ML20247D346
| ML20247D346 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/31/1988 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8907250173 | |
| Download: ML20247D346 (143) | |
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10CFR50.59(b)(25 PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955-65 CHESTERBROOK BLVD.
WAYNE, PA 19087-5691 (215) 640 6000 July 17, 1989 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555
SUBJECT:
Peach Bottom Atomic Power Station, Units 2 and 3 Annual 10 CFR 50.59 Report for 1988
Dear Sir:
Enclosed is the 1988 Annual 10 CFR 50.59. Report,
" Changes, Tests and Experiments", for Peach Bottom Atomic Power Station, Units 2 and 3.
If you have any questions or require additional information, please do not hesitate to contact us.
Very truly yours, of.
G. A. Hunger, Jr.
Director Licensing.Section Nuclear Support Division Enclosure cc: W. T. Russell, Administrator, Region I, USNRC T. P. Johnson, USNRC Senior Resident Inspector R. E. Martin, Project Manager, USNRC T. E. Magette, State of Maryland J. Urban, Delmarva Power R. A. Burricelli, Public Service Electric & Gas H. C. Schwemm, Atlantic Electric T. M. Gerusky, Director, PA Bureau of Radiological Protection 8907250173 881231 :+
{DR ADOCK 05000277 PDC
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'O raaom'en1^ st=cra c coara=1 PEACH BOTTOM ATOMIC POWER STATION .j i
UNITS 2 AND 3-DOCKET NOS. 50-277; 50-278 J
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1988' J ANNUAL 10 CFR~50.59 REPORT 1
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Docket Nos. 50-277 50-278 l
O 1988 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT 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 DPR-44 and DPR-56, respectively). This report addresses, but is not limited to, changes to the facility and procedures as they are described in the Updated. Final Safety Analysis Report. This. report consists primarily of plant modifications that were installed in 1988. Also included are special operations or tests,-not described in the safety analysis report, conducted in 1988.- A summary of the safety evaluation for each item, concluding that an unreviewed safety question as defined in 10 CFR 50.59 (a)(2) was not involved, is included.
Some plant modifications described in this report'were completed on one'of the units in 1988, and the same modification was completed on the other unit in 1989. Modifications in this category were reported for both units in this' report to avoid duplication in next year's report. Some items described in this- .
report were completed or conducted pr*.t to 1988.
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TABLE OF CONTENTS- -
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l Modifications to Unit-2 l
'Pages 1 through 70 :i Number. System.
'219- Varmious ' 'I
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542A. High Pressure. Coolant Injection 'l
.633- Automatic Depressurization 664 1 Primary' Containment .
1029E Various
'1200' Radiation-Monitoring 1240 .High; Pressure Coolant Injection 3
-1324 -High Pressure Service. Water i 1351A Diesel Generators 1351C- Fire Systems .
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- g 1352A High Pressure Coolant Injection 11 1352B- Miscellaneous Instruments- 'j l'359 Uninterruptable AC 1 "1636- Fire Systems )
1669 Structural 1684 Feedwater.
1724 Primary Containment 1790 Feedwater 1905 ' Structural' :
'1916 . Reactor Protection 1950 , Automatic Depressurization I}
(N ^)'1950A Automatic Depressurization ' ;l 1965 Condensate 1982 Reactor Recirculation )
2006 Primary Containment 2078 Fire' Systems 2079 Various -1 2080 High, Pressure Coolant Injection 1 2081 Securityf(lighting) 2083 'Various 2084 Residual Heat Removal 2189A Residual Heat Removal .i 2235 ' Structural !
2249 Miscellaneous 2275 Security "!
2281 Feedwater 2318 Fire Systems )
2329 High Pressure 4 Service.: Water 2353 Primary Containment ;
' 2355 Various 2371- Emergency Service' Water j 2383 Reactor Protection )
2389 Feedwater '
'2391 Feedwater 2489A Secondary Containment. .
.. 2578- Residual Heat Removal
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- Number- System:
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?2580 Cooling-Towers-
- ;-4102- LReactor Water Clean-up.
34112A- Residual Heat: Removal 5001 . Reactor Core Isolation Cooling-5030 ' oxygen' Analyzers '
5033: Standby Liquid Control;80-155 Feedwater-83-18 . Main' Steam 85-99 Standby Gas; Treatment 86-04 Electro-Hydraulic Control,
- 86-42 Structural 86-60 L480 Volt Emergency Load < Centers i 86-76' Instrument ~ Nitrogen 87-33 Control Rod-Drives 87-45 Core Spray-87-47 480 Volt MCCs-87-63 . Process Computer 87-92' ' Structural 87-100 High Pressure Coolant. Injection 88-35 Control Rod Drives-I' 188-38 Reactor: Protection.
88-51 -Off-Gas'& Recombiner-88-57 Feedwater Modifications'to-Unit 3' Pages -71 through 80 Number' System 1958 Miscellaneous 1967A Various 2042 13;KV Power 86-118 Reactor: Recirculation M-G Sets 86-156* Process Computer' 87-32* Process, Computer.
87-78 Structural i 87-105 . Fuel Handling 88-13 Radwaste-88-14 Radwaste 88-45 Main Steam Turbine Modifications to Both Units Pages 81 through 105-Number- System 643 Annunciators 687 ~ Reactor' Water Clean-up-810B Various '
'1078 Residual Heat Refaovall 1228 Fire: Protection
' ()l1352I** Security ^(lighting / communications) 1353I** Security (lighting / communications) 1571 Fire-Protection 1879 Reactor Water Sampling Notes:(*)(**):. Report entry for these modifications is combined.
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1 Number System i
1915 Various j
(~T 2091 Fire Protection l
\/- 2096* Ventilation 2096A* Ventilation 2215 Process Computer j 2339 Fire Systems 1 2376 Structural 1 5045 Reactor Protection -l 85-132 Residual Heat Removal and High Pressure )
Coolant Injection 1 86-51 Turbine Building Ventilation i 86-121 Process Computer j 86-127 Condensate i 86-132** Electrical Heat Tracing 86-136** Electrical Heat Tracing 87-04** Electrical Heat Tracing i 87-13 Feedwater Pump Turbine !
87-27 Core Spray l 87-73 Ventilation .
i .88-21 High Pressure Coolant Injection Modifications on Systems t Common to Both Units Pages 106 through 119 ,
I Number System' 1
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1205 Diesel Generators 5007 Fire Systems85-106 Diesel Generators .
86-79 Main Stack Radiation Monitoring j 87-12 Radwaste i 87-26 Radwaste l 87-83 Control Room Ventilation 87-102 Diesel Generators87-113 Emergency Service Water 88-10 Diesel Generators 88-18 Diesel Generators 88-31 Computerized History and Maintenance Planning 88-41 Off-Gas 88-61 Security
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'J Report entry for these modifications is combined.
l Notes (*)(**):
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Procedure Changes - Pages 120 through 125 l_ ' Number Title v- ST-9.17-3, Revs. 5,6,7 Reactor Coolant' Leakage Test-- Unit 3 ST-9.17-3, Rev. 8 Reactor Coolant. Leakage Test - Unit 3 ST-9.17-2, Rev. 5 . ~
Reactor Coola.nt Leakage Test - Unit 2 ...
ST-13.8-1, Revs. 6,7 Unit 2-Excessz Flow Check Valve Operability HPO/Co-18, Rev.'19 ._
Processing Liquid Radioactive Waste l- GP-2, Rev. 48 Normal ~ Plant Startup'
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Miscellaneous Changes and Procedures Pages 126 through 136 Reference Description SP-1109 'De-energize E-33 Bus' SP-ll46 Special Reactor Cavity. Water Cleanup System.
SP-ll76 Running a D/G'Without< Normal-Ventilation SP-1203 Mud Removal.from Pump. Bays SP-5000 Refueling' Bridge Interlock Removal SP-5001 Crosstying S/U: Sources t< :PS-2(3)380 A&B =Setpoint, Change
.FS-0470 A&B Setpoint Change Bkrs 52-3785 & 52-3803. Setpoint Change
-FS-0760 A&B _'Setpoint Change FS-20008 Setpoint Change .j 1
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i P:cch Bottom Ato2ic Power Stntion {
Unit 2 Docket No. 50-277 i Annual 10 CFR 50.59 Report !
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1 Resetting of MCC Feed Breakers in Load Centers to )
Prevent Selective Tripping l Modification No.: 219 ,
j A. System: Various q B. Description and Reason for Change: ;
1 This modification changed the magnetic trip settings on selected 480V molded case circuit breakers. Changing the j trip settings will allow the circuit breakers to operate i l
correctly with respect to the in-rush currents of the loada being supplied. Where the new magnetic trip setting could j not be obtained on the existing HFA circuit breaker, either a spare HFA circuit breaker or a new HFB circuit breaker was l j
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-installed. {
i C. Safety Evaluation Summary: j This modification will not have an adverse effect on the operation of the plant. The new circuit breaker settings 1
( will allow their correct operation based on nameplate data of i
(_3 / the supplied loads. The replacement circuit breakers are i qualified for their intended use where required. No equipment important to safety as described in the UFSAR is !
adversely affected. .
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I Prech Bottom Atomic Pow 3r Station Unit 2 !
Docket No. 50-277
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Annual 10 CPR 50.59 Report HPCI Suction Pressure Switch Replacement 1 i
Modification No.: 0542A ;
i A. System: HPCI Instrumentation B. Description and Reason for Change: )
l The purpose of this modification was to replace a pressure i switch, which due to a spare parts problem, was difficult to '
maintain. The modification also replaced a pressure switch I and pressure transmitter with instruments of higher proof i pressure and added a pulsation dampener in a pressure sensing i line to allow the instruments to operate under transient i conditions without instrument failure.
C. Safety Evaluation Summary,:
This modification did involve safety-related equipment; however, no unreviewed safety questions were raised because the new instruments perform the same function as the instruments previously installed. This modification did not affect operation of the HPCI system.
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Pacch Bottom Atomic Powsr Station Unit 2 Docket No. 50-277 Annual 10 CPR 50.59 Report C:)
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High Drywell Precsure Bypass and Manual Inhibit f
Modification No.: 633 I
A. System: Automatic Depressurization System (ADS)
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An automatic bypass of the high drywell pressure ADS j initiation signal, and manual inhibit of ADS were provided. 1 A timer is actuated on Lo-Lo-Lo reactor water level. When the timer ' times out', the high drywell pressure signal will i be bypassed and ADS will be initiated on coincidental Lo-Lo- i Lo reactor water level and low pressure ECCS' pumps' running !
signals. A keylocked switch is used for the manual inhibit
, of each ADS trip logic. Alarms were provided in the Main f Control 1 Room for bypass timer initiation and manual inhibit ]
activation. j j
This modification was implemented to satisfy Item II.K.3.18 of NUREG-0737, " Clarification 'of '?MI Action Plan r~ Requirements", to reduce the dependence on operator action.
(-) Previously, automatic initiation of ADS required coincidental ,
low reactor water level, low pressure ECCS pumps running, and !
high drywell pressure signals. For transient and accident !
events which would not directly. produce high drywell pressure l and are degraded by a loss of all high pressure injection systems, adequate core cooling depended upon manual depressurization of the reactor vessel, followed by injection from the low pressure systems.
C. Safety Evaluation Summary:
An analysis demonstrated that a setting of ten minutes on the bypass timer will assure adequate core cooling for transient or accident events which previously required manual depressurization as previously evaluated, and a setting of nine minutes was chosen for conservatism.. The calculation of ,
the timer setting was based on: 1) the avoidance of excessive fuel cladding heatup using 10 CFR'50, Appendix K models; and
- 2) providing sufficient time to recover reactor water level above level 1 during an ATWS event. In the event of an ATWS, this setting combined with the existing two-minute timer in the ADS-logic allows sufficient time for the operators to diagnose plant conditions and follow emergency procedures to inhibit ADS.
The keylocked switch used for the manual inhibit, and the annunciator minimize the probability of inadvertent inhibit of ADS. Also, all design requirements. applicable to the original equipment and circuitry were applied to this change.
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Peach Bottom Atomic Power Station
' Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
High Drywel'. Pressure Bypass and Manual Inhibit (Continued)
The operational basis'of the ADS was not changed and the basis for the Technical Specifications was not affected.
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Peach Bottom At'cmic. Power' Station. 1 Unit 2
.DocketENo. 50-277 )
Annual 10.CFR.50.59 Report Hi Rad Trip of Containment Vent / Purge Valves Modification No.:. 664 A. ' System:: . Primary Containment B.- . Description'and Reasonifor Change:
This' modification changes the existing PCISslogic for ;
i containmentcvent and purge valves.by adding'a trip signal from-the off-gas-stack' radiation monitors. This. change was implemented to meet:.the requirements.of Item'll.E.4.2(7) of.
NOREG 0737, " Clarification of TMI Action = Plan Requirements." :'
C. Safety Evaluation Summary: 1
' tThis? modification,. adds an additional. containment isolation signa) not previously part of.the plant design. This new-signal has no effect-on' existing. trip settings. . Failure of-this signal leaves"the plant in.the:same. condition as evaluated'in the UFSAR. Therefore, no~unreviewed safety 1 questionLis involved.or-created. l O
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Peach Bottom Atomic Power Station j Unit 2 Docket No. 50-277 Annual 10 CPR 50.59-Report (O
w Installation of Process and Diagnostic Instrumentation Modification No.: 1029E A. System: Reactor Core 1 solation Cooling (RCIC), Residual Heat Removal (RHR), High Pressure Service Water (HPSW), l Emergency Service Water (ESW), Core Spray (CS), d Condensate Storage Tank (CST) l B. Description and Reason for Change:
Diagnostic instrumentation has been provided for the systems associated with the three methous of safe shutdown, at the control room. In addition, process instrumentation for reactor water level, reactor pressure, drywell pressure, !
, drywell temperature, Suppression Pool water level, L
' Suppression Pool temperature, and Ccndensate Storage. Tank {
1 (CST) water level for each fire area has been made available in the control room. In the event of a fire in the Turbine Building, control room indication for the CST water level and Emergency Service Water (ESW) discharge pressure would not be available. Therefore, two complete instrument loops were installed for CST water level and ESW discharge pressure ,
() monitoring. Numerous instruments were relocated and cables were rerouted, some additional indicators were installed, and power feeds were changed. These changes were implemented in accordance with the requirements of 10 CFR 50, Appendix R, to 3 ensure safe plant shutdewn capability in the event of a l design basis fire.
C. Safety Evaluation Summary:
All of the components and equipment affected by this modification are of the appropriate safety grade and meet the proper qualifications for the functions that they serve.
Cables and wires associated with safety-related systems were I
rerouted using the appropriate electrical separation requirements. Only indicating instrumentation was changed by this modification; no controls were changed. Thus, no safety functions were degraded. Overall plant safety was improved !
by enhancing the operator's ability to safely shut down the
, plant in the event of a fire.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10-CFR 50 59 Report i Control Room Radiation Monitoring System Flow Transmitter Replacement j Modification No.: 1200 A. System: Radiation Monitoring B. Description and Reason for Change: )
l This modification replaces Control Room Ventilation Radiation '
Monitoring System, flow transmitters and removes square root ccaverters which control flow control valves to keep the air j flow inside the particulate sampling lines constant. The new {
flow transmitter vill transmit a linear signal to the flow controllers /inficaters, therefore, there will be no need for squar.? root converters. The previous transmitters have {
proven unreliable and have had a history of failures.
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C. _ Safety Evaluation Summary:
Where this modification interfaces with the Control Room Ventilation Radiation Monitoring System, original design criteria have been applied. No functional or logic changes j
were made to the system. The transmitters do not perform an active safety function. This modification does not affect the performance or operation of any other safety or non-safety related systems or equipment.
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1 Peach, Bottom Atomic Pownr. Station; Unit 2_
Docket'No. 50-277 ,
Annual ~10'CFR 50.59 Report :
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Replacement.of HPCI Shaft Driven Oil Pump
. Modification'No.: -1240
. A. - System: High Pressure Coolant Injection (HPCI).
- B. : Description and Reason for Change:- 3
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A new.model_ main oi16 pump.was' installed to replace 5the.
obsolt ia HPCI turbine oil" pump. . Theinew: pump design ..
necessitated shortening of the pump shaft-by 5/16"fand.
changitag approximately 14"'of the inlet p39ing to 211/2" from
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! . C. Safety Evaluation Snamary
, No unreviewed" safety question"is. involved because the:two pumps are nearly identical and'therefore, interchangeable. 3 The new pump / piping arrangement. will improve the existing. l conditions and willinot have ator overall'effect on the HPCI 1 turbine.- The new pump does not affect the operationfofsthe HPCI system, nor does-it2 create any new failure. modes of the system.
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Pacch Bottom' Atomic'Powar. Station j Unit 2 j Docket No. 50-277 l Annual.10.CFR'50.59 Report' Of High-Pressure Service Water Bypass'Line j
.i Modification No.: .1324 A. System: .High Pressure Service Water (HPSW) ;j B.. - Description 1and Reason for Change: .1 A three-inch;manualDvalve and'approximately 20 feet of piping.
with' orifices welded. into pipe spocas were installed 'around L tht. MO-2-10-89D'HPSW vaJve.- The MO-2-10-BSD valve'.is on the HPSW outlet of the 'D' Residual Heau Removal ~ heat exchanger. 1 l
1 Successful completion of the primary containment Integrated Leak Rate. Test-(ILRT) depends greatly on stable drywell air
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-^: temperature. ' oThisz change was made to provide a means for 1 I
. maintaining more stable-drywell air temperature-during the ILRT by more closely matching. shutdown cooling. capacity to decay heat' generation rate.
C. Safety Evaluation' Summary:
The:ato.; i ion of this valve.does not adversely impact the- -
safety-related function of the HPSW system orLthe, Residual
-Heat Removal system because'it can notz significantly degrade flow through the heat exchanger,.and the MO-2-10-89D valve does not' perform a safety-related isolation function.
This valve will be normally closed, as is the MO-2-10-89D'
~v alve. A-high-radiation signal downstream of.these valves will automatically trip 1the HPSW kamp'and therefis.an isolation valve downstream to prevent: the release of -
radioactivity. j 1
The new valve can'not degrade the' safety-related; function of -l the EPSW system. The design of:the system, as discussed-in. l i
the bases for the Technical Specifications,-is predicated on-the need for,only ene pump in accident-conditions.. j i
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report Alternative Control Stations for Diesel Generators Modification No.: 1351A A. System: Diesel Generators B. Description and Reason for Change:
Circuits were rerouted to a transfer / isolation signal switch which will transfer the control location of two diesel generators from their normal control panel to an alternative control station and isolate the system circuits that could adversely affect safe shutdown due to cable or panel damage resulting from a fire in either the Main Control Room, Cable Spreading Room, or Emergency Shutdown Panel Area. This modification was implemented to satisfy the requirements of
. Appendix: R to dLO CFR 50 to ensure safe shutdown capability in the event of a design basis fire.
C. Safety Evaluation Summary:
The safety function of the diesel generators and their 4kV reakers is maintained for normal plant operations, loss of O coolant accidents, and seismic avents. Rerouting of circuits and the addition of safety related transfer / isolation switches and the automatic initiation / trip features of the system are not affected whea the transfer switches are in the
" Normal" or " Test" positions. " Test" and " Emergency" switch positions are annunciated in the Main Control'to alert the operator of these abnormal conditions so that the system can be returned to " Normal" if there is no fire.
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l Patch Bottom Atomic Pownr Station ,.
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Unit 2 Docket No. 50-277 1 Annual.10:CFR 50.59 Report
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i Raceway Encapsulation and Circuit Reroutes Modification No.: 1351C A. System: Fire Systems B. Description and Reason for Change:
Raceways were encapsulated and safety-related circuits were rerouted to support alternative shutdown capability required )
for a design basis fire in the Main Control Room, Cable ]
Spreading Room, or' Emergency Shutdown Panel Area. Cable ampacities (theoretical current carrying capacity of cables) ,
were derated because the encapsulation reduces the raceways' j ability to disperse htat generated from resistive power 1 losses.
This change was implemented to satisfy the requirements'of Appendix R to 10 CFR 50, to ensure safe shutdown capability in the event of a design basis fire.
C. Safety Evaluation Summary:
() The integrity of the raceways was enhanced without degrading the reliability of the safety-related cables. The
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probability of malfunction of equipment was decreased by-satisfying the requirements of Appendix R_to 10 CFR'50.
The proper electrical separation criteria was followed and the control of' safety-related equipment was not affected. j Also, raceway supports were strengthened as nece.dsary to l 3s ,
,.supportsthe. additional weight ofJrerouted cables and i encapsulation material.
The reliability of associated Technical Specification equipment was increased by the protection of the cables.
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. i Peach Bottom Atomic Pow 3r Station Unit 2 1 Docket No. 50-277 Annual 10.CFr. 50.59 Report i
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High Pressure Coolant Injection Alternative Control l
Modification No.: 1352A A. System: High Pressure Coolant Injection (HPCI)
B. Description and Reason for Change:
1 An Alternative Control Station (ACS) was established for !
operating the HPCI system to supply coolant to the reactor l and remove heat from the reactor if a fire occurs in the Control Room, Cable Spreading Room or Remote Shutdown Panel area. Circuits were rerouted to a transfer / isolation switch ;
at the ACS which transfers control from the normal Control 1 Room Panel to the ACS and isolates the system from the j j
effects of the fire. The purpose of this modification is to comply with-requirements of Appendix R to 10 CFR 50. ]
C. Safety Evaluation Summary:
The safety objective of HPCI is to provide coolant to the i reactor if a loss of coolant accident that doesn't result in )
rw rapid depressurization of the reactor vessel occurs. The 1
(,) safety objective of the Primary Containment Isolation System j is to minimize the release of radioactive material from the l primary system. 10 CFR 50, Appendix R annuls the l requirements to postulate a LOCA or seismic event coincident 1 with loss of offsite power in the event of a fire. ;
Additionally, postulating an event that would result in gross i release of radioactive material from the primary system is not required during a fire requiring alternative shutdown control. Therefore, defeating the automatic initiation of HPCI and the PCIS function of HPCI valves during a fire does not pose an unreviewed safety question. There would be sufficient time to manually operate HPCI from the ACS to keep '
the reactor covered with coolant. The operation of the affected systems as described in the UFSAR is not changed when the transfer / isolation switches are in the " Normal" and
" Test" positions. Transfer / isolation switch in the
" Emergency" position causes an alarm in the Control Room and opening of the normally locked roll-up door that secures the ACS also causes an alarm in the Control Room.
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Prach Bottom Atomic Power Station I Unit 2 Docket No. 50-277
- Annual'10tCFR 50.59 Report r~x
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i Diagnostic Monitoring Instrumentation Modification No.: -1352B ,
I A.. System: Miscellaneous Instrument Systems !
I B. Description and Reason for Change: j i
Modification 1352B provides diagnostic monitoring. !
instrumentation to meet 10 CFR 50 Appendix R . .
1 L (Alternative / Safe Shutdown) requirements at Peach Bottom Unit 4
- 2. The scope of this modification entails the installa? ion- j of complete instrument loops for RHR, HPSW and HPCI flow, ESW i l and HPCI discharge pressure, and RBR Heat Exchanger l differential' pressure indication. Also installed is a i l potentiometer to control HPCI turbine speed, annunciation !
l ealarms--(foraHPCIs turbine exhaust line high discharge l pressure), and pump suction low pressure. ;,
C. Safety Evaluation Summary: I l The addition of the instrument loops will not affect the I operation of any systems. In addition,.the wiring changes to f the HPCI system including the addition of a potentiometer to
. control HPCI turbine speed wil not affect the separation of l the HPCI system because applicable cable and raceway I
separation specifications are being followed...Since the operation and separation of the plant systems will not be l affected by this modification, the plant safety-related l systems will perform their safety function.
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. Parch Bottom Atomic Powar Statior.
' Unit ;? 1 Docket No.-50-277 '
. Annual 10 CFR 50.59 Report
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Replacement?of the Static Inverter l
Modification No.:'1359 i q
A. . System: Uninterruptable AC )
B. Description and Reason for Change:.
This modification replaced the static' inverter and selected 120'VAC power cabless relocated the Reactor-Protection System )
(RPS). alternate feed to the distribution panel supplied by the new inverter, and added'an adjustLble voltage transformer between the inverter and the RPS protection panel. .The purpose was to increase the reliability of the RPS alternate !
supply and the uninterruptable power supply-systems.
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p -C.. . Safety, Evaluation Summary: 1 J
The isolation of the non-safety related inverter _from its i' safety related power supplies.is maintained. The modification increases the reliability of the inverter to the.
connected uninterruptable loads. The voltage regulation of the RPS alternate feed is improved. There ere no changes to
() the RPS. power supply trip setpoints. {
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" Peach Bottom Atomic Powar Station -
Unit 2 )
Docket No. 50-277 )
Annual.10.CFR.50.59 Report
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Recirc. M-G Set Room Sprinkler Expansion Modification No.: 1636' i
A. System:- Fire Systems ]
B. Description and Reason for Change: !
The Reactor Recirc. Pumps M-G set sprinkler system was expanded to provide tctal room coverage.- Previously,'only the M-G sets were covered. The valve station was acved, additional la ger piping-and larger: valves, and additional-sprinkler heads were installed.' This' change was'made to protect structural steel' members which support 10 CFR 50 Appendix R safe shutdown' fire barriers.
>- C.- Safety-EvaluationiSummary:
a The overall result of this modification-is increased assurance'of reactor' safety by improving plant fire protection and safe shutdown capability in accordance with Appendix R to-10 CFR 50 without adversely affecting any safety features. Safetyfrelated electrical equipment-which-
. \' could-be adversely affected in the unlikely occurrence of inadvertent-~ water discharge was protected from direct water impingement-by sealing the raceway-penetrations.into the top of the equipment and by providing deflector hoods over the equipment. The. potential. damage to_ electrical equipment that >
may be caused by actuation of the sprinkler-system is j enveloped by the Appendix R Safe Shutdown Analysis described in the Fire Protection Program. Therefore,'there.is no
, a adverse'effect on any FSAR.or. Technical Specification safety '
i considerations.
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Prach Bottom At'omic Power Station '
I Unit 2. )
Docket No. 50-277- a
' Annual 410'CFR 50.59 Report !
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Platform & Ladder Installation-Modification No.: 1669 A. System: Structural B. Description and Reason for Change:
l A Steel grate personnel platform andTladder were. installed in
/
.the Reactor Water Clean-upTIsolation Pit at. elevation'174'-
l 9". This eliminates the need to erect-temporaryEscaffolding l
to facilitate testing and maintenance of valves in the area,_
which is in the interest of ALARA.
C. ' Safety Evaluation Summary:
TheDnew platform and ladder were designed and installed such l that they can withstand design basis earthquakes'and would I
not damage equipment important to safety. This modification does not change the plant as described in the FSAR orfimpact any safety features.
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. R Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report
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Installation of Control Valves in the Drain Lines of the 3rd and 4th Heater Extraction Steam Lines Modification No.: 1684 A. System: Feedwater B. Description and Reason for Change:
Air-operated control valves were installed on each of the I three 3rd and three 4th-feedwater heater extraction steam drain lines. These drain lines run parallel to and bypass the extraction steam isolation valves. The drain control valves eliminate a bypass flowpath around the extraction steam isolaFion valves. The new valves close automatically on heater high-high level indication to prevent turbine water induction.
C. Safety Evaluation Summary:
The addition of the valves only affects the feedwater heater extraction steam drain lines which are non-safety related.
The valves do not affect the extraction steam supply to the feedwater heaters, and therefore do not affect feedwater O heating capability.
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Peach Bottoa Atomic-Power Station Unit 2 Docket No. 50-277 Annual = 10 '5'R 50.59 Report O-Replacement of Torus Vacuum Breaker Pressure Switches Modification No.: 1724 A. System: Primary Containment B. Description and Reason for Change:
Differential Pressure' Indicating Switches (DPIS) 2503 A'and B ,
were replaced with DPIS's specifically designed and tested to j meet the requirements of Class lE service, in accordance with l
) IEEE 323-1974 and IEEE 3d4-1975. The installation of the new l l DPIS's will strengthen the documentation of environmental !
l qualification.
l C. Safety Evaluation Summary:
i In addition to replacing the DPIS's an interposing relay was j added to the control circuit of each torus vacuum breaker !
valve. The interposing relay adds a negligible (25 1 milliseconds) delay to the operation-(opening)'of the torus vacuum breaker valves. The operation of the control circuit vacuum breaker valve and DPIS, as modified, is fail-safe as l 3]
i originally designed with improvements in equipment !
l performance and reliability.
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. l Prach Bottom Atomic Power Station '
Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report
( i Replacement of the 1A Feedwater Heater !
Modification No.: 1790 A. System: Feedwater j l
j B. Description and Reason for Change:
l The 1A Feedwater heater tube bundle'was replaced, and a 14-
. inch' drain line wascadded to assist in keeping'the liquid level in the Gheam side of - the heater to a minimum. The-replacement of the tube bundle.was necessary due to the high number of, plugged tubes in the. previous bundle.
C. Safety Evaluation Summary:
This modification does not involve safety-related equipment.
The replacement tube bundle performs the same function as the original tube bundle and meets the same design criteria.as the previous burdle. The changes do not adversely affect the heating capabilities of the feedwater system. Therefore, no-new accidents are introduced, and there are no adverse affects to existing accident analyses.
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Pacch Bottom' Atomic Power. Station Unit 2 DocketfNo. 50-277
. Annual 10 CFR'50.59 Report O J Personnel Contamination Monitors Modification No.: 1905
-A. System: Structural B. Description and Reason for~ Change:
L i-Personnel Contamination':(Eberline-PCM-1s) were installed at' various -locations, and : shielding wasi-installed for the PCMs.
and existing friskers as necessary. This improves frisking capabilities-for personnel.
C. Safety Evaluation Summary:
- The PCM-1 monitors and shielding do not-perform a safety
. function but may.be;1ocated'near equipment important to safety. A seismic analysis was performed to assure.that the PCM-l's and shielding located-near equipment important to-safety will have no effect on the' operation of this equipment. Malfunctions such asLleakage of'the P-10 gas,-
potential missile of the gas cylinder, and'the effects of an-earthquake have been evaluated and they do not cause an
({} unreviewed malfunction. This modification does not: adversely 1
' affect the plant as described in:the PSAR or. Technical' Specifications or reduce the level of fire protection.
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j Pacch Bottom AtomiciPower Stdtion Unit.2 :
Docket.No. 50-277
' Annual 10 CFR 50.59 Report LRPS Protection Panel' Modification i
Modification No.: 1916 i l
-A." System: Reactor Protection System (RPS) 1 B. Description and Reason for Change:
This' modification wires.the spare' pole;on.each. breaker.in RPS {
protection. panels:2AC757, 2BC757 and 2CC757Asin aTresult series with
'the. breaker 'a' switch and shunt trip coil. of, this modification, the' shunt trip coil circuit.will'open . !
= whenever the breaker is gpen or tripped .thus assuring that j the coil's.momentaryfrating.is.not exceeded. 'The shunt: trip j i
local indicating lamp was' replaced.with a high intensity lamp and the breaker trip control room alarm circuit was improved.
C. Safety 4 Evaluation Summary:
This modification will. provide additional protection against exceeding the momentary rating of the shunt trip coil without inhibiting the existing level of protective relay trip. I protection. Operator awareness of an existing RPS power
'() supply breaker trip condition.is improved both'in the control
~
room and locally.at the RPS protection panels.
undervoltage, overvoltage and underfrequency trip setpoints s
.The for the RPS breakers are not changed by this modification, and the reliability of the shunt trip coil, circuit is ;
improved. -l 1
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. Pacch BottomiAtomic,Powar Stationif , .3 UnitJ2:
Docket No. 50-277- H
' Annual:10.'CFR 50.59 Report l
(k Reroute of Control Cables for SRVs-Modification No.: 1950 y.
A.- System: Automatic Depressurization' j
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B.. Description and: Reason for Change:
.The control: cables for? safety relief valves (SRVs) RV-2 ~071E, H,:and.J were rerouted...The purpose of.this .
modification::was'to meet'.10;CFRL50 Appendix R requirements.
1 C. Safety Evaluation Summary: ,
' Prior to completing this' modification, all 11 SRVs were routed through a common area and subject to damage from'a
.singleifire.; -Rerouting the. cables to different fire areas ensureu the' ability to depressurize'the reactor vessel.: -The safety' function of the SRVs'is.'not.affected by the relocation of' cables because the relocated cables'are installed in accordance-with approved-installation specifications.
Relocating the control cables shortens the; cable length and'
' thereby reduces the voltage drop.-
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Prach Botton Atomic Pow 3r Stetion Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Rnport A-Installation of Back-ibp Nitrogen Supply to Non-ADS MSRVs Modification No.: 1950A A. . System: Automatic Depress:irization D. Description and Reason for Changer A backup pneumatic nitrogen supply was and installed J. This for safety relief valves (SRVs)'RV-2-02-071E, H, modification was required for compliance with Appendix R to 10 CFR 50.
C. Safety Evaluation Summary:
The backup nitrogen system was installed as a Q-listed seismic system. The back-up system complies with. primary j
containmentcisolation requirements, and is designed for Appendix J. leak testing requirements. The installation of
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the system does not alter the function of the SRVs, and therefore, does not introduce any new accident scenarios.
The new system provides-the capacity to operate the three SRVs for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an Appendix'R fire.
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i, Prach' Bottom Atomic Powar Station Unit 2 Docket No. 50-277 j
> Annual.10 CFR 50.59 Report i O
Investigate Design of CST Moat l
Modification No.: 1965 i A. Sirstem: Condensate B. Description and Reason for Change:
The dikes surrounding.the condensate Storage Tanks'(CST),
Refuel Water ^ Storage Tanks (RWST) and the Torus Water-Storage Tank were waterproofed. The purpose was to-prevent.an a uncontrolled release.of radioactive water to the' environment
'in case of an overflow or tank rupture.
C. Safety Evaluation Summary:
-This-modification has..no effect'on any plant safety related s.
systems or equipment. The waterproofing involved the application of asphaltic concrete overlay, coating the asphalt with an aluminizer and sealing all joints in the piping insulation with caulk. The work maintains the water tightness of the dikes, and has no effect on the seismic analysis of the dike walls. !
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- 1 Prach Bottom Atomic:Powar Station .
Unit 2 "
1 Docket No. 50-277 Annual 10 CFR 50.59 Report Reactor Recirculation Pump Shaft Crack Detection Modification No.: 1982 A. . System: Reactor Recirculation B. Description and Reason for Change:
A pump shaft crack detection system wasEinstalled.for.the .
reactor recirc. pumps. This is a data acquisition system 1 that collects vibration data, stores and transmits the data- i offsite-for analysis. Two additional proximity probes were f installed on each. pump, along with one monitor amplifier for {
each. pump. This modification was implemented to avoid a' i
shaft failure as addressed by I.E. Information Notice 86-19 ]
and General Electric SIL No. 459. j i
1 C. Safety Evaluation Summary:
The previously installed monitors will still provide vibration level indication and alarms-to the operators.- This .
modification does not effect the performance or operation of q any safety or non-safety related equipment previously evaluated in the FSAR, and does not effect any Technical,
'( ) Specifications.
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l Peach Bottom Atomic Power. Station Unit 2 Docket No. 50-27'
. Annual 10 CFR 50.59 Report. j
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Temperature Indicator Replacements -q Modification No.: 2006 A. System: Primary Contai'nment B. ' Description and Reason for Change: ]
l Temperature Indicators 2501 and 2100 were replaced with new digital models. These-indicators lr>nitor various area and D process temperatures throughout the plant, including the-drywell. This modification was implemented because the previous indicators were obsolete and~ spare parts for them: )
were no longer available. ]
i C. Safety Evaluation Summary:- J Overall plant reliability and maintainability was improved by.
replacing obsolete. equipment with functionally similar new equipment. The replacement of these instruments with new instruments which perform the same function has no effect on the probability or consequences of an accident, and no effect r- on margins of safety. The malfunction of the new instruments i has the same effect as that'of the existing instruments.
Reg. Guide 1.97 and proper seismic mounting considerations were addressed in the design of this modification.
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. Pea'ch Bottom Atomic Power' Station UnitL2 ] s
. Docket No. 50-277 Annual'10 CFR 50.59 Report
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Encapsulation and/or Reroute of' Safe Shutdown Cables i
.j Modification No.: 2078 j A. System: Fire Systems- d :l B. Description _and Reason.for Change: ,
This modification involves the encapsulation and/or reroute-of. safe shutdown ~ cables. Thisimodification brings PBAPS in compliance with criteria /provided in Appendix R/to 10 CFR 50. .)
C. . Safety Evaluation Summary:
This modification involves encapsulating and/or rerouting :,
raceways that contain cables connected to, safety-related equipment. As a result of this modification, the safety-related' function.of the systems, which are associated with the affected cables, is not changed. _This modification
~
brings PBAPS into compliance with' criteria provided in Appendix R to 10 CFR'50.
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Prach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report
'O Protection of Safe Shutdown Valves Modification No.: 2079 A. System: High Pressure Coolant Injection, Reactor Core Isolation Cooling and Emergency Service Water B. Description and Reason for Change:-
This modification corrected three motor-operated valve-circuits which were not in compliance with Appendix R to 10 CPR 50. In each case, spurious operation of the valves could have resulted from a fire in areas through which valve control cables were routed and in which operation of the valves was required for safe shutdown. This modification corrected these problems.by either relocating valve control circuits.to. fire areas in which operation of the valves is not relied upon for safe shutdown or by making logic changes in the control circuits to prevent' spurious operation.
The valves affected by this modification are:
MO-2-23-15: High Pressure Coolant Injection (HPCI)
() Inboard Steam Supply Isolation Valve MO-2-13-15: Reactor Core Isolation Cooling (RCIC)
Steam Line Isolation Valve MO-0498: Emergency Service Water (ESW) Discharge J Valve C. Safety Evaluation Summary: i This modification does not involve an unreviewed safety qu'astion.. The function of the HPCI and RCIC systems is-unchanged by relocating the controls for the steam isolation valves. Automatic closing capabilities of MO-0498 will be eliminated; however, annunciation of conditions which require this valve to be closed will provide sufficient time for the s operators to close the valve. Therefore, the margin:of safety'is improved by providing a means of supporting EPCI, !
RCIC, and ESW operation for those fire areas where they are !
relied upon for safe shutdown.
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Peach Bottom Atomic Power Station Unit 2 i Docket No. 50-277 Annual 10 CFR 50.59 Report j
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HPCI Turbine Pushbutton and MO-23-24 Logic Isolation Switch Modification No.: 2080 A. System: High Pressure Coolant Injection (HPCI)
B. Description and Reason for Change:
This modification involves the addition of a set of contacts j to the HPCI trip pushbutton in the control room to ensure HPCI can be manually tripped in an Appendix R fire scenario.
Also, an isolation switch is added to the control circuit for l valve MO-2-23-24 to isolate a sustained close signal which j could prevent HPCI recirculation flow in the event of a fire- I induced cable failure, l l
t C. . Safety. Evaluation Summary: 1 Manual trip logic has been added to the system while maintaining existing trip logic, to provide system control during a 10 CFR 50 Appendix R fire. Accident events discussed in the UFSAR are not affected by this modification.
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Perch Bottom Atomic Powar Station Unit'2 Docket No. 50-277 Annual-10,CFR.50.59 Report .
Safe Shutdown Lighting Modification No.: 2081 !
I A. System: Security (lighting)
B.- Description and Reason for Change: .
Emergency lighting was installed in various locations of the l l plant-to illuminate areas where safe shutdown manual j operations may need to be performed (included the routes of !l l
access and egress). The lighting is powered-by 8-hour j
' battery packs which are continuously charged by local AC {
lighting circuits. Loss of AC power causes the lighting to automatically. switch on. j 1
- p. ' C.-vSafety.--Evaluation Summary: {
The emergency lighting does not adversely affect any safety related equipment including that equipment's ability to perform during design basis events. The lighting fixtures were seismically qualified to preclude interaction with a nearby safety related equipment. No Technical Specification l
.( ) Bases involve emergency lighting and no other Technical l Specification Bases were affected.
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Peach Bottom Atomic Powsr Station f Unit 2 Docket No. 50-277 i Annual 10 CFR 50.59 Report I~')
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Circuit Breaker Re-Calibration and Replacement l l
Modification No.: 2083 i System: Various l A. 1 B. Description and Reason for Change:
The settings of 12 adjustable 480 VAC circuit breakers were changed, and two 120 VAC fuses were replaced. The purpose )
was to achieve better coordination of safe shutdown equipment !
to meet Appendix R criteria. )
C .. Safety Evaluation Snmmary:
No additional electrical loads resulted from this The ir.,reased breaker settings do not degrade j j
modification.
the level of safety of the plant. This modification enhances !
safety by providing better electrical reliability through )
improved coordination between electrical devices. Fault current levels are within the ratings of protective l equipment. )
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277-Annual 10 P.FR 50.59 Report O
RHR Pump Minimum Flow Bypass Valve Modification No.: 2084 A. System: Residual Heat Removal (RHR)
B. Description and Reason for Change:
The circuit of minimum flow bypass valve MO-2-10-16D was modified so that it normally remains open. The purpose of this modification was to resolve a 10 CPR 50 Appendix R scenario that would cause a short in the minimum flow v31ve, causing it to close and possibly failing the 'D' RHR pump on loss of minimum flow capability.
C$ Safety Evaluation Sumnary:
The consequences'of MO-2-10-16D failing to close on a safety injection signal have been addressed. The flow diversion through this line would not prevent the RHR system from providing adequate core cooling. Although the reduction in flow to the vessel could result in a slight increase in peak-clad temperature, there would still exist a substantial llh margin with respect to the peak clad temperature limit.
The consequences of MO-2-10-16D remaining open and providing a drainage path from the reactor has also been evaluated.
Whenever in shutdown cooling mode, MO-2-10-16D will be closed under administrative control thereby preventing a drainage ,
path. Additionally, MO-2-10-16D is interlocked closed prior to isolation valve MO-2-10-17 opening.
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Pancii Bottom Atomic- Powsr Station Unit 2 Docket No. 50-277-Annual 10_.CFR 50.59 Report RHR Motor Space Heater Upgrade-Modification No.: 2189A' A.- System: Residual. Heat Removal-(RHR)
B. Description and Reason for Change:.
A second space heater.was added-to each RHR pump-motor.. The
~
second space heater was needed to maintain motor stator windingLtemperatures 20 degrees F higher than ambient to prevent condensation on the motor winding.
C. Safety-Evaluation Summary:.
The space heaters and 1associated' circuits are'non-safety
, irelated. eThis. modification does'not changexthe function or
- operability-of the~RHR system.- The additional heater' raises the motor ~ stator winding temperature to a previously described level. Failure of an RHR pump motor is already considered in existing accident analyses.
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Pacch Bottom Atomic Power: Station-Unit 2 ;
Docket No. 50-277 {;
Annual-10 CFR 50.59 Report I ,
Reinspection and Reanalysis of Masonry Block Walls l
Modification No.: 2235 A. System: Structural i Description and Reason for Change:
B.
This modification involved the reinspection, reanalysis and ] J modification of masonry (concrete block) walls. The purpose was to assure compliance with IE Bulletin 80-11 " Masonry Wall Design".
C. Safety Evaluation Summary:
Neither the resulting individual modifications nor the exploratory core, boring adversely affect the function or !!
structural integrity of the block walls. The individual
~
modifications ensure the structural integrity of concrete. I block walls and blockouts, and therefore have no adverse effect on plant equipment installed in proximity to the walls.
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1 P:Ich Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report' i H
Replacement of Safety-Related Relays R Modification No.: 2249 1 l
j A. System: Miscellaneous l '
l B. Description and Reason for Change:
A number of safety-related circuits were equipped with time ]
delay relays which can no longer be purchased as 1 l environmental and seismic qualified replacement parts. This modification replaced these relays with another model that was environmental and seismically qualified.
C. Safety Evaluation Summary: l This modification replaced time delay-relays, which could.not longer be obtained, with new time delay relays which meet or 1 exceed the performance of the replaced parts. This modification did not change the plant as described in the UFSAR.
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a Parch: Bottom'Ato:2ic Pow 3rLStation l
' Unit:2.
Docket!No. 50-277-
= Annual 10 CFR:50.59 neport' a
Inverter Feed for SAS and CAS Security System ,
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Modification No.: 2275 A. System: Security i B. Description and Reason:for Change:
.An alternate-electrical' feed was provided from.00B79 through l
' transformer,20X150 which was-upgraded to aL30KVA, regulated type. The bypass / isolation switch.!20S315 was moved'to:the:
Cable Spread-Room. These were done'to satisfy 10 CFR.73.55 which requires selected power supplies to bealocated within.
1 vital areas.- 1 C. Safety Evaluation Summary:
1The relocation'of'the transformer.and the maintenance bypass- 'l switch to vital areas makes the uninterruptible!120VAC power? '
systems more secure. This modification increases'the-capacity of the~ single phase.480V-120V transformer. There- 1 are no electrical loads added byfthis modification'. ']
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a Peach Bottom Atomic Power Station '!
Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report A
b Replacement of Feedwater Long Path Recirculation Valve Modification No.: 2281 i
A. System: Feedwater B. Description and Reason for Change:
This modification involves replacement of Feedwater Long: Path' Recirculation Valve MO-2663. This valve is not a safety related component; however, safety related piping, anchors, and supports are affected by this modification because the replacement valve assembly is heavier that the original valve assembly. One pipe support has been modified to maintain piping system integrity. Severe erosion of the valve ~ body.
and seat, and a histery of maintenance problems with thL original valve.are the reasons for this modification.
C. Safety Evaluation Summary:
This modification does not create an unreview?d safety question. No functional changes have been made to the system and the system itself serves no safety related function. The replacement valve / operator assembly meets or exceeds all requirements of the original valve / operator. The system's l ability to function as designed is maintained.
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l Peach. Bottom Atomic Power Station Unit 2 ;
Docket No. 50-277 j Annual 10'CFR S0.59 Report' ]
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Replacement and Repair of Fire Resistant Doors and Frames i
Modification No.:.2318 l A. System: Fire Systems B.. Description and Reason for Change:
This modification repairs and replaces fire resistant doors, door frames and door hardware which currently do not meet plant requirements C. Safety Evaluation Summary:
The modification will enhance fire protection capability by ,
reducing the effects of a fire on the plant by insuring that "
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- fire door 4 assemblies provide the required' fire resistance.
Safety-related, secondary containment fire door assemblies-will be replaced with equipment which is capable of' performing the same function as the existing assemblies.. The modification provides the required fire resistance of the fire door assemblies.
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Peach Bottom Atomic Power Station. >
Unit 2 Docket No. 50-277 Annual 10..CFR,50.59 Report '1 1
HPSW Pump Packing Box Modification Modification No.: 2329 ;
1 A. System: High Pressure Service Water.(HPSW) i D. Description and Reason for Change: a The packing' configuration and leak-off piping'of the HPSW pumps have been modified to reduce pressure in the packing i box, thereby improving packing and shaft sleeve life. High '
pressure in the packing box previously resulted in short l packing life and accelerated shaft sleeve wear. 1 C. Safety Evaluation Summary:
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I This. packing box modification does.not impact the operation !!
of equipment 'important to safety, nor does not it impact any .l analysis of equipment important to safety as previously identified in the UFSAR. This modification increases pump i availability by extending packing service life. No 'l unreviewed safety queron is involved.
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. 'i Peach Bottom Atomic'Powar. Station Unit 2 Docket No. 50-277
. Annual 10~CFR 50.59 Report l 7"s i Q) 1 Addition of Terminal Blocks to the SPOTMOS System j i
Modification No.: 2353 I
- A.- System: Primary Contains.ent B.- Description and' Reason for Change:
Thisimodification added four terminal strips-in the SPOTMOS system panel 20Cl24. .Six foot cables were connected-to
~ temperature indicator TIS-2-2-71A&B and terminated at the new terminal strips.. The new terminal' strips will allow a quick lI '
replacement of only the six foot sect' ion.
C. Safety Evaluation Summary:
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.Thic; modification /willradd'four terminal strips in SPOTMOS system' PANEL-20Cl24'so that the cable-replacement can be faciliated by requiring only cables from the temperature indicator to the terminal strip. In addition, the terminal
' strips will permit voltage measurements which are not now s possible. In the process of mounting and demounting the
]
f- currently installed 1 cables / connectors, wear occurs-in the ~
l . connectors requiring replacement of the entire cable. There are no control or indicating functions added or deleted by this modification. ,
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- , s Prach. Bottom Atomic Power Station Unit'2' Docket No. 50-277 Annual 10.CFR-50.59 Report m'
. U,.l Inspection / Rework of Splices Modification No.: 2355 A. Sys!. em: Various B.. Description 1and Reason for Change:
. Field-installed equipment lead wire splices were-visually inspected for proper ' installation per splice manufacturer's installation instructions. . Any-improperly installed splices were reworked. This modification was prompted'by the results of a previous sample inspection and I.E. Information Notice 86-53.
,, C.-l Safety ~ Evaluation Summary:
This work had no effect on the plant _as discussed'in the.FSAR or Technical Specification Bases, and did not. change any equipment or system design. Proper installation was merely ensured.
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Peach Bottom Atomic' Power Station Unit 2 Docket No. 50-277 Annual 10~CFR'50.59 Report Emergency Service Water Small Piping Replacement Modification No.: 2371 A. System: Emergency. Service Water (ESW)
B. _D_escription and Reason for Change. .;
Two-inch diameter and smaller ESW pipe in the Core Standby Cooling System Rooms (RHR, Core Spray and HPCI) was replaced with the same class of pipe. The pipe routing and the valve placement was duplicated. However, valves that were purely for redundancy were eliminated and other valves were replaced with new valves. Some globe valves were replaced with lower res 'tance plug valves and the valve immediately downstream of each cooler was replaced with a plug type balancing valve. <
Low poirt drains.and lateral c' "-aut fittings were added d for-improved' maintainability.- 'inese replacements and changes are the result of pipe wall corrosion and fouling due to use of untreated river water. Flow blockage was being experienced. )
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C. Safety Evaluation Summary:
\
This modification ensures that the system can function as designed and as described in the UPSAR. The changes do not change the system's operation as described in the UFSAR. The replacement pipe and valves were procured, installed and ,
tested in accordance with the same, or more stringent, codes and specifications as those governing the former components.
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Psech Bottom' Atomic Power. Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
Main Steam Line Radiation Monitor Cable Replacement Modification No.: 2383 A. System: Reactor Protection B. Description and Reason for Change:
The' main steam The oldlineorganic tunnel radiation monitor cables were cables replac" were
' with silicon
' replaced.
dioxide insulated Old conduitcable wasthat has aand removed stainless sceel outer a new junction box and sheath.
new connectors were installed. This modification was implemented!to provide more resistance to'the high ,
radiation /high temperature environment.
, C.- safety Evaluation Summary:
This change does not alter the plant as' described in the FSAR or Technical Specification Bases, and does not change theThe performance of any safety or non-safety related system.
new cable and connectors are environmentally' qualified for
> nuclear safety related applications and do not increase the
() power loading of the plant- The new material is superior to the old material. i l
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j Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 -'
Annual 10.CFR 50.59 Report 8
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l Separate Power Supplies for Feedwater Heaters l and Feedwater Recirc Valves q Modification No.: 2389 A. System: Feedwrter D. Description and ReaLon for Change:
i 3eparate 120V power supplies were provided for-the. extraction i steam block valves on the three strings of the 3rd, 4th, and 5th feedwater heaters and the reactor feedpump (RFP) recirculation valve solenoids. This modification was required to restore the feedwater system to within the bounds of the Loss of Feedwater Beating event as discussed in UFSAR 14.5.2.2.
l C. . Safety-Evaluation Summary:- {
l This modification ensures that a loss of a single Y panel will not result.in the closure of all three strings of extraction steam block valves, thereby causing a'feedwater.
heating transient beyond that analyzed in UFSAR 14.5.2.2. ;
({} The addition of twc 120V power supplies into the feedwater and RFP recirculation systems will not directly affect the l function or operation of any safety related equipment.
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Peach Botto3' Atomic Power Station
~y Unit 2 Docket No. 50-277
' Annual 10 CFR 50.59 Report f'
b] .
Feedwater Long Path Orifice Replacement Modification No.: 2391 A. System: Feedwater B. Description'and Reason for Change:
The feedwater long_ path orifice has been replaced with multiple flat plate orifices to eliminate vibration which has-been experienced in the long path recirculation piping downstream of'the orifice.
C. Safety Evaluation Summary:
This. modification does not involve safety related material or ;
v .
, equipment and'will,not'effect any safety related portions of the feedwater system. The plant as. described in the UFSAR~is not changed. No unreviewed safety question is involved.
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Perch Bottom Atomic Powar Station
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Unit-2 Docket No. 50-277 Annual.10 CFR 50.59 Report O
Equipment Access Lock Doors (Railroad Doorsl Modification No.:,2489A A.. System: Secondary Containment B. Description and Reason for Change:
The secondary containment doors were upgraded. This included-strengthening welds on the guide pipe,. adjusting.the strike rod and inspecting / upgrading door hardware. This was donecto upgrade.the railroad door requirements to meet seismic design cri te.ria.
C. Safety Evaluation Summary:
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This-modification does notfaffect safety-related equipment-other'thar. t.he doors. The repair work restores the structural qualification of the doors to within design requiremenic for Seismic Category I. structures. This work has no impact on the probability of analyzed accidents or malfunctions'of interfacing safety related equipment.
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Peach Bottom' Atomic Power Station j Unit 2 Docket No. 50-277 Annual 10 CPR 50.59 Report Modify Start Logic for RHR Compartment Cooler Fans Modification No.: 2578 A. System: Residual Heat Removal System (RHR)
B. Description and Reason for Change:
The starting logic for the RER compartment coolers was
-changed. The coolers will start on either an RHR pump start signal (present design) or when the RHR system receives a LOCA signal (this modification). With the present design, the RER compartment coolers may not start during.a LOCA without a loss of offsite power (LOOP), in particular, when only one offsite power source is available. This is a result of potential low bus voltage caused by attempting to start an RHR pump,.the-RER compartment cooler fans, and other 480 volt
'lcads at the'same time.
C. Safety Evaluation Summary:
Tnis modification starts the RER compartment coolers sooner j than the previous design during a LOCA to prevent possible j starting problems due to degraded bus voltage. Previous '
start conditions are maintained. The modification meets the design. criteria of the RER and secondary containment BVAC systems.
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,1 b P :ch Bottom Atonic'Powar Station 'l Unit 2 Docket'No.'50-277 Annual 10 CFR 50.59 Report: ;
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-Trip Cooling Tower Loads on LOCA i Modification No.: 2580 ;
1 A. System: Cooling. Tower B. Description and Reason for Change: 1 1
2he cooling tower loads (pumps and fans) will be j automatically shed upon: receipt of a LOCA signal from Unit 2
.cr 3. . The signal:will be received from the Residual Heat j Removal (RHR) or Core Spray-(CS) relays. .The purpose of shedding these loads is to improve the voltage levels on'.the 13.8kV unit auxiliary buses and the.4.16kV emergency buses.
C. Safety Evaluation Summary:
1 The cooling tower-loads are non-safety related. Shedding- !
these loads accommodates the motor starting transients of the j l
Emergency Core Cooling System pumps. Cooling tower operation I is covered by NPDES permit No. PA 0009733 Rev. 9/1985. As indicated in the permit, cooling tower operation is not. 1
. required during various emergencies, including a LOCA.-
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-Paach. Bottom AtomicLPowar Station Unit 2 Docket No. 50-277 Annual;10 CFR 50.59 Report
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Pipe Hangers on-RWCU Dump" Lines to Condenser Modification No.: 4102 A. System: . Reactor Water Cleanup (RWCD)
~
B. Description'and Reason for Change:
Pipe restraints have been'added and-exiting supports and' Fire Barrier Penetration-Seals have been modified to restrain the large internal. flow loads that exist in the non-O 4. inch RWCD' dump lines to the condenser.
C. Safety Evaluation Summary:
This modification involves-no equipment important-to safety.
.The~ pipe loads.on safety related structures were reviewed and found acceptable. The support attachments <were minor and -l installed in accordance with existing safety related ]
procedures.
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. Patch Bottol Atonic Powar Station Unit 2 Docket No.:50-277 !
Annual 10 CFR 50.59 Report ~
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Replacement of Relief Valves on RHR Suction Lines, Modification No.: 4112A '
A. System:. Residual. Heat Renioval J q
/
B. Description and Reason for Change:
The thermal relief valves and valve: connections for the Residual. Heat Removal (RHR)' pump suction lines were replaced. .
.The purpose was to replace the previous non-Q relief valves q with Q relief valves.
C. . Safety Evaluation Summary:
The replacement relief valves and modified' piping' fulfill the f- requirements.that the original relief valves and piping were .
required'to meet. The new valves are. essentially the.same as 1 the original valve except for the type of end' connections, the materials of construction and set pressure. .These- i variances will not compromise the pressure boundary integrity j I
or safety function of the system. ,
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Prach Bottom Atomic Pow r Station Unit 2 Docket No. 50-277 Anaual 10.CFR 50.59 Report g
Removal of RCIC Electrical Overspeed Device l
Modification No.: 5001 i
A. System: Reactor Core Isolation Cooling Systam (RCIC)
B. Description and Reason for Change:
This modification removes the RCIC turbine electronic overspeed tripping device including the associated cables and connections. The removal is performed as recommended in GE Service Information Letter (SIL) No. 382.
C. Safety Evaluation Summary:
GE SIL No. 382 recommends the removal of the electronic overspeed trip feature from the RCIC turbine control system.
'The original reason for incorporating the device into the system was to provide a means of resetting the system remotely following an overspeed trip. But, operating experience has shown that when an overspeed electronic trip does not terminate the transient before the mechanical trip
, is actuated requiring the mechanical overspeed trip to be (j reset at the turbine. SIL 392 provides justification for
'~
removal of the devices. The RCIC will continue to be protected from overspeed transients by the mechanical overspeed trip device e i
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Psach Bottom Atomic Pownr StationL
-Unit ^2 Docket No; 50-277
..Annuale 10;CFR 50.59 Report
.( Y f Underrated Relay Contacts Modification No.: 5030 p . .'
A. System:- Oxygen'> Analyzers.
i B. Description and Reason for Change:
The contacts forLtwo' relays (63-ISO-2A and 63-ISO-2B) in-the.
drywell oxygen analyzer channel BLsolenoid valve control-circuitry were rewired. . The purpose was to resolve excess current on the relay contacts.
C.- Safety Evaluation' Summary:
-This. modification corrects .a condition of nonconformance te
.where arrelayscontact'sicurrent-rating was exceeded. This modification >does not' functionally change-the operability of the drywell oxygen analyzer or~the. primary' containment-isolation valves. . Itl simply wires ~two contacts in series to attain the required current ratings to support the. holding current for the eight solenoidLcoils. .
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Pacch Bottom Atomic Power' Station. I Unit 2 Docket No. 50-277
. Annual 10 CPR 50.59 Report; y
~ Containment 1 solation Check Valve-Replacement Modification No.: 5033 ,
A. System: Standby Liquid Control.(SBLC)-
B .-
Description:
and Reason for Change:~ ;
The SBLC Syst'em inboard isolation check valve (2-ll-16) was'-
replaced with a new type of valve because the existing valve !
-would notipass a local leak rate test. The new: check valve is an Anchor / Darling dual seat: disc. valve with.a soft elastomer seat. .An additional pipe support was=also added based on. piping calculations.
1 C. Safety Evaluation Summary: ij JThe' Anchor / Darling ~dualiseat[ check valve will perform the same safety function and is constructed to meet-all the l design requirements of: the original valve: and the: potential l increase.in pressure drop to the SBLC system is acceptable.
Therefore,~the Anchor / Darling valve is deemed an acceptable
- replacement for the existing swing check' valve.
- O The~ replacement piping used in the-modification-is a. lower
. carbon content austentic stainless steel with-better 1 corrosion resistance, the same hardness-and slightly-lower 'l tensile strength. The increased mass of the new valve,and j the difference.in the material. properties'were considered'in j kc the' piping calcul& tion and were' determined' to lua acceptable. >
The variances,will not compromise the pressure boundary integrity or safety function of theisystem-and will have no adverse impact on any other' safety related. systems, and will a not create any new or different' type of potential.'
malfunctions or accidents.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report
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i Reactor Feedpump Turbine (RFPT) Linkage Improvements Modification No.: 80-155 A. System: Feedwater B. Description and Reason for Change:
This modification placed reactor feedpump turbine (RFPT) relative motion linkages on bearing surfaces. Specifically, clevis / lever linkages were modified to accept bushings and new pins sized for a light press fit on the lever; and heim joints / lever linkages were modified to accept new sized pins for a light press fit in the holen of the heim joint and clevis. The reason for this modification is to prevent unstable oscillations in the RFPT control system that have been caused, in.part, by loose, worn linkages.
C. Safety Evaluation Summary:
This modification does not alter the function of the RFPT controls, either hydraulic or electrical. No safety related
. material or equipment is affected by this modification. This l,,l l
modification reduced the free play in the control valve l linkage pivot points, thereby reducing the possibility of l unstable oscillations in the RFPT control system.
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Peach Bottom Atomic Power Station 3 Unit 2 l Docket No. 50-277 '
Annual 10 CFR 50.59 Report
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l Main Steam Relief Valves Base to Body _ Flange Modification Modification No.: 83-018 l A. System: Main Steam B. Description and Reason for Change.
l This modification enlarged the base-to-body flange gasket !
groove outside diameter from 7 15/32" to 7 3/4" to !
accommodate a new flexitallic gasket wound with vinc plated 1 carbon steel to inhibit base material corrosion. The purpose of this modification is to eliminate external steam leaks common to the valves.
( C. Safety Evaluation Summary:
This modification does not affect the ability of the main steam safety / relief valves to perform their design function.
It merely is intended to eliminate external steam leaks from the valve. No other safety related equipment is affected by I
this modification.
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Parch Bottom Atomic Powar Station i Unit 2 l Docket No. 50-277 Annual 10 CFR 50.59 Report
,e Standby Gas Treatment (SBGT) System Plenum / Fire Deluge System Interface Modification Modification No.: 85-099 A. System: Standby Gas Treatment '
B. Description and Reason for Change:
1 This modification removed the existing 2" brass female j adapter, which was brazed to the 10 gauge steel plate i ventilation plenum on the outside and sweated to the copper tube on the inside. The adapter was replaced with a 2" carbon steel pipe threaded at both ends to make a nipple.
The nipple was welded to the plenum. A copper 2" female ,
I adapter was sweated to the 2" copper tube on the insidet This design will increase the strength of the interface by a factor of four. The reason for this modification was to prevent breakage of SBGT/ Fire Deluge adapter fittings.
C. S,afety Evaluation Summary:
()
This modification does not affect the function of any safety rele.ted systems or equipment. The operation of the SBGTS/ Fire Deluge System is not changed. The new design increases the reliability of the system.
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' Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
EBC Hydraulic Modification in Response to T'L-841-3A Modification No.: 86-004 A. System: Electro-Hydreulic Control (EHC) System B. Description asad Reason for Change:
This is a modification to the Electro-Hydraulic Control (EBC) system in response to TIL-841-3A. According to TIL-841-3A, two out of overy three tubing failures on General Electric Company high pressurc EBC systems have occurred on the 1/2" OD lines for certain EHC systems. This modification upgraded the EBC system and allow it to operate more reliably.'
C. Safety Evaluation Summary:
According to TIL-841-3A, two out of every three tubing failures on General Electric Company high pressure EBC systems have occurred on the 2/2" OD 1 nes provided for certain EBC systems. These failures resulted in forced outages. In addition to the 1/2" tubing, Pelded fittings were found to be much more reliable than the 37 degree flared ll) fittings currently used at Peach Bottom. The RBC system does not perform any safety functions; however, this modification reduced the probability of unexpected plant shutdowns, which enhances the overall safety of the plant.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
CRD Scram Discharge Volume Instrumentation Work Platform Modification No.: 86-042 A. System: Structural B. Description and Reason for Change:
This modification insta.lled a work platform around the Unit 2 '
CRD scram discharge volume instrumentation ,similar to Unit
- 3) to allow maintenanc-3 to be performed safely.
C. Safety Evaluation Summary:
The platform is not in physical contact with the scram discharge volume instrumentation. It allows work to be performed in a safer environment. This modification does not impact equipment important to safety.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
Installation of a Load Center Hot Spot Temperature Indicator, Modification No.: 86-060 A. System: 480V Load Centers / Emergency Load Centers B.
~
, Description and Reason for Chance:
This modification installed a hot spot temperature indicator on load center transformer E-324 to replace a defective temperature indicator installed under MOD 459.
C. Safety Evaluation Summary:
The installation of the load center temperature indicator was merely a like replacement of the temperature indicator installed under MOD 459. The function of the load center has not been changed. Malfunction of the temperature indicators would not affect the ability of the load centers to supply their safety-related loads. The operation of the transformers will be unaffected by this modification.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
Replacement of Instrument Nitrogen Gas Dryer '
Modification No.: 86-076 A. System: Instrument Nitrogen B. Description and Reason for Change:
This modification replaced the Instrument Nitrogen Gas Dryer '
(2AS207). This modification was required because one of the dryers failed, and that model is no longer manufactured. The new model is comparable to the previous model, but will use pressure switches and indicators instead ,of temperature switches and indicators.
C. Safety Evaluation Summary:
This modification involves the replacement of equipment with equipment that performs the-same function. The operation of the Instrument Nitrogen System will not be affected. This modification is not nuclear safety related.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
Replacement of Differential Pressure Transmitter Modification No.: 87-33 A. System: Control Rod Drive (CRD)
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B. Description and Reason for Change:
The differential pressure transmitter (DPT-2-3-218) between '
the reactor vessel and the CRD cooling water header was replaced with a new Rosemount model. The old transmitter, Barton model 268, was defective and is no longer manufactured. An isolated power supply was installed to convert the 120 VAC souce to 68 VDC for the new transmitter.
C. Safety Evaluation Summary:
No safety functions were affected by this modification. DPT-2-3-218 provides indication- only and serves no control function. The new transmitter fulfills the function of the old transmitter and is an acceptable replacement.
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j Peach Bottom Atomic Power Station "i Unit 2 Docket No. 50-277 l
1 Annual 10 CFR 50.59 Report i
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I Pressure Transmitter Replacement Modification No.: 87-45 j i
A. System: Core Spray l I,
~
B. Description and Reason for Change: a
~
The pressure transmitter on the discharge of the 'A' Core Spray Pump (PT-2-14-38A) was replaced with a new Rosemount l model. The old transmitter, GE model 551, had failed and is no longer manufactured.
C. Safety Evaluation Summary: 1
.i This modification has no impact on the performance of Core l Spray as discussed in the UFSAR. This transmitter is only j designed to monitor pump performance and does not serve any ;
. control function. The new model is of better quality and .;
fulfills the same function as the old model. l I
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
Temporary Replacement of Feeder Breaker to Battery Charger Modification No.: 87-047 A. System: 480 Volt Motor Control Centers B. Description and Reason for Change:
1 During preventive maintenance on MCC #52-3893 a 100 amp !
breaker was found to be damaged and in need of replacement.
The exact replacement part was unavailable and the eqtdpment '
needed to be put back in service immediately because Of a drain on the station 125 volt batteries. A new 100 amp j breaker with a higher range was used temporarily. This breaker was calibrated at 560 amps which was the setting on j the original 100 amp breaker. The temporary breaker was i subsequently replaced with an exact replacement. ! l C. Safety Evaluation Summary: - )
l This modification did not degrade safe operation of the plant because it involved a temporary replacement of a 100 amp ;
breaker with a qualifir- 100 amp replacement breaker. The J
7-).
u temporary replacement n~eaker has a higher range but the jl breaker was calibrated to the same setpoint as the original I breaker.
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Peach Bottom Atomic Power Station Unit 2 {
Docket No. 50-277 j Annual 10 CFR 50.59 Report Beginning of Cycle 8 Process Computer Update ,
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Modification No: 87-63 j i'
A. System: Process Computer ,
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B. Description and Reason for Change:
The Process Computer Data bank was updated to reflect the new core configuration for the beginning of Cycle 8. This update was made to assure that the Process Computer makes accurate core thermal calculations.
C. Safety Evaluation Summary:
)
-I This modification merely enables the Process Computer to do I the calculations 1ecessary to evaluate core thermal limits i evaluated in the FSAR. This update does not change the way 2 the calculations are done by the Process Computer; the data j for the calculations was merely updated to reflect the new core configuration. The revised Unit 2 Technical Specification fuel limits, such as MFLPD, MAPLHGR, and MCPR, were addressed by this update. 1
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Docket No. 50-277 i Annual 10 CFR 50.59 Report O
Installation of Lift Pads for Valve No. MO-2663 j Modification No.: 87-092 l l
A. System: Structural B. Description and Reason fo'r' Change:
]
This modification installed two lift pads above valve MO-
~
l 2663, feedwater long path recirculation valve and two lift !
pads for the operator for the valve. This will facilitate removal and reinstallation of the valve and operator during .
9 maintenance. The reason for this modification is that there 8 previously was no safe means of rigging the valve / operator for removal and reinstallation.
C. Safety Evaluation Summary:
This modification does not affect the function of any safety related systems or equipment. It provides a safe means of rigging the valve / operator for removal and maintenance. ]
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report HPCI Aux Oil Pump Pressure Switch Replacement Modification No.: 87-100 A. System: High Pressure Coolant Injection System B. Description and Reason fo'r' Change:
~
This modification replaced the existing auxiliary oil pump pressure switch, PS-4541, for the HPCI system with a new pressure switch. The failure of the existing swtich was discovered while performing preventive maintenance work. The existing switch model is no longer available, and a new safety-related switch was required.
C. Safety. Evaluation Summary: ,
The function of the pressure switch is to start the motor driven auxiliary oil pump, which provides oil pressure to the HPCI turbine. This switch senses the oil pressure at the discharge of the turbine (mechanical) driven oil pump, and starts the auxiliary oil pump whenever low oil pressure is
- () sensed and the HPCI system is in operation. The auxiliary i pump operates during startup and shutdown of the HPCI turbine. This modification only replaced the existing inoperable pressure switch with a new model switch that meets the system design requirements. The operation of the HPCI system has not been changed.
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_ nes'4 hme Peach Bottom Atomic Power Station Unit 2 Docket No.-50-277 Annual 10 CFR 50.59 Report'
-Replacement of Defective Pressure Transmitter Modification.No.: 88-035 A. System: Control Rod Drive (CRD)
, B. Description and Reason for' Change: i j
This modification replaced a defective GEMAC pressure
'i transmitter, which provided charging header pressure indication and high pressure alarms for the CRD system, with l a Rosemount transmitter. GEMAC no longer manufacturers the {
subject transmitter.
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C. Safety Evaluation Summary: -j i
.The-subject pressure transmitter does not-perform a safety j function and does'not interfere with equipment or devices '
that perform safety functions. The modification replaced the j defective transmitter with an equivalent transmitter. j i
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Peach Bottom Atomic Power Station' Unit 2 Docket No. 50-277 Annual.10 CFR 50.59 Report t'
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Replacement of RPS MG Set Voltage Potentiometer Modification No.: 88-038 A. System: Reactor Protection System B. Description and Reason fo'~r Change:
This modification changed the M-G Set voltage regulator
~
potentiometer from a 1-turn to a 10-turn adjustment potentiometer. This modification'provides finer control of manual voltage adjustments. The previous potentiometer was too sensitive and several M-G set trips had occurred as a result of using it to adjust voltage.
C. Safety Evaluation Summary:
The functions of the new potentiometer are the same as the
. previous one. No change in-mounting or installation methods were required. The new model potentiometer meets the system design requirements.
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Peach Bottom Atomic Power Station -
' Unit 2 Docket No. 50-277 i r m Annual 10 CPR 50.59 Report 1
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Replacement of Recombiner Transmitter
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Modification No.: 88-051 A. System: Off gas and Recombiner
\
_ B. Description and Reason for Change: j This modification replaced the recombiner pressure i transmitter for the glycol pump discharge pressure. The j previous transmitter was defective and it is no longer j manufactured. The new transmitter meets the system j requirements using the existing 24V DC power supply. j s
C. Safety Evaluation Summary:
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.This modification replaces a component with a like component j that performs the same function. No equipment important to ^
safety is affected and no other system changes or additional electrical loading was required.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 n Annual 10 CFR 50.59 Report N-]
Install Anti-Rotation Pins in the Reactor Feedpump Discharge Check Valves (AO-2(3)l47A, B, C)
Modification No.: 88-057 A. System: Feedwater System B. Description and Reason for Change:
This modification installed an anti-rotation pin in the Reactor Feedpump Discharge Check Valve and also a solid pin instead of a cotter pin on the disc nut. The modification was recommended by the valve vendor, Schutte & Koerting.
This modification will help prevent the possibility of separation of the disc from the tail link by preventing the disc nut from loosening by the rotation of the disc. The cotter pin and disc nut were missing from valve AO-2147A when it was inspected. -
C. Safety Evaluation Summary:
) The Reactor Feedpump Discharge Check Valves (AO-2147 A, B, C) were inspected to assess the effect.s of previous system transients which caused slamming of the valves on several occasions. This modification implements the recommendation made as a conclusion of the inspection, and decreasca the possibility of a feedwater transient which, in turn, decreases the probability of a scram.
The check valve disc and tail link was drilled anij an anti-rotation pin installed. This pin will help prevent the '
possibility of the separation of the disc from the tail link, by preventing the disc nut from loosening by the rotation of the disc.
II Additionally, the cotter pin holding the disc nut was replaced with a solid pin which will provide better holding power.
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Peach Bottom Atomic Power Station '
Unit 3 Docket No. 50-278 Annual'10 CFR 50.59 Report I
Remote Shutdown Panel / Human Factor Enhancement ,
Modification No.: 1958 ,
A. System: Miscellaneous
- B. Description and Reason for Change:
1 a
~ The background of the remote shutdown panel was painted beige, j color fields and outlines were added to group related instruments j and controls, and the panel was re-labeled using a hierarchial !
labeling scheme. These enhancements were made in response to !
deficiencies identified during the Control Room Design Review 1 (CRDR). j J
C. Safety Evaluation Summary: .I These modifications are merely enhancements to the remote shutdown panel to improve human factors. These enhancements will l
not affect the function of any instruments. They will enhance :
safety by reducing the overall probability of operater errors. !
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Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual 10 CPR 50.59 Report O
Recirculation Piping Chemical Decontamination ,
Modification No.: 1967A 1 A. System: Reactor, Recirculation System, Residual Heat Removal System and the Reactor Wat_er Cleanup System i
' 1 B. Description and Reason for Change: ,
This Safety Evaluation addresses the chemical decontamination of the Peach Bottom Atomic Power Station Unit 3 Recirculation l System, portions of the Residual Heat Removal (RHR) System and '
portions of the Reactor Water Cleanup (RWCD) systems, Prior to the removal of pipe during the Unit 3 pipe replacement d outage, chemical decontamination will be performed on the .j Recirculation System.and portions of the RHR and RWCU systems. .;
Radiation fields on the recirculation, RHR, and RWCU piping will be the source of most of the-exposure during pipe replacement, 1
therefore, removal of the radioactive corrosion products that cause these fields will be one of the cost effective ways of controlling radiation dose.
-( ) C . Safety Evaluation Summary- l l
Although the chemical decontamination affects safety-related components, the decontamination process itself is not considered ;
safety-related. Sufficient quality control will be provided to
.l assure that all critical aspects of the decontamination process "j will be maintained so that there will be no deleterious effect on the remaining Recirculation System piping components. -
During the decontamination there will be no fuel in the reactor vessel. The Recirculation System piping will be replaced ";
following the decontamination. From a structural integrity standpoint, exposure to the decontainment for those components which are to remain in service presents no significant concerns.
Coupon evaluation and inspection may dictate additional inspection or replacement requirements prior to returning the decontaminated components to service.
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Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual 10 CFR 50.59 Report C
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13kV Outage Power Source Modification No.: 2042 -
A. System: 13 KV B. Description and Reason for ' Change: 1 This modification installs two 1000 kVA, 13.2 kV/480 volt i transformers to provide permanent outage power facilities. The purpose of the modification is to provide permanent outage power for construction and maintenance to facilitate large outages. j i
C. Safety Evaluation Summary: ,
1 This modification makes changes to the Remote Miscellaneous Building power system. A temporary installation of two 750kVA q wye-wye transformers is being replaced by the permanent installation of two 1000 kVA- delta-wye transformers. These 13kV i to 480 volt transformers will supply the outage power needs of 4 future maintenance and construction projects.
[~N t i) No changes to the 30A01-11 13kV feeder circuit breaker settings i are required. These breakers are presently set to the current ]
f limit of the cables that ace connected to the breakers.
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Peach Bottom Atomic Power Station !
Unit 3 '
Docket No. 50-278- i Annual 10 CPR 50.59 Report )
$) i Recirculation Motor / Generator-Fluid Drive Scoop' Tube Fix !
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Modification No.:L86-118 A. System: Recirculation Motor / Generator j
( B. Descriptio'n and Reason for'Dhange:
4 This' modification replaced the oval shaped scoop tube on'the 3B {
Recirculation M/G set with a scoop tube of trapezoidal shape.
.j The' internal Heim joints were also replaced. The purpose of this 1 modification failures, was to reduce oscillations which have caused linkage w
Safety Evaluation Summary: .$
C.
]
4 This modification did not. change the movement of the_ scoop tube f or the function of the fluid drive.- The modificationiinvolved- j only non safety-related equipment and did not affect M/G Set operation or recirculation pump operation. The plant as described-in DFSAR Sections 4.3, 14.55, and 14.5.6 was not ]
changed.
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Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual 10 CFR 50.59 Report Update of Shutdown Rod North Minimizer (RWM) Sequence 46.42 Modification No.: 86-156 and 87-32 46.44 A. System: Process Computer 46.46
',_ B. Description and Reason for'dhange: 46.48
~
The RWMSEQ program of the process computer was used to update the 46.51' shutdown RWM sequence arrays. The modifications were necessary 46.53 to reflect new control rod patterns. 46.54 C. Safety Evaluation Summary: 46.57.
These modifications enabled the process computer to enforce the 46.60 proper RWM sequence, augmenting the RSCS rod worth control as 47.2 described in-the Fina Safety Analysis Report. The modifications 47.5 did not change the sc.,se or function of the process computer RWM 47.7 j sequence. They only revised-the control rod sequence to reflect 47.8 the current control rod pattern. Therefore, control rod worths- 47.11 L were limited, which minimizes the consequences of a control rod 47.12 drop accident as defined in the technical specification bases. 47.13 L
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Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual 10 CFR 50.59 Report
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Erection of Temporary Facility for Unit 3 Pipe Replacement Modification No.: 87-078 A. System: Structural B. Description and Reason for' Change:
This modification erected temporary facilit,ies west of the reactor building needed to support the Unit 3 pipe replacement outage. These facilities were used as the construction access building (CAD), dosimetry and access control card issue area, and a variety of Health Physics office facilities. These facilities
' contained all necessary fire suppression and alarm systems. The use of electrical power to the facilities was addressed in the Safety Evaluation for Mod 2042.
C. Safety Evaluation Summary:
Equipment important to safety is not affected by this modification, and the temporary facilities and their plant interfaces are non-safety related. The loading caused by the CAB on the Reactor Building slab west of the Reactor Building has
-(])- been evaluated and determined to be acceptable. The connections to the fire water system'do not degrade the system capability. , ,
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i Peach Bottom Atomic Power Station !
Unit 3 1 Docket No. 50-278 j Annual 10 CFR 50.59 Report i r
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I Revise Unit 3 Refuel Platform Speed Control Circuit )
Modification No.: 87-105 A. System: Fuel Handling
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Description and Reason for Change
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B.
This modification removed a jumper applied to the Refuel Platform speed control circuit. This. jumper added a 1k ohm resistor which i caused c feedback circuit to initiate in 4 milliseconds. The d system will not operate properly with only a 4 millisecond time delay. An overcurrent situation would result, causing a fault i{
lockout. By removing the jumper, the feedback circuit initiates in 300 milliseconds. The reason for this modification was to j '
allow the refuel platform to operate as designed with no sporadic fault lockouts.
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. _ .i C. Safety Evaluation Summary: l This modification does not alter the function of any safety related systems or equipment. It merely removes an unneeded i
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i jumper in the speed control circuit. This modification will enable the refuel crane circuit to perform its intended desion function.
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Peach Bottom Atomic Power Station Unit 3 l Docket No. 50-278 Annual 10 CFR 50.59 Report n
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Installation of Drain Line from Condensate Storage (CSTl
' Refuel Water, and Torus Water Storage to Backwash j Receiver Tank 1
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Modification No.: 98-013
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A. System: Ra'dwaste i
B. Description and Reason for Change:
This modification installed a 3" pipe with a shut-off valve between the Unit 2 CST, Refuel Water Storage Tank, the Unit 3 CST q j
and Torus Water Storage Tank Drain Header to the Unit 3 Backwash 1 Receiver Tank. This will provide a means of cleaning the Unit 3 J CST and Torus Water Storage Tanks by processing the effluent from these tanks to the Unit 3 Backwash Receiver Tank. '
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The purpose of the modification is to provide a means of cleaning '
the Unit 2 CST and Torus Water Storage Tanks by processing the effluent from these tanks to the Unit 3 Backwash Receiver Tank. )
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C. Gafety Evaluation Summary:
R7_)g This modification will not alter the operation of the CST, torus water tank, the condensate backwash receiver tank, or any safety ;
related system. The drain line will only provide a means to clean the Unit 3 CST and torus water storage tanks. Cleaning of i these tanks will not have a major impact on the radwaste system.
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Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual 10 CFR 50.59 Report f3
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Installation of Penetration From Reactor Core Isolation Cooling (RCIC) Room to Water Sludge Tank Room Modification No.: 88-014 A. , System: Radwaste B. Description and Reason for Change:
This modification installed a 3" penetration from the Unit 3 RCIC Room to the Waste Sludge Tank Room. This penetration provides a direct route for the removal of highly contaminated sludge from the Reactor Building Sump Room to the Waste Sludge Tank. The penetration is sealed when not in use.
C. Safety Evaluation Summary:
Although the wall will retain its waterproof, radiation proof, and air-proof integrity after sealing, failure of the seal will not adversely impact safe shutdown of the plant. Failure of the seal will be bounded by the FSAR accident analysis. The integrity of the wall is not degraded. No safety related systems
-(]) or equipment are affected by this modification.
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Pench Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual 10 CFR 50.59 Report O
Installation of Insulated Spacer in the Main Turbine Front Standard Coupling
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I Modification No.: 88-045 J A. System: Main Steam Turbine ,
B. Description and Reason for Change:.
This modification installed an insulated spacer in the front standard coupling. The insulated spacer, recommended by G.E. via TIL 973, protects the turbine front standard coupling from the harmful effects of shaft current discharge.
9 C. Safety Evaluation Summary:
J This modification vas needed due to the results of a recent inspection of the front standard coupling on Unit 3. During this inspection, the front standard bearings were found with electrolysis damage due to stray shaft current. This modification is strongly recommended by General Electric via TIL 973. The insulated coupling will protect the gear shaft teeth,
-() journals and bearings frou electrical pitting. This modification has no direct safety impact; however, improving the durability and reliability of the main turbine decreases the probability of an up xpected plant shutdown.
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I Peach Bottom Atomic Power Station l Units 2 & 3 1 Docket Nos. 50-277; 50-278 i Annual 10 CFR 50.59 Report j
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.1 First-In Alarm to Control Room Annunciators Modification No.: 0643 j l
A. System: Annunciators _
Description $and Reason for Change:
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The purpose of this modification was to eliminate, relocate, and' i group'together on a "first alarm in" basis' selected annunciator ;
l alarms in the Main Control Room'. The modification was designed to aid control room operators in diagnosing operational transients involving many alarms. 1 C. Safety Evaluation Summary: j
- The ' modification..did not ' involve " safety related . equipment. The f alarms affected by this modification were non Q-listed. No .
wiring changes were made at the devices being annunciated, j 1
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Peach Bottom Atomic Power Station Units 2 & 3 ,
DGeket Nos. 50-277; 50-278 Annual 10 CPR 50.59 Report )
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j Reactor Water Clean-up Pressure Instruments !
i Modification No.: 687 {
A. System: Reactor Water Clean-up (RWCU) -
B. Description'and Reason for Criange: )
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A differential pressure indicator was installed in parallel with' l the existing differential pressure indicator for each RWCU post
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strainer. The post strainers are in the outlet piping of the '
A&B RWCU filter domineralizers. The existing indicators are DPIS 12-4-72 A&B and the new indicators were designated DPI 12-4-98 A&B. Also, a pressure gauge was installed on the low ;
pressure leg of each instrument loop. j L
DPIS 12-4-72 A&B are in high radiation areas. The additional instruments were installed in lower radiation areas to reduce ,
I personnel exposure when operators use the instrumentation during d demineralized valving operations, i 1
C. Safety Evaluation Summary: ,
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) The installation of these instruments does not affect safety l related equipment or change the operation of RWCU as described I in the PSAR. The instruments involved in this modification do s not have any control function; they are merely for indication. l The piping and instruments involved are not safety-related, l n
The presence of these instruments does not introduce any new reactor safety considerations or new equipment failure modes. j The RWCU piping involved is separated from the primary system -
pressure boundary by automatic containment isolation valves. l
. Also, the new instruments are typical of instruments in the plant and described in the FSAR.
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Perch Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report Innta11ation of Block Valves and Test Connections Modification No.: 810B A. System: Turbine Building Coolin'g Water (TBCW), Drywell l Chilled Water (DCW),.and Reactor Building Cooling Water (RBCW)
B. Description and Reason for Chance: .
Block valves and test connections were installed to facilitate local leak rate testing of certain RBCW and DCW containment isolation valves. Valves added to the RBCW side of the system interfaces also serve to prevent leakage of RBCW water to clean systems (TBCW and DCW), which had been a problem. Associated valve indication circuitry was provided for these cooling loops.
C. Safety Evaluation Summary:
Failure of any of the new valves will not block flow of RBCW to essential equipment when required as described in the UFSAR.
! Adequate instrumentation was provided to ensure that for the f3 possible combination of valve failures, true indication of flow
() would be provided. Thus, this modification did not adversely affect the operation of safety-related equipment.
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L . Peach' Bottom ~ Atomic Powar Station Units 2 & 3 Docket Nos.- 50-277; 50-278
~ Annual.10.CFR 50.59 Report-10
' Flush and Drain Connections on Flow Elements Modification No.: 1078-A.- System: Residual' Heat Removal'(RHR)
B. Description and Reason for'Ch'ange:
~
One-inch. flush-and drain connections were installed on flow elements FE-2(3)-108 AEB and FE-2(3)-110 A&B ofl24 inch RHR pipe. .
This facilitates flushing of nrud that accumulates-in'the annular gap between the leading-edge of the flow element nozzle and the inside wall of the pipe and will reduce radiation-exposure to personnel.
C. Safety Evaluation Summary:
" These connections will not affect the safety' functions of RHR or any other system as described in the FSAR or Technical Specification. The connections were installed such that the pressure integrity.of the pipe will not be affected, and the'
([) small mass'of the connections is insignif; cant with respect to' seismic consideration for the pipe. The connections will not ,
-1 effect the differential pressure across the flow element and, 3 thus, will not affect flow measurement. i, Failure of the piping is covered by LOCA' analyses described in-the FSAR. The flow elements are upstream of automatic l a
containment isolation valves. l l
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Peach Bottom Atomic Power Station 1 Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report l
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l Fire Protection - Coating of Structural Steel Modification No.: 1228 l 3
A. System: Fire Protection
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B. Description and Reason for Change:
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A fireproof coating was installed on safety related structural _j steel members and covers were instal 3ed on cable trays j containing cables connected to safety related equipment. These {
modifications were done to conform to 10 CFR 50 Appendix R .j requirements. j 1
C. Safety Evaluation Summ*.ry: '
The need for the fireproofing was identified by a structural j steel survivability analysis. The fireproof coating was
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installed in sufficient thickness to ensure the beam's j capability to maintain its structural integrity daring a fire.
Cable derating caused by the installation of the tray covers has I been evaluated and will not create a safety hazard or adversely 7 affect equipment performance. This modification does not affect
- ) the function of the steel members or cable trays. It merely ]
enhances the fire protection provided. 1 i
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Per.ch Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report n
%I Alternative Shutdown Lighting and Communicatic p Modification No.: 1352I (Unit 2) and 1353I (Unit 3)
A. System: Secur.ity (lighting / communications)
- B. Description and Reason for Change:
2 Emergency lighting was installed in various locations on elevation 135. feet of the Radwaste Building and Control Structure to illuminate Alternative Control Stations (ACSs) for 10 CFR 50, Appendix R equipment (including the routes of access and egress). The lighting is powered by local 8-hour battery packs which are continuously charged from Class lE 120 VAC sources. Loss of AC power causes the lighting to automatically switch on; however, a manual switch was provided.
Emergency communications equipment was also installed to support use of the ACSs. Radio cot. coles were placed in the ACS areas and a radio desk console was placed in the Control Room. The battery powered radio base station can maintain communications for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after loss of its 110 VAC continuous charging fs source.
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C. Safety Evaluation Summary: j 1
Alternative shutdown lighting and communications are totally independent of other normal and emergency lighting and '
communications. Seismic Class 2 over 1 considerations and ;
approved mounting procedures were followed for mounting any lighting or communication boxes to any seismically qualified walls. The additional loads to the Class lE batteries have been .
i analyzed and do not degrade the safety function of the batteries.
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Peach Bottom Atomic Power Station 1 Units 2 & 3 )
Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report t'~'s
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Fire Damper Replacement l 1
Modification No.: 1571 i A. System: Fire Protection l a
_ B. Descriptic., and Reason for Change: j
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^ Fire' damper assemblies above doors in the Cable Spreading Room j were not. installed in accordance with the damper manufacturer's e installation instructions. The existing configuration could have impeded proper functioning of the dampers during a fire, i The affected dampers were replaced with dampers meeting the fire j protection requirements. j C. Safety Evaluation Summary. l This modification enhances fire protection capability by R 1
reducing the effects of a fire on the plant by ensuring proper f fire damper operation. The modification is not associated with d safety related equipment and does not affect the plant as i described in the UFSAR, Section 10.12. 1
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Peach Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report
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Upgrade of Chemistry On-Line Monitoring System Modification No.: 1879 A. System: Reactor Water Sampling .
._ B. Description and Reason for Change:
Conductivity monitoring equipment has been replaced with new temperature compensated conductivity monitoring equipment at the following sample points:
Feedwater, Reactor Building Sample Station Reactor Water Cleanup (RWCU) Inlet RWCU Filter Demineralized Discharge (A, B)
Condensate Pump Discharge (A, B, C)
Condensate. Filter Demineralized Outlet Make-Up Demineralized, Water Treatment Building C. Safety Evaluation Summary:
Changes to the existing sampling systems associated with this
/3 modification have been designed to maintain the existing flow i') control and pressure relief capability. These changes do not alter the design of the process sampling system as described in UFSAR Sections 4.9, 10.19, 10.20, 11.7, and 11.8. The design of the Makeup Water Treatment System as shown in Figure 10.16.2 of the UFSAR has been changed to include two additional conductivity recorders. The modification improves the ability a to monitor plant water chemistry and does not adversely affect ;
any systems important.to safety.
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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report MOVATS of HPCI and RCIC Motor CDerated Valves (IE Bulletin 85-03)
Modification No.: 1915 A. System: Various
- l B. Description and Reason for Change:
A safety-related motor-operated valve modification and test i program is being implemented in order to comply with the q requirements of I.E. Bulletin 85-03. The testing program is limited to motor-operated valves in the High Pressure Coolant j s
Injection (HPCI) System and Reactor Core Isolation Cooling 1 (RCIC) System. '
C. . Safety Evaluation Summary:
i The IEB 85-03 Test Program will demonstrate that the as-left j switch settings are adequate to support design basis operation as described in the UFSAR. Therefore, the end result of the ;
test program will be to increase motor operated valve r.nd system i
< reliabilities. The kotor-operator modifications are of either j
(_1) no consequence to safe >y or serve to increase motor-operated q valve reliability. j i
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Peach Bottom Atonic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report
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Plant Yard Manhole Fire Protection ,
4 Modification No.: 2091 4
System: Fire Protection o A. ;
J B. Description and Reason for Change: H This modification provided fire protection for electrical manholes located in the plant yard area which, if affected by a fire, could have adversely affected safe shutdown capability )
j required by Appendix R of 10'CFR 50. There is no fixed '
combustible naterial in the immediate. vicinity of the manhholes; however, combustible liquid from a transient source, if ignited j
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and allowed to enter selected manholes simultaneously, could 1 adversely affect safe shutdown. The modification raised the (
. concrete walls around the manholes to 6 inches above grade, replaced manhole covers with gasketed checker plate covers, and sealed.open piping penetrations between manholes to prevent q combustible liquids from entering the manholes. ~1 C. Safety Evaluation Summary: i
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~ (_]- The modification does not affect the plant as described in UFSAR .1 j~
Section 10.12. The protection provided for the manholes assures ;
that safe shutdown capability is maintained during a fire as ,
described in Section 6.3 of the Peach Bottom Fire Protection "
Program. The modification does not affect the function of any 4 safety rela *.ed equipment. J,a i
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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report Security Closures on HVAC Ductwork and Air Intake Openiggs Modification No.: 2096 and 2096A A. System: Ventilation B. Description and Reason for Change:
Security screens were installed in three air intak'e openings in )
each Reactor Building and in ten ventilation openings in the Diesel Generator Building to achieve conformance with Regulatory Guide 5.65.
l Tamper proof locking arrangements were installed in eleven fire damper access doors on HVAC ducts to secure the openings to i provide a delay to forced entry. The access doors are outside of vital arearboundaries but are in ductwork which cross vital .;
area boundaries. The locking c:rangements consist of steel -1 angles bolted to ductwork and a steel bar spanning the width of I the door with a padlock on one end and we.ded bar stock on the i other end.
1 r~3 C. Safety Evaluation Summary:
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The security screens and locking arrangements are i._ safety I related. The modifications did not adversely affect the ;
capabilities of the affected intake structures and ductwork. '
The modifications did not affect the plant as described in the 1 UFSAR Sections 5.3.2, 7.12.5, 10.14, 10.15, 13.6.5, and Appendix 4 O. The modification maintains the level of safeguard '
effectiveness required by the Security Plan. ;
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Peach Bottom Atomic' Power Station !
/ Units 2 & 3 Docket Nos.- 50-277; 50-278 ;
Annual 10'CFR 50.59. Report '
O i Power Supply to Support New Plant Computer Modification No.:l2215 A. System: Computer -
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_ B. Description and Reason for Change: J
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480 volt, 3 phase AC power. distribution panels, circuit breakers j and cables ~were installed in the Administration Building to. O provide. power for the new plant process computer and its. 1 associated equipment, all of which will be installed at a later date.
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C. Safety Evaluation Summary-
.The power circuits added by:this modification do not affect n plant safety. No safety related equipment was involved and this' i modification conforms to the safety guidelines described.in the ,
Auxiliary Power Systems section of the UFSAR.
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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277;-50-278 Annual 10 CFR 50.59 Report, Fire Rated Seals in Seismic Gaps Modification No.: 2339 A. System: . Fire Systeks "
. B. Description and Reason for Change:
Some of the unqualified material in the seismic expansion joints of the Radwaste, Turbine and Reactor Buildings was removed and '
3 replaced with a three-hour fire rated material. These gaps were found to be filled with a combustible building joint material and-had to be upgraded to satisfy 10 CFn 50, Appendix R fire barrier requirements.
C. Safety Evaluation Summary:
This modification did not change any equipment important:to '
safety as described in the UFSAR and Fire Protection Program (FPP). Rather, this modification corrected a deficiency and brought the plant configuration into conformance with the UFSAR and FPP. The modification does not affect the operation of any 7s safety related systems, and the modification is consistent with
~( the Fire Protection Technical Specification Bases. ,
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Peach Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 1
_ Annual 10 CFR 50.59 leport L]
Control Panel Anchorage Modification No.: 2376 i
A. System: Structural
_ B. Description and Reason for Change:
Stitch welds and new bolts were added to panel base channels and floor embeds. The existing bolts that connect panel frames to the floor base were tightened. Brackets and supports were installed to prevent non-Q panels from turning over on Q panels.
The purpose of this modification was to ensure that the seismic qualifications of the control panels in the control room, cable spreading room and balance of plant confc m to the configuration specification.
C. Safety Evaluation Summary:
This modification involves no functional or operational changes or additions, but rather, upgrades the features of the previous control panel anchorage design. The structural enhancements r~s provide resistance for peak seismic accelerations at the maximum i-) credible earthquake. This modification does not affect the 1
human factors or arrangements of indications on the control 1 panels. J 1
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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10'CFR 50.59 Report
/~'s V
Change to the Instrument Trips Settings Modification No.: 5045 A. System: Reactor Protection System (RPS)
- B. Description and Reason for Change:
. 1 This modification revises the instantaneous trip settings for' '
the circuit breakers feeding the RPS M/G sets 2AG02, 2BG02, 3AG02, 3BG02. The trip settings are being revised to prevent circuit breakers from tripping instantaneously on motor generator set starting due to increased bus voltage.
C. Safety Evaluation Summary:
This modification. corrects a condition where the RPS M-G sets were tripping instantaneously upon M-G sets starting. This modification will ensure the proper starting of the Reactor Protection System Motor Generator Sets while also providing e protection for the M-G set feeder cables with respect to short circuit rating. Selective coordination with the upstream load -,
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\_,) center breaker is maintained to preclude tripping of the j unfaulted portion of the electrical system.
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Peach Bottom Atomic Powar Station- i Units 2 & 3 Docket Nos. 50-277; 50-278
' Annual 10 CFR 50.59 Report b,,,
Installation of Threaded Caps on Stem Leak-Off Ports on 7 Valves Modification No.:'85-132 i
f A. . System: Residual Heat Removal (RHR) and High Pressure 1 Coolant Injection (HPCI) '
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B. Description and Reason for Change- {
This modification installed threaded fittings and caps on1the i stem leak-off ports of 6 RHR valves and one HPCI valve. The modification involved cutting the existing welded caps off.the
] i stem leak-off ports, threading the pipe with standard threads, and capping with a' threaded cap. The reason for this modification is to' facilitate leak testing of the valve packing of these valves.- Previously, the weld caps had to be cut off then rewelded.each~ time packing leak testing was performed. j C. Safety Evaluation Summary: -
1 The function of the RHR/EPCI. valves is not affected by this.
modification. The valve stem leak-off ports are not part of the
(~g valves' pressure boundary. It-is desired to test these valves' 3 j
\d packings to eliminate packing leakage as a potential acurce of 1 leakage.during performance of the Integrated Leak Rate Test. ]
This modification will merely provide a means to pressurize the ;
packing chamber to detect any packing leaks.
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Prach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278
,. Annual 10 CFR 50.59 Report U
Reactor Feedpump (RFP) Motor Generator Unit (MGU) Cabinet Ventilation Supply Modification No.: 86-051 A. System $ Turbine Building Ventilation B. Description and Reason for Change:
This modification involves tapping into the existing RFP room ventilation supply and installing ventilation ductwork to supply cool clean air to the RFP turbine MGU control cabinets. The purpose of this modification is to prevent failure of the electronic components within the cabinet due to excessive room temperature. As a result, temperatures inside the RFP turbine
! MGU cabinets remain below 50 degress C, as recommended by the l vendor.
C. Safety Evaluation Summary: -
I This modification does not involve safety related equipment.
The function of the MGU control cabinets remains unchanged.
(s This modification will decrease the probability of electronic 1_) equipment failure due to excessive heat in the cabinets, q
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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report O
V New Nuclear Steam Supply (NSS) Software In Plant Process Computer Modification No.: 55-121 A. System: Process Computer
_ B. Description and Reason for Change:
~ A new General Electric software package was loaded into the process computer to. support operation with GE 8X8 multiple lattice fuel. The new coftware uses the J-Factor Method.
C. Safety Evaluation Summary:
The new software installation only affects the NSS software and l
NSS databank of the process computer. No hardware changes were l required. The accuracy of the new software has been confirmed to be comparable to that of the old software and the computational time is substantially the same. The software was installed and tested in accordance with detailed procedures.
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' Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual ~10'CFR 50.59 Report.
Installation of Small Flush Valve and Sample Line at Outlet Header of Condensate Filter / Demineralized Vessel Modification No.: 86-127 A. -System: Condensate Filter / Demineralized i B. Description and Reason for Change: ,
This modification involved the following changes to the-Condensate Filter / Demineralized System: 1) Installation of an elbow to existing 2" tap at each of the 10 demineralizers for
.both units, 2) installation of a reducer toL3/4" after'the elbow,'3) installation of a 3/4" sample valve, 4) and installation of.3/4" block drain valve and' piping to drain header.
The purpose of this modification is to ensure safe, controlled flushing of a condensate filter demineralized.and to reduce'the possibility of resin intrusion into the reactor. .i C. Safety Evaluation Summary:
) .This Filter modification does
/ Demineralized not affect the. function of the' Condensate s system. It provides a means of safely controlling the flushing of the condensate demineralized and sampling. No safety related system'or equipment is affected by this modification.
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P :ch Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report Beat Tracing Replacements
.j Modification No.: 86-132,86-136 and 87-04 4
A. System: Electrical Beat Tracing a
a B. Description and Reason for Change: !
_ Faulty Brisk heat tracing on the Post Accident Sampling (PAS)
System (Unit 2 only) was replaced with Chemlex self limiting ~
heat tracing because Brisk heat tracing is no longer 1 manufactured. The power feed was changed from 240 VAC to 120 q
VAC at a lower wattage. H y
Faulty Electrowrap heat tracing on the outdoor Condensate Storage Tank (CST), Refuel Water Storage Tank and Auxiliary j Boiler fuel oil piping was replaced with Autotrace self limiting heat tracing because Electrowrap is no longer manufactured. Two a surveillance panels were also installed so that this heat tracing can be monitored. d
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C. Safety Evaluation Summary:
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~k-) The heat tracing is not evaluated in the UFSAR nor are the heat I tracing specifications addressed. Replacing the heat tracing with a new type will not change the performance of the systems because the new type serves the same function as the old type and will be more reliable. Because the PAS System is needed after an accident has occurred, the system does not affect the probability of occurence of any type of accident. The new heat tracing increases the reliability of the systems to fulfill y their design functions as described in the FSAR. With the "
surveillance panels, the operators can better monitor the status of the CST heat tracing to assure its availability to provide water to emergency cooling systems. No safety related equipment was affected by these modifications.
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t Y Peach Bottom Atomic Power Station Units 2-& 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 5.0.59 ReporE Venting of RFPT Lube Oil Pressure Regulating Valves 1
Modification No.: 87-013 i
- A. System: Reactor Feed Pump Turbine.
. B. Description and Reason for'Ch'ange:
l The purpose of this modification was to reduce failures in the 1 Reactor Feed Pump Turbine (RFPT)' lube oil pressure regulating) j valves that are due- to valve diaphragm leakage. The .
i modification involved venting of the valve bonnets by drilling a 3/32 inch hole in the bonnets. .TheLvalves regulate oil. pressure i by balancing'the forces of-downstream oil pressure and an- ;
adjustable spring assembly across the valve diaphragm. ~ i i
- C. Safety' Evaluation Summary:
1 The modification did not involve safety related aquipment. The 1
- modification did not change the design of the,RFP?-controls. ;
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1 Peach Bottom Atomic. Power Station Units 2 & 3 i Docket Nos.. 50-277; 50-278 Annual 10 CFR 50.59 Report (9
(/ .g Permanent Drain Lines for Core Spray Stayfull System Modification No.: 87-027 ,
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A. System: Core Spray j u
. B. Description and Reason for Change: j 4
This modification installed permanent stainless steel drain Ji lines from the core spray pressurizing system to a local floor 3 drain. The reason for this modification.was to remove a i tripping hazard that existed due to the use of tygon tubing for drainage purposes. ]
sa C. Safety Evaluation Summary: 'l
.]
i1 This modification does not affect the functL.. of the core spray system or any other safety related equipment. The only consequence of a malfunction of this modification is an j
p additional input to the liquid radwaste system, which has been 3 previously evaluated in the FSAR. 1 h
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Peach Bottom Atomic Power Station !
Units 2 & 3 l Docket Nos. 50-277; 50-278 j Annual 10 CFR 50.59 Report I"')
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q Provide Access Openings in M/G Set and Drywell Exhaust Fans d Modification No.: 87-073 l
1 A. System: Ventilation j
)
- B. Description and Reason for Change: F)
.i
. This modification provided access openings in the M/G set fans I and drywell recirculation fans. This mod will allow vibration readings and bearing inspections. This modification is not safety-related.
j
.i C. Safety Evaluation Smumary:
q To facilitate preventive maintenance (vibration readings, bearing inspection)ta 9"x9", access opening was provided on M/G exhaust fans, and drywell exhaust fans. These fans are not safety-related. There was no operational affect from the addition of access doors. '
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103
Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report Replacement of 125V Station Battery Charger Undervoltage Relay Sockets Modification No.: 87-112 A. System: DC. System B. Description and Reason for Change:
This modification installed the proper rated sockets for undervoltage relays installed under MOD 411. The previous sockets were underrated for their application. This modification required drilling a new hole in the battery charger door, as the new sockets are a different size. It was determined that this would not impact the seismic response of the battery chargers. The component is acceptable for all plant conditions.
C. Safety Evaluation Summary:
This modification involved replacement of sockets with a model compatible with the relay. This will decrease the possibility of an equipment malfunction, and allow for more reliable r~g operation of the system as described in the FSAR. No other id safety related system or equipment is affected by this modification. <
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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report O
High Pressure Coolant Injection (HPCI) Stop Valve Cover Eye Bolts Modification No.: 88-021
'A. System: HPCI B. Description and Reason for Change:
Two 5/8"-11 threaded holes were added to the HPCI turbine steam supply stop valve cover. Located in the flange bolt circle area 180 degrees apart, the holes will be used for eye bolts while rigging the cover off and also for jacking the cover off the flange when necessary. The holes are located so as not to interfere with the pressure chamber adjustment ports.
C. Safety Evaluation Summary:
This modification does not affect the function of any safety related equipment or systems, nor does it affect the pressure boundary. The HPCI stop valve will maintain its ability to operate as designed.
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Peach Bottom. Atomic Power Station Common Docket Nos. . 50-277; 50-278 f Annual 10 CFR 50.59. Report Replacement of Diesel Generator Differential Protection Relays ,
i Modification No.: 1205 A. ' System: Diesel Generators..
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B. Description and Reason for Change: ' . ]
4 Twelve diesel generator differential protection relays in the1C .-,
29A, B, C, and D panels in the' control room were replaced. The. j
. purpose of'the modification.was to replace the seismically 1 unqualified GE.CFD type relays with' seismically' qualified IJD type 1 relays. J C. . Safety Evaluation Summary:
Three relays are' required for each diesel l generator, one per phase. The purpose of these relays is to respond toLthe-differential between generator output current and bus current. ]
The replacement relays perform the same function'as the previous !
relays. Tiie reliability of differential protection is enhanced because the replacement relays are seismically qualified. This O modification does not affect plant loading. i 1
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Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278
('s V
Annual 10 CFR 50.59 Report
-Fire Pump Replacement Modification No.: 5007 A. Syntem: F. ire Systems - Motor-Driven Pump
- 3. Description and Reason for Change: .
The motor driven fire pump was replaced with a different model from the same manufacturer because the pump failed and a replacement of the same model was not available. A pressure i regulating relief valve was also installed on the pump discharge to avoid exceeding the maximum allowable system pressure since the new pump develops more head than the old pump. The old pump was a 3-stage pump and the new pump is a 4-stage pump with similar performance characteristics.
C. Safety Evaluation Summary: -
The replacement pump has performance characteristics similar to that of the old pump. The materials and design are essentially r- the same as the old pump with the exception of minor differences
\_)s which do not effect system functionality. The design and installation of this pump satisfy the applicable requirements of the Peach Bottom Fire Protection Program, UFSAR, NFPA Code 20 (1987 Edition) and NRC Peach Bottom Fire Protection Safety Evaluation Report dated May 23, 1979. The relief valve discharge piping was installed such that it does not compromise the leak tightness or structural integrity of the Emergency Service Water Pump Room in which it is located. The higher running current of :'
the new pump'was also determined not to have any adverse effects.
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Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278
(~} Annual 10 CPR 50.59 Report ;
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Replacement of Air Start Valves on Dmergency Diesel Generators Modification No.: 85-106 I System: Diesel Generators. l A.
B. Description and Reason for Change: j k
This modification replaced Diesel Air Start Valves with a newer )
model and reworked the pipe hangers to maintain seismic i qualification. The purpose of this modification was to replace the valves with the newer model since the old model is no longer made and is not available.
c C. Safety Evaluation Summary:
i q
This modification replaced an obsolete component with a qualified replacement component that performs the same function. The function of the Diesel Generators is not changed by this modification, nor is the function of any other safety related system or equipment. Seismic qualification is maintained.
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l Peach Bottom Atomic Power Station Common i Docket Nos. 50-277; 50-278
} Annual 10 CFR 50.59 Report Chanceout of Stack Gas Radiation Monitoring Pressure Switches 3
Modification No.: 86-079 A. System: Main Stack Radiation Monitoring B. Description and Reason for Change: .
This modification replaced two pressure switches employed to j provide control room alarm in the event of high or low main stack '
sample gas flow. One of the switches also starts an auxiliary '
sample pump in the event of low-low sample flow. The previous j switches were mercury bulb type which in the past caused spurious (
l alarms due to local vibration. The new switches use snap action j switches in place of mercury bulbs. No mechanical or logic "
l changes were needed.
j C. Safety Evaluation Summary: -
This modification is not nuclear safety related. It merely involved removing an old switch and replacing it with a new switch which meets or exceeds specifications of the original switch.
O- This modification increases reliability of the stack radiation monitoring system due to enhanced performance of the sampling system.
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Peach Bottom Atomic Power Station Common Docket-Nos. 50-277; 50-278-q , Annual 10 CFR 50.59 Report 1
Internal' Standpipe Installation Modification No.: 87-012 d
A. System: Radwastel , _ .
~
B. description and Reason for Change: ,
This modification installed internal standpipe drains'in the waste Tj collector tank and the floor drain: collector tank. The previous ]
configuration' allowed sludge buildup to destroy collector drain filters. The installed standpipes allow a' longer operating cycle of the collector tanks and lessen the possibility of destroying ]j the drain filters.
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C. Safety Evaluation Summary: J This modification is not nuclear safety related.- No safety related systems are impacted by this change. No safety functions discussed in the UFSAR were affected.
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Peach Bottom Atonic Power Station' 1 Common .I Docket Nos. 50-277; 50-278 '{
Annual 10 CFR 50.59 Report
)
I Radwaste Flow Roto-meter Replacements Modification No.:'87-26
~
A. System: Radwaste ._
j B. Description and Reason for Change: ,
Two flow roto-meters, FI-0-20-505 and FI-0-20-742, were replaced !
with new meters. These meters are on the Service Air supply line to the Waste Demineralized (FI-505) and Floor Drain Demineralized (FI-742). The previous FI-742 meter failed and it was determined
.that both meters were being continuously used at or near their i rated pressure (150 psig). The new meters are rated'at 600 psig. l C. Safety Evaluation Summary:
l-b l
The new roto-meters serve the same' purpose as the previous roto- #
meters and are less likely to fail. The instruments are not safety related and the modification has no impact on nuclear- '
safety. There is no adverse effect on the plant as deucribed in the UFSAR. !
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g Peach Bottom Atomic Power Stetion Common l Docket Nos. 50-277; 50-278 I
' v. (} - Annual 10 CFR 50.59 Report j
J Permanent Drain Lines for Control Room A/C Cooling Coil Modification No.: 87-083 !
J A. System: Control Room Ventilation B. Description and Reason for Change: . !!
.1 This modification installed stainless steel tubing to drain'the d condensation from the control room A/C supply cooling coil to the nearest floor drain. The installation embedded the stainleus ,:
steel tubing in the floor and eliminated a tripping hazard created i l by the temporary copper and tygon tubing previously ._n use. g C. Safety Evaluation Summary: '
This modification did not change the function or operation of the control room A/C supply. The modification only replaced temporary I copper and tygon tubing with permanent stainless steel drain lines.
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i Peach Bottom Atomic Power Station Common Docket Nos.150-277; 50-278 l
(} Annual 10 CFR 50.59 Report l i
Access Opening Installation t' -~ Assist in Installation of.
Expansion Joint' Material for the Emergency Diesel Generator Ventilation Ductwork l Modification No.: 87-102- ..
i A. System: Diesel Generator Building Ventilation .
~
B. Description and Reason for Change:
This modification installe.d a 25" x 25" access opening in the ductwork for the Emergency Diesel Generator Building Ventilation i ' Supply "A" to gain access needed to secure expansion joint l material to the ductwork. The reason for replacing the expansion joint material was because of excessive work. The expansion joint
.was . replaced and secured to the ductwork by fastening 1" x 1/8" j bar over the material with 1/4" sheet metal screws. '
C. Safety Evaluation Summary:
This modification does not affect the function of any nuclear
' O. safety.related equipment.
changed and its integrity is The function of the ductwork'is not not compromised. Therefore, this i
modification does not increase the possibility of an accident or malfunction the FSAR.
of a different type than any evaluated previously.in
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P ach Bottom Atomic Pow 3r Station i Common Docket Nos. 50-277; 50-278 g Annual 10 CFR 50.59 Report
'k V Beader Pressure Gauge Replacement Modification No.: 87-113 A. System: Emergency Service Water (ESW)
B. Description and Reason for Change-The ESW pump discharge header pressure gauges, PI-0240 A&B, were rreplaced with a new model indicator better suited for the ESW application. The old model had a phospher bronte bourdon tube, 3
and a scale of 0-60 psig. The new gauges have a more corrosion !
resistant stainless steel bourdon tube, and a scala of 0-100 psig. !
The ESW header often experiences pressure greater than 60 psig. {
C. Safety Evaluation Summary:
The performance of ESW is not affected by this modification. !
These pressure indicators have no interlocks or automatic functions and, thus, do not affect system operation. The new (
model is consistent with the equipment specification for these indicators with the exception of on the scale range, which is !
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,, 'r not significant. !
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Peach Bottom Atomic Power Station Common Docket Nos. 50-277;-50-278 Annual 10 CFR 50.59 Report
.O Replacement of Diesel Generator Fuel Oil Crossover Lines Modification No.: 88-010 A. System: Diesel Generators B. Description and Reason for Change:
This modification replaced the diesel generator fuel oil crossover piping with a Fairbanks Morse replacement parts kit. The previous diesel generator fuel oil crossover piping material was copper and subject to cavitational erosion. The replacement piping material was s eel.
C. Safety Evaluation Summary:
1, The diesel generator manufacturer recommended replacement of the diesel generator fuel oil crossover pipe due to pinhole leaks experienced at other commericial sites. The modification did not affect the performance of the diesel generators.
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Peach Bottom Atomic Power Station Common Docket Nos. 50-277 50-278 Annual 10 CPR 50.59 Report Replacement of Diesel Generator Air Coolant Flexible Hoses Modification No.:-88-018 A. System: Diesel Generators B. Description and Reason for Change:
This modification replaced the presently installed diesel generator air coolant flexible hoses.
C. Safety Evaluation Summary:
This modification replaced nonconforming air coolant vent hoses
.with Fairbanks. Morse supplied components. The PECo mechanical engineering group has reviewed and-agreed with replacing the diesel generator air coolant flexible vent hoses with the Fairbanks Morse replacement parts which are considated to be qualified replacements.
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'i Peach Eottom Atomic Powar Station Common j Docket Nos. 50-277; 50-278 !
Annual 10 CFR 50.59 Report O
Addition of 4 Computer Terminals to' Chemistry Lab
~
Modification No.: 88-031
'A.
System: Computerized History And Maintenance Planning System-(CHAMPS) 3 B. Description and Reas.'n for Change:
This modification instrAled 4 computer circuits between the q Chemistry Lab and the Cable Spreading Room / Computer Room.
controller'and the pcwer circuits needed to' operate the' terminals.
The purpose of this modification.is to support the Chemistry Group's input into the CHAMPS system.
C. Safety Evaluation Summary: I This modification' simply provides additional terminals for tha :j existing non-safety related CHAMPS system. The design ensures i that the non-safety related cable raceways are.not loaded beyond design. criteria.
I Circuits are fed from non-safety related, non-vital panels within the capacity of the distribution system. q I
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Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278
, Annual 10 CPR 50.59 Report A
Off-Gas Sample Flow Instrument Replacement and Tubing / Instrument Mounting Upgrades Modification No.: 88-41 A. System: Off-Gas B. Description and Reason for Change:
The flow indicating transmitter (FIT-7341) on the suction line of the Off-Gas Stack Sample Pumps was replaced with a new model from the same manufacturer. The instrument was defective and the same model is no longer manufactured. The corresponding flow element was also replaced with a new compatible model. FIT-7341 is located on Panel 00C101. Tubing for and the mounting.of instruments on this panel were also upgraded with new materials.
l C. Safety Evaluation Summary:
1 The new instruments perform the same function as the old ones and meet the same system requirements, including accuracy. The operation and performance of the system are unchanged. The
(]) mounting and tubing upgrades also have no adverse impact and improve the reliability of the components. No safety-related equipment was affected.
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Patch Bottoin Atomic Powar Station Common Docket'Nos. 50-277; 50-278 .l f Annual 10 CFR 50.59 Report i
Security System Improvements a
Modification No.: 88-61 A. System: Security.
).
B. Description and Reason for Change:
E-field 6 was shortened and E-field 5 was lengthened to permit )
coverage of E-field.5. Cameras 5 and 8 were repositioned to improve coverage. A removable fence was installed along the Reactor Recirculation M-G Set transformer pads and lighting improvements'were also made.
-)
C.. Safety Evaluation Summary:
This modification had no adverse impact on any plant' equipment nor did it affect plant operations. These security 1 improvements i
correct i.screpancies between the security system and.the Security Plan and, thus, improve the physical protection of the plant. {
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Prech Bottom Atoraic Power Station Unit 3 Docket No. 50-278 Annual.10 CFR 50.59 Report n
(_) i Procedure No: ST-9.17-3, Revisions 5, 6, and 7 i
A. Procedure
Title:
Reactor Coolant Leakage Test - Unit 3 l i
B. Description and Reason for Change: I Floor drain unidentified leakage rate limit was reduced to reflect the elimination of a known source of leakage, the Reactor ]
Core Isolation Cooling (RCIC) system MO-15 valve. The valve j packing leak had been repaired. A step to calculate a 24-hoor 1 running average floor drain leakage rate was added and the 24 j hour averaoe was incorporated into the leakage limit criteria. !
C. Safety E5;aluation Summary:
l
.These.cnanges did not affect any of the leakage monitoring steps described in the UFSAR. The change resulting from the RCIC valve repair returns the leaki.ge limit to that specified in the UFSAR, and the running eturage is in addition to the limits outlined in j the UFSAR.
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Peach; Bottom' Atomic Powar. Station l Unit'3 1 Docket.No.L50-278 Annual 10.CPR 50.59 Report 1 ,
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Procedure No.: ST 9.17-3, Revision 8 -
f A. Procedure
Title:
Reactor Coo) ant Leakage' Test Unit 3 B. Description and Reason'for Change:
The leakageLlimits in1this procedure were-restored to the Technical Specification limits.,More. restrictive limits had been applied'during: Cycle 7.due to primary'sytsem pipe crack.
indications'(ref: 11-27-85 letter ~to NEC froa.S.1L. Daltroff,)
PEco). A requirement was"added)to shut down the unit if leakage-reaches specified limits. -Directions were added to initiate-other sampling procedures when'the flow system is out of service or when increased leakage. is' identified. ~
t C. Safety Evaluation Summary i
' These changes do not alterdthe intentiof'the test as described'in
-the FSAR, nor do the_ changes alter:the procedure-for calculating the drywell pump-out rates' described in the.FSAR. 'The additional !
sampling 1 requirements and the shutdown requirement. improve the '
effectiveness-of the test >in responding to unacceptable. leakage.
rates. .The Technical q
Specification leakage. limits are now-
- 5s (- acceptable since pipe' replacement' modifications have.been completed. . Margins of safety are,.thus, as discussed in the-
' Technical Specification Bases.
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1 Peach Bottom Atomic Power Station j Unit 2 '
Docket No. 50-277 Annual 10 CFR 50.59 Report O
Procedure No.: ST 9.17-2, Revision 5 I
A. Procedure
Title:
Reactor Coolant Leakage Test - Unit 2 B. Description and Reason for Change:
A requirement was added to shut down the unit if leakage reaches.specified limits. Directions were added to initiate other sampling procedures when the flow system is out of .
)
service or when increased leakage is identified.
C. Safety Evaluation Summary:
These changes'do not alter the intent of the test as described in the FSAR, nor do the changes alter the procedure for calculating the drywell pump-out rates described in the FSAR. The. additional sampling requirements and the shutdown requirement improve the effectiveness of the test in responding to unacceptable leakage rates. i I
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122 6
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual.10 CFR 50.59 Report 0-Procedure No.: ST 13.8-1, Revisions 6 and 7 A.
Procedure
Title:
Unit 2 Excess Flow Check Valve Operability B. Description and Reason for Change:
Provisions were incorporated to permit partial performance of the test and steps were clarified. Errors in the l
identification of valves and instrument racks were corrected, and a precautionary note about planned and unplanned safety e/ stem actuations was added.
C. Safety Evaluation Summary:
None of the changes affect the leakage rate criteria or test method as discussed in the UFSAR or Technical Specifications.
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1 Pacch Bot' tom Atomic-PowerfStation {l Units 2 & 3 1 Docket'Nos.- 50-277; 50-278 1
-AnnualL10 CFR 50.59 Report L(f~
Procedure No.r HPO/CO-18, Revision 19 A. -Procedure-Title: Processing Liquid. Radioactive Waste:
B. Description and Reason for' Change:
iThe activity-limit:for release =from a laundry drain' tank was p increased.-.Also, Administrative changes were made to1 incorporate the revised requirements 5of' Technical Specification 1
3.8.B.4 for the discharge of laundry drain tank contents,fto i
. incorporate.BWR. Chemistry Control Program limits, and to make '
otherLinstructional improvements.' ..
C. Safety Evaluation' Summary:-
The 'echnical' Specification lrelated change.merely. reflects a
" license change found. acceptable by,theJNRC,,and the other-
' administrative changes'are of no safety? significance.:-
'The FSAR lists a maximum: expected concentration'of 0.00001' uCi/cc in a laundry drain-tank. However,Lrelease of 0.0001 uCi/cc concentration liquid from a? laundry tank with the.
dilution. flow:of one Circulating ~ Water Pump-was: evaluated and-
,. (]).
., ~ determined to'r.onstitute only a fraction of'the maximum permissible concentration-(MPC) of"all isotopes.
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Parch Bottom Atomic Powar Station i
Units 2 4 3--
-l Docket: Nos. 50-277; 50-278 :
Annual 10 CFR 50.59 Report 'I
- O' Procedure No: - GP-2, R9,tsion'48 !
A. Procedure
Title:
NormalLPlant Startup B. Description-and Reason for Change:
i
.The s'teps of GP-2A, " Reactor Startup and'Beatup" and'GP-1, " Pre-- i Startup Checkoff Lists";were incorporated into GP-2 so that tl'etc is only one procedure for plant startup. Also, operator. i training' comments, and new system and new Technical Specification requirements were incorporated.
C, Safety EvaluetloT Summary:
1 This procedure revision did not alter startup using the Rod l Sequence ~ Control System,'wlsich is a-procedure described in the.
FSAR. - - Primarily,; this .we.s an. administrative : revision. - No l]
margins of safety were reduced since the necessary requirements for a safe, orderly startup are still' procedurally imposed.
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Peach Botton Atomic Powar Station' Unit 3 Docket No. 50-278 Annual 10 CFri 50.59 Report O
De-energize E-33 Emergency Bus A.
Reference:
'Special Procedure (SP) 1109 B. Description and Reason for Change:
The E-33 4kV emergency bus was de-energized to facilitate removal of a relay'needed for the replacement of the corresponding relay on th: E-22 bus which had failed. Certain E-33 loads, such as
'A' Iteactor. Protection. System bus and 'C' 125 VDC~ Battery Charger were powered from alternate sources. The E-33' bus was re-energized.after a replacement relay was.obtained and installed.
C. ff fety Evaluation Summary:
This special condition'was_ acceptable because Unit 3 was.defueled and temporary. power.was.provided to essential ~ loads. Equipment necessary:to available.
fulfill Technical Specification-requirements were The E-33 bus powers Unit 3 loads'and Unit 2 was not adversely supply ~theaffected essential by'using Unit 2 emergency buses to temporarily E-33 loads.
condition. Unit 2 was in the cold ~
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Peach' Bottom Atomic- Power? Station .l Unit 3 {
Docket No. 50-278 j
, Annual JOs CFR 50.59 Report . :j
)
Gpecial Use of'a Temporary Equipment Pool Water Cleanup l System with Auxiliary Pumps :
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Reference:
- Special~ Procedure (SP) 1146 j B. j Description and Reason for Change:
}
Because the Reactor Water Clean-up: System had to~be removed!from service during;the'. major Unit 3 pipe replacement project, an' auxiliary system for maintaining. acceptable water clarity intthe' reactor-cavity and equipment pool was established to-faciliate underwater work' .
The Fuel-Pool Cooling and Clean-up System was used to clean the water. Water was. transferred by a submersible pump'fromtthel equipment pool'to the Fuel Pool Skimmer Surge. Tank, land water"was
, returned..to the equipment pool by.another submersible pump.
' Protective controls were included in'this' auxiliary system to. J prevent any uncontrolled. transfer.of' water..
C. jafety Evaluation S2mmary: 1 1
}
This auxiliary system did not affect any safety-related equipment J
^ and protecd.ve measures precluded-inadvertent' drainage of the Fuel-Pool.
sources. The 'iLO pumps were powered from;non-safety relatedl system-underwent'an in-service hydrostatic' pressure test, and protective devices would have. tripped the pumps if~ flow became imbalanced. Postulated leakage could have. adequately'been handled by the Reactor Building Floor Drain Sump, as evaluated in the FSAR.
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i Peach Bottom Atomic Power Station .
Common ;
Docket Nos. 50-277;:50-278 1 Annual 10 CFR 50.59 Report l Running An Emergency Diesel Generator Without Hormal Ventilation A..
Reference:
Special Procedure ~-(SP) 1176 l
B. Description and Reason for Change:
One diesel generator at a time was operated for approximately 30 minutes in accordance with normal operating procedures, but with !
the room supply'and supplemental supply fans off and'the dampers in the full recirculation mode. This test vas performed to collect data necessary to resolve uncertainties surrounding a proposed modification that would delay the auto-start of.the fans.
C. Safety Evaluation Summary:
This test!was, performed at a time when only-two of the four-diesel generators were required to be:available, and adequate _ precautions l were taken such that the_ ventilation system could-have been placed- '1 in service promptly if degraded diesel engine performance .!
occurred. _The other three diesel generator rooms were not affected by conduct of the test. With respect to mitigating the consequences of an acci6ent,-the nominal risk associated with this
(() test (lack of' diesel generator independence from off-site ~ power I during the test,-due to electrical line up parallel to grid) is similar'to that of the routine diesel generator operability test.
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l Pacch Bottom Atomic Powsr Station
'Unita 2 & 3 Docket Nos. 50-277; 50-278
. Annual 10 CFR 50.59 Report
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' . Mud Removal From Pump Bays - Postulated Release of Low i i
Level Activity Mud to Discharge Canal l
A. .
Reference:
Special Procedure'(SP) 1203 j D. Description and Reason for Change:
Mud was removed from the bottom of the Pump Structure bays and '
transferred to tanker trucks. ' Appropriate sampling of the mud was performed prior to removal.. This is done periodically to properly maintain the cooling systems.
C. -Safety Evaluation Summary:
]
The postulated release of low level activity mud-to the discharge canal was. evaluated and determined to'be negligible relative to releases previously evaluated in the FSAR. This l procedure'did not adversely affect any safety related activity or reduce any margins of safety. >
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Peach Bottom Atomic Power Station Unit 12 Docket No. 50-277 Annual 10 CFR 50.59 Report O
Refueling Bridge Interlock Removal to Facilitate In-Vessel Work A.
Reference:
Special' Procedure (SP) 5000 l l
B. Description and Reason for Change: j The-logic requirement that the Mode Switch be in REFUEL to y move the refueling bridge over the core was temporarily bypassed. 1 This was done to permit use of the bridge for reactor vessel manway inspection / repair while the Mode Switch:
was in SHUTDOWN.
C. Safety Evaluation Summary:
This change did not decrease the protection from inadvertent criticality described in the Updated FSAR. No core alterations,,as defined in the Technical Specifications, were performed while this interlock was bypassed. A continuous rod block was in place. Thus, no margins of safety were reduced.
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130' l
Peach Bottom Atomic Powar Station I Units ~2 & 3 Docket Nos. 50-277; 50-278
' Annual 10 CFR 50.59 Report' O
Temporarily Crosstying Startup Sources A.
Reference:
Special Procedure (SP) 5001 B. Description and Reason for Change:
The No. 3 Startup Switchgear was momentarily crosstied to the )
.No. 2 Startup Source through the 2SU-A, A-1 and 3SU-l breakers.
This was done to maintain' power to the 4kV emergency switchgear'
-while the 3435 breaker (No.'343 S/U Source Supply) was.
temporarily removed from service due to transmission system problems. The 2SU-E breaker was out of service and, thus, the j
l No. 2 startup source could not' supply the emergency switchgear "
in the normal line-up. ,
I C. Safety Evaluation Summary: !
The duration of the crosstie was very brief and the units were ;
l in the shutdown or refuel mode. Unit 3 was defueled. The ;
objective of having two independent power _ sources is to ensure '
that power will be available to safely shutdown the units. This special operation ensured that power was available to support the safeguards loads.
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Parch' Bottom Atomic Powcr Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report Setpoint Change on High Pressure fervice Water (HPSW)
Pressure Switch (Change #88-32 ,
A.
Reference:
PS-2380 A&B and PS-3380 A&B B. Description and Reason for Change:
The setpoint on the HPSW pumps header pressure switches PS-2(3)380 A&D was lowered (250 to 235 psig).because it was causing-a low pressure alarm when the HPSW system was at normal pressure.
C. Safety Evaluation Summary:
This switch only provides a low pressure alarm in the control room and does not have any control' function. Based on the manufacturers. pump. curve the setpoint was too high, thus'an-unwarranted alarm could. mask an actual low pressure condition.
This setpoint change does not affect.the system as described in the FSAR. These switches have no Technical Specification l related function.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report O
Setpoint Change on Offgas Stack Flow Switches (Change #88-8-6)
A.
Reference:
Flow Switches FS-0470 A&B B. Description and Reason for Change:
The setpoint on FS-0470 A&B was increased slightly to ensure that the Technical Specifications required minimum dilution flow in the Offgas Main Stack would be maintained.
C. Safety Evaluation Summary:
This change ensures that the Offgas Main Stack low flow alarm annunciated at a flow greater than or equal to the minimum dilution flow required by the Technical Specifications. The previous setpoint was non-conservative given the tolerance of
- r. the instrument. These flow switches only have an alarm function and, therefore,'the setpoint change has no impact on system operation.
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Peach Bottom Atomic Power Station !
Unit 2 Docket No. 50-277 Annual 10 CFR 50.59 Report l
)
Setpoint Change on Reactor Protection System (RPS) M-G Set Breakers '
A.
Reference:
Breakers 52-3785 and 52-3803 B. Description and Reason for Change:
The instantaneous trip setpoint on the RPS M-G set feeder breakers (52-3785 and 52-3803) in motor control centers 20B37 and 20B38 was increased from 490 amps to 885 amps to improve i the protection capability of the breaker.
l C. Safety Evaluation Summary: l
{
This change does not constitute a change to the system as described in the UFS AR. The plant is designed for RPS M-G set failure. . The nasw setpoint was evaluated and verified to
. provide.better feed source and load protection.
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134
Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual 10 CFR 50.59 Report Setpoint Changes on Control Room Ventilation Flow Switch (Change #88-63-1) l A.
Reference:
FS-0760A and B B. Description and Reason for Change:
The low trip setpoint on the flow switch at the suction of each Control Room Vent Radiation Monitor Sample Pump (OAP188 and OBP188) was changed from 5" Hg vac. to 1" Hg vac. The 5" setting created a low flow signal when flow was not low. Also, a 10" Hg vac.
' trip setting was established on these switches to initiate an alarm indicating that the upstream iodine and particulate filters may be clogged.
C. Safety Evaluation Summary:
These changes corrected a configuration that was not consistent with the UFSAR system description. With these changes the system will fulfilloperate more its safety closely to that of its original design and will objective.
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i Patch Bottom Atomic Power Station Unit 2 j Docket No. 50-277 )
. Annual 10,CFR 50.59 Report r^v ,
I
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Setpoint Change on Standby Gas Flow Switch A.
Reference:
Flow Switch (FS) 20008 B. Description and_ Reason-for Change:
The setpoint on FS-20008 was lowered from 8000 cfm to 2000 cfm. This switch is supposed to trip _when it senses that' ;
Standby Gas Treatment System flow erists, permitting a partial Group III primary containment isolation signal on offgas stack hi-hi radiation. The 8000 cfm' setting, original setting when the switch was installed during a recent i i
modification, was too high. 1 C. Safety Evaluation Summary:
1 The new setpoint will. ensure that flow is sensed, whereas the 1 1
original setpoint was too high to. trip the switch with only one fan in operation. Thus, the new setpoint permits the switch to perform its specified function.
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