size and is located to
' minimize bypass leakage. The function of the valve is not affected.-
2)
. Does this modification create the posslbility for an accident or malfunction of a different type than any evaluated previously In the safety analysis report?
j Answer:
No. The valve does not perform an active safety function and is only required to maintain pressure integrity.. The upstream disk required to maintain pressure Integrity was not modified, t,
d:
6)
- 3) J
.Does this modification reduce the rnargin of safety as defined in the basis fcr the Technical Specifications?
n"'
~
Answer:
No. Since the disk that was modified does not function as a pressure boundary daring any ncrmal-or accident condition, the hole in the disk does not affect the safety of the system.
o d
(
3,,
45
.~
7-j PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No.' 50-277 & 50-278 198910 CFR 50.59 REPORT
_ Install Control Vafves on 3rd and 4th Feedwater Heater Extraction Steam Drain Lines
_Mooirication No.: 1684 A. SysisM: Feedwater
' B. OcseniprioN:'
Install control valves on the 3rd and 4th feedwater heater extraction steam drain lines.
C. RsAsow Fon CHAnos:
This modification is necessary to prevent turbine water induction.
D. Sarrry EvAtuanow
SUMMARY
1)-
Does this modification increase the probabnity of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safetyarclysis repo Answer:
i No. The ' addition of valves only affects the feedwater heater extraction steam draln lines which are not safety related._. The power, instrumentation and control utilities for the new valves were constructed in accordance with applicable separation and fire protection criteria.
y 2)
Does this modification create the possibility for an accident or malfunction of a different type tha any evaluated previously in the safety analysis report?
Answer:
No. The new valves are not required for safe shutdown and are not capable of impalring the a to achieve safe shutdown.
3).
Does this-modification reduce the margin-of safety as defined in the basis for the Technical m
Specifications? :
Answer:
No. The feedwater extraction steam drain lines are not addressed in the Peach Bottom Specifications.
l y
l E
46 F
.m PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 198910 CFR 50.59 REPORT' Reactor Feed Pumo Discharoe Valve Motor MooineAnow No.: 1713 A. Systru: Feedwater
. ~. >.
B.pesemPUoN:
This modification involved replacing the motor on a reactor feed pump discharge valve (MO 31498).
C. REAsow Fon CHawor:
The existing motor needed repair. There was a lengthy period required for that repair. The existing motor
- was therefore temporarily replaced with an equivalent one.
D. SAFrTY EvAtuanow Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident -
or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
I No.. Although the replacement motor increases the total weight of the valve operator by 56%, the.
r added weight.does not signllicantly increase stresses in the valve and valve operator. The replacement.ls functionally identical to the original.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evalua' previously in the safety analysis report?
E Answer:
.No. Although the replacement motor increases the total weight of the valve operator by 56%, the
- i
- added weight does not significantly. increase stresses in the valve and valve operator. The
l replacement is functionally identical to the original.
3)-
Does this modification reduce the marg!n of safety as defined in the basis for the Technical Specifications?.
8,nswer:
No. The valve has no safety related function and is not listed in the Technical Specification as a-i
- containment isolation valve for the feedwater line.-
l _. /
I l.
1 47-i.
PEACH BOTTOM ATOMIC POWER STATION y
UNIT 3 DOCKET No. 50-278 198910 CFR 50.59 REPORT P
Reolace Existina Flow Transmitters MooiricArioN No.: 1891 A. Systru: Various
~.
- 6. DESCRIPfloN:
This modification replaces existing flow transmitters with environmentally qualified flow transmittdrs In various systems. It also replaces existing signal conditioning instruments associated with these flow transmitters to accommodate the new transmitter with 4 20 ma output.
o C. PEASON FoM CNANor:
This modification is necessary to meet a Reg. Guide 1.97 commitment.
D.- SarrTy Evaluation
SUMMARY
1)
Does this modification increase the probability of occurrence or the consequences of an accident.
or malfunction of equipment important to safetyas previously evaluated in the safety analysis report?
Answer:
No. -This modification replaces existing flow transmitters with environmentally qualified flow transmitters in various safety systems. This modification does not change the operation of the safety systems involved.
?
- 2)
' Does this modific9 tion create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
lp, Answer:
No. This modification will help the operators read the flow Indications more accurately.-
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
,1 3 Answer:
No. Technical Specifications 3.5.A,3.5.C,3.5.D, and Table B were reviewed to determine that this modification does not require a change to the Technical Specifications.
l h
lf^
a I
m3 48 f
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 s
198910 CFR 50.59 REPORT Uoarade Instrumentation for Emeroency Ventilation System Mooirication No.: 1892 A. SystrMt Ventilation r - -
B pesemiption:
Upgrade accident monitoring instrumentation for various safety related nmergency ventilation systems.
.f C. RcAson Fon CHAwor:
This modification is necessary to comply with Regulatory Guide 1.97.
D.' SartrY EVALUATION SUMMAnY:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. The addition of the monitoring equipment will enhance the operator's ability to mitigate the i
consequences of an accident.
l 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 1 Answer:
.No. This modification meets the requirements of Regulatory Guide 1.97 and the design requirements L
of the affected ventilation systems.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answeh:
No. Technical Specification bases 3.5.A H; 3.7.B,C; 3.11.A; 4.5.A,H; 4.7.B,C and 4.11.A have been reviewed and are not affected by this modification.
\\
i l
49
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 i
DOCKET No. 5&278 1
198910 CFR 50.59 REPORT I
i installation of Back Un Nitrocen Sunoiv to Non Automatte Deoressurization System Main Steam Reflef Valves MoowicATion No.: 1950A A. SysTru; Automatic Depresurization
.o
~
B. Descamriow:
0
~ Backup pneumatic nitrogen supp'y was Installed for safety relief valves RV3 02-71E, H, J.
C. REAsen Fon CHANOE:
t This modification was required to comply with Appendix R to 10 CFR 50.
D. SAnTY EVALUATION SUMMAnY!
1)
Does this modification increase the probability of occurrence or the consequences of an accident f
or malfunction of equipment Important to safety as previously evaluated in the safety analysis rep Answer:
No. The backup nitrogen system was installed as a 0 listed selsmic system and compiles with primary containment isolation requirements.
2)
Does this modification create the possibility for an accident or malfunction of a different type than
-l any evaluated previously in the safety analysis report?
~
t; Answer:
t No. The Installation of the system does not alter the function of the safety tellef valves.
+
3)
Does this modification reduce the margin of safety as defined in the bas!s for the Technical j
- Specifications?
Answer:
No. The new system increases the reliability and operability of the safety relief valves.
n
-)
50
a PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT.
t Relocation of Motor Controls
_Moorrication No.: 2079 A. SYSTEM: High Pressure Coolant injection (HPCI) and Reactor Core Isolation Cooling (RCIC)
' B. Deseniirm!g
~
Relocation of power and/or motor control cables for motor operated valves, MO-3 2315 (High Pressure Coolant injection System) and MO 31315 (Reactor Core Isolation Cooling System).
C. RsAsow Fon CwAwar:
This modification is necessary to bring the 2 motor operated valve circuits into compilance with Appendix R of 10CFR50.
D. SAFETY EvAtuaTrow SuuuanY:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment irnportant to safety as previously evaluated in the safety analysis report?
Answer:
No. Modification 2079 relocated power and control cables for specliic critical valves, thus assuring redundant power and control to these valves in case of an Appendix R fire situation. The functions of the HPCI and RCIC systems are unchanged by this relocation.
' 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 operability of the HPCI and RCIC systems is unchanged by this modification.
' 3) '
Does this modification reduce the margin of. safety as defined in the basis for the Technical-L Specifications?
Answer:
No. 'This modification' improves the margin of safety by providing a more reliable method of redundancy for the HPIC and RCIC systems as they are routed through Appendix R fire areas.
a 51 1,
1 PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 '
1 198910 CFR 50.59 REPORT-Reolace Hloh Pressure Coolant inlection fHPCO Turbine Trio Pushbuttons On Control Room P t
MoomcatioN No.: 2080 A. SysTeu: HPCI B. DEsenwrioN:
Replacement of HPCI Turbine Trip Pushbuttons on Control Room Panels to provide the ability to trip from Control Room Panel in the event of an Append!x R fire, lastallation of an isolation switch in the cable spreading room to isolate all automatic close signals to MO-3-23 24 (HPCI shut off to CST valve).
C, REASON Fon CHANoE:
This modification is necessary to comply with 10 CFR 50 Appendix R.
D. SAFETY EVALUATION
SUMMARY
1).
Does this modification increase the probability of occurrence or the consequences of an accident
't or malfunction of equipment important to safety as previously evaluated in the safety analysis rep!
Answer:
No. This modification was designed and constructed in accordance with criterla applicable to the -
safety related HPCI system.
= 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
m, Answer:
No. The present automatic and manual HPCI turbine trlp capabilities are maintained.'
,.w 3)
Does this modification reduce thimargin of safety as defined in the basis for the Technical' L$
Spccifications?:
i L-Answer:
No.
a"r The HPCI system's initiation and operability requirements as described.in the Technical i
Specifications. The isolation switch does not effect the HPCI nystem when placed in " Normal' W
position. Although the unit may enter a limiting condition for oporation when the switch is placed.
j I-
.. in ' emergency position *,.this action'is acceptable to achieve safe shutdown in the event of an-emergency such as an Appendix R fire.
l s 'y i;-.
I1 b
r i
52
1 PEACH BOTTOM ATOMIC POWER STATION-UNIT 3
}
DOCKET No. 50 278
{
198910 CFR 50.59 REPORT I
I Revise Trio Settinos of Circuit Breakers and Reolace Fuses
.I MoDWICATION No.: 2083
& SyJytfuI 480 Volt B. DesenipuoN:
y This modification changes the setting of 12 adjustable 480 volt circuit breakers and replaces 2120 volt fuses to achieve better coordination and consistency in the power distribution system.
C. fhAsow Fon CHAnos:
This modification.was performed to achieve better coordination of safe shutdown electrical isolation equipment required by 10 CFR 50 Appendix R criteria.
D. Sarrry Evatuanon SuuuAny:
l 1)
Does this modification increase the probability of occurrence or the consequences of an accident i
or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
t Answer:
No.' This modification only improves the coordination between electrical protective devices.-
i 2)
Does this modification create the possibility for an accident or malfunction of a different type than a!
4 any evaluated previously in the safety analysis report?
i
-Answer:
1
'No. Fault current levels are within the ratings of the protective equipment. The approved level of
. fire protection is being maintained.
~
- 3) '
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. The reliability of the electrical power system is improved by this modification.
1 1
o 53 i~
sg
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 198910 CFR 50.59 REPORT Accendix R Additional Process instruments and Unarade Mooincarion No.: 2085 A. Systeu: - Process and Diagnostic Instrumentation si B. Desemioview:
This modification reroutes the existing instrument and power cables, changes power feeds and adds alternate power feed to existing instrument loops. It also replaces a condensate storage tank water leve e
transmitter, a suppression pool temperature element, and installs a complete reactor water level Instrume loop.
C. RcAson Fon Cuawas:
This modification is necessary to comply with 10 CFR 50 Appendix R " Safe Shutdown Instrumentatio Requirements".
D. LarsTv Evawarion Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident -
- or malfunction of equipment important to safety as previously evaluated in the safety analys Answer:
No.. Where this modification interfaces with existing safety-related systems and equipment, original design criteria applicable to those systems and equipment have been applied.
I 2).
Does this modification create the possibility for an accident or malfunction of a different ty$
/
any evaluated previously in the safety analysis report?
' Answer:
u.
T l
- No. This modification is for Indication only, it will not affect the operation or separation system.
(
'Does this-modification reduce the margin of safety.as defhied in the bads for the 0
3)
Specifications?
a Answer:
s No. Section 3 and section 4 of the Technical Specifications have been reviewed and it has bee
- determined that this modification will not affect or change the applicable bases or tables..
l 1
A l5 0
'f.
,-l.
54 4,
q l
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 i
198910 CFR 50.59 REPORT l
Emeroenev Service Water Ploe Rectacement MoomcatioN No.: 2106 A. Systru: Emergency Service Water (ESW)
B. DesenwnoN:
i' This modification replaces the ESW Torus Room ring headers and associated branch piping and provides f ring header segmentation valves for maintenance purposes. It also provides balancing valves with differentia!
pressure taps and Indicators, to allow direct flow measurement. Small bypass lines with plug valves allow continuous flow of corrosion inhibitor and blocide chemicals through all ESW piping and components. Also -
l the Residual Heat Removal (RHR) Pump Room Unit Cooler control switches are being relocated.
C. RFASoN Fon CHANoe:
This modification is necessary because of past problems with piping corrosion and flow balancing.
Relocating the RHR Pump Room Unit Cooler switches will decrease personnel exposure.
D. SareTY Evatuanow
SUMMARY
4 t;
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 evaluated in the safety analysis report?
{
L Answer:
No. This modification makes no functional changes to the ESW system. The modification will i
maintain the system's capability of functioning as presently designed. The ' environmental and seismic quallfications of the RHR switches will not be affected by their relocation. The replacement -
piping and valves were designed,' purchased, installed, and tested to the same or more recent requirements as the existlag piping and valves.
j
- 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No. This modification does not introduce any new hazards, since there are no functional changesi to the system.
3)
Does this modification reduce the margin of safety =as defined in the basis for the Technical:
Specifications? :
Answer:
No. The Technical Specifications require that the ESW system be operable during plant operations.
This modification will not decrease the efficiency or affect the operability of the ESW system.-
h a
I; i!
55 u
L i
s
F e
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 54278 198910 CFR 50.59 REPORT Reolace 3 3B Feedwater Heater and Associated shell side Relief Valve Mooincation No.: 2224 A. SYSMu: Feedwater Heater-4 B Desemption:
Replacement of the 3-3B Feedwater Heater and associated shell side reflef valve while the unit w shutdown.
C. Asason Fon CHawoe:
Tl is modification was necessary because 15% of the heater tubes were plugged and an addition extensive damage.
D. Sarrry Evatuatiow Suuuanz I
'1)
Does this modification increase the probability of occurrence or the consequences of an acc or malfunction of equipment important to safety as previously evaluated in the safety an Answer:
No,' The design of the new equipment is In accordance with existing plant criteria.
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 implementation of this modification was completed under controlled condition equipment is consistent with existing plant standards.
' 3)
D'oes this modification reduce the margin of safety as defined in the basis for the T
~
Specifications?
Answer:
' No. Technical Specifications 3.8.C.Sc,3.14.A.4,3.14.A.6,3.9,4.8.C.5b, and 4.8.C.5b,4.8 C reviewed..
1 i
l
-)
56
+
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 198910 CFR 50.59 REPORT Relocate Uninterruotlble Power System's (UPS) cnd Uoarade Inverter Bvnass Switch Transformer To Vital Areas MooincAtioN No.: 2275 A. System: Instrument Altemating Current (AC) System
~ B. DeseniptioN:
Relocate UPS transformer to emergency 4KV switchgear room. Upgrade the transformer KVA rating fro 25 KVA to 50 KVA. Relocate inverter maintenance bypass switch to the cable spreading room.
C. REASON Fon CHANOE:
This modification is necessary to conform to 10 CFR 73.55.
D. SurTY EVALUATION SUMMAnY
' )'
' Does this modification increase the probability of occurrence or the consequences of an acciden 1
or malfunction of equipment important to safety as previously evaluated in the safety analysis repor' /
Answer:
No. This modification meets the design criteria of the uninterruptible 120 VAC power system and Class IE 480 voit power system; 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
r Answer:
l p-No. The secondary power source remains unchanged. The capacity of the single phase 480-120V '
transformer is increased. The new locations of the transformer and the maintenance bypass switch
-r are vital areas making the uninterruptible 120 VAC power system more secure.' No additional -
electrical loads are added by this modification.
L$
y' 3)
Does this modification reduce the margin of safety as defined in the basis for the Technical' Specifications?
Answer:
No. Technical Specifications 3.9 and 4.'9 were reviewed. This modification does not change the design of electrical systems described in the Technical Specificationc i
' tf Y,
i.!+
.[t 57 p
7
,N sk
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 198910 CFR 50.59 REPORT RHR M'in Flow Protection MooiricATion No.: 2285 A. Sysisu: Residual Heat Removal B. DESCRWTioN:
This modification provides minimum flow protection to RHR Pump 3AP35 in the event of an Appendix R fire irr Fire Area 13 North by lastalling an alternate power supply to RHR Minimum Flow Bypass Valve MO 31016A.~
C. RsAsow FoM CHANOEI An Appendix R fire !n Fire Area 13N requires RHR Pump 3AP35 for suppression pool cooling, approxim 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> into the fire scenario, to maintain suppression pool temperature within RHR Pump NPSH limits.
Normally, when the pump is running without a sufficient discharge path, the min flow bypass valve opens on a high differential pressure signal to prevent pump damage. The load center which feeds the bypass valve is located in Firo Area 13N and could become de-energized during an Appendix R fire in this fire area.
Minimum flow protection for Pump 3AP35 is lost when the pump starts automatically on valid signats with -
no power available to the min flow bypass valves. Since there may not be enough time (cr an operator to recognize the si'uation and trip the pump, the alternate power supply was Installed.
D. SAFETY EVALUATtoN
SUMMARY
1)
Does this modification increase the probability of occurrence or the cor' sequences of an accident
- or malfunction of equipment important to safety as previously evaluated in the screty analysis report?
Answer:
No. The min flow bypass valve operates as previously designed to provide minimum flow protection t
for RHR Pump 3AP35 during normal operation or design basis events. This moditteflon adds a -
backup power source to this valve for an Appendix R fire..
l
- 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?
D Answer:
No. There is no adverse impact on plant bus loading or voltage regulation. The auto-transfer switch,
- auxillary motor controller which controls the bypass valve, and associated cable rerouting are 1
designed with criterla applicable to safety related systems. The cable routing associated with the-backup source is designed and constructed in accordance with Appendix R safe shutdown criteria.
t 3)
.Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
u>
No.' This modification does not reduce the margin of safety as defined in the basis for any Technica Specification.
The RHR system initiation and operability requirements are unchanged by this modification. The operability of the bypass valve is maintained to provide minimum flow protection
. for RHR Pump 3AP35 in the event of a fire in Fire Area 13N.
o PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50-278 i
1:
198910 CFR 50.59 REPORT l
Relocate 2 Relavs ll MooiricAriow No.: 2354 l
l-A.Sysisu: Reactor Core Isolation Cooling (RCIC)
\\'
B. Desemipriow:
This modification relocates relays 10A K167 and 10A K168 C. Reasow Fon CHANoE:
To improve the capability to safely shutdown the plant in the event of an Appendix R Fire D. SArrry EVALUATION SUMMAnY:
1)
Does this modification increase the probability of occurrence or the consequences of an accident
. or malfunction of equipment importent to safety as previously evaluated in the safety analysis report? -
l-Answer:
No. The operability of the RCIC system will not be comprised by this modification.
2)
Does this modification create the possibility for an accident or malfunction of a different type than 3
any evaluated previously in tha safety analysis report?
Answer:
No. The Relocation of the relays will not prevent the RCIC system from performing its designed-l safety function.
- 3) :
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?.:
Answer:
-1 No. This modification improves the abilty to shutdown the plant in the event of an Appendix R Fire.
0 L
1 4
.l 59 O
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKEl' No. 50 278 198910 CFR 50.69 REPORT 2g:noe/Sotit Power Feods to Foodwater H-ators 3 4 and 5 Moowiention No.: 2389 A.,$ystru: Feedwater B. Deseminigst This modification changes the power supply to the third, fourth, and fifth Feedwater Heater extraction steam block valves from one source to three sources.
C. Maasow Fon Cnanos:
This modification is tsquired to ensure compilance with feodwater transient analyses contained In the final UFSAR.
D. Santy EvAtuatioN Suuuamv:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Anewer:
No. The diversity of the power supplies will Icwer the probability of future feedwater translents caused by single power source failures.
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. Thore is no change in the overall loading of the plant's electrical system as a result of this modification.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
l Answer:
f i
No. Cables added by this modification are fire resistant and are routed such that they will have no adverse effect on the safe shutdown capability of the plant. The Technical Specifications have been reviewed and the circuitry involved in this modification is not specifically discussed.
l e,
o PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT Railroad Airlock Door Renalr Metaganow No.: 2489A A. Sysnu: Secondary Containment B. Desemnow:
This modification involved strengthening the top cnd bottom strike rod guide support welds on tre equipment access lock door, commonly known as the railroad airlock doors. The door strikes were also adjusted.
C. Epsow Fo.a Qswqn This modification was made so that the door would meet the minlmum specified margins of safety for extreme environmental conditions.
D. SAnty Evatut.noN SuuuAmy:
1)
Does this modification increase th9 probability of occurrence or the consequences of en accident or malfunction of equipment important to safety as previously evaluated la the safety analysis report?
Anmen No. This work has nc impact on the probability of analyzed accidents or malfunctions of interfacing safety related equipment. Also, the consequences of previously analyzed accidents or malfunctions are not aoV9tsely affected by this modification.
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?
&QtEtU No. The repair work does not create a possibility for an accident or malfunction of a different type than any evaluated previously. The repair work was needed to increase the margin of safety under severe environmental conditions. The doors have sufficient strength and stiffness to withstand extreme environmental conditions, which includes Maximum Credible Earthquake loads, without failure.
3)
Does this modificat!on reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.- The repair did not reduce the margin of safety. The modification was made so that the door
- would meet the minimum specified margins of safety for extreme environmental conditions.
61
i e-PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 60 278 198910 CFR 60.69 REPORT Jnstallation Of Cables Between Motor Control Centers (MCC)
Moomcatiow No.: 25178 A. S.r.ntal 480 Volt MCC B, Desewriow:
This modification installs cables between MCC's 30859 and 30B49,30B00 and 30850,30B36 and 30B30.
C, Reasow Fon CHaN0Q The cables will act as swing buses between the pairs of MCC's to provide power to selected criticalloads during MCC bus outages.
D. SamY EvAtuarow Summany:
1) 000 this modification increase the probability of occurrence or the consequences of an accident 1
or maNnction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This temporary connection will be used only when the unit is in cold shutdown or refuel mode.
M equipment powered from the swing bus will remain functional as well as all equipment remaining connected to the MCC supplying the swing bus.
i 2)
Does this modification create the possbility for an accident or malfunction of a different type than any evaluated previously in the safety anaiysis report?
AtitEtG No. The existing MCC will continue to provide electrical protection for each of the loadt ',elng supplied.
t 3)
Does this modification reduce the trargin of safety as defined in the basis for the Technical Specifications?
Answer:
No. While the cable installation Is permanent, the cables will only be connected on a temporary basis.
o.
s.
I i
62 I
j
PEACH BOTTOM ATOMIC P3WER STATION UNIT 3 DOCKET No. 50 278 196910 CFR 50.59 REPORT Xctgl Manway Fix
.Mooineation Nc.: 2532 A. Systsu: Reector Vessel and internals B. Drsemmtiow; This modification replaced 2 wolded access hole covers located in the shroud support plate at the bottom of the Peach Bottom Unit 3 reactor vessel annulus. These covers seal holes which we to the bottom of the vessel during construction. This modification also reduces the gap between the access hole cover and the shroud support plate.
C. Reason Fon CHAN0tl This modification was necessary because of Intergrannular stress corrosion cracking (IGSCC).
D. Sarrry Evatuation SuuuAny:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis rep Answer:
No. By replacing existing covers. this modification reduces the probability of failure.
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?
AGEMtti No. The raaterials have been selected for IGSCC resistance and are compatible with BWR.
environment. There is a small potential of loose parts In the future, but the existence of these pa would not jeopardize safe reactor operation.
1.
1 3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Spacifications?
Answer:
No. There are no Technical Specifications associated with access hole covers.
1 l
N
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 60 278 198910 CFR 50.69 REPORT T
Start Loolc for Residual Heat Bemoval (RHR) Comnartment Cooler Fans
{
Moowiention No.: 2578 A. Systru: RHR B. Desewtion:
The starting logic for the RHR compartment coolers was changed.
C.Rgasow Fon CHAwar:
The coolers will now start on either an RHR pump start signal, or when the RHR receives a Loss Of Coolant Accident signal. Before this modification, the RHR compartment coolers risked a no %11~!r." a Loss of Coolant Accident without a loss of offsite power, in particular, when only one offsite power source vculd be available, because of low bus voltage. Low bus voltage could be caused by attempting to start an RHR pump, the RHR compartment cooler fans and other 480 loads at the same time.
D Sarrry Evatuattow Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This modification starts the RHR compartment coolers sooner than previous design during a Loss of Coolant Accident to prevent possible starting problems due to degraded bus voltage.
2)
Doei, this modification create the possibility for an accident or malfunction of a d;fferent type than any evaluated previously in the safety ana!ysis report?
Answer:
No. Previous start conditions are maintained.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Atlanta No. This modification meets the design criteria of the RHR and secondary containment HVAC systems. Technical Speelfication sections 3.4A,4.4A,3.7C, and 4.70 were reviewed in making this decision.
=
b.
m.
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT Removal of Reactor Core Isolation Coolina (RCIC) Electrical Oversneed Device Mooiricatiow No.: 5001 A. SYstru: RCIC B. Desemmtion:
This modification removes the RCIC turbine electronic overspeed trip device including the associated cables and connections.
L C. Rrasow Fon CHawor:
This change was recommended by General Electric Service Information Letter 382.
D. SarrTy Evatuatiow Summany:
1)
Does this modification increase the probabDity of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
~
Answen No. Service Information Letter 382 provides justification for removal of these devices. The RCIC turbine will continue to be protected from overspeed translents by the mechanical overspeed trip device.
2)
Does thls modification create the possibi!)ty for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No. The affected safety systems will continue to perform their Intended functions.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. Technical Specification 3.5 sections D. G, and H were reviewed. This modification removes an overspeed trip device which was originally installed for operator convenlence, but which has not operated per its design and is no longer supported as a O device by its manufacturer.
l 1
65
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 60 278 198910 CFR 60.59 REPORT Move Turbine Generator Turnina Gear Oil Pumo (TGOP) Feed Moowientiou No.:
5156 A. Aylitu; Turblne Generator B.DesenwrioN:
This modification moves the feed supplying the turbine generator TGOP from the B Diesel Generator to the D Diesel Generator.
C.Ihasow Fon CHawot:
This modification is necessary to help balance the loading of the diesel generators and load centers and their transformers.
D. Sarrry Evatuatiow Suuuany:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This modification only relocates the power feed for the Turbine Generator TGOP from a safety related Motor Control Center to a non safety related Motor Control Center. The TGOP is not being physically changed or repositioned. No new challenges are presented to the safety system.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No Relocating the power feed to the Turbine Generator TGOP from a safety related Motor Control Center will not overload the *D' Diesel Generator. The additional load applied to the Diesel Generator Is within the loading requirements as determined by the Diesel Generator loading and voltage regulation study.
3)
Does this modification reduce the rnargin of safety as defined in the basis for the Technical Specifications?
Answer:
No Relocating the Turbine Generator TGCP feed to the Diesel Generator is within the Diesel Generator's load limits as determined by the Diesel Gere'r.'.or loading and voltage regulation study.
The Tech Specs. do not address the operation of the rurbine Generator TGOP or Motor Control Centers; therefore, no margin of safety addressed in the Tech Specs. Is affected.
4 PEACH B01 TOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT N2 Instrument Comoressor Low Lube Oil Alarm Moomicatiew No.: 79 016 A. Systrut Instrument Nitrogen B. Desenettow:
This modification installs a defeat switch on the A and B N2 Instrument compressors' control panels.
The switch will defeat the low lube oil alarm when the compressors' feed is blocked.
C. Rusow Fon CHAwar:
This will prevent masking out the common control room trouble alarm on the operating compressor when the other is blocked. This allows the alarm to work properly for the working compressor.
D. _Surry Evaluation SuwwAny:
1)
Does this modifica!!on increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No. This modification will allow the alarm to work properly.
2)
Does this modification create the possibility for an accident or malfunction of a different type than any ev0luated previously in the safety analysis report?
_A_ nswer:
No. This allows the control room operator to better monitor the operating compressor.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No. The Technical Speelfications have not been ci snged.
I-l l
l' 1
1 67
w PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 50 278 198910 CFR 50.59 REPORT jnstallation Of Calibration Head Chambers Mooineation No.: 83-037 1
A. Svetru: Miscellaneous Instrumentation B,Qtscamtiow:
This modtfication adds calibration head chambers on LT-61 (Rx wide range level transmitter) vla double ven valves.
C. RaAsow Fon CHANot!
This will prevent spurious serams and PCIS isolations that may occur while backfilling LT-61 after res D. Sarrry Evaluation SuuuAny:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment irnportant to safety as previously evaluated in the safety analysis report?
Answer:
i No. This modification will double isolate the reactor water in the transmitter from the envir 2)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated provlously in the safety analysis report?
AlllWAG No. This modification improves reliability because it prevents air from entering the reactor instrumentation lines.
3)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
~
Answer:
l
]
No. The Technical Specifications are not affected, n
)
PEACH BOTTOM ATOMIC POWER STATION UNIT 3 DOCKET No. 60 278 198910 CFR 50.59 REPORT 03A Reelte Pumo ard Valves Mooineatiou No.: 86-095 A. Systru: Rectre Pump and Valves B.Rusamiton' For each of 4 circuit setters, CS-8(8)029 on P&lD M 353, an upstream and downstream isolation valve (manual) and bypass piping with manual throttling valve was Installed. Valves and piping are equivalent to those used in the original seal purge Installation.
C. Reasow Fon CManor:
This will allow removal of circult seal setter for maintenance while maintaining normal seal purge flow path.
D. Sarrry Evaluation SuuuAny:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answen No. If seal purge flow were to stop, seal cooling flow would be supplied by reactor water through the thermal sleeve. No impact to recirc pump operation would occur.
2)
Oces thls 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 only possible malfunction would be blocking seal purge flow as discussed above.
l-3)
' Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
l AGREtu l
l-No. Seal purge is a recent addition and is not the basis for any Technical Specifications.
l t
69
-