ML20196K031
| ML20196K031 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/31/1987 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20196J978 | List: |
| References | |
| NUDOCS 8807060498 | |
| Download: ML20196K031 (167) | |
Text
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PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 i
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DOCKET NOS. 50-277;50-278
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1987 ANNUAL PLANT MODIFICATION REPORT PURSUANT TO 10 CFR 50.59
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8807060490 800630 ADOCK05000{7 DR
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Docket Nos. 50-277 50-278 L
1987 c
L PEACH BOTTOM ATOMIC POWER STATION ANNUAL PLANT MODIFICATION 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, License
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Numbers DPR-44 and DPR-56, respectively.
This report includes L
primarily modifications that were completed in 1987, including changes made to the facility and procedures as described in the safety analysis report.
A summary of the safety evaluation for each change,
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concluding that an unreviewed safety question (as defined in 10 CFR 50.59 (a)(2)) was not involved, is included.
There are several modifications described in this report which were completed on one of the units in 1987, and the same modification was p
L completed on the other unit in 1988.
Modifications in this category were reported for both units in this report to avoid duplication in next year's report.
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L ANNUAL PLANT MODIFICATZON REPORT PEACH BOTTOM ATOMIC POWER STATION
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1987 TABLE OF CONTENTS
{ Modification System Page 0421 Core Spray 76
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0494 Primary Containment 2
0556 Reactor Building Cooling Water 100 0625 Instrument Air and Instrument Nitrogen 77 0682 Dryer / Separator 101 0800 Cranes, Elevators, Rigging Equipment, Tools 102
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0800D Cranes, Elevators, Rigging Equipment, Tools 3
0842A Structural 4
0843 Ventilation 103 Standby Liquid Control 5
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0867 0945 Recirculation and Residual Heat Removal 105
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0927 Reactor Core Isolation Cooling and 7
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0959A DC System 107 1029 J and K 480V Motor Control Centers 108 1110 Fire Barriers 110 lll5A Miscellaneous 111 1180 Containment Atmospheric Dilution 9
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1181 Cranes, Hoists, Elevators, Rigging Equipment 79 1236 Turbine Generator 81 1243 High Pressure Service Water 83 1265 Critical Equipment Monitoring System 137 y
1' 1268 HPCI Governor Control System 112 1273 Main Steam 11 1351D Emergency Service Water 13
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1352C Residual Heat Removal 15 1352D High Pressure Service Water 17 I
L 1352E 4 KV Circuit 114 l
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AnnuS1 Plant Modification RSport P:r_ch Bottom Atomic Powar Station
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1987 Table of Contents (Continued)
Modification System Page 1352F DC System 19 1352G Automatic Depressurization System 21 1352H Miscellaneous Instrument Systems 23 1371 Cooling Towers 138 1404 Reactor Protection System and Turbine Generator 25 1419 Reactor Protection System and Residual 116 Heat Removal System 1449 Core Spray 26 1457 Plant Protection Instrumentation 118 1544 Structural (General) 120 1548 Main Steam 28 1549C Hydrogen Water Chemistry 140 1583 DC Power Supply 30 1584 480 Volt Motor Control Centers 32 1599 Radiation Monitoring 142 m
1L 1609 Security System 85 Control Rod Drive 87
{ 1620 1626 Main Turbine 34 r
l 1647 Transversing Incore Probe 35 1649 Miscellaneous 143 1677 High Pressure Service Water 37
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1695 Feedwater 38 1716 Primary Containment 39
[ 1744 Main Steam 41 1768 Hydraulic Snubbers 89 F
1834 Main Generator 43 iii
Annual Plant Modification Report Peach Bottom Atomic Power Station 1987 Table of Contents (Continued)
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Modification System Page N
1935 Containment Atmospheric Control 44 F
1939 Process Computer 122 1958 Miscellaneous Instruments 46 2015 Emergency Service Water 47 2029 Radwaste 144 2082 High Pressure Coolant Injection 49 f
75-048 480V Motor Control Centers 146 77-036 Transversing Incore Probe 51 83-145 Reactor Building Closed Cooling Water Pumps 124 83-152 Process Computer 148 83-159 Ventilation 90 l
84-006 HPCI Turbine 126 H'84-018 Control Rod Drive 92 e
k84-033 Incident Detection Instrumentation 52 s85-007 Reactor Recirculation 53 85-048 Reactor Recirculation 55 85-066 Residual Heat Removal 56 85-071 Control Rod Drive 58 85-121 Control Room Ventilation 150 85-127 Computer 59 85-138 Turbine Lube Oil 61 85-146 Residual Heat Removal 62 85-149 Residual Heat Removal 64
(86-006 Electrohydraulic Control 128 86-014 Reactor Feedwater 65 86-027 Control Rod Drive 94 iv l
Annual Plant Modification Report Peach Bchtom Atomic Power Station 1987-Table of Contents (Continued) h Modification.
System
_Page h
..-86-030 Recirculation 66 86-031 Cardox Fire Protection 151 t-86-047 Service Water 129 86-052' Instrument Air 152
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86-054 Structural 68 84-056 Feedwater 69 86-070 Condensate Return 70 86-073' Auxiliary Steam 154 86-089 Main Steam 72 86-095 Recirculation Pump and Valves 130 86-103 Process Computer 156 86-117 Reactor Vessel and Internals 132 86-139 Rod Sequence Control 96
)86-142 Diesel Generator 157 86-148 Primary Containment 73 C
87-006 Process Computer 74 87-020 Instrument Nitrogen 134 87-022 Process Computer 98 Procedure Title Page
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GP-3 Normal Plant Shutdown 160 S.2.3.1.A Startup of a Recirculation Pump 161 y
l UNIT 2
5
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Revision to Range of Drywell Temperature Indication Modification No.:
0494 A.
System:
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Description:==
The range of drywell temperature indication was revised to 40
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degrees F - 440 degrees F and TR-4805(5805) was replaced.
C.
Reason for Change:
This modification eliminated certain discrepancies which existed between the range of instrumentation as required by t
Technical Specifications and the actual ranges installed in the plant.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, this modification merely increases the range of drywell temperature indication so that temperature excursions which could occur during accident conditions are observable by the operator.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any
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evaluated previously in the safety analysis report?
r Answer.
No, it is not possible for an instrument range change used for indication purposes only to cause a different type of tecident or malfunction.
L iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, this modification eliminates a discrepancy between the Technical Specifications and installed equipment, thereby ensuring the defined margin of safety.
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Peach Bottom Atomic Powsr Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Dryer / Separator Lifting Device l
Modification No.:
08000 A.
System:,
Cranes, Hoists, Elev, Rigging Equipment, Tools B.
Description This modifics..... involved replacing the (2 3/16") socket
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pins with 2" socket pins.
C.
Reason for Change:
.The lifting beam had been modified to comply with the heavy loads single failure criteria of NUREG-0612, but was left misaligned.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of l
occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The modification does not impact nuclear safety related equipment.
The 2" socket pins will continue to meet the criteria of NUREG 0612.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The 2" socket pins meet the same design requirements of the 2 3/16" pins.
- Thus, their reduced size will not introduce new accident precursors or failure mechanisms, iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
The smaller socket pins meet the same design requirements.
Neither their use nor failure will impact the operation of equipment addressed by the Technical Specifications.
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-Peach Bottom Atomic Power Station
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Unit 2 Docket No. 50-277-Annual Plant Modification Report Torus A,ttached Piping, Supports and Restraints l[c 3.
tion No.:
0842A Structural
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ption:
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.a attached piping, piping supports and piping restraints were added, relocated, or deleted.
Also, angle struts or yoke stiffenern were added to several valves, and yoked bolts j
were replaced on several valves.
C.
Reason for Change:
These changes were made to enabla torus attached piping and valve operatocs to withstand predicted hydrodynamic loads due to safety / relief valve discharges and a loss-of-coolant accident.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipmeat important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The changes were designed to increase the reliabilicy of the torus attached piping / supports and valves without changing the operation of the equipment as addressed in the original safety evaluation.
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ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluatec previously in the safety analysis report?
Answer:
No.
The design of these changes made the stresses in the piping, the support, and structural loads within the final safety analysis report requirements, and accelerations on valves were made within qualified levels.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
Only the location of equipment and the J
support / restraint of equipment was changed to increase its reliability without changing the safety function.
Also, appropriate surveillance requirements were added to the Technical Specifications for newly installed l
components.
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Peach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Enriched Boron Solution to Standby Liquid Control System Modification No.:
0867 A.
System:
Standby Liquid Control-(SLCS)
B.
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Description:==
By using a new control solution enriched in Boron-10, the SLCS was modified to provide a minimum flow capacity and a Poron content equivalent in control capacity to 86 GPM of 13 weight percent sodium pentaborate.
As a result, the SLCS can now perform its function of shutting down the reactor approximately twice as quickly.
C.
Reason for Change:
E The purpose of this modification was to comply with 10 CFR 50.62.
D.
' Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, the use of a new standby liquid control solution does not introduce any new accident precursors because it is only used in the event of a transient without scram.
The solution change has no impact on other safety related equipment or the primary system boundary.
The use of the SLCS is not specifically required for any transient or accident evaluated in Section 14 of the t
Further, the enriched boron solution will enhance the effectiveness of the SLCS to shutdown the reactor, 11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, this modification is required to meet 10 CFR 50.62 and has already been subject to accident and safety analyses associated with the proposed rulemaking procedures.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, this modification will e: hance the margin of safety.
The enriched Boron concentration will allow the SLCS to shutdown the reactor 5
Peach Bottom Atomic Power Station Unit 2 j
l Docket No. 50-277 Annual Plant Modification Report faster.
The solution tank heater and heat tracing on the pump and suction piping will operate automatically to maintain solution temperature well above the saturation temperature and prevent Boron from
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precipitnting out of solution.
Answer:
No, calculations indicate that the charcoal adsorption capabilities of the new offgas
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systrm provide a system release rate and offsite whole body dose rate lower than those of the existing system.
The margin of safety
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is improved by this modification.
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Peach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Ramoval of Fuses from Neutral Feeds
- Modification No.:
0957
'A.
System:
Re' actor Core Isolation Cooling (RCIC) and Residual Heat Removal (RHR)
-B.
Description:
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The 10 amp-fuses on the neutral feeds of the control circuitry for the RCIC steam supply isolation valve, MO-2 15, and the RHR shutdown cooling suction isolation valve, MO-(-
2-10-18, were removed.
These fuses were in the Remote Shutdown Panel control circuitry and were replaced with solid
, links.
C.
Reason for Change:
These fuses were removed to reduce the potential for a failure of the circuitry due to a blown fuse.
It is not a common practice to install a fuse on the neutral feed of ac circuits.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in t.ne safety analysis report?
L Answer:
No.
The elimination of the fuses will not I
adversely affect the operation of the valves.
In fact, the absence of the fuses should r
decrease the probability of a malfunction because they are not necessary and could blow during a transient which otherwise would not disable the circuit.
il)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
J Answer:
No.
The fuses were not necessary and their absence does not create any new potential failure modes because the circuits are I
protected by other fuses.
Also, the design of this change was in accordance with all criteria applicable to the original design, as well as 10CFR50, Appendix R, fire protection considerations.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report i
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Does this modification reduce the margin of safety as
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defined in the basis for the Technical Specifications?
Answer:
No.
Removal of the fuses does not affect the operational basis of the circuits or che reliability of the Remote Shutdown Panel.
No Technical Specifications were affected.
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Peach Jottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report CAD Oxygen Analyzer Reliability Improvement
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Modification No.:
1180 A.
' System:
Containment Atmospheric Dilution B.
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Description:==
This is a modification to the CAD Oxygen Analyzers.
The water reservoirs will be replaced with reservoirs of a larger capacity.
The analyzer compartment front cover will be equipped with a locking mechanism to prevent unauthorized tampering with the Analyzers.
This modification involves installing a 3 month supply water reservoir on the CAD Oxygen Analyzers and installing a locking mechanism on the front cover of the Analyzer.
C.
Reason for Change:
The purpose of the modification is to improve the security and reliability of these Analyzers and to ensure a 62 day supply of water to the Analyzers in the event of a LOCA.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluatet in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence or the consequences ofan accident or malfunction of equipment important to safety as previously evaluated because the modification will increase the size of the reservoir and not adversely L
affect the operation of the system.
Therefore, no accident as previously analyzed in Chapter 14 will be affected.
4 11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
J Answer:
This modification does not cre
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the possibility for an accident or malfunction of a different type than any evaluated previously because the modification will not
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affect the operation of the system except to improve the security and reliability of the Analyzers, iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
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Peach Bottom Atomic Power Station Unit 2
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Docket No. 50-277 (E
Annual Plant Modification Report' Answer:
The margin of safety as defined in the bases N
of the Technical Specifications is not reduced by this modification because the operation of the system will not be affected by this modification except to improve.the
{E security and reliability of the Analyzers.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modificatior. Report Removal of Diodes Across DC Solenoids of MSIVs Modification No.:
1273 A.
System:
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B.
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Description:==
The diodes across the DC solenoids of the Main Pteam Isolation Valves (MSIVs) were removed.
C.
Reason for Chance:
The purpose of the modification was to minimize the unnecessary closure of the MSIVs and the loss of control room indication of the MSIVs due to a diode failure.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No this modification does not affect the operation of the MSIVs.
The function of the diodes was to protect the relay contacts in the circuit from transient overvoltages.
Testing of a MSIV DC solenoid without a diode indicated that the worst transient voltage was within the maximum dielectric strength of the solenoid circuit.
Also, Unit 3 does not have installed diodes, and there are no known circuit failures.
Review of the MSIV vendor instruction manual revealed that diodes are required only for 250 volts DC solenoid.
The DC solenoids of the MSIVs operate on 125 volts DC.
Because the operation of the MSIVs is not affected, neither are the FSAR accident analyses.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, in addition to the operation of the MSIVs being unchanged, the elimination of the diodes from the control circuits did not result in a significant reduction in the bus loading.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
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Peach Bottom Atomic Power Station Unit.2
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Docket No. 50-277 Annual Plant Modification Report t
Answer:
No, because although the MSIVs are addressed F
in'the Technical Specification,-there were no L'
functional changes as a result of the' modification.
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Peach Botton Atomic Power Station Unit 2 i
Docket No. 50-277 Annual Plant Modification Report Emergency Service Water Alternative Control Station Modification No.:
1351D A.
System:
Emergency Service Water (ESW)
B.
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Description:==
An Alternative Control Station (ACS) was established for the OAP57 ESW pump.
The ACS helps assure a supply of cooling water to the diesel generators during a fire in the main control room, cable spreading room or the remote shutdown panel area.
Circuit changes were performed for the OAP57 and OBP57 ESW pumps to prevent spurious operation of the pumps during fires in other areas.
C.
Reason for Change:
This modification is necessary to conform to 10CFR50, Appendix R.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of p
occurrence or the consequences of an accident or L
malfunction of equipment impcrtant to safety as previously evaluated in the safecy analysis report?
/
f Answer:
No.
The safety objective of the ESW system; to provide a reliable supply of cooling water to diesel generators and selected equipment coolers during a loss of offsite power; is enhanced.
Defeating the automatic initiation of the ESW pump for a fire does not increase the probability of or consequences of an accident, because the time available to manually operate the pump is sufficient to establish cooling water for safe shutdown.
j ii)
Does this modification create the possibility for an
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accident or malfunction of a different type than any evaluated previously in the safety analysis report?
J Answer:
No.
The ESW pump operations described in the Updated Final Safety Analysis Report, when transfer / isolation switches are in the "Normal" or "Test" modes, are not changed.
The "Emergency" and "Test" switch positions are individually alarmed in the main control room.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report iii)
Does this modification reduce the margin of safety as
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defined in the basis for the Technical Specifications?
Answer:
No.
The safety function of~the ESW system is f-not affected by adding safety-related transfer / isolation switches.
All automatic start and trip features of the ESW pumps are maintained when the transfer / isolation switches are in the "Normal" and "Test" positions.
This modification ensures control of the "A" ESW pump for a fire in the main control room, cable spreading r.com or the emergency shutdown panel areas for a design basis fire.
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Peach-Bottom Atomic Power' Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Establish a Residual Heat Removal (RHR) Alternate
. Control Station (ACS)
Modification No.:
1352C
.A.
System:
Description,:
This modification provides alternative control capabilities for the B loop of the Unit 2 RHR System.
One Alternate Control Station (ACS) is established for the Unit 2B safeguard channel RHR pump at the 20A1602, 4KV emergency switchgear cubicle.
A second Unit 2 RHR ACS is established at the Unit 2 HPCI ACS panel for the B loop valves needed to assure torus cooling capabilities.
C.
Reason for Change:
The purpose of the modification is to bring PBAPS into compliance with requirements of Appendix R to 10CFR50.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased because the new transfer / isolation switches as well as the new cable routings to I
the switches that are part of the normal operating circuits and systems will be L
constructed in accordance with applicable safety-related equipment criteria.
The safety objective of the RHR System is
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maintained when the transfer / isolation switches are in the normal and test positions.
The switches are assured of remaining in the proper position because the access to them is administratively J
controlled.
Although placing the transfer / isolation switches in the emergency mode may bypass certain RHR System functions, t
this condition will only occur during the mitigation of a fire, i.e. when a concurrent
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design basis accident or severe natural phenomenon is not postulated.
No Chapter 14 safety analyses are affected by this change.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report 11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification does not change the function of the RHR System as described in the UFSAR when the transfer / isolation switches are in the Normal or Test modes.
The transfer / isolation switches are administratively controlled to ensure the switches remain in their power positions.
Although placing the transfer / isolation switches in the emergency mode may bypass certain RHR System functions, this condition will only occur during the mitigation of a
- fire, i.e. when a concurrent design basis accident or severe natural phenomenon is not postulated; therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created.
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iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
r Answer:
The margin of safety as defined in the basis L
for any Technical Specification is not reduced.
This modification provides a means of RHR System control for a fire in the Main g
l Control Room, the Cable Spreading Room or the Emergency Shutdown Panel Area and does not change the normal operation of the RHR System as described in the Technical Specifications.
1 10CFR50.54(x) allows a departure from the Technical Specifications in an emergency such as an Appendix R fire.
In this case, placing L
the transfer / isolation switches in the Emergency position are acceptable because this mode will only be used for a fire in Fire Area 25.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Establish a High Pressure Service Water (HPSW) System Alternate Control Station (ACS)
Modification No.:
1352D A.
System:
High Pressure Service Water B.
==
Description:==
This modification installs an alternate control station (ACS) for the 4 KV emergency circuit breaker for the Unit 2, B safeguard channel, High Pressure Service Water pump 2BP42.
A
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three position transfer / isolation switch is installed to isolate circuits subject to fire damage and to transfer breaker control from the control room to the HPSW ACS.
This modification involves installing the ACS at the 20A1607, 4 KV emergency switchgear cubicle, b
C.
Reason for Change:
The purpose of this modification is to bring PBAPS into
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compliance with the requirements of Appendix R to-10CFR50.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased.
The safety objective of the HPSW System is to provide a reliable source of cooling water for the RHR under post accident conditions.
The HPSW components
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which will have alternative control capabilities by virtue of this modification, do not automatically start or reposition for any design basis accident.
Sufficient time L
is available to start HPSW to support RHR for an alternative shutdown scenario.
Therefore, bypassing the Main Plant Control Room, the i
Cable Shutdown Spreading Room and the Emergency Shutdown Panel areas to take manual control of HPSW for alternative shutdown does not pose an unreviewed safety question nor does it create an accident different than previously evaluated in Chapter 14.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
The probability for an accident or a malfunction of a different type than any evaluated previously in the safety analysis report is not created since the safety function of HPSW is not affected by rerouting circuits or installing safety-related transfer / isolation switches.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
The margin of safety as defined in the basis for any Technical Specification is not reduced. 'Providing a means of HPSW System control for a fire in the Main Control Room, the Cable Spreading Room, or the Emergency Shutdown Panel Area does not change the normal operation of the system as described in the Technical Specifications.
The use of this panel in the alternative shutdown mode is not described in the Technical Specifications.
The use of this panel to respond to an Appendix R fire is required to ensure safe shutdown.
10CFR50.54(x) allows departure from the Technical Specifications for an emerger.cy such as an Appendix R fire.
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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report
- Installation of 125V Distribution Panels and Feeds
. Modification No.:
1352F l
A.
System:
DC System B.
==
Description:==
This modification installs two new safety-related, 125V D.C.
power distribution panels (2BD306 and 3DD306), associated fuse boxes, and power circuits to supply D.C.
control power to alternative shutdown loads for systems and components
(
modified by modifications 1351A and D,
1352A thru H, and modifications 1353A thru H.
C.
Reason for Change:
The purpose of the modification is to bring the Peach Bottom Atomic Power Station into compliance with the requirements of Appendix R to 10CFR50.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as
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previously evaluated in the safety analysis report?
Answer:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased by this modification.
This modification colocates D.C. power distribution for safety-related loads.
Where new loads are added, the capacity of the
(.
station batteries is sufficient to support the LOCA duty cycle with new load added.
ii)
Does this modification create the possibility for an
[
accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
The possibility for an accident or malfunction of a different tyr. than any evaluated previously in the rafety analysis report is not created by th!a modification.
The loads that the alternatn e shutdown modifications add to the e.aergency batteries do not prevent them from performing their safety-related function.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
19
Peach Bottom Atomic Power Station Unit'2
-Docket No. 50-277 Annual Plant Modification Report
- l Answer: ~ The margin of safety as defined in the-basis for the Technical Specifications is not reduced by this modification.
This modification providea -D.C.
power for alternative shutdown while' maintaining the
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. margin of safety specified.for the emergency batteries.
.The Tect:nical Specifications are not affected by this modification.
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Peach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Automatic Depressurization System (ADS) Alternate Shutdown Station Modification No.:
1352G A.
System:
Automatic Depressurization System B.
==
Description:==
This modification provides alternative shutdown capability for three Automatic Depressurization System (ADS) valves and control isolation capability for all other safety relief f
valves (SRV's) to meet the requirements of Appendix R to 10CFR50.
C.
Reason for Change:
The purpose of the modification is to bring PBAPS into compliance with requirements of Appendix R to 10CFR50.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is i
not increased.
This modification does not change the operation of the SRV's or the nitrogen supply valves when the isolation and transfer / isolation switches are in the Normal or Test positions.
Since, for an Appendix R fire, it is not necessary to postulate a LOCA or seismic event coincident with a fire and a concurrent loss of offsite power, defeating the control or all SRV's except A, B and K and transferring control of these and the nitrogen supply valves to the Alternate Control Station (ACS) does not pose an unreviewed safety question.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
r Answer:
The possibility for an accident or a
[
malfunction of a different type than any evaluated previously in the safety analysis report is not created.
The safety functions
[
of the ADS and PCIS systems are not affected by rerouting cables to safety-related 21
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report transfer / isolation switches to provide alternative shutdown control.
Isolation of
[.
SRV cables from their normal controls is done only for a fire in fire area 25, in which case control of the SRV's could not be
(
guaranteed.
111).
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
{
Answer:
The margin of safety as defined in the basis for any Technical Specification is not h
reduced.
Providing a means of SRV and nitrogen supply valve control necessary for safe shutdown method D (Alternative Shutdown) does not affect the margin of safety for the
[~
10CFR50.54(x) allows departure from the Technical Specifications for emergencies such as an Appendix R fire.
E ra P'
L r
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L E
L I
L 22 r
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report J
-Installation of Process Monitor Instrumentation
.)
Modification No.:
1352H A.
System:
Miscellaneous Instrument Systems
'E.
==
Description:==
F Modification 1352H provides process instrumentation necessary
)
to assure safe alternative shutdown following a fire in either the Main Control Room (fire area 29), the Cable Spreading Room (fire area 28), or the emergency shutdown area (fire area 25).
The alternative process instrumentation to be provided includes reactor vessel water level and pressure, suppression pool water level.and temperature, drywell pressure and temperature, safety relief valve discharge temperature and condensate storage tank water level.
Indication for this instrumentation will be provided at the Unit 2 HPCI Alternative Control Station (ACS) which will be located in the recirculation M-G set room on elevation 135' of the Radwaste Building of Unit 2.
C.
Reason for Change:
This modification is necessary to satisfy requirements as set forth in Appendix R to 10CFR50.
[
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated because the modification will not affect the operation of the equipment to be monitored.
No Chapter 14 analyses as previously evaluated will be altered as a result of this modification.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any
{
evaluated previously in the safety analysis report?
Answer:
This modification does not create the possibility for an accident or malfunction of
[
a different type than evaluated previously because the modification will not affect the operation of equipment to be monitored.
The
{
modification is being installed to meet the 23 F.
Peach' Bottom Atomic Power Station f'-
Unit 2 Docket No. 50-277 Annual Plant Modification Report fire protection requirements of 10CFR50 Appendix R.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
The modification does not reduce the margin of safety as. defined in the basis for the r
Technical Specifications because the
(
modification will not affect.the operation of the systems to be monitored.
The monitoring instrumentation is not discussed in the
(.
Technical Specifications.
(
L
[-
[
[
[-
[
[
[
[
[
[
24
Peach Bottom Atomic Power Station
(
Unit 2
(
Docket No. 50-277 Annual Plant Modification Report
[
Removal of Low Vacuum Bypass Interlocks Modification No.:
1404 A.
System:
Reactor Protection System and Turbine Ger.arator B.
==
Description:==
The 600 psig RPS interlocks for main steam isolation and main condenser vacuum were removed.
The NRC approved the Technical Specification Amendment (3/14/86) to bypass these interlocks when in modes other than RUN.
C.
Reason for Change:
This modification improves station efficiency by facilitating low pressure (<600 psig) turbine pre-warming.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The setpoint was not designed as a
[
safety limit setting.
The bypassing of the interlocks was evaluated against the UFSAR and determined not to be essential to safety.
{
The NRC concluded that no credit is taken for a scram initiated by these signals in Section 14 of the UFSAR.
Removal of the interlocks
(
also reduces the probability of unnecessary L
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The subject interlocks were not designed L
to perform safety functions.
General Electric performed tests, the results of which (NEDO-20697) revealed no unacceptable r(
operating regions.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
The Technical Specifications were amended to reflect bypassing these interlocks.
The overall safety margir. is increased by reducing the possibility of unnecessary scrams.
25 l
Peach' Bottom Atomic Power Station Unit 2 h
Docket No. 50-277 Annual Plant Modification Report
[
Core Spray Injection Line' Check Valve Test Tap Installation
. Modification No.:
1449,
{
A.
System:
Core Spray B.
Descriotion:
This modification relocates the existing 1" test connection from the upstream side of the Core Spray valves AO-13A,B to the downstream (reactor side) of the valves.
In addition, a
(-
new 1" manual block valve will be installed in the equalizer line downstream of the equalizer valves AO-15A,B.
C.
Reason for Change:
The purpose of the modification is to facilitate testing of these check valves which have recently become the inner h
containment isolation boundary for Core Spray penetrations per Technical Specification Amendment 83-18.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or 7L malfunction of equipment important to safety as previously evaluated in the safety analysis report?
E Answer:
This modification does not increase the L
probability of occurrence or the consequences of an accident or malfunction of equipment e
important to safety as previously evaluated L
because the modifi u tion relocates test taps associated with the ' ore Spray valves (AO-13A,B) and installs a new 1" manual block
~
l valve in the equalizer line downstream of equalizer valves AO-15A,B.
The installation of the test connection and block valve does not affect the operation of the system nor
~'
any analyses as previously evaluated in w
Chapter 14.
ii)
Does this modification create the possibility for an
~
accident or malfunction of a different type than any evaluated previously in the safety analysis report?
E L
Answer:
This modification does not create the possibility for an accident or malfunction of p
a different type than any evaluated j
previously because the modification will not affect the operation of valves AO-13A,B and
~
AO-15A,B.
The test connection and block valve will only be used in the testing of AO-13A,B and AO-15A,B.
The modification does 26
Paach Bottom Atomic Power Station Unit 2 h.
Docket No. 50-277 Annual Plant Modification Report
[L not effect any accident analyses:as previously. evaluated in Chapter 14.
111).
Does this modification reduce the margin of safety as defined in the basis for the Technical ~ Specifications?
[-
Answer:
This modification does not reduce the margin of safety as defined in tne basis of the Technical Specifications because the
[..
installation of the test connection and block valve does not affect the operation of the Core Spray System.
The discussion of the
[.
test connection and block valve it not a part-of the Technical Specifications.
[.
[
E E
E c
L W
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27 F
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report h
Installation of Stem and Stem Connector on MSIVs
(
Modification No.:
'1548 A.
System:
(
B.
==
Description:==
r A valve stem and stem connector with keyslots were installed
(
on one Main Steam Isolation Valve (MSIV).
The stem connections on the remaining valves are to be upgraded as part of refurbishment of internal parts as those repairs are
(
required.
C.
Reason for Change:
This installation was made to provide a more positive method
.of preventing stem rotation at the threaded connection to the air operator than that originally supplied with the valve.
D.
Safety Evaluation Summary:
[
1)
Does this modification increase the probability of h
occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, the new design of stem connector has keyslots in the respective threaded portions
{
to accept two stepped keys.
The step in the key prevents the key from sliding axially along the stem.
The key is otherwise prevented from leaving the assembly by the head of the air cylinder rod retaining bolt.
Thus, the stem will be positively locked in place.
This modification enhances the reliability of the MSIV, and does not adversely affect any accidents previously evaluated.
r l
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
L Answer:
No, because although this modification involved a material substitution for the stem
[
and the stem connector, the materials substituted were equivalent or better than those originally supplied.
An engineering review of the modification concluded that the
(
original Design Report for the MSIV was not affected.
Stress in the stem threads was not significantly increased and is within design limits.
Because the function and operability of the MSIV was not impaired, the possibility 28
Poach Bottom Atomic Power Station T,
Unit 2 Docket No. 50-277 Annual Plant Modification Report for a different type of accident was not I
created.
iii)-
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
3 Answer:
No, because the function or functional-
{^
testing of the MSIV was not. impaired.
This modification enhances the reliability of the MSIV by preventing stem rotation at the j-threaded connection to the air operator.
Therefore, the margin of safety is not decreased.
i 1
(
[
29
Peach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Replacement of 49X DC Relays Modification No.:
1583 A.
System:
DC Power Eupply B.
==
Description:==
The 120 VDC thermal overload auxiliary relays in the 250 VDC motor operated valve motor control center (MCC) compartments were replaced with 137 VDC relays.
C.
Reason for Change:
The previous Cutler-Hammer, Type M relays designated as 49X relays in the DC motor-operated valve control circuits required replacement.
Two of these relays had failed while in service due to premature coil aging.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because t.'
DC power system reliability is actually increased because the 137 VDC relays are less likely to fail.
The previous 120 VDC relays were used in control circuits where the voltage is approximately 133 VDC.
Continuously energizing coils at a higher voltage shortclied coil life.
The replacement 137 VDC coils are better able to withstand this voltage.
Therefore, there are no adve:se impacts on any accidents previously evaluated.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, because the control logic of the motor operated valves is not affected because the 137 VDC relay is better able to withstand continuous energization at 133 VDC.
Additionally, the replacement relays are seismically and environmentally qualified for use in the subject locations, and perform the same function as the previous relays.
No new failure modes were introduced by this modification.
[
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
30
(
Peach Bottom Atomic Power Station'-
Unit 2
~
Docket No. 50-277-Annual Plant Modification Report Answer:
No, because the 49X-relay does not perform any. safety-related. function.
Although
(
connected to the safety-related motor operated valve (MOV) control circuits, its only safety-related requirement is that the
{-
relay coil must maintain control circuit electrical integrity.
The relay contact positions do not affect safety-related MOV operations.
Therefore, the margi" of safety
[-
is not reduced.
rL.
I f
L r
H I
31
-Pacch-Bottom Atomic Powar Station
-Unit 2 (I
Docket No. 50-277 Annual Plant Modification Report
(
Thermal Overload-Annunciation for Primary Containment-
- Isolation Valves Modification No.
- -
1584
' A.
System:
480 Volt MotorLControl Centers B.
==
Description:==
When the thermal' overload trip device operates on a motor-operated isolation valve, the newly installed annunciator alerts control room personnel.
An existing design feature bypasses the thermal overload trip device when valve operation is automatically required by the Primary.
-Containment Isolation ~ System.
t C.
Reason for Change:
Implementation of this modification corrects the der.f.gn deficiency (lack of control room annunciation) identified in NRC Information Notice 84-13.
D.
Safety Evaluation Summary:
W 1)
Does this modification increase the probability of occurrence or the consequences of an accident or f-malfunction of equipment important to safety as previously evaluated in the safety analysis report?
p Answer:
No.
Installation of additional control room indication of safety functions will not adversely impact the fulfillment of that function.
This annunciation was installed in L
the same manner as other annunciations on other safety related motor operated valves and corrects the identified design f.
deficiency.
u li)
Does this modification create the possibility for an accident or malfunction of a different type than any e-evaluated previously in the safety analysis report?
Answer:
No.
Neither the additional control room l
indication nor its installation will affect the function or control of any safety related equipment.
k 32 7
P&ach Bottom Atomic ~Powar Station Unit 2 n
Docket No. 50-277~
Annual Plant Modification Report lii)
Does_this' modification reduce the margin of safety as defined in the basis for the Technical Specifications?
h
~
No.
The installation of this modification is Answer:
consistent with other similar annunciators on
{.
similar equipment.
The increased annunciation will correct-an identified design deficiency-thereby improving the operator's ability to respond early and
[. -
correctly to potential or actual failures.
(:.
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L P
Y 33 IL....
,n L
Daach Bottom Atomic'Powar Station-Unit 2 Docket No. 50-277 Annual Plant Modification Report 1
4-Power-to-Load Unbalance Circuit Blocking
~ Modification No.:
1626
'A.
. System:~
Main Turbine B.
==
Description:==
(
Trip blocking circuitry was installed in the turbine Lo generator electrohydraulic control system to block operation of the power-to-load unbalance during turbine shell prewarming.
C.
Reason for Change:
This circuit prevents reactor scrams, and associated transients, during turbine shell warming, which are induced by the power-to-load unbalance circuit.
L D.
Safety Evaluation Sun. mary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment'important to safety as previously evaluated in the safety analysis report?
h Answer:
No.
The modification does not involve any changes in the operation of equipment Jmportant to safety and reduces the possibility of unnecessary scrams.
The 4
nuclear safety-related circuitry on the main steam line isolation valves, main steam stop valves and turbine control valves are not involved in this modification.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
This modification does not involve any changes in the operation of equipment important to safety.
The safety-related circuitry on the main steam isolation valves, main steam stop valves and turbine control valves are not involved in this modification.
5 H
iii)
Does this modification reduce the margin of safety as
(
defined in the basis for the Technical Specifications?
Answer:
No.
This modification does not involve any
[
changes in the operation of equipment important to safety or addreased by the Technical Specifications or their bases, r
k 34
Patch Bottom Atomic Power Station Unit 2
[
Docket No. 50-277 Annual Plant Modification Report Rerouting TIP Guide Tubing h
Modification No.:
1647 A.
System:.
-Transversing Incore Probe (TIP)
.B.
==
Description:==
The TIP guide tubing was rerouted from the indexer machines to the bottom of the reactor vessel.
The modification involved rerouting all 43 TIP guide tubes through seven reactor pedestal penetrations instead of the previoLs routing which used six penetrations.
C.
Reason for Change:
The purpose of this modification was to facilitate calibration, maintenance, removal and reinstallation of neutron monitoring sensors and tubing;-and to alleviate the previous congested tubing installation.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as
[-
previously evaluated in the safety analysis report?
Answer:
No, because this modification does not affect any safety related portions of the TIP system.
The removable tubing sections are not safety related.
The only portions of the TIP system that are safety related are the guide tubes and penetrations into the reactor pressure vessel and the primary containment penetration, including tubing and associated shear and ball isolation valves and supports.
The modification does not affect the operation of the TIF system.
11)
Does this modification create the pcssibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, relocating the tubing also involved relocating the tube supports.
The supports are of the same type originally installed and are also attached to shootout steel.
There are no seismic requirements for the removable tubing or associated supports.
The installation was performed in accordance with FSAR requirements for minimum allowable bend radii and support spacing of a maximum five feet.
35 l
l:?
l Peach Bottom Atomic Powar Station
-Unit 2 Docket No. 50-277 Annual. Plant Modification Report i.
[.e
.Does this modification reduce the margin of safety as ili) defined in the basis for tb.: Technical Specifications?
-Answer:
No, this modification elocates the tubing to provide straighter tube runs an providec better access to the undervessel area for-inspections and minor maintenance.
This is
.in accordance with the ALARA philosophy.
The
(:
-modification does not affect the operation of L-any safety-related equipment, and therefore does not increase or decrease any margin of safety.
[
[
[
F L:
[
E E
36 I
h _
I' P3ach Bottom Atomic Power Station Unit 2
(:-
Docket No. 50-277 Annual Plant Modification Report
[
High Pressure Service Water Pipe Restraint Repair Modification No.:
1677 A.
_ System:
High Pressure Service Water (HPSW)
B.
Descripticn:
Three of the four bolts on-the anchor plate on the HPSW pipe restraint 32-6B-S63 were modified.
C.
Reason for Change:
The anchor plate was found to be pulled approximately 1/16 inches away from the wall.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because the reliability of the anchor plate was enhanced by the modification to three of the four bolts.
The upper right
[
hand bolt and the lower left hand bolt were reinstalled with washers to provide a snug fit.
The lower right hand bolt was replaced f
with a Hilti Kwik bolt.
The enhanced reliability of the anchor plate does not reduce any margin of safety.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, because the functions of the bolts, anchor plate and pipe restraint were not changed.
A failure of the modified bolts would result in the same consequences as a failure of the previous bolts.
{-
lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, because this modification enhanced the reliability of the anchor plate thereby enhancing the reliability of the High Pressure Service Water System.
The modified bolts enable the anchor plate to better withstand the vibrations which are inherent to system operation.
37 J
.Paach Bottom Atomic Powar Station Unit 2
[9 Docket No. 50-277
?
Annual Plant Modification Report 1.
iw-Replacament and Relocation of Reactor Feedpump REcirctlation Valves Modification'No.:
1695 A.
System:
Feedwater B.-
==
Description:==
(
The three reactor feedpump recirculation air operated valves were replaced with new high pressure drop, low recovery air-operated valves.
The new valves were installed in a new location adjacent to the condensar and the existing downstream orifices were removed.
C.
Reason for Change:
The new design and location eliminate high maintenance requirements and improved reliability.
L D.
Saeety evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, the previous valves were replaced with valves of an improved design, and the new valves will still perform the same function as the previous valves, while reducing the probability of valve failure and reactor feedpump damage.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, this modification does not alter the design of the feedwater system.
The g
I consequences of the failure of these valves are the same as those of the previous valves, and do not affect the analyses in the UFSAR.
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
r(
Answer:
No, although the feedwater pumps are addressed in the Technical Specifications, the recirculation valves are not,
- Purther,
[
the new design of the recirculation valves improves the reliability of the feedwater
- system, s
38 E
P3ach Bottom Atomic Powar Station Unit 2 (i
Docket No. 50-277 l.
Annual Plant Modification Report L
Capping of One-inch Pipe Stubs in Torus Air Space Modification No.:
1716 A.
System:
==
Description:==
Four caps were welded to one-inch pipe stubs located in the (L
Torus Air Space.
The lines are "A" and "B" RHR test and suppression pool cooling return lines, RCIC vacuum pump discharge line and HPCI turbine exhaust drain line.
[.
C.
Reason for Change:
This modification satisfied the criteria of Regulatory Guide 1.141 and ANSI N271-1976 defining containment isolation.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
Capping these lines is in conformance
['
with Regulatory Guide 1.141 and ANSI N271-1976 requirements for isolation of containment piping terminating below the minimum suppression pool water level.
A design analysis has determined that capping the anti-siphon devices on this piping will have no effect on the operation or functions L
of the safety systems, 11)
Does this modification create the possibility for an F
accident or malfunction of a different type than any L
evaluated previously in the safety analysis report?
Answer:
No.
The design analysis and review of the r(
Updated Final Safety Analysis Report demonstrates that capping the anti-siphon devices will not impact the operation or function of the safety systems.
E E
e L
39 r
~
Phach Bottom-Atomic Powar Station
~
. Unit 2
{:
Docket No. 50-277 Annual Plant Modification Report
..iii)
Does this modification reduce the margin of safety as defined'in~the basis for the Technical Specifications?
Answer:
No.
The capping of the anti-siphon devices will-not impact the operation.or function of
[
the safety systems.
The caps do not impact t
the primary containment environmental, dynamic, isolation or pressure boundary parameters used as the bases for the Technical Specifications.
g
(:
[
[.
[
[
L
[
(
[
40 L_
-Pcach Bottom Atomic:Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Replacement of Main Steam Drain Valve / Operator Assemblies Modification No.:
1744 A.--System:
Main Steam B.'
==
Description:==
-The main steam drain valve / operator assemblies MO-2-1-79 and MO-3-1-79 were replaced.
The previous MO-79 assembly was a 3",
600# ANSI rated, carbon steel globe valve with a Limitorque-SMB-00-7 1/2 motor operator.
The replacement assembly is a 3",
900# ANSI rated, ASME section III, Class 1, carbon steel, double disc gate valve with a Limitorque SMB-000-5 motor operator.
C.
Reason for Change:
Thia modification replaced a valve which had in-body seat
{
damage and generic problems associated with valve maintenance and spare parts.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, although the MO-79 valve is safety related, its function or operation is not affected by this modification.
The required function of MO-79 is open/ closed, with no throttling requirements.
A double disc gate valve provides the required isolation and therefore fulfills all system requirements.
The portion of main steam drain piping that contains valve MO-79 is non-seismic and is classified as Group III piping per Appendix A of the UFSAR.
The replacement valve was procured as a seismically qualified ASME Section III Class 1 valve, thereby surpassing the original requirements.
Additionally, the MO-79 operator is not required to be Class lE (IEEE) qualified.
However, as a conservative measure, the replacement Limitorque operator l
was procured as an outside containment Class t
lE safety-related operator.
Because the replacement valve / operator assembly meets or exceeds all of the requirements of the original valve /cperator assembly and the function of the valve is not changed, there is no increase in the probability of any ace'. dent or malfunction previously evaluated.
41
(
Poach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report
[
ii)
Does this modification create the possibility for an accident or malfonction of a different type than any
(-
evaluated-previously in the safety analysis report?
Answet:
No, a reanalysis of the existing piping, and pipe support evaluation has been performed
[' '
considering the increased weight of the replacement assembly.
As a result, two pipe support structures were modified in order to
(;
maintain piping and support stresses within ANSI B31.1 allowable limits.
The power requirements of the replacement operator have been reviewed and found to be satisfied by the existing operator-power supply with no increase in load on the plant electrical system.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, because the replacement valve / operator assembly meets or exceeds all requirements of f
the previous valve / operator assembly.
The L-required function of the MO-79 valve is not affected.
Therefore, the margin of safety is not reduced.
E rm L
H u
I i
IL 42
?
1
Peach Bottom Atomic Powar Station Unit 2 (1
Docket No. 50-277 Annual Plant Modification Report b
Rtplacement of Main Generator Seal Oil Vacuum Gear Motor h
Modification No.:
1834 A.
System:
Main Generator B.
-Description:
The-main generator seal oil vacuum gear 2 HP motor was
~
[-
replaced with a 3HP motor, and a new solenoid valve was installed into the motor circuit.
{
C.
. Reason for Change:
The previous motor'was unreliable, and replacement parts were r
not available.
The solenoid valve was added to break vacuum
(
upon-de-energization of the mctor.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurronce or the consequences of an accident or r
malfunction of equipment important to safety as L-previously evaluated in the safety analysis report?
Answer:
No, because the main generator does not serve any safety-related purpose; thus, modifications to the seal oil vacuum gear motor do not affect the probabilir'; of an accident.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any L
evaluated previously in the safety analysis report?
Answer:
No, the power requirements of the replacement I
motor and solenoid valve were reviewed and L
found suitable for the existing motor power supply.
The slight increase in load does not result in bus overloading.
Failure of the r
L new motor or solenoid valve would not be a different type than previously evaluated because the functions have not been changed.
~
l 111)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
m L
Answer:
No, because this modification does not affect the operation of any safety-related equipment.
There is no impact on the Technical Specifications or their bases.
m 43
PGach Bottom Atomic Powar Station Unit 2
[
Docket No. 50-277 Annual Plant Modification Report t
Bronze Bearing Replacement for Clow Butterfly Valves in the Containment Atmospheric Control System Modification No.:
1935 A.
~ System:
Containment Atmospheric Control B.
-Description:
b This modification replaces carbon sleeve bearings with bronze bearings in the Containment Atmospheric Control System butterfly valves manufactured by Clow Corporation.
C.
Reason for Change:
(
The replacement is being done based on the failure of two
(
Clow butterfly valves installed in the CAC system.
The shaft of each valve had bonded to the upper and lower sleeve bearings such that disc movement was severely rec'-i ted.
The root cause of the failure was found to be chloride contamination in the carbon (graphite) bearings.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or
('
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
{
Answer:
The replacement of the carbon sleeve bearings with bronze bearings will not adversely affect the operation of the valve.
Bronze has been determined by Clow to be an
(
acceptable alternative material.
The va?ves operation will not be altered; therefore, Chapter 14 safety analyses will not be
{
impacted by this modification.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
The replacement of the carbon sleeve bearings with bronze bearings will not create an accident or malfunction of a different type than evaluated previously.
Clow has y
determined that the use of bronze as an
(
alternate bearing material is acceptable.
Conversations with Clow determined that the bronze bearings will not affect stroke times.
[
The valves will perform their safety function as previously analyzed.
iii)
Does this modification reduce the margin of safety as
[
defined in the basis for the Technical Specifications?
44
' Peach Bottom Atomic Powar Station-Unit 2
({
Docket No.-50-277 Annual Plant Modification Report
(-
Answer:
The replacement of the carbon sleeve bearings with bronze bearings will not reduce the
(;'.
. margin of safety as defined in the basis fo'r the Technical Specifications.._The bronze
- bearings have been determined by Clow to be an acceptable alternative.
[
[;
[L
[
E.
E E
E E
m
' 1 L
FL 45
PGach Bottom Atomic Pow 0r Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Human Factors Enhancement of Remote Shutdown Panels Modification No.:
1958 A.
System:
Miscellaneous Instrument Systems B.
==
Description:==
The panels' backgrounds were painted beige.
Color fields and outlines were added to group-related instrument and controls.
The panels were also re-labeled using a hierarchical labeling scheme.
C.
Reason for Change:
The need for man-machine interface improvement was identified during the Control Room Design Review.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or i
malfunction of equipment important to safety as
(
previously evaluated in the sar'ety analysis report?
Answer:
No.
The electrical bus loading is not impacted.
The man / machine interface is enhanced, thereby enhancing his ability to properly respond during an accident.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
(
Answer:
No.
Painting and relabeling the panels will not impact the ability of the equipment to
[
perfota its function.
The human factors t
enhancements will serve to reduce the overall probability of operator errors, iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
The human factors enhancements will not physically impede the functioning of the panels.
'.nproving the man-machine interf ace will not reduce a margin of safety.
(
46
b Peach Bottom Atomic Powar Stotion Unit 2
(
Dockot No. 50-277 Annual Plant Modifi' cation Report Addition of Block Valves on Sukoly and Return Branch Lines to the ECCS Rooms Modification No.:
2015 A.
Sy.s t em :
Emergency Service Water B.
Dpscription:
Block valves and vent taps will be installed on emergency service Nater (ESW) system supply and return branch lines to various ECCS rooms.
C.
Reason for Change:
This modification is being performed to facilitate future
(
repairs and cleaning of ESW piping through the use of isolation capabilities provided by block valves and vent taps.
This modification will provide the flexibility of performing this maintenance with the unit at ocwer.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as
[
previously evaluated in the safety analysis report?
Answer:
The probability of occurrence or the consequences of an accident or malfunction of
[
equipment important to safety previously described in the Safety Analysis Report is not increased.
Installation of the manual
(
block valves are to improve the maintainability of the system.
Proper valve position following maintenance is confirmed
[
by preservice testing.
Whenever the ESW L
system is required to be operable, these valves will be administratively controlled by being locked in their open position.
The use of threaded couplings for vent connections, flanged joints, and socket weld fittings for 4" and under piping are permitted under the design code ASME Sectien III, Class 3 piping, Article ND-3600.
A review of the piping stress analysis indicates that the additional weight of the socket weld fittings does not
[
significantly affect the piping stress or support loadings.
The surface examination methods and hydrostatic test pressure for
{
this r.odification are consistent with the requirements of the Peach Bottom Inservice Inspection Program.
47 E
Peach Bottom Atomic Powar Station Unit 2
('
Docket No. 50-277 Annual Plant Modification Report
{-
11)
Does this modification create the possibility for an accident or malfunction of a different. type than any
(-.
evaluated previously in the safety analysis report?
Answer:
The possibility of an accident or malfunction of a different type than previously evaluated
{-
in the Safety Analysis Report is not created.
The capability of supplying safety-related equipment with cooling water is not
[
jeopardized by the installation, examination and testing of these valves and vent taps in accordance with the requirements of ASME
(
Section III for-Class 3 nuclear components and the Peach Bottom Inservice Inspection Program.
['
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
(
Answer:
The margin of safety as defined in the basis for the operating license is not reduced, since the design, installation and testing associated with this modification in no way
[
restricts the ESW system from performing its intended safety function.
A change to the PBAPS Technical Specifications is not
(
required.
The function of the existing plant systems are no affected by the installation of these block valves.
The Technical
[
Specifications do not address the operability of valves installed for tne purpose of performing equipment maintenance.
I L
r L
F 48 F
Perch Bottom Atomic Powar Station Unit 2 Docket No.-50-277 Annual Plant Modification Report
_Ralocation of Power Feed for MCC 20DllA-
~
Modification No.:
2082 A.
S/ stem:
High Pressure Coolant Injection B.
==
Description:==
This modification reroutes safety-related power cables from a fuaed disconnect located in a compartment of the 20Dll D.C.
power MCC to provide power to the 20DllA auxiliary MCC.
Presently, D.C. MCC 20Dll, located in the Unit 2 Reactor Building on elevation 135', shares a 250V D.C. Bus with the 20D11A MCC.
20DllA is located in the Unit 2 Reactor Building on elevation 165'.
C.
Reason for Change:
The purpose of the modification is to provide electrical isolation in accordance with Appendix R to prevent a fire-initiated fault on the 20DllA Bus from tripping the feeder breaker that powers the 20Dll MCC.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
f Answer:
The probability of occurrence or the
- onsequences of an accident or malfunction of equipment important to safety previousij evaluated in the safety analysis report is not increased.
This modification maintains the Division II emergency safeguard D.C.
power supply to MO-4245, the HPCI turbine exhaust valve.
This valve is the only load on the 20DllA MCC.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
The probability of an accident or malfunction of a different type than previously evaluated in the safety analysis report is not created.
(
The emergency safeguard D.C. power or HPCI systems are unchanged as described in the UFSAR.
The new cable from 20Dll to 20DllA is sized to accommodate the load of the MO-4245 motor operator, the only load on the 20DllA MCC.
Also, the length of this cable route has been considered in sizing the cable to avoid significant voltage drops and to assure 49
[
(l:
I Peach Bottom Atomic Powar; Station Unit 2 Docket No. 50-277' Annual-Plant tiodification Report adequate voltage levels at the MO-4245 motor
/
operator.
The fuse at the 20Dll compartment l
U powering 20DllA is sized to protect the new l
cable against fault currents and to provide i
proper coordination with the breaker that powers the 20D11 MCC.
iii)
Does this modification reduce the margin of safety as i
defined in the basis for the Technical Specifications?
i
('
Answer:
The margin of safety as defined in the basis t
of any Technical Specification is not reduced i
by this modification.
This modification does not change the emergency safeguard D.C. power or HPCI systems as described in the Technical Specifications.
f 0
(
50 F
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Self-Locking Setscrews Modification No.:
77-036 A.
System:
Transversing Incore Probe (TIP)
B.
Descriotion:
Self-locking setscrews were installed in various 90rtions of the TIP drive mechanism.
C.
Reascn for Change:
This change was recommended by General Electric Con:pany Service Information Letter 211 (Supplement 1) to prevent failure of the drive mechanism.
D.
Safety Evaluation Summary:
I i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The self-locking setscrews provide more assurance that the drive mechanism will perform without malfunctioning, but the safety-related explosive cable shearing valves and the automatic isolation ball valves were not affected.
ii)
Does this modification create the possibility for an
\\
accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The original safety analysis addressed failure of the drive mechanism.
The explosive cable shearing valves would provide containment isolation in the event of an accident with the TIP stuck in the tube.
1 iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
k Answer:
No.
The drive mechanisc was made more reliable, thereby increasing the margin of safety.
The details of the dr:.ve t.echanism
~
h are not discussed the baces for the Technical Specif l
51 1
1 1
MI$0
PeachDBottom Atomic Powar Station Unit 2 Docket No. 50-277
~ Annual Plant Modification Report
-Replacement of Containment Spray Flow Transmitter-Modification No.:
84-033 A.
System:
Incident Detection Instrumentation B.
==
Description:==
L
~The RHR containment spray flow transmitter, Barton Model No.
368, was replaced with a Rosemount Model No. llSlDP6B22.
This transmitter provides flow indication in the control 3
room.
C.
Reason for Change:
The Barton transmitter is no longer manufactured, and the Rosemount was determined to be an equivalent component.
D.
Safety Evaluation Summary:
1).
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The Rosemount transmitter performs the same non-safety related function in the same manner as the Barton.
The replacement transmitter meets the same design criteria as the Barton for maintaining the RHR system pressure boundary.
O 11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
This modification introduces no new design criteria or features.
Tests show that the Rosemount transmitter will maintain the pressure boundary integrity and not impact the performance of nuclear safety related equipment.
[
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
The transmitters only provide control p
room indication containment spray flow.
u Thus, replacement with an equivalent model does not reduce any safety margin.
s F
52
Pstch Bottom Atomic Powsr Station F
Unit 2 Docket No. 50-277-Annual Plant Modification Report Install a Lock Pin on the '2A' Recirculation Pump Impeller Locking Nut Modification No.:
85-007 A.
System:
Reactor Recirculation
.B.
==
Description:==
The purpose of this modification is-to inctall a lock pin on the 2A recirculation pump impeller lock nut set screw.
When removing the impeller for shaft replacement, it was discovered that the tack weld on the impeller lock nut set screw was broken.
The tack weld prevents the set screw from vibrating loose.
To prevent this, a lock pin will be installed on the set screw as recommended by the manufacturer, Gotg Warner.
All new impellers have this design. incorporated.
The
'2B' impoller has already beer.
reinstalled with the new lock pin design.
The modification
' involves drilling a lock pin hole 1/4" x 9/16", inserting the 5/16" long 304 stainless steel lock pin, and welding over the lock pin.
C.
Reason for Change:
The purpose of the modification is to replace the old design of tack welding the set screw with the new design of a lock pin to prevent the impeller lock pin set screw from vibrating loose.
l D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence of an accident or malfunction of equipment important to safety because the modification will not alter the operation of the
'2A' recirculation pump as previously evaluated in the Chapter 14 Safety
(
Analyses.
11)
Does this modification create the possibility for an y
accident or malfunction of a different type than any
(
evaluated previously in the safety analysis report?
Answer:
This modification does not create the i
possibility for an accident or malfunction of a different type than any evaluated previously because the modification will not change the operation of the
'2A' recirculation pump as evaluated in the 53 L
j
Patch Bottom Atomic Powar Station Unit 2~
. Docket-No.-50-277 Annual Plant Modification Report y
Chapter 14 Safety Analyses.
The modification has been approved by the pump manufacturer, Borg Warner.
iii)
Does this modification' reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
This modification does not reduce the margin of safety as defir.ed in the basis for the Technical Specifications because-the modification-will not alter the operation of the '2A' recirculation pump as previously evaluated in the Chapter 14 Safety Analyses.
The Technical Specifications are not changed f
by this modification.
l l
(
54 f
Y Bottom Atomic Powar Station-Unit 2 f
Docket No. 50-277
/lant Modification Report
?-
Racirculation Pump Machining h
Modification No.:
85-048 A.
.cystem:
Reactor Recirculation B.-
==
Description:==
The Unit 3 A and B Recirculation pumps were machined to h
remove approximately 1/8 inch from the half coupling lower ledge and approximately 3/32 inch off the lock plate in order to align pump shaft sleeve replacement parts.
Reason for Change:
.C.
(
New shaft sleeves were installed and because of manufacturing
(
machine tolerances, minor plant site machine work was required for parts alignment.
=
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The machine work corrected the alignment problems on the pump parts following repairs l
and without the repairs and the resulting proper clearance, the pump might not be reliable.
The internal parts of the recirculation pumps are not evaluated in the FSAR.
Also, the modifications to the pump L
were recommended, reviewed, and approved by the pump manufacturer.
f 11)
Does this modification create the possibility for an accident or malfunction of a different type than any ecaluated previously in the safety analysis report?
Answer:
No.
Following the modification, the pumps y
were tested to verify that the repairs and modifications were acceptable.
The
[
modification, including machine work to align
?
the pump parts, did not include any changes to the function of the purt.p.
Iw lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
[
Answer:
No.
The machine work to align the pump parts allows the pump to perform its original design function.
55 C
\\
P:ach Bottom Atomic Power Station i
Unit 2 Dockot No. 50-277 Annual Flant Modification Report Rrplacement of Residual Heat Removal System Differential Pressure Transmitter DPT-2-10-91D Modification No.:
85-066 A.
System:
==
Description:==
This modification replaces the RHR
'D' heat exchanger tube to shell differential pressure transmitter.
The modification involves replacing the existing GE Model 552 transmitter with a Rosemount Model ll51DP7B22 transmitter.
C.
Reason for Change:
The purpose of this modification is to replace a leaking transmitter which is no longer manufactured with a similar transmitter.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence of an accident or malfunction of equipment important to safety
(
because the current transmitter will be replaced with a transmitter that performs the same function.
The safety-related equipment
(
associated with this modification will not be adversely affected.
The accident analyses as described in Chapter 14 are not affected by this modification.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification does not create the possibility for an accident or malfunction of a different type than evaluated previously because the transmitter will be replaced with a transmitter that performs the same function.
The modification will not cdversely affect the operation of the RHR System.
This modification will not affect any previously evaluated accident as described in Chapter 14.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
56 l
[ \\ y-Pscch' Bottom Atomic Powar Station f'-
Unit 2 Docket No.50-27k Annual Plant Modification Report Answer:
The margin of safety as defined in the basis-jf for the Technical Specifications will not be l
reduced because.the transmitter will be replaced with a transmitter.that performs the same function.
The modification will not adversely affect the operation of the RHR Systen..
The Technical Specification are not impacted as a result of.this change.
N
(
[
[
57 i
E Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report
.Rcplacement of CRD Differential Pressure Transmitters Modification No.:
85-071 A.
System Control Rod Drive B.
-Description:
The Barton model 296 differettial pressure transmitter (DPT 2-3-211) was replaced with a newer vintage Rosemount model, 1151.
This instrument provides control room indication of the-differential pressure between the reactor vessel and the control rod drive water header.
C.
Reason for Change:
The original transmitter had failed and is no longer manufactured, so a substitute was installed.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The Rosemount transmitter performs the same, non safety related function in the same
-manner as the Barton; and it meets or exceeds
{
the original design cpecifications, 11)
Does this modification create the possibility for an accident or malfunction of a different type than any
[
evaluated previously in the safety analysis report?
Answer:
No.
Since the two transmitters perform the same non safety related f'inctions in the same manner, and the manufacturing and installation standards are the same, the former safety analyses hold true.
No safety corcerns are introduced.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
Neither the Technical Specificat'.ons nor their bases define a margin of saferf associated with this instrument or its function.
Replacing a failed mumponent with an engineering equivalent will not reduce a L
margin of safety.
58
Pocch Bottom Atomic Powar Station Unit 2 Docket No.-50-277 Annual Plant Modification Report Unit 2 Cycle 6 Shutdown Rod Worth Minimizer (RWM) Sequence Update
{
Modification No.:
85-127 A.
System:
Computer B.
==
Description:==
This modification updates the computer Rod Worth Minimizer (RWM) shutdown sequence arrays to the current rod pattern.
C.
Reason for Change:
This modification is necessary to update the RWM shutdown sequence arrays to the current rod pattern resulting from the new cycle.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence or malfunction of equipment important to safety because the modification updates computer software for
(
the shutdown RWM sequence.
This update has L
been analyzed and determined to have no adverse effect to plant safety nor a change to any safety analysis as analyzed in Chapter p
L 14.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis raport?
Answer:
Thin modification will not create the L
possibility for an accident or malfunction not previously evaluated because th Unit 2 Cycle 6 shutdown RWM sequence update has been l
analyzed and determined to have no adverse effect to plant safety nor impact on any safety analysis in Chapter 14.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
The Unit 2 Cycle 6 shutdown RWM sequence update has been analyzed and determined to have no adverse effect on plant safety and therefore, would not reduce the margin of safety as defined in the basis for the 59
_ ____________________ A
)
Panch Bottom Atomic Pow 0r Station
)
Unit 2 Docket No. 50-277 Annual Plant Modification Report Technical Specifications.
This update is necessary for the Unit 2 Cycle 6 rod pattern.
M F
60
[
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Install Bypass Line from Dirty Lube Oil Storage Tank to Suction of Lube Oil Transfer
'lodification No.:
85-138 A.
Syytem:
Turbine Lube Oil B.
==
Description:==
A 3/4 inch bypass line and block valve was installed between the drain valve of the "dirty oil" tank and a pressure tap on the suction side of the lube oil transfer pump.
C.
Reason for Change:
This bypass line facilitates the cleaning of the "dirty-oil" tank.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to sr..fety as previously evaluated in the safety analysis report?
Answer:
No.
The bypass line is attached to the turbine lube oil system, which is non safety-related.
The bypass line is installed within the spill containment dike, thus not affecting the fire safety analysis, 11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The safety analysis report does not address this equipment, and failure of this line will not affect equipment or systems which are described, iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
f Answer:
No.
The equipment which is affected by the modification is not addressed by the Technical Specifications.
Its performance or failure will not impact a margin of safety as defined in the Technical Specifications.
61
________ A
Poach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Residual Heat Removal (RHR) System Pump Wear Ring Retaining Dowels Modification No.:
85-146 A.
System:
==
Description:==
This modification replaces the Resi 11 Heat Removal (RHR)
System pump impeller wear rings.with 6 dowel holes instead of the originally supplied equipment with 8 dowel holes.
C.
Reason for Change:
The purpose of this modification is that the manufacturer has changed the number of dowel holes in the rings from 8 to 6 in order to standardize all of the rings in the pumps supplied by the manufacturer, Bingham-Williamette Company.
D.-
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification will not increase the probability of an occurrence or malfunction of equipment important to safety because the change to the number of dowel holes in the rings from 8 to 6 will in no way affect the integrity of the ring being staked.
Therefore, pump operation will not be l-affected.
No Chapter 14 safety analysis is impacted as a result of this modification.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
L Answer:
This modification will not create the possibility of an accident or malfunction of a different type because the modification to change the dowel holes in the pump rings from 8 to 6 will not change the operation of the pumps.
The pumps will perform as analyzed in the Chapter 14 safety analysis without change.
iii)
Does this modification reduce the margin of safety as g
defined in the basis for the Technical Specifications?
Answer:
This modification will not reduce the margin of safety as defined in the basis for the F
Technical Specifications because the 62 1
- t I
Paach Bottom Atomic Powar Station
~
Onit 2 Docket No. 50-277-
' Annual Plant Modification. Report-x l.
modification to change the dowel holes in the pump rings from 8 to 6.will not change ~the-operation of the pumps.
This change does not affect the Technical Specifications.
(
63
Peach Bottom Atomic Powar Station Unit 2
{
Docket No. 50-277 Annual Plant Modification Report MO-154 Lock Nut Set Screw Installation f
~ Modification No.:
85-149
.A.
System:
B.
==
Description:==
The two existing set screws in the upper bearing locknut of
-s the MO-154 valve were replaced with three equally spaced setscrews (120 degrees apart).
The MO-154 valves are Walworth 24" right angle globe valves.
The valve stem is threaded within a yoke nut which is keyed into the drive gear of the Limitorque.
Vertical movement of the yokenut is prevented by the upper bearing lock nut securing the yoke nut bearing assembly.
C.
Reason for Change:
This modification provides a more reliable method of securing the upper bearing locknut to the bearing assembly.
A malfunction of the upper bearing locknut on the MO-2-10-154A resulted in a complete valve failure.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because this modification decreases the probability of failure of the injection valve which increases the reliability of the RHR system.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, because the only consequence of a malfunction resulting from this modification is the failure of the MO-154 which is already evaluated in the FSAR.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, because the modification increases the reliability of the valve by the addition of a third set screw.
Therefore, the margin of safety as defined in the Technical Specification basis is not reduced.
f 64 l.
)
Peach Bottom Atomic Powar Station, Unit 2 Docket No. 50-277 Annual Plant Modification Report
~ Removal of Reactor Feedpump Min-Flow Line Test Taps Modification No.:
86-014 A.
' System:
Reactor Feedwater B.-
==
Description:==
The pressure taps on the reactor feedpump minimum flow recirculation line were removed and replaced with welded caps.
i C.
Reason for Change:
The pressure taps were not used and were frequently damaged.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The min-flow pressure taps were not used and frequently required repair.
Their removal and replacement with a weld cap does not affect min-flow operation, and the pressure taps do not contribute to the accident analyses in the Updated Final Safety Analysis Report.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The removal of the pressure taps which were not used and frequently required repair will remove safety risks of that line.
Further, the minflow operation of the feedwater system is not affected and the previous safety analyses remain valid.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
No basis for Technical Specifications address the pressure taps.
Removal of unnecessary equipment which requires frequent repair improves the overall margin of plant safety.
65
. Peach Bottom Atomic Power Station Unit.2 Docket No. 50-277 Annual Plant Modification Report Installation of Isolation Links on Recirc Pump Switch f
Modification No.:
86-030 A.
System:
Reactor Recirculation B.
==
Description:==
Isolation links were installed on the dual contact flow switch, FS-2-26A(B), which annunciates a high or low flow condition at a single alarm window.
C.
Reason for Change:
Isolation of the flow switch contacts protides a method of determining whether a high or a low flow condition exists.
D.
Safety Evaluation Summary:
'i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, flow switch FS-2-26A(B) is a dual contact switch that annunciates a hi or lo flow condition at a single alarm window.
Neither the flow switch nor the instrument loop is safety related.
In order to assess the pump seals condition when the alarm annunciates, it is necessary to know whether a hi or lo flow condition exists.
This modification enhances troubleshooting.
It does not affect the operation of any safety related equipment, and therefore does not affect the probability of any previously evaluated accidents.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the.afety analysis report?
Answer:
No, because solid links are used to isolate each flow switch contact.
Therefore the modification will not increase the possibility of the alarm becoming inoperable due to a circuit fault.
This modification is not safety related and does not create the possibility for any different accident than evaluated in the UFSAR.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
66 I
I
I Peach Bottom Atomic Powar' Station 4
Unit 2 h
Docket No. 50-277 Annual Plant Modification Report Answer:
No,.because the operation ~or' failure of the flow switch is. considered-in the basis'for
( =.
the-Technical. Specifications.. Additionally, the installation of the. isolation' links does
'not-adversely affect the operation of-the flow switch.
i
[;
L ru J
67 P
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Peach Bottom Atomic Power Station Unit 2 f-Docket No. 50-277 Annual Plant Modification Report Installation of Shield-Wall in Source Vault / Storage Area Modification No.:
86-054 A.
System:
Structural B.
==
Description:==
A wall was constructed in the Unit 2, Elevation 195' source vault / storage area to provide an area to store radioactive sources.
The room previously used to store the sources will be used as a dosimeter calibration area.
C.
Reason for Change:
An area was needed for dosimeter calibration.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment impar' ant to safety as previously evaluated in the safety analysis report?
Answer:
No, the wal'. was constructed of 8" high density coacrete blocks.
The source is enclosed within the room by an additional shield wall which runs perpendicular to the existing wall and the new wall to provide shielding for the corridor leading to the dosimeter calibration area.
The additional wall is not located near any safety related equipmenc.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, because failure of this wall will not affect any safety related equipment.
Loss of the shield wall will result in an increase in the dose rate in the area, which is accessed only through A-84, "Control of High Radiation Area Keys".
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, because the wall is not located near any safety related equipment.
The new wall or the previous shield wall are not addressed in the Technical Specifications or their bases.
68
Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Permanent Shielding for Feedwater Heater Drain Lines Modification No.:
86-056 A.
System:
Feedwater B.
==
Description:==
An additional layer of 1/4 inch lead shielding was added to existing shielding; covered with wallboard and painted.
C Reason for Change:
The shielding reduces the radiation level in the health physics field office lunchroom.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The design, function and operation of nuclear safety-related equipment and systems are not impacted by the addition of this permanent shielding.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The permanent shielding wb.ich was installed does not impact the design, or interfere with operation of safety related equipment, iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
The modification is consistent with the ALARA concept which the Technical Specifications do address.
The shielding does not affect plant operations or equipment.
L 69 l
Peach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Installation of Auxiliary' Steam Cor.densate Return Pump Isolation Valves
(
-Modification No.:
86-070 A.
System:
Condensate Return B.
Descriotion:
Isolation valves were installed on the suction lines of the auxiliary steam condensate return pumps.
C.
Reason for Change:
This modification permits individual isolation of each pump to permit maintenance, without isolating the entire system.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The auxiliary steam system is required during plant startup to cupport condensate deaeration and turbine shaft seal steam.
This condensate is returned to the main condenser.
The condensate return system is used for building heating.
The addition of individual isolation valves allows each pump or line to be isolated for maintenance without rendering the entire system inoperable.
Installation operation, or failure will not impact nuclear safety-L related s/ stems.
,/
11)
Does this modification create the possibility for an L
accident or malfunction of a different type than any evaluated previously in the safety analysis report?
e
(
Answer:
No.
The Updated Final Safety Analysis Report does not address the auxiliary steam system.
The modification enhances operation and
[
maintenance of the system without affecting nuclear safety related systems.
N 70
Peach Bottom Atomic Power Station Unit 2 Docket'No. 50-277 Annual Plant Modification Report i
~
lii)
.Does this~ modification reduce the1 margin of safety as defined in the basis for.the Technical Specifications?
Answer:
No.
The auxiliary steam condensate return-system is not-a nuclear safety related
. system.
Neither the system or its operability are addressed'by the Technical-
-Specifications.
Thus, improving the ability to perform maintenance will not reduce a margin of safety.
/
L I
3 I
71 f
Poach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report
. Leak Repal: on Manway Flange on Main Steam Cross Around Pipe Modification No.~:
86-089 A.
System:
Main Steam B.
==
Description:==
A clamp and chamber fixture were attached into the flange of f
a 30" r.anway on the 42" cross-around pipe between the HP turbine and the "C" moisture separator.
An elastromeric sealant compound was also injected into the chamber.
C.
Reason for Change:
This modification temporarily repairs the leaking flange.
.D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to Fafety as previously evaluated in the safety analysis report?
[.
Answer:
No.
Failure of the sealant or the clamp will not affect the ability of the plant to achieve safe shutdown.
The affected equipment is not nuclear safety related.
11)
Does this modification create the possibility for an L
accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
Although the injected sealant compenent will satisfactorily withstand all anticipated plant operating conditions and the added mass I
is within acceptable limits for the-support L
system, the cross-around piping is not required to achieve safe shutdown of the plant.
Failure of the seal would be bounded by the FSAR accident analyses for a main steam leak.
~
111)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
The main steam cross-around pipe is not nuclear safety related and not addressed by g.
the Technical Specifications.
Temporary sealing of a leak on this piping will not adversely impact safety.
s 72
Peach Bottom Atomic Power Station Unit 2 Docket No.'50-277 Annual' Plant Modification Report Installation of Fill Valves on Torus Level Indication M o d i f i c a t i o n N o. :-
86-148 A.
System:
==
Description:==
As part of a repair of torus water level transmitter LT-8123A, valves were installed on the high side of the level
. transmitter to permit the filling the sealed sensing line with fluid and remove air from it.
C.
Reason for Change:
The valve permits filling of sealed system line with transmitter fluid.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or
[,
malfunction of equipment important to safety as t
previously evaluated in the safety analysis report?
Answer:
No.
The modification meets the original design criteria for the system.
The equipment was repaired to meet the original design criteria.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The modificati0n does not change the existing design or operation of the torus f.
water level indication.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
J Answer:
No.
No functional design changes were made.
)
No part of the Technical Specifications or
[
equipment th6y address are adversely impacted.
73
Peach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report
-Rod Worth Minimizer (RWM) Shutdown Sequence Update h
Modification Nos.:
87-006 A.
System:
Process Computer B.
==
Description:==
(
y The Process Computer Shutdown RWM sequence was updated to
[
reflect changes to the control rod patterns.
C.
Reason for Change:
This update was made to assure that the RWM would enforce proper group notch operation during shutdown.
h D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of
)
[
occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer No.
The update enables the Process Computer to enforce the proper RWM sequence, augment.ing the Rod Sequence Control System
(
reactivity worth control as described in the UFSAR.
ii)
Does this modification create the possibility for an
~
accident or malfunction of a different type than any evaluated previously in the safety analysis report?
h Answer:
No.
The update does not change the scope or function of the Process Computer RWM sequence.
The update merely revises the RWM
{
control rod sequence to reflect the current control rod pattern.
iii)
Does this modification reduce the margin of safety as
[
defined in the basis for the Technical Specifications?
Answer:
No.
This update enables the RWM to enforce
{
the proper control rod sequence, reducing control rod worth to minimize the effect of a drop accident as defined in the bases for Technical Specifications.
c 74 I
UNIT 3
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L 75 r
Peach Bottom Atomic Powar Station Unit 3 l
Docket No. 50-278 Annual Plant Modification Report Rnplacement of Core Spray Relief Valves f'
Modification No.:
0421 i
A.
System:
Core Spray B.
==
Description:==
The Crosby Style JMC-C-E relief valves were replaced with Lonergan Style LCT-20 relief valves.
The Lonergan valve was determined to be an acceptable, equivalent, substitute.
{
C.
Reason for Change:
The Crosby relief valves and replacement parts are difficult to procure and the Lonegran valves are used at the Limerick Generating Station.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previ.ously evaluated in the safety analysis report?
Answer:
No.
Lonergan valves tueet the same design specifications as the Crosby valves.
They both perform the identical functions in the same manner.
r L
11)
Does this modification create the possibility for an accident or malfunction of a different type than any e
evaluated previously in the safety analysis report?
l' Answer:
No.
The Lonergran valves were determined to be equivalent to the Crosby valves, so than
[
installation of the newer valves does not impact system operation or safety analysis, lii)
Does this modification reduce the margin of safety as r
defined in the basis for the Technical Specifications?
Answer:
No.
Neither the Technical Specifications nor the bases addresses the relief valve design.
The performance or reliability of the system is not reduced, and no margins of safety are reduced.
I 76
Peach Bottom Atomic'Powar~ Station r'
Unit 3 t
Docket No. 50-278 Annual Plant Modification Report
()
Modification to the Safety Grade Air Supplies Modification No.:
0625 A.
System:
Instrument Air and Instrument Nitrogen B.
==
Description:==
[L This modification consists of the replacement-of check valves and I
the addition of test and vent connections.
This work is being done to improve the reliability and testability of portions of the Instrument Nitrogen System which are safety-related.
Also
['
-included in this modification is the installation of a permanent seismic pneumatic supply to the ADS valve accumulators inside primary containment.
This backup supply is being provided to
[~
assure ADS valve operability for a period of 100 days following E
an accident per NUREG-0737, item II.K.3.28.
Reason for Change:
C.
This modification will improve the reliability and testability of
-safety-related pvrtions of the Instrument Air and Nitrogen f
Systems.
rae installation of the safety grade pneumatic supply L
-will assure the long-term operability of the ADS valves.
D.
Safety Evaluation Sur.tnary,:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or
[
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
F Answer:
This modification does not increase the L
probability of occurrence or the consequences of an accident or malfunction of equipment important to-safety because the modification r
L.
will improve the reliability and testability of safety-related portions of the Instrument
~
Air and Nitrogen Systems.
This modification will not adversely impact the operation the these systems nor will it alter any Chapter 14 safety analyses.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification does not create the possibility for an accident or malfunction of a different type than evaluated previously because this modification will improve the reliability and testability of the safety-related portions of the Instrument Air and
Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report' Nitrogen Systems.
This modification will not adversely impact these. systems nor'will it alter any Chapter 14 safety analyses.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
The margin of safety as defined in the basis for the Technical Specifications is not reduced because this modification will improve the reliability and testability of the safety-related portions of the Instrument Air and Nitrogen Systems, p
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78 f
Paach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Installation of Heat Exchanger Lifting Beam Modification No.:
1181 A.
System:
Cranes, Hoists, Elevators, Rigging Equipment E.
==
Description:==
A lifting beam, consisting of two beams attached to the existing building steel above the "C" and "D" heat exchangers was installed.
C.
Reason for Change:
The beam permits the performance of maintenance on the heat exchanger during power operation and while maintaining secondary containment integrity.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or
(
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The safety analyses in the UFSAR reveal that the unit can be brought to a safe shutdown condition without one of the heat exchangers.
Neither the installation and use of the lifting beams nor the failure thereof will impact the safety analyses.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No. Only the failure of the lifting beam will adversely impact nuclear safety-related equipment.
The plant and system designs incorporate dual and redundant shutdown paths.
The elimination of a heat exchanger due to a beam failure will not result in an unanalyzed condition.
f I
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79 r
l
Peach Bottom Atomic Power Station-Unit 3 Docket No. 50-278 Annual Plant Modification Report lii)
Does-this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
. Answer:
No.
Operation or failure of the beam would render only a single heat exchanger inoperable; however, this remains within the design bases since only three RHR loops are I
required.
Furthermore, implementation permits performance at-power maintenance without breaching secondary containment, which enhances the overall margin-of-safety.
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80
Peach Bottom Atomic Powcr Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Installation of LP Turbine Pre-warming Monitoring System Modification No.:
1236 A.
System:
Turbine Generator
.B.
==
Description:==
Four thermocouples were installed in the low pressure (LP) turbines.
Two were installed in the 12th stage diaphragm, generator end of
'A' LP turbine, and two were installed in the 12th stage turbine end of 'C' LP turbine.
Thermocouple wiring was run through the turbine casing to the external side of the casing, through conduit to a chart recorder in the control room.
C.
Reason for Change:
The purpose of the modification is to provide a pre-warming monitoring system for the
'A' and
'C' LP turbines.
D.
Safety Evaluation Summary:
i)
Does this modification increase _the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because the monitoring system will not affect the operation of the low pressure turbines.
The system will allow turbine temperatures to be monitored during startup to ensure that they do not drop below 125 degrees F.
With data confirming startup temperatures, the inspection period for the turbine wheels can be extended and the replacement of the wheels delayed.
This modification has no affect on any of the previously evaluated accident analyses, 11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, the modification simply monitors low pressure turbine temperatures.
It cannot affect the temperatures.
The modification utilizes existing recorders, and no additional loads are added to the electrical buses.
Therefore, this modification does not create f
the possibility for any different type of I
accident or malfunction.
81
Peach Bottom-Atomic Power Station Unit 3' Docket No. 50-278 Annual Plant' Modification Report lii)
Doe.3 this-modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, this modification has no~ affect on the operation of the low' pressure turbines.
The monitoring system is not required or assumed to be operable by the Technical I
Specifications.
Its function is merely to
-record startup temperatures for inspection scheduling.
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82
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Peach Bottom Atomic Power Station Unit'3 Docket No. 50-278-Annual Plant Modification Report
.[
l
~ Alternate High Pressure Service Water Alternate Flow Pattern Modification No.:
1243 A.-
System:
High Pressure Service Water (HPSW)
B.
==
Description:==
Flanges sealed with blind flanges were installed on the HPSW
/
discharge line of the Residual Heat Removal (RHR) heat exchangers for both loops.
An 8" diameter hose will be connected to the flange when the alternate flow path is in use.
The alternate HPSW flow path will be used only during Cold Shutdown and when the fuel cask or irra6iated fuel is not being moved in the reactor building.
C.
Reason for Chance:
This change was made to supply an alternate discharge flow path for the High Precsure Service Water to remove decay heat from the reactor in the event that the block valves or common HPSW line are removed f rcm service.
D.
Safety Evaluation 9ummary:
t i)
Does this modification increase the probability of occurrence or the conse.cuences of an accident or malfunction of equipment ~ important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The system design accounts for the possibility of flooding in the event of a HPSW line break.
Room flood alarm and a second RHR heat exchanger loop are available should a flood occur.
Therefore, this modification does not increase the probability of occurrence or the consequences of an accident or malfunction.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
This modification provides an alternative flow path for HPSW to maintain shutdown cooling capability of the RHR heat exchangers.
The design meets the appropriate FSAR I
I requirements to protect against RHR room flooding.
Furthermore, the alternate HPSW flow path is to be used only during Cold Shutdown and when the fuel cask is irradiated, fuel is not being moved in the reactor building.
83 b
- Peach'. Bottom. Atomic Power Station Unit 3-Docket No. 50-278
^
Annual Plant Modification Report Therefore, this modificatior.does'not create the. possibility of a new accident or malfunction.
- il i-). Does this modification reduce the margin of. safety'as defined in the basis'for the Technical Specifications?
y Answer:
.No.
Neither the function of-the HPSW system
[
has been cha.nged, nor will secondary.
containment be breached with the installation or operation of this modificatior.in cold shutdown.
Therefore, no margins of safety as defined in the bases of the Techn'ical Specifications were reduced.
)
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84
4 Peach Bottom Atomic Power Station-Unit'3' Docket No.' 50-278 Annual Plant Modification Report Install'a Personnel Door in the Unit 3 Railroad Door
--Modification No.:
1609 1
A..
System:
Security Syst'em B.
Description:
The modification involves installing a 3' x 7' personnel door in the interior railroad door on Unit 3 providing access to the equipment airlock from the Reactor Building.
This door will be part of the Secondary Containment and will be provided with a blue ligat system.
C.
Reason for Change:
The purpose of the modification is to provide access to the equipment airlock tiithout opening the railroad doors.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
I Answer:
This modification does not increase the probability of occurrence or the consequences I
of an accident or malfunction of equipment important to safety because the installation
)
of the personnel door has been 0-listed and j
was performed in accordance with appropriate procedures.
No safety related equipment would j
be adversely affected and no safety analyses as previously evaluated in Chapter 14 will be E
affected by this modification.
Appropriate measures will be taken to ensure that Secondary Containment integrity will be i
maintained, li)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
L Answer:
This modification does not create the possibility for an accident of malfunction of a different type Decause the installation of the personnel door was done in accordance with the appropriate quality procedures.
No safety related equipment would be adversely affected by the installation of the door.
Appropriate measures will be taken to ensure that secondary containment is maintained.
(
85
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t
Paach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
. Answer:
The modification does not reduce the margin of safety as defined in the basis of the
(
Technical Specifications because the access door is not disccssed in the Technical Specifications.
All appropriate quality procedures were utilized in the installation of the door thus ensuring that no safety 5
related equipment would be adversely affected by this modification.
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86
Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Replacement of Control Rod Drive Flow Control Valves Modification No.:
1620 A.
System:
Control Rod Drive B.
==
Description:==
The flow control valves of the control rod drive system were replaced.
C.
Reason for Chance:
Old valves were damaged beyond repair and identical replacements were unavailable.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because the modification did not affect any safety related equipment.
The replacement valves are of equal or better quality than the previous valves, and their function remains unchanged.
The operation of the control rod drive system was not affected.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, the modification involved non-safety related equipment which is not addressed in any of the accident analyses.
No design or piping configurations were changed.
A review of the previous evaluation indicates that associated piping is non-seismic; and because the replacement valves are approximately the same weight, no piping reanalysis was required.
Therefore the possibility for a different type of accident does not exist.
iii)
Does this modification reduce the margin of safety as defined in the basis for thc Technical Specifications?
Answer:
No, because flow control valves are not
[
addressed by the Technical Specifications, and
(
the operation of the control rod drive system was not adversely impacted.
The modification did not affect the intended function or l
(
i 1 M
Patch Bottom Atomic Powar Station
~
~
Unit ~3 Docket No. 50-278
' Annual' Plant Modificatic.1 Report-functional testing of the' valves, and all
- f. -
- replacement materials and components were of L-equal or better quality.
3
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r 88
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r Peach Bottom Atomic 1 Power Station 2
Unit-3 Docket No.,50-278-g 3
Annual Plant Modification' Report Parmanent Removal of'Four Snubbers from Main Steam
- Safety Relief Valve Discharge Lines Modification No.
1768
-.'A.
- System:'
Hydraulic Snu!bers f3 B.
. Descriptlon:
J
~ Two snubbers from.each the 71B and 71F-SRV lines were permanently removed.
j; C.
-Reason for Change:-
Other pipe support. modifications required the size of the subject r
snubbers to be reduced.-
Upon evaluating the effect of the size change, it was determined that the snubbers were not required.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of-occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
Removal of the snubbers will not result in postulated piping stresses greater than the established allowable limits which are assumed
)
in the UFSAR.
(
11)
Joes this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The snubbers would not significantly contribute to the support of the discharge piping and the piping stresses would remain within the allowable limits.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
=
Answer:
No.
The piping stresses will remain within L
the allowable limits, so the unit's overall margin of safety is not impacted,,
89
Pasch Bottom Atomic Powar Station Unit 3
^
Docket'No. 50-278 Annual Plant Modification Report
. Fuel Pin' Puncturing Process Area Exhaust Line
(
-Modification No.:
83-159 A.
. System:
Ventilation Bs
~
Description:
LA 2-inch copper pipe was installed from the fuel pin puncturing station in the' fuel pool area, to the Standby Gas Treatment
- System.(SBGTS).
A 3/4-inch manual block _ valve (MK-130) was supplied for the copper pipe at the connection to the SBGTS piping, to isolate the line when not in use.
C.
Reason for Change:
This line provides a direct exhaust path to the SBGTS uo eliminate frequent airborne radioactive releases to tae refuel floor _ area during the fuel pin puncturing process.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously
. evaluated in the safety analysis report?
Answer:
No.
The installation of this exhaust line does not adversely effect the operation or
[
function of the SBGTS.
Should a line break occur during operation, the exhaust line can be manually isolated from the SBGTS at the f
block valve, 11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The exhaust line is designed specifically l
for use during fuel pin puncturing operations only; the line is blocked at all other times.
Should the line break during fuel sipping activities, the SBGTS would actuate on high L
radiation level, which is a condition bounded by the UFSAR.
90
Peach Bottom-Atomic Powar Station Unit 3 Docket No. 50-278 Annual Plant Modification Report 111)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
~
-Answer:
.No.
Although the SBGTS is addressed in the Technical Specifications, the installation of this exhaust line is not covered within its scope.
This modification does not alter the
'+ '
operation or safety function of the SBGTS.
It merely provides a direct exhaust path from the fuel sipping area to the SBGTS to reduce airborne activity levels of the fuel pool area during fuel pin puncturing process.
i 91 L
Patch Bottom Atomic Powar Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Installation of Acoustic Sensors to Scram Pilot Valves Modification No.:
84-018 A.
. System:
Control Rod Drive B.
==
Description:==
A Piezoelectric-crystal sensing element was installed on the instrument control air tube batween the scram pilot solenoid valves on each of the cont:ol rad drives.
C.
Reason for Change:
The sensing elements were installed to allow monitoring of the scram pilot solenoid valves.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the. safety analysis report?
Answer:
No, because this modification is not safety related even though it involves the attachment of sensors to safety related equipment.
The sensors will be used only to verify the operation of the individual scram pilot valves on the control rod drives.
The sensors will not affect the operation of the valves.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, because the sensors' weights (less than 50 grams) are small enough not to adversely affect the seismic evaluation for the control rod drives.
In addition, the sensors require no electrical power.
During the periodic monitoring of the operation of the solenoid valves, power will be supplied from local station light and will not adversely affect panel loading.
Since the modification has no affect on the operation of the valves, the
(
possibility of a different type of accident is not created.
(
iii)
Does this modification reduce the margin of safety as i
defined in the basis for the Technical Specifications?
92 b
Pa ch Bottom' Atomic Powar Station
. Unit 3' Docket No. 50-278-Annual Plant Modification-Report 4,
~ Answer:
No, because the modification does not affect the' operation.of function:of any safety h.'
Specification basis.
The sensors merely related equipment considered in the Technical
. verify the operation of the individual scram pilot valves on the control _ rod drives.
f 5
93
Peach Bottom Atomic Power' Station Unit'3 l
l5 Docket No. 50-278 l
Annual Plant Modification Report Replacement of Equalizer Valve on Control Rod Drive Differential
' Pressure Transmitter DPT-3-3-218 f
Modification-No.:
86-027 A.
System:
Control Rod Drive
.B.
==
Description:==
1 This modification' replaces the "Hoke" manufactured equalizer valve (RTV-3-3-86) on Control Rod Drive differential-pressure transmitter DPT-3-3-218 with a valve manufactured by Dragon.
Although.the valve is not Q-listed, the sensing line for reactor pressure is common with equipment which is Q-listed and would affect the Q-listed component.
C.
. Reason for Chance:
The existing "Hoke" manufactured valve, which was leaking, is no longer being manufactured.
Therefore, the valve must be replaced with a suitable valve.
In this case, a suitable replacement was determined to be manufactured by Dragon.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification will not increase the probability of an accident or malfunction of equipment as previously evaluated in the safety analysis report.
The new valve meets or exceeds the specifications of the existing valve except for the pressure rating.
The existing valve is rated for 6000 psi and the new valve is rated for 3600 psi.
The nominal pressure in the drive water header is 1300 psi and the maximum discharge pressure for the CRD pump is 1600 psi, therefore, the new valve is sufficient for this use.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
h Answer:
This modification does not create the possibility of an accident or malfunction of a different type than previously evaluated in
(
the safety analysis report.
The new valve will serve the same function as the existing valve and meets or exceeds the specifications 94
.Psach-Bottom' Atomic'PowariStation
- n Unit 3 s
Docket No. 50-278-Annual Plant Modification Report i
of the. existing valve except for the pressure
. rating, 111)
Does thisLmodification reduce the margin of safety as defined in'the basis forothe Technical Specifications?
Answer: '
4'he replacement valve meets or. exceeds the f,.
specifications.'of the existing valve.except
'for the pressure rating.
This valve is not addressed.n the Technical Specifications.
r L
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1
)
95
P Ech Bottom Atomic Powar Station Unit 3 Docket No. 50-278 Annual Plant Modification Report By-passing Rod 10-47 on Unit 3 Rod Sequence Control System Modification No.:
86-138 A.
System:
Rod Sequence Control System (RSCS)
B.
==
Description:==
A jumper was used to indicate to the RSCS that control rod 10-47 was in the fully withdrawn position while it was fully inserted.
The associated hydcaulic control unit was electrically disabled, to prevent inadvertent control rod motion.
C.
Reason for Change:
This temporary modification was installed to permit reactor startup with control rod 10-47 inoperable.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
Installation of the jumper in the RSCS is permissible if no high worth rod pattern is created and Technical Specifications are met.
The General Electric safety evaluation concluded that bypassing control rod 10-47 is within the bounds of the rod drop analysis and that no high worth will be created, 11)
Does this modification create the possibility for an ac(
int or malfunction of a different type than any
[
evaluated previously in the safety analysis report?
I Answer:
No.
Bypassing control rod 10-47 is within the i
bounds of the current rod drop analysis.
General Electric re-evaluated this analysis for omission of control rod 10-47 from the RSCS and determined that the core would remain within the bounds of this analysis.
A jumper was used, rather than the bypass switch since a review of the UFSAR indicates that the purpose of the bypass switches is to bypass rod switch indications on operable control rods, and not for bypassing inoperable ones.
96
Patch Bottom Atomic Pownr' Station Unit 3 Docket No. 50-278
~
-Annual Plant Modification Report lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
f:-
Answer:
No.
Limiting Condition for Operation 3.3.A was reviewed, and control rod 10-47 was
[.
determined to be inoperable.. Electrically i
disabling a fully inserted control. rod does not reduce the margin of safety as described in this section.
(
[
[
L
[
97
1 PSEch Bottom Atomic Pow 3r Station Unit 3 Docket No. 50-278 Annual Plant Modification Report L
h
~ Update of Shutdown Rod Worth Minimizer Sequence Modification No.:
87-022 A.
Systemt Process Computer p(
B.
==
Description:==
The RWMSEO program of the process computer was used to update the shutdown RWM sequence arrays.
C.
Reason for Change:
The modification was needed to update the RWM sequence according to the new control rod pattern.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
h Answer:
No, because this modification enables the process computer to enforce the proper rod worth minimizer (RWM) sequence, augmenting the RSCS control rod worth as described in the Final Safety Analysis Report.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
(
Answer:
No, because the modification does not change the scope or function of the process computer RWM sequence.
It only revises the control rod sequence to reflect the current control rod pattern.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, because this modification enables the process computer to enforce the proper control rod sequence to limit control rod worths which minimizes the consequences of a control rod drop accident as defined in tbr technical specification bases.
98 f
I l
l UNITS 2 & 3 3
t t
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5 N
99
PaGch Bottom Atomic Powar. Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Upgrade-of Emergency Service Water / Reactor Building Cooling Water Interface Modification No.:
0556 A.
System:
Reactor Building Cooling Water B.
==
Description:==
External reinforcement was installed on the reactor building cooling water (RBCW) heat exchanger.
Pipe supports were also added.
C.
Reason for Change:
This modification corrects design' deficiencies upgrading the RBCW heat exchanger and associated piping to seismic and nuclear safety related requirements.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of I
occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
This modificaticn was designed and installed to reduce the consequences of an accident or seismic event.
t l
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
Upgrading the ESW and RBCW systems to meet or exceed established safety criteria l
will assure that the equipment will perform as designed during an accident or seismic event.
This will also reduce the negative impact of its failure on other equipment and systems, iii)
Does this modification reduce the margin of safety as de#ined in the basis for the Technical Specifications?
Answer:
No.
The design upgrades will increase margins of safety for the RBCW and ESW systems.
t wo 1
Peach Bottom Atomic Powar Station
-Units 2 & 3
{
Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of a Lifting Device for Dryer / Separator Pool Plugs _
('
LModification No.:
0682
-A.-
System:
Dryer / Separator B.
==
Description:==
f A remotely-operated lifting device for the removal and reinstallation of dryer / separator pool plugs was installed.
.C.
Reason for Change:
The lifting device eliminates the need for a diver to remove equipment pit plugs when the plugs are under water and eliminates
(
the. possibility of dropping the pool plugs during removal or reinstallation.
{
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of r.
occurrence or the consequences of an accident or
(
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
(
Answer:
No, this modification eliminates concern for possible damage to adjacent equipment caused by dropping of the plug during removal of reinstallation.
This modification has no
[.
adverse impact on equipment important to safety as previously evaluated in the UFSAR.
(
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, this modification does not alter the function of the dryer / separator pool plugs.
It only affects the method in which they are
(
removed or reinstalled.
111)
Does this modification reduce the margin of safety as
{
defined in the basis for the Technical Specifications?
Answer:
No, the function of the dryer / separator pool r
plugs is not changed.
The furnishing and use
(
of the lifting device required by this modification does not involve safety-related equipment.
101
-Peach Bottom-Atomic Powar Station Units 2 & 3 f"
' Docket Nos. 50-277; 50-278 Annual Plant Modification Report Lifting Device Strengthening
[
, Modification No.:
0800 A.
System:
Cranes, Elevators,-Rigging Equipment, Tools B.
==
Description:==
The lifting lugs and load bearing members of the Units 2 and 3
('
reactor pressure vessel-head strongbacks, dryer separator slings, service platform lifting rings, and the Unit 2 hydraulic tensioner lifting attachment were modified to increase their strength.
C.
Reason for Change:
(-
This modification was performed to assure compliance NUREG-0612 D.-
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis-report?
Answer:
No.
Increasing the strength of heavy load
{
lifting devices will not impact accident analyses.
The modification does not involve increasing the loads which are lifted, only increasing the reliability and margin of
[.
safety.
11)
Does this modification create the possibility for an I~
accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The lcads which are lifted and their paths remain unchanged.
Increasing the capacity of the devices does not introduce new accident precursors.
/*
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
r L
Answer:
No.
The margins of safety associated with each lifting device and task are increased by the increased strength and reliability of the devices.
L L
102 w
Peach Bottom Atomic Powsr Station Units 2 & 3 j
Dockst Nos. 50-277; 50-278 Annual Plant Modification Report S'ilsmic Upgrading of Various Fans
(
Modificatioa No.:
0843 A.
System:
Ventilation f
B.
==
Description:==
This modification seismically upgrades ventilation fans in
(-
several fan systems to meet the structural requirements necessary to withstand a design basis earthquake and maximum credible ea.:hquake.
The modification involves structural reinforcement to several fan bases and installation of deflection limiters on all fans involved.
C.
Reason for Change:
The purpose of the modification is to seismically upgrade various ventilation fans to meet an FSAR commitment.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of
(
occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment
[
important to safety as previously evaluated because the modification will not affect the operation of ventilation system.
This
{
modification only upgrades the structural requirements for the fans 3.n order that they survive a seismic event.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification does not create the possibility for an accident or malfunction of r
a different type than any evaluated previously beccuse the modification only affects the L
structural reinforcement of several fans a
bases.
The operation of the fans will not be
(
altered by this modification.
iii)
Does this modification reduce the margin of safety as
[
defined in the basis for the Technical Specifications?
Answer:
The margin of safety as defined in the basis for the Technical Specifications is not
(
reduced because the modification only affects 103
.1 P3ach; Bottom' Atomic Powar Station-Unita 2 &,3
{-
Docket Nos.-50-277; 50-278 Annual Plant Modification ~ Report' the-structural reinforcement of the fans and will not. affect the operation of.the fans.
' The fans are not discussed in the Technical'-
Specifications.
L l
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(-
[
[.
[;
[
[
[
104
(
Paach Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification _ Report Induction Heating Stress Improvements Modification No.:-
0945 A.
Systems:
Recirculation and Residual Heat Removal B.
==
Description:==
Induction Heating Stress Improvements were applied to 23 welds in the non-isolatable sections of the recirculation suction piping (loops A and B) and Residual Heat Removal Shutdown cooling suction piping up to the first motor operated isolation valve in each line.
This modification also involved temporary supply of dimineralized cooling water to the selected piping systems.
C.
Reason for Change:
The induction heating stress improvements significantly reduce the susceptibility of these welds to Intergrannular Stress Corrosien Cracking.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously
(
evaluated in the safety analysis report?
Answer:
No, because the application of Induction Heating Stress Improvements (IHSI) enhances the
(
reliability of the welds by reducing their susceptibility to Intergrannular Stress Corrosion Cracking (IGSCC).
The IHSI process generates
(
permanent compressive stresses on the inside surtaces of the pipe by rapidly heating the outside surface with specially designed electric
[
induction heating coils while the inside surface is kept cool by water flowing through the pipe.
L 11)
Does this modification create the possibility for an accident or malfunction of a different type than any y
evaluated previously in the safety analysis report?
f Answer:
No, because this modification did not alter the piping configurations as described in the FSAR.
It only enhanced the reliability of the welds, f
The welds and adjacent piping areas were
(
nondestructively examined before and after the IHSI treatment.
(
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
r Answer:
No, because the reliability of the welds is L
enhanced by this modification.
No changes were 105
i; Patch-Bottom Atomic Powsr Staticn i
Units 12 &~3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report made to any_ safety related components or the Technical Specifications.
The margin of. safety was not' reduced.
(L (i
L L
[
[:
L L
F L
L r
L I
L I
7 Peach Bottom Atowic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report HFA Relay Coil Replacement l
Modification No.:
0959A A.
System:
DC System B.
==
Description:==
This modification was initiated to replace the 120 VDC HFA relays and coils, and the replacement of the 115 VAC scram contactor coils with 120 VAC HFA relay coils.
Previously, 115 VAC HFA relay coils were replaced in the RPS and PCIS system, r
C.
Reason for Chance:
This modification is in response to NRC Bulletin No. 84-02 and will improve the reliability of the RPS, PCIS and other safety-related circuits that use the GE type HFA relay.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This mcdification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety because the coils will be replaced with a coils that will perform the same function and not affect the operation of the relay.
Therefore, the system operation will not be affected by this modification.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
The modification does not create the possibility for an accident or malfunction of
]
a different type than evaluated previously J
because the operation of the relay will not be affected by the change of the coil.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specificat..ur?
Answer:
The margin of safety as defined in the ossis for the Technical Specifications is not reduced because the replacement coils will perform the same function as the previous coils; therefore, the operation of the system will not be affected.
107
-Peach Bottom Atomic' Power Station Units 2 4-3 l.
Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Thermal-Magnetic Breakers or Power Fuses in Various-AC MCC Compartments Modification No.:
1029 J and K
'A.-
System:
480V Motor Control Centers B.
==
Description:==
This modification installs either thermal-magnetic breakers or power fuses in various AC motor control center compartments to provide coordination to meet associated circuit requirements as L
specified in Appendix R to 10CFR50.
Modification 1029J covers i
common plant MCC's, and 1029K covers Unit 2 MCC's.
This modification involves either installing power fuses in safety-related electrical circuits or replacing 480V magnetic-trip only
(.--
Also, two load center breakers for safety-related MCC's are circuit breakers with 480V thermal-magnetic circuit creakers.
calibrated to revised trip settings, b
C.
Reason for Change:
The purpose of the modification is to bring PBAPS into compliance f
with Appendix R to 10CFR50.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously
(
evaluated in the safety analysis report?
Answer:
This modification does not increase the
[-
probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated because the operation of the electric system
[
will not be affected as a result of this modification.
This modification will provide additional circuit protection as required by
(
Appendix R requirements.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any
[-
evaluated previously in the safety analysis report?
Answer:
This modification does not create the
(~
possibility for an accident or malfunction of a different type because the operation of the electrical system will not be affected as e result of this modification.
This
{
modification will provide additional circui t protection as required by Appendix R requirements.
[
108 I
fi Psach' Bottom Atomic Powar Station Unite 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report 1N iii)
Does this modification reduce the n.argin of safety as defined in the basis for the Technical Specifications?
Answer:
This modification doe.-not reduce the margin of safety as defined in the basis for the Technical Specifications because the modification will not affect the operation of the electrical system except-to provide additional protection to the electrical system in accordance with Appendix R requice.nents.
l lr r
(
I L
i
(
l 109 l
Peach Bottom Atomic ~Powar StEtion Units 2 & 3 L '
Docket Nos. 50-277; 50-278 Annual Plant Modification Report
~ Installation and Repair of Fire Barrier Penetration Seals Modification No.:
1110 f
A, System:
-B.
==
Description:==
Penetration seals for 3-hour fire barriers were installed, repaired or reworked.
C.
Reason for Change:
This modification is required to meet 10 CFR 50, Appendix R for protection of safe shutdown equipment.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of
(
occurrence or the consequences of an accident or i
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, the reliability of the fire barriers is improved by ensuring that the fire resistance rating of the penetration seals is equivalent to that of the barrier.
Shutdown systems requiring a 3-hour fire resistance rating were identified by the Peach Bottom Safe Shutdown Report.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, this modification does not result in any unanalyzed conditions.
Upgrading the fire resistance rating of fire barrier penetration seals does not have an adverse impact on any safety-related equipment.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, this modification enhances the reliability of the fire barriers which consequently enhances the reliability of the equipment necessary for shutting down the plant in the event of a fire.
The margin of safety is not
- reduced, i
110
b w
i Poach Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Lighting Improvements in Reactor Building Modification No.:
lll5A j
l A,
System:
Miscellaneous l
B.
==
Description:==
Two lighting fixtures were added, and nine existing fixtures were relocated in the Unit 2 reactor building.
One lighting fixture was relocated in the Unit 3 reactor building.
Three lighting fixtures were relocated in the
'2B' and in the '3C' core spray I
rooms.
C.
Reason for Change:
Lighting was improved in poorly lit areas of the reactor buildings.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer No.
The changes in electrical load were determined to not adversely affect the load centers and supplies.
The installation of the lights do not otherwise impact operation of the stations or the components or systems, therein.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The installation, operation or failure of the additional lights will not impact nuclear safety related equipment or systems.
iii)
Does this modification reduce the margin of safety as defined in the basic for the Technical Specifications?
k Answer:
No.
The additional lights enhance safety by better illuminating areas where operators may have to perform manual operations to respond to emergency situations and during normal operations.
Installation, operation or failure of the modified light system will not adversely impact systems or components which are described in the Technical Specifications.
111
[
PGach Bottom Atomic Powsr Station-
' Unite 2 & 3
{.
Docket Nos. 50-277; 50-278 Annual Plant Modification Report l'
Modification to the High Pressure Coolant Injection (HPCI) System Governor Control System
' Modification No.:
1268 j
A.
System:
HPCI Governor Control System B.
Descriotion:
f This is a modification to the HPCI Governor Control System so that the valves will be partially shut when steam is admitted.
This will provide a more controlled opening of the turbine control valves increasing the stability of HPCI operation.
The modification involves the installation of a check valve in a line which will bypass the EG-R hydraulic pump during the time when the turbine is not running.
This will allow control oil at reduced pressure (80 psig) to be supplied to the control valve remote servo from the auxiliary oil pump for control actuation i
during startup.
After the turbine starts, the internal oil pump in the EG-R will take over and supply the control oil at the normal pressure (400 psig).
The normal oil pressure will close the check valve and the system will operate normally.
C.
Reason for Change:
The purpose of the modification is to provide a smooth startup acceleration ramp and increased stability of HPCI operation.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
I Answer:
This modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety because the modification will only affect the HPCI System during startup.
Once normal oil pressure is established, the system will operate normally.
No safety analyses as previously analyzed in Chapter 14 is affected by this modification.
{
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification does not create the possibility for ar. accident or malfunction of a different type than any evaluated previously because the operation of the HPCI Governor Control System is only affected during 112
l- ~~
Pasch Bottom Atomic Powar Station
).
Units 2 & 3 Docket-Nos. 50-277; 50-278-Annual Plant Modification Report startup.
The modification will not affect k-normal operation.
iii)
Does this_ modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer::
The margin of safety as defined in the basis for the Technical Specifications is not reduced because the modification only affects h-the HPCI Governor Control System during startup.
The operation of the HPCI System is not affected by this modification.
t 113
Peach Bottom Atomic Powor Station j-Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report AC Power Distribution Alternative Control Stations Modification No.:
1352E uA.
System:
4 KV Circuit B.
==
Description:==
1 This modification provides alternative control capabilities for the-20A1605, 20A1806, 30A1605, and 30A1806 4KV emergency circuit breakers to assure AC power for alternative shutdown.
C.
Reason for Change:
The purpose of the modification is to bring PBAPS into compliance with requirements of Appendix R to 10CFR50.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased.
This modification does not change the emergency AC power system, as described in the UFSAR, when the transfer / isolation switches are in the Normal or Test modes.
(
Also the design basis fire does not postulate a LOCA or seismic event coincident with a fire and the coincident loss of offsite power.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the rafety analysis report?
Answer:
The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created.
This modification does not change the emergency AC power system described in the UFSAR, when the transfer / isolation switches are in the Normal or Test modes.
Also the design basis fire does not postulate a LOCA, or seismic event coincident with a fire and coincident loss of offsite power.
Test and Emergency transfer / isolation switch modes are individually annunciated in the Main Control Room.
114
Peach Bottoin Atomic Powar Station 9"
Units 2.& 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
The margin of safety as defined in the basis for any Technical Specification is not reduced.
This modification provides a means of emergency power system control for a fire in the Main Control Room, the Cable Spreading Room or the Emergency Shutdown Panel Area and does not change the normal operation of the j
system as described in the Technical Specifications.
The use of this panel in the alternative shutdown mode is not described in the Technical Specificationa.
The use of this panel to respond to an Appendix R fire is f
required to ensure safe shutdown.
10CFR50.54(X) allows departure from the Technical Specifications in an emergency such ao an Appendix R fire.
c i
(
115
Poach Bottom Atomic Po'ar Station w
Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Replace Safety-Related Rosemount 1151 Transmitters
- with Rosemount 1153 Transmitters Modification No.:
1419 A.
. System:
Reactor Protection System and Residual Heat R1moval System B.
==
Description:==
This modification replaces the Rosemount 1151 transmitters on Units 2 and 3 with Rosemount 1153 transmitters.
C.
Reason for Change:
Th_s modification is necessary because the equipment qualification life expectancy of the transmitters is due to expire.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety because the 1151 transmitters vill be replaced with 1153
)
transmitters which have been determined to
[
perform on an equal basis.
The operation of the RPS System or RHR Systera is not affected by the modification.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
The modification does not create the possibility for an accident or malfunction of a different type than evaluated previously because the replacement transmitters (1153) will not alter the operation of the RPS or RHR Logic.
The replacement transmitters have been determined te be an equal replacement.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
This modification does not reduce the margin of safety as defined in the besis for the Technical Specifications because the RPS or RHR Logic will not be affected by this 116
h Peach Bottom Atomic Pown: Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report l
i modification.
The Rosemount transmitters are not discussed in the Technical Specifications.
t
[
[
l '. 7
Peach Bottom Atomic Power Station ~
Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report i
Rocctor Water Level (Yarway) Instrument-Leg Reroute Modification No.:
1457 A.
System:
Plant Protection Instrumentation B.
Descriotion:
This modification removes the Yarway temperature compensated reference column, and reroutes the associated reactor level measurement lines through different drywell penetrations, minimizing the elevation drop.
Four independent pressure compensated instrument systems will be installed to reduce the effects of high drywell temperature on level measurement.
This modification involves installing four independent pressure compensated systems, re.noving the Yarway temperature compensating column, and rerouting the instrument lines through a dlfferent penetration.
C.
Reason for Change:
The purpc3e of the modificatic, is to minimize che effects of high drywell temperature on el indication.
D.
Safety Evaluation Summar" i)
Does this modificacion increase the probability cf occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis report is not increased.
This modification improves the reactor water level measurement for both tne wide-range and the fuel-zone ranges.
The pressure compensation equipment replaces the Rosemount trip units with Foxboro equipment that maintains or increases the level of reliability of the previous equipment and provides the same actuation functions as the previous equipment.
The pressure compensation increases the accuracy of the coatrol room water level indication for pressures less than the normal 1000 PSIG operating pressure, provides the operator with correct level indication, and reduces the need for the operator to interpret the indications at lower reactor pressures.
This reduces the probability of occurrences of an accident or malfunctions of equipment important to safety.
118
Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 I
Annual Plant Modification Report t
t -
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
The possibility of an accident or malfunction of a different type than any previously evaluated in the safety analysis is not created.
This modification only adds pressure compensation to the reactor water level measurement, changes the type of equipment used for wide and fuel-zone range reactor water level measurements, and reroutes the level measurement instrument sensing li.es.
The new equipment provides the same actuation functions at the same levels as the previous equipment and increases or maintains the same level of reliability as the previous measurement equipment.
The piping reroute does not create the possibility of an accident or malfunction of a different type than any previously evaluated since it merely reroutes existing instrument piping.
iii i Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
The margin of safety as described in the Technical Specifications is not reduced.
The existing actuations at levels 8, 2,
1 and 0 are all maintained.
With the pressure compensation, the reactor water level generated actuations will occur near the
[
designated setpoints for all operating reactor pressures instead of only at the calibration conditions as occurs with the present design.
The reliability of the new instruments meets the reliability goals established in the bases for the Technical Specifications.
119
Peach Bottom Atomic Power Station j.
Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report l
Installation of Inboard Main Steam Isolation Valve Rigging Pick Points
[
Modification No.:
1544
'A.
System:
Structural (General)
B.
==
Description:==
This modification installs new pick points required to remove the various components of the inboard MSIV in the Unit 2 drywell.
These pick points have been designed to meet the single failure proof requiremente of NUREG-0612.
The modification involves welding several steel support plates threaded to accommodate various rated safety hoist rings at elevations 154' - 10,5" and 135' - 00" in the drywell.
A structural tube will be installed for one of the pick points.
Several existing structural beams will be reenforced with plating.
The lifting rings presently attached to the MSIV air cylinders and bonnets will be removed and replaced with 4,000 pound safety hoist lifting rings.
C.
Reason for Change:
The purpose of the modification is to ensure that rigging and pick points used for lifting the inboard MSIV's comply with the requirements of NUREG-0612, Control of Heavy Loads.
D.
Safety Evaluation Srmmary:
i)
Does this modification increase the probabJlity of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the l
probability of an accident or malfunction of equipment important to safety as previously evaluated because the modification will in no way affect the operation of the MSIV as previously evaluated in Chapter 14.
This modification will ensure the safe movement of the MSIV when necessary.
The previous lifting lugs and pick points do not comply with the single failure proof criteria of NUREG-0612.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification dces not create the possibility of an accident or malfunction of a different type than evaluated previously because this modification only installs pick points wqich will be utilized in the movement 120
Peach Bottom Atomic Power Station-
- r.
Units 2.& 3 Docket Nos. 50-277; 50-278 Annual ~ Plant Modification Report of the MSIV when necessary.
The operation of the MSIV will not be affected.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
The modification does not reduce the margin of F
safety as defined in the basis of the Technical Specifications because the installation'of the pick points is not discussed in any Technical Specification, nor will it affect the operation of the MSIV.
l 121 l
h t
Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report
' Additional Process Computer Alarm Typers Modification No.:
1939 A.
System:
Process Computer B.
==
Description:==
One additional printer was provided for the process computer for each unit.
The new printers are located in the cable spreading room.
The process computer alarms were segregated into high_and f-low priority categories.
The lower priority alarms were reprogrammed to print out in the cable spreading room.
This modification also reinstates the audible alarm on the alarm
/
typer.
C.
Reason for Change:
This modification provides two additional printers for the process computer in order segregate the h!gher priority _ alarms from the lesser priority alarms and thereby providing more pertinent information to the operators.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because the equipment affected is not safety related.
In addition, this modification allows the process computer to provide more pertinent information to the operators by screening the lower priority alarms.
Previously, the alarm typer printed out those alarms not annunciated on the main control board.
The audible alarm connected to the printer annunciated each time an alarm was printed.
This alarm had been disconnected because it was a nuisance to the operators.
Reinstating the audible alarm for only those high priority alarms enhances the effectiveness of the operators.
ii)
D0es this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, because the significant alarms which require rapid operator response are still annunciated on the panels in the main control room.
This modification does not affect any safety-related equipment or circuits.
The modification does not affect any automatic 122
Peach Bottom Atomic Powar Station-Units 2 & 3 Docket Nos.-50-277; 50-278 Annual Plant Modification Report g
actuations which result from parameters reaching'their alarm setpoints.
iii )
Does this modification reduce the margin of safety as defined in the basis for the Technical Specif'. 4ons?
Answer:
No, the function of the process computer alarm f
is not addressed in the Technical Specifications.
This modification enhances the effectiveness of the operators by screening the information provided to them by the process computer alarm typer.
f L
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123
)
4 Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report g
? Installation of Cyclone Separetors to the RBCCW Pump Seals Modification No.:
83-145 A.
System:
Reactor Building Closed Cooling Water (RBCCW) Pumps B.
==
Description:==
This modification installs cyclone separators on the seal injection lines of the RBCCW pumps.
The separators are manufa tured by Dorr-Oliver Inc. and have been provided in a predesigned kit package from the pump manufacturer, Ingersoll-Rand Co.
The modification involves removing'the existing seal injection lines, installing the separators, and routing the drainage of the separators to an existing tap on the pump's casing suction side.
In addition, local pressure indicators will tua installed on the seal injection lines.
C.
Reason for Change:
The purpose of the modification is to ensure that the seals in the RBCCW pumps remain free from dirt and rust.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluate in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence of an accident or malfunction of equipment important to safety because the RBCCW pumps and piping are not Q-
[-
listed; therefore, their failure would not
\\-
alter the performance of safety-related equipment as previously analyzed in the
/
Chapter 14 safety analysis.
The separators
(
will increase the lifetime of the RBCCW pump seals.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification will not create the possibi:.ity for an accident or maltunction of a different type than evaluated previously because the failure of the RBCCW pumps would not create an accident not previously evaluated in Chapter 14.
The RBCCW pumps and the separators are non 0-listed.
124
Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Ar.nual Plant Modification Report
(
.i i i )
Does this modification reduce the margin of safety as defined in the basis for the' Technical Specifications?
k
- Answer:
The addition of the cyclone separators on the seal' injection lines of the RBCCW-pumps, does not reduce the margin of safety as defined in the basis for the Technical Specifications.
The addition of the-cyclone separators does not affect any Technical Specification i
125
1 u
Peach Bottom Atomic Power Station-Units 2 & 3-Docket Nos. 50-277; 50-278 10 Annual Plant Modification Report
~
LInstall Improved High Pressure Coolant Injection (HPCI)
Mnchanical Overspeed Trip Delay
' Modification No.:
84-005 A.
System:
HPCI Turbine B.
==
Description:==
The purpose of this modification is to install an improved HPCI turbine mechanical overspeed trip assembly with an assembly'of improved materials.
The modification replaces the existing
' tappet assembly with a redesigned assembly from Terry Corp.
This modification was recommendeo by SIL 392.
The HPCI turbine overspeed trip was successfully tested at the manufacturer's facility.
C..
Reason for Change:
This modification will improve the operability and reliability of the HPCI System by providing an improved mechanical overspeed trip assembly.- This modification was recommended by SIL 392.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or
}
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety becaut.e the redesigned tappet assembly with improved materials will provide greater reliability to the mechanical overspeed trip assembly.
No safety analysis as previously analyzed in Chapter 14 would be affected.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
This modification will not create the possibility of an accident or malfunction of a different type than any evaluated previously because the existing tappet assembly was replaced with a redesigned tappet assembly of improved materials that will perform the same function.
The modification will eliminate the reset problem on the previous trip mechanism.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
126
l' : s t.
,,,+
l.-
kb '
Peach Bottom Atomic Power Station Units 2 & 3' Docket Nos. 50-277; 50-278-n.
p; d-Annual Plant Modification' Report
. Answer:
This modificat' ion does,not reduce the margin of safety as defined in the basic for the Technical Specifications because the. existing
[
tappet assembly was replaced with an improved tappet assembly-.which will eliminate reset problems.. The tappet assembly is.not
[
discuksed in the Technical Specifications..
5 I'
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m
Peach Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report i
Electrohydraulic Control Filter Telltale Modification No.:
86-006 A.
System:
==
Description:==
A telltale was installed on the T-box pressure indicator line upstream of the EHC hydraulic fluid pump filters.
A cap on the end of the telltale will prevent loss of fluid if the valve fails.
C.
Reason for Change:
The modification provides maintenance personnel a checkpoint for verification of system pressurization.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
Implementation of this modification permits system operation within the bounds of the UFSAR evaluation.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
The equipment impacted by the modification is not safety-related.
Its installation, operation or failure will not introduce precursors for an accident or malfunction of safety related equipment.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
This minor change in the electrohydraulic control system will not impact operation requirements as defined in the Technical Specifications.
The overall margin of safety may be increased, as the telltale will provide indication of a pressurized system to prevent attempted maintenance in that condition.
128
]
-Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Addition of Air Relief Valves at Generator Stator Cooling Standpipe Modification No.:
86-047 A.
System:
-Service Water B.
==
Description:==
Air relief valves were installed on the generator stator cooling standpipe.
The valves allow the maximum amount of air to be displaced, and break vacuum in the event the piping is drained.
C.
Reason for Change:
The standpipe gooseneck overflowed into a temporary 55 gallon drum when service water pumps were started or swapped during operation.
The air relief valves eliminate the overflow.
D.
Safety Evaluation Sumuary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
The installation of the air relief valves does not affect system operation, but will prevent the overflow of water when service water pumps are started or swapped.
It was determined that this modification will not impact equipment important to safety as described in Section 10.6 of the UFSAR.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
(.
Answer:
No.
The installation of the air relief valves does not impact system operation.
- Further,
/
the service water system is not a nuclear i
safety related system.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
Neither operation nor failure of the air relief valves will impact the operability of tha service water system or a system which is addressed by the Technical Specifications.
129
i-Peach Bottom Atomic Powar' Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report t
Installation of Seal Purge Circuit Setter Isolation Valves and Seal Purge Supply High Point Vents Modification No.:
86-095 A.
System:
Recirculation Pump and Valves
-B.
==
Description:==
This modification adds an upstream, downstream and a bypass valve for each seal purge circuit setter so that the circuit setter can be removed for maintenance while maintaining normal seal purge flow path.
This modification is non-safety related.
Additionally, high point vent valves will be installed between the circuit setters and the seal purge supply manual isolation valves for Unit 3.
C.
Reason for Change The purpose of the modification is to aid in extending recirculation pump seal life.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
This modification allows for circuit setter removal for maintenance while maintaining a normal seal purge flow path.
If seal purge r
flow were to be interrupted, seal cooling flow
(
would be supplied by reactor water through the thermal sleeve.
Recirculation pump operation can continue without seal purge.
If it was
[
decided to stop the recirculation pump, single loop operation could be utilized.
Therefore, plant safety as previously analyzed would not be adversely affected.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
If seal purge flow were to be interrupted, seal cooling flow would be supplied by reactor water through the thermal sleeve.
Recirculation pump operation can continue without seal purge.
This modification does not create the possibility of an accident or malfunction not previously evaluated in Chapter 14.
230
Peach Bottom Atomic-Power' Station Units-2 &-3
. Docket Nos.- 50-277; 50-278 Annual Plant Modification Report lii),
Does this modification: reduce the margin of safety as
. defined'in the basis for the Technical Specifications?
Answer:.
The modification does not. reduce the margin of safety as defined in the. basis for any Technical Specification.
The Technical Specifications'.would not be changed as a-result-of this modification.
(
I
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g 131 i
Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report RAplacement of Reactor Vessel Shroud Head Bolts Modification No.:
86-117 A.
System:.
Reactor Vessel and Internals B.
==
Description:==
The shroud head (moisture separator hold down) bolts were replaced with a crevice-free design.
Seventy-nine new bolt assemblies were obtained and 34 uncontaminated bolt assemblies stored onsite were modified.
C.
Reason for Change:
The purpose of the modification is to replace the existing type bolts which were.known to be susceptible to intergranular stress corrosion cracking (IGSCO).
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because this modification makes the shroud head bolts more reliable.
The modified bolts have a split collar with vertical full penetration weld versus a collar with a horizontal weld which is susceptible to cracking due to stressing by the preloading of the bolt and contact loads during bolt latching operation.
The newly purchased bolts are also of a crevice-free design.
In addition, the new assemblies have other IGSCC mitigating improvements including heat treatment and flow circulation orifices in the bolt assembly sleeve.
The design of new bolts and the modified uncontaminated bolts eliminates the crevice which was the site of the previously identified shroud head bolt failures.
Because the reliability of the bolts is increased, the probability of any previously evaluated accidents is not adversely affected.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No, because the new design of bolts is more reliable and performs the same function as the previous design.
For the modification of the 34 uncontaminated bolts, no load bearing 132
6:
Peach Bottom Atomic'PoWor Station i
Units.2 &-3 j
Docket Nos. 50-277; 50-278
' Annual Plant Modification Report
)[
lt ti material was removed.
The newly purchased' bolt assemblies are the standard GE design supplied with the BWR/6 reactor assembly.
O..
Therefore, the possibility for a different type.of accident is not created.
'111)
.Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, there are no technical specifications related to the shroud head bolts or moisture separator assembly.
The function of the bolts is unchanged.
Only the reliability is enhanced.
Failure of the bolts has already been analyzed and realized.
The new design will not reduce the margin of safety, f
o
[
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{
133
Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Instrument Nitrogen Bypass Lines Modification No.:
87-020 A.
System:
Instrument Nitrogen System B.
_D,escription:
A bypass line with a check valve for each of the backup Automatic Depressurization System solenoid valves (SV-8130A&B and SV-9130A&B) was installed.
C.
Reason for Change:
The bypass lines provide a means of supplying the solenoid valves with an adequate nitrogen supply during a station blackout or a design basis fire.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No.
In the event of a station blackout or a design basis fire, the TRIP procedures provide guidance which will ensure the availability of the minimum equipment to achieve a stable shutdown.
Neither the use of the equipment as prescribed by these procedures nor its presence will increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer:
No.
In the event of a station blackout or design basis fire, the TRIP procedures provide guidance to ensure the availability of the
[
equipment required to achieve a stable shutdown.
Although use of the bypass line will disable the solenoid valve isolation function, the check valve will perform that function.
Operability of the safety relief valves required for safe shutdown will be maintained.
134
Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 g
Annual Plant Modification Report lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No.
Use of the bypass lines as prescribed by the TRIP procedures will ensure the operability of the minimum required equipment to achieve and maintain safe shutdown.
This s
will, in turn, result in an overall increase in the station's margin of safety by improving the ability to respond to a station blackout.
l
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135
s COMMOM i
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136
Peacn Bottom Atomic Power $tation Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Critical Equipment Monitoring System Installation Modification No.:
1265 A.
System:
Critical Equipment Monitoring System (CEMS)
B.
==
Description:==
g l
CEMS is a computer-based equipment labeling system.
Fifteen dish-type antennae were installed on the refueling floor, pump structure and water treatment building.
The computer system hardware is on the fourth floor of the administration building.
1 C.
Reason for Change:
This modification upgraded the radio communication between the CEMS computer and the hand-held terminals.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: No.
The CEMS system is not nuclear safety related.
The antennae installation and operation do not impact the design or operability nuclear safety related equipment.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer: No.
The CEMS installation and operation do not impact the operability of nuclear safety related equipment.
Existence, operation or failure of the CEMS will not impact equipment or systems which contribute to the operation or safety of the units.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No.
Installation, operation or failure of the CEMS will not impact systems described in or addressed by the Technical Specifications.
Consequently, no l
margins of safety will be impacted.
l I
137
Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Replacement of Distribution Headers on Cooling Towers Modification No.:
1371 A.
System:
==
Description:==
The redwood distribution headers on the original three cooling towers (A, B and C) were replaced with reinforced thermosetting resin pipe.
Both inboard and outboard edges of the distribution basis decks were bolted to the seal stops.
C.
Reason for Change:
The previous redwood distribution headers were leaking.
This modification eliminates the potential for structural damage and personnel hazards resulting from continuous leakage from the headers.
D.
Safety Evaluation Summary:
i)
Does this niodification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: No, because the distribution header does not affect the operation of the cooling towers, and the cooling towers are not safety related.
The resin pipe will prevent potential hazards from continuous leakage from the headers, and bolting the edges of the distribution basin decks will eliminate water leakage from the basins.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
[
Answer: No, because replacement of the distribution headers and bolting the edges of the distribution basin decks conforms to the original design and function f
of the cooling towers.
No new failure modes are introduced.
Therefore, it does not create the possibility of a new type of accident or malfunction.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No, the operation of the cooling towers is not affected by this modification.
Further, the cooling towers are not safety related, and do not 138
Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report affect any safety margins defined in the bases for the Technical Specifications.
l 139 r
\\
Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Hydrogen and Oxygen Storage Systems Modification No.:
1549C A.
System:
Hydrogen Water Chemistry B.
Descriotion:
A cryogenic storage, vaporization and gaseous hydrogen and oxygen storage system was installed outside the plant boundary north of the emergency cooling tower.
C.
Reason for Change:
This modification provides storage systems for the supply of hydrogen and oxygen which are required for implementation and maintenance of the hydrogen water chemistry system.
The installation of the hydrogen water chemistry system will prevent intergranular stress corrosion cracking by reducing the dissolved oxygen in the reactor coolant by adding hydrogen gas to the feedwater.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
/
Answer: No.
This modification invokes the installation of
(
new equipment which is physically located exterior to the station.
Existing station equipment and procedures are not affected by the installation,
(
operation or failure of the facility or system.
ii)
Does this modification create the possibility for an r
accident or malfunction of a different type than any
[
evaluated previously in the safety analysis report?
Answer: No.
The facility and system are located at a
{
distance from the station, such that no credible accidents at the facility could have adverse effects upon the station or any nuclear safety-
[
related equipment located therein in excess of any L
effects previously analyzed.
[
140 I
l-
Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No.
The hydrogen and oxygen storage facilities have been located at sufficient distance from the station so that no credible accidents could have adverse effects upon the station or any nuclear safety related equipment located therein.
This includes consideration of all accident scenarios included in the UFSAR.
i I
[
141
Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 s
Annual Plant Modification Reporr Replacement of Control Room Ventilation Radiation Monitor Valve Modification No.:
1599 A.
System:
Radiation Monitoring B.
==
Description:==
The control room ventilation radiation monitor valve (SV-0760B) was replaced.
The original valve model, circle seal #SV415-9032, is no longer manufactured.
The replacement valvo, ASCO #210-036-5F was evaluated and determined to be an equivalent replacement.
C.
Reason for Change:
The valve was replaced because it had failed.
The original valve model is no longer manufactured.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of
\\
occurrence or the consequences of an accident or s
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: No.
The Control Room Ventilation Radiation
(
Monitoring System function is not altered.
The
/
replacemer.t valve was determined to meet the same design, operating, seismic (IEEE 344-1975) and environmental qualification (IEEE 323-1974) criteria as the original valves.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer: No.
The new valve will perform the same function s
in the same manner as the original equipment.
The valve also meets all previous design requirements.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No.
The procurement, installation and testing will be controlled by the PBAPS Quality Assurance Plan and performed in accordance with the original design specifications.
Since the replacement valve
i is an equivalent replacement, a margin of safety is not reduced.
142 1
Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Site Emergency Communication System Modification No.:
1649
~
A.
System:
Miscellaneous g
B.
==
Description:==
A GTE Omni Sl-telephone switch was installed in the communications
~
room, replacing the private voice lift line facilities.
This modification also includes establishing electrical service for the system.
C.
Reason for Change:
This modification pravides a dedicated emergency telephone system.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: No.
Nuclear safety related design systems or equipment are not impacted by the installation or operation.
The accident scenarios of Section 14 of the UFSAR do not take credit for the emergency response, which will be enhanced.
Finally, no safety-related electrical load er supply is impacted.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
o Answer: No.
The design and function of nuclear safety related systems and their components are maintained.
The installation, operation or failure of the phone system will not contribute to the event or precursors of a credible accident.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No.
Installation, operation or failure of telephone system will not adversely affect equipment or systems addressed by the UFSAR.
The l
improved emergency communications which will 7
result, will increase the overall level of plant safety.
143
Peach 8v ;om Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report
-Redwaste Trash Compactor Mo,dification No.:
2029
'A.
System:'
Radwaste B.
==
Description:==
A new radwaste drum compactor for dry active waste (DAW) was installed on the 135' elevation of the radwaste enclosure in the area of the existing box compactor.
L C..
Reason for Change:
The previous drum was inoperable and irrep d.rable.
A drum compaction system is required for use with uhe procured General Electric super-compaction service.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or.
malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: No, because the replacement compactor is more reliable than the previous compactor.
The enclosed loading chamber of the compactor is evacuated by a built in fan to prevent the escape of air-borne contaminants during compaction.
The fan draws air through a roughing filter followed by HEPA filtration.
Differer.tial pressure gauges allow the operator to determine when filter changeout is required.
Used filters are dropped into the compactor drum without being touched by hand.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer: No, because this modification does not create any new failure modes.
The fan exhaust is connected to the radwaste building equipment compartment exhaust duct.
There is no adverse impact on the building HVAC system from this additional intermittent load.
The compactor is fed from the same compartment as the old compactor in 480V MCC B44, which is a non-safeguard power supply.
The load on the plant electrical system is not significantly changed.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
144
~,
?
.Petch Bottom' Atomic Powar Station Common Docket ~Nos. 50-277; 50-278 Annual Plant Modification Report Answer: No, because the: operability of the old or the new-compactor is not assumed in the basis ~ for any
' Technical Specification.
The modification involves 4
the installation of a more reliable compactor.
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f 50:
Paach' Bottom Atomic Powar Station
,W h
Common Docket.Nos 50-277; 50-278 Annual Plant Modification Report FI 1
'Rocetting of-MCC Feed Breakers to Prevent Nuisance Trippings
- Modification No.:
L75-048-t
'A; System:
480V Motor Control Centers LB.
==
Description:==
7 The purpose of.this modification is to inspect and reset the MCC
. feed breakers' thermal overload ratings and magnetic trip settings.
7C.. Reason for Change:
. This modification will provide better equipment protection and
- reduce nuisance trippings.
Bechtel. originally chose the-trip values based on nominal full load current while PECo Electrical Engineering used the actual nameplate data.
D.
Safety Evaluation Summary:
1)
Does this modification increase the' probability of occurrence-or the: consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: This modification does not-increase the probability of occurrence or malfunction of equipment important to safety because the modification only inspects and resets the 480V MCC feed breakers' thermal overloadstatings and magnetic trip settings.
These settings have bean analyzed by PECo Engineering as being acceptable for safe plant operation.
No Chapter 14 safety analyses is affected by this p
modification.
ii)
Does this modification create the possibility for an t
accident or malfunction of a different type than any
[
evaluated previously in the safety analysis report?
Answer: This modification will not create the possibility for an accident or malfunction not previously evaluated because the 480V MCC feed breakers will function as d^ signed.
The modification only effects the trip settings fot Lhermal ooerload and magnetic trip.
These settings iave been analyzed by PECo Engineering as sing acceptable for safe plant operation.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
146
77 ru g e Peach Bottom Atomic Power' Station Common Docket Nos. 50-277; 50-278 l
Annual Plant Modification Report Answer: Tnis modification will not reduce the margin of safetyLas defined in the Technical Specifications.
The trip settings have been anelyzed by PECo-Engineering as being acceptable for safe shutdown operation.
The breakers' trip settings were modified to reduce nuisance' trips and provide better equipment protection.
The trip settings are not included in the Technical Specifications.
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t 147
Psach Bcttom Atomic Powar' Station
-Common-Docket Nos. 50-277; 50-278 Annual Plant Modification Report
?Ramoval of Jumpers on Process Computer Inputs Modification No.:-
.83-152 A.
System:
ProcesJ Computer B.-
==
Description:==
The temporary jumpers on 57 process points in the isolation cabinets in the computer room were replaced with permanent wiring.
Associated electrical prints were revised to reflect this modification.
C..
Reason for Change:
During.the startup of the process computer in 1974, it was found that some sensors would not provide valid computer inputs because of ground loop induced noise.
To correct this problem, cable shields were either rerouted to the negative terminal or removed.
This is.an acceptable technique as ground loops are difficult to predict during design stages when low level, noise susceptible signals are involved.
D.
Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: No, this modification reduces ground loop induced noise by acceptable techniques.
This original modification was temporarily completed in 1974 via
(
This modification updates
}
applicable prints to reflect field conditions.
The method of redecing ground loop induced noise incorporated by this modification has proven effective where as no additional induced noise problems have occurred.
This modification does not involve cafety-related equipment.
These computer inputs are not evaluated in the FSAR nor does this modification increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety as evaluated in the FSAR.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer: No, the method used to reduce ground loop induced noise is a customarily acceptable practice.
The inputs involved in this modification are reactor feed pump turbine lube oil and bearing temperature 148 f
.)
h' Poach Bottom Atomic Powar Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report indications.
These inputs are not interlocked.
They provide temperature indication only (reactor feed pump 1 turbine lube oil temperature and thrust bearing temperature recorder TR-2492 and process computer analog inputs).
This modification does not. create the. possibility of an accident or malfunction of a'different type than.any evaluated previously in the FSAR, iii)
Does chis modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No,:the inputs to the process computer involved in this modification are not a basis for any Technical Specification; therefore, the n.argin of safety as defined in the basis for any T:chnical Specification is not reduced.
f 149
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-Paach Bott'om Atomic Powsr Station Common Docket'~Nos. 50-277;-50-278 Annual-Plant Modification Report Installation of Control Room Ventilation Ductwork Access Doors
'Modificatica No.:
85-121 A..
System:
Control Room Ventilation B.
Description:
An' access door (18"x9") was installed in the ductwork near the plenum of the control room fresh air fans.
C.
. Reason for Change:
The door was installed to provide access to facilitate cleaning and maintenancelof the fans.
1 D.- ' Safety Evaluation Gummary:
.1)
Does this modification increase.the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated _in the safety analysis report?
Answer:. No.
The modification facilitates cleaning and mainter,ance without preventing the system from operating within the bou is of the previous evaluation.
Both the design and safety evaluation of the emergency control room ventilation system remain unaffected.
ii)
Does this modification create the poss'.bility for an accident or malfunction of a different e.ype than any L
evaluated previously in the safety analysis report?
Answer: No.
The emergency control room vantilation system's operation is unchanged.
The UFSAR L
evaluations'of an accident or malfunction to main car'rol room ventilation also remain unaffected.
iii)
Does this modification reduce the_margi.i of safety as defined in the basis for the Technical Specifications?
L Answer: No.
The access dcor will not impact the operation of the emergency control room ventilation system, and not impact an associated margin of safety.
L s
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Pasch'BottomAtom[pc Powsr Station s
' Common Docket Nos. 50-277;J50-278-1 Annual Plant-Modification Report.
Improvement ~ofECardox Circuitry Indication
~
zModification No.':
86-031' A '.
Systemt.
Cardox. Fire Protection.
B '. LDescription:
!The Cardox' disarm' circuit was modified such that the. red
-indicating light is-lit directly-from a spare contact on the encrgizing relay which blocks'Cardox initiation.
C.
' Reason-for Change:
This-modificetion increases personnel safety by direct indication
- of the Cardox blocking. relay condition.
D.. Safety Evaluation Summary:
i)
Does this modification increase the probability of occurrence or the consequences of an accident or
- malfunction of equipment important to safety as previously evaluated in the safety analysis report?-
b
. Answer: No.
Changing the signal source to improve the
. reliability of an indication light does not in itself-impact accident scenarios or' consequences.
Further, the contact used f.*r the signal was a spare.
11)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously:in the safety analysis report?
t.
Answer: No.
Using a spare contact to provide a signal to an indicator is not a potential accident precursor.
The design and implementation of this modification are not nuclear safety related and only impact the indicator light.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No.
The improvement in indicator reliability will serve to enhance personnel safety, and not affect the operation of nuclear safety-related equipment or systems as described in the Technical Specifications, b
151
4 Psach Bottom' Atomic Powsr Station-Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report
_ Flow Instrumentation Taps to the Instrument Air Dryers Modification No.:
86-052 A.
System:-
' Instrument Air B.
==
Description:==
- This modification provides permanent flow instrumentation taps to the-instrument air dryers.
These taps were previously installed on a temporary basis.
The installation of-these instrumentation taps.is to provide a convenient method to determine (trend)-the
-flow through the instrument air dryers.
'C.
Reason for Change:
1This modification will allow permanent installation of flow instrumentation taps to the instrument air dryers.
D..
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: This modification does not increase the probability l
of occurrence of an accident or malfunction of equipment important to safety as previously evaluated because a failure of these instrument air taps would not affect the ability of the instrument air dryers to function because the taps are down".tream of the air dryers.
Additionally, the l
equipment is nonsafety-related.
This modification doen not affect any Chapter 14 safety analysis.
1 11)
'Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer: This modification does not create the possibility l
for an accident of malfunction of a different type L
than evaluated previously because this mod 3fication
[
does not affect the operability of the instrement air system, it only provides a convenient method to measure the flow from the instrument air dryers.
iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: This modification does not reduce the margin of safety as defined in the basis for the Technical Specifications because the instrument air dryer flow taps are not discussed in the Technical 152
Pacch Bottom Atomic Powsr Station Common Docket Nos. 50-277, 50-278 Annual Plant Modification Report Specifications and do not affect the operation of the Instrument Air System.
I 153
P@tch Bottom. Atomic Pow 3r Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Chemical Injection'Line to Auxiliary Boilers and
-Auxiliary Deaerator Modification No.:
86-073 A.-
System:
Auxiliary' Steam B.
Descriotion:
This modification relocates the chemical addition valve.
It has been determined that the present location of the valve is in an inconvenient and unsafe location.
The valve was relocated to a more appropriate location with a block valve added.
The chemical addition to either/or the deaerator or auxiliary boilers would prove more safe, convenient and improve both operation and maintenance.
This modification is nonsafety-related.
C; Reason for Change:
.The purpose of the modification is to improve areas of safety, operation and maintenance.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in :he safety analysis report?
Answer: This modification does not increase the probability of occurrence of an eccident or malfunction of equipment important to safety as previously evaluated because the modification only relocates the chemical addi. tion valve and adds a block valve.
This modification is being performed to nonsafety-related equipment.
The normal operation of associated systems are not impacted by this modification.
No chapter 14 safety analysis is affected as a result of this modification.
11)
Does this modification create the possibility for an accident or malfunction of a different type thc any evaluated previously in the safety analysis report?
Answer: This modification does not create the possibility for an accident or malfunction of a different type than evaluated previously oecause the modification only relocates the chemical addition valve and adds a break valve.
The normal operation of the system would not be affected.
This modification is nonsafety-related, j
lii)
Does this modification reduce the margin of safety as F
defined in the basis for the Technical Specifications?
154 f
Peach Bottom-LAtomic Powar Station Common Dockat Nos.'50-277;-50-278 Annual Plant Modification Report l-Answer: This modification does not reduce the margin of' safety as defined in the basis for the Technical Speci*1 cations.because the modification is being performed.on nonsafety-related equipment which is.-
not, mentioned in any basis _of the Technical-Specifications.
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155
Pecch Bottom Atomic Pow $r Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Transformer Temperature Alarm Changes Modification No.:
86-103 A.
System:
Process Computer B.
==
Description:==
The computer descriptions and seupoints of the main transformer temperature alarms were updated to improve the clarity of the description and to reflect actual transformer temperatures.
C.
Reason for Change:
The modification permits tracking of main transformer winding temperatures and oil temperatures.
D.
Safety Evaluation Summary:
1)
Does this modification increase the probability of occurrence or the consequences of _n accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer: No.
This modification permits more accurate and easily understood control room readings of main transformer temperatures.
No nuclear safety related functions are impacted by this modification.
ii)
Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
Answer: No.
This modification improves the data which is available in the control room.
The data is not nuclear safety related and it's improved format and accuracy is not a new accident precursor, iii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer: No.
The alteration of computer points associated with the main transformers does not affect any systems discussed the Technical Specifications, and therefore this modifications does not reduce any margins of safety.
156
m Paach! Bottom Atomic Power Station Common Docket 5.
50-277; 50-278 Annual Plant
,dification Report Rchlace 460 VAC Power Alarm Relay on E-4 Diesel Modification No.:
86-142 A.-
System:
Diesel Generator
)B.
==
Description:==
This is a temporary modification to replace the existing, irrepairable Westinghouse control relay with a GE control relay.
This relay monitors the 460 VAC Bus to various diesel skid mounted equipment and alarms on loss of 460 VAC power.
This modification involves removing the old relay, mounting the new relay, lug crimping and bolting the old 460 VAC feeds, and wiring the new relay coil to transformer XF7 and its contacts into existing alarm circuitry.
C.
-Reason for Change:
The purpose of the modification is to temporarily replace an alarm relay, for which a spare relay was not available in time to support the E-4 diesel outage.
D.
Safety Evaluation Summary:
i)
Does this modificaticn increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety ana'.jsis report?
l Answer: This modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety because the modification will not affect the operation of the diesel generator.
The modification is temporary and replaces an old relay.
No safety analyses as analyzed in Chapter 14 are affected by this modification, 11)
Does this modification create the possibility for an j
accident or malfunction of a different type than any l
evaluated previously in the safety analysis report?
L Answer: This modification does not create the possibility I
for an accident or malfunction of a different type than any evaluated previously because the modification will not affect the operaticn of the diesel generator.
The modification replaces an existing irreparable relay.
^
lii)
Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?
l 157
r
-e.
-Pcach Bottom Atomic Powar~; Station
. Common Docket Nos. 50-277; 50-278-Annual Plant Modification' Report Answer: This modification does-not-reduce the margin of safety.as. defined in'the basis for-the Technical
. Specifications and does not affect the~ operation of Lthe' Diesel Generator.
The relay is not discussed in the Technical Specifications.
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PROCEDURES t
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Patch Bottom Atomic Powar-Station
-Units 2 &-3 Docket Nos.-50-277; 50-278 Annual Plant Modification Report-4 Procedure No.:
GP-3, Revision 36
- A.. Procedure Titlet.
Normal Plant-Shutdown B..: Description of Change:
-The' requirement to reconnect the power feeds to the reactor head vent valves before opening the valves after a shutdown was
' deleted.
C.
Reason for Change:
The requirement to disconnect the power feeds to the reactor head
-vent valves during power operation was a temporary compensatory measure in response to an Appendix R design problem.
A permanent modification to correct the Appendix R design problem was recently completed thereby eliminated the need for disconnecting, and in turn for reconnecting the power feeds.
D.-
Safety Evaluation ~ Summary:
i)
Does this change increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because disconnection /reconnection of the reactor head vent valve power feeds is no longer of any safety significance since completion of the associated Appendix R modification.
11)
Does this change create the possibility for an accident or malfunction'of a different type'than any evaluated previously in the safety analysis report?
Answer:
No, becausa this change returns operation of the head vent valves to normal; the power feeds will no longer be disconnected and reconnected.
- Thus, no new accident precursors or types of malfunctions were introduced by this procedure change.
iii)
Does this change reduce the margin of safety as defined in the basis for the Technical Specifications?
Answer:
No, because the head vent valves are not addressed by any Technical Specifications and discontinuing to disconnect and reconnect the power feeds has no impact on any safety-related equipment or safety-related activity.
160
m Porch Bottom Atomic Powar Station Units 2 & 3 Dockst Ncs. 50-277; 50-278 Annual Plant Modification Report Procedure No.:
S.2.3.1.A, Revision 12 A.
Procedure Title-Startup of a Recirculation Pump B.
Descriotion:
Format changes were made to more clearly associate prerequisites with the applicable steps.
Also, the procedure was updated to reflect changes to the pump shaft seal purge piping (Mod. 86-95).
C.
Reason for Change:
These changes were made to improve the procedure by providing more specific operator guidance and clearer instructions, thereby reducing the possibi.lity of an operator error.
D.
Safety Evaluation Supmary:
1)
Does this change increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the safety analysis report?
Answer:
No, because these changes do not alter the method for starting the pumps; the procedure was merely improved from a human factors standpoint.
The procedural steps are still as described in the FSAR.
Thus, this procedure revision does not affect tae probability of or consequences of an accident.
ii)
Does this change create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?
No, because these changes involve items which are Answer:
beyond the level of detail discussed in the FSAR and do not contradict any of the pump startup steps specified in the FSAR.
No new procedural steps were added.
These changes do not create any new potentia' failure modes or introduce any new type of poter.tial transient.
iii)
Does this change reduce the margin of safety as defined in the basis for the Techrical Specifications?
Answer:
No, because these changes do not affect any Technical Specifications.
The startup of a recirculation p2mp is not discussed in the Technical Specit'ication Bases and these changes do not adversely impact any safety-related systems that are discussed in the Technical Specification Bases.
The thermal-hydraulic stability provisions of the Technical Specifications (3.6.E, 3.6.F) are not affected.
161 1