Text

Annual Plant Mod Rept for 1986
	ML20235W294
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1986
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20235W267	List: ML20235W263; ML20235W294;
References
	NUDOCS 8707230519
	Download: ML20235W294 (141)
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PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 '

S 1986-ANNUAL PLANT MODIFICATION-REPORT' DOCKET NOS.- 50-277;50-278 ~~ ~~ ~ '

9707230519 870714 PDR ADOCK 05000277 P PDR 1

t Docket ~Nos. 50-277 50-278-PEACH BOTTOM ATOMIC POWER STATION ANNUAL PLANT MODIFICATION REPORT 1986 Thisireport for Peach Bottom Atomic Power-Station Units 2 and 3, "

License-Numbers DPR-44 and DPR-56, respectively, is issued pursuant to the reporting requirements of 10 CFR 50.59. This report includes-

. codifications that were completed in 1986, including changes made to the facility and procedures as described in the safety analysis i report.

For each of the modifications and procedures included in this report, a summary of the safety evaluation performed indicates that an unreviewed safety question as defined in 10 CPR 50.59(a)(2) was not created in that: (i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report was not increased; or (ii) a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report was

. not created, or (iii) the margin of safety as defined in the basis for any technical specification was not reduced.

ANNUAL PLANT MODIFICATION REFORT i PEACH BOTTOM ATOMIC POWER STATION 1986 i TABLE OF CONTENTS SYSTEM Modification System Page 00021C High Pressure Service Water 70

'00257B Oily Waste System 72 l 00575 Safety-Relief Valves, Safety Valves 2 i 00625 Instrument Nitrogen 39 '

00632 Primary Containnant 73 00693 Radwarte 109 d 00821 Structural 75 00825 Core Spray, Residual Heat Removal, 76 Reactor Recirculation 00826 Primary Containment' 78 j 00829 Water Treatment ~ 80 00832 Standby Gas Treatment 111 00841 High Pressure Coolant Injection, Residual 82 Heat Removal, Condensate 00880 Main Generator 4 00964 Control Rod Drive' 83 00983 Ventilation 85 01007 Containment Atmosphere Control 41 01029B Main Steam, Ventilation '

87 01029D Diesel Generators, Emergency Service: Water 89 01029E Reactor Core Isolation Cooling, Residual 43 Beat Removal, High Prrasure Service Water, Emergency Service Water, Core Spray, Condensate Storage Tank 01029G 4kV and 480V AC Power 6 01029H 4kV and 480V AC Power 45 01029N 4kV AC Power 112 01029R Emergency Lighting DC 47 01029S Miscellaneous 91 01140 Spent Puel Pool 8 01171 Structural ,

92 Oll81A Structural 93 01187 Reactor Water Cleanup 10 01192 Reactor Motor-Generator Lube Oil 12 01243 High Pressure Service Water 14 01246 Reactor Motor-Generator Lube Oil 16 Ol351B Miscellaneous 95 01353B Residual Heat Removal, High Pressure 48 Service Water, High Pressure Coolant Injection 01353G Automatic Depressurization 50 Ol353H Miscellaneous Instruments 52 01398 13k V Transformer 114 01413A Circulating Water 116 :

01427 Structural 97 (

01533 Fuel Har.dling 54 01603 High Pressure Coolant Injection 56 01763 Reactor Water Sampling 58 ;

01774 Containment Atmosphere Dilution 98 01784 Structural 93 01850 Residual Heat Removal 60 01861 Structural 93 {

01904 Structural 93 01926 Condensate 100 01927A Diesel Generators 118 01945 Residual Heat Removal 18

Annual Plant Modification Report Peach Bottom Atomic Power Station 1986 Table of Contents (Continued)-

Modification System Page

.4 02062 Various 20

'80-007 Condensate Demineralized 22 ..

~ 8 3-0 41' Reactor Recirculation 23 83-159 Ventilation 24

84-037 Containment Atmospheric Control 120 85-008 Standby Gas Treatment 122 85-074 Reactor Recirculation 26 l 85-104 Fire Systems .

101 85-118 Condenstte Storage Tank, Radwaste 28 85-137; Circulating Water Sampling 124 85-141 Process Computer 62 l

85-145 Reactor Recirculation 102 86-001 Residual Heat Removal 30 86-003' Reactor Recirculation 63 86-008 Miscellaneous 32 86-009 Reactor Water Cleanup 33 86-019 Rod Worth Minimizer 103 86-020 Rod Worth Minimizer 35 86-021 Process Computer 36-86-032 Primary Containment 104 86-036 Rod Worth Minimizer 64 048 Rod Worth Minimizer 35 86-049 Rod Worth Minimizer 64 86-067 Rod Worth Minimizer 35 86-074 Process Computer 65 86-080 Main Generator Transformer 105 86-093 Process Computer 125 i 86-099 Condensate 66 86-105 Process Computer 107 86-109 Rod Worth Minimizer 35 86-111 Process Computer 37 l 86-113 Diesel Generator Controls 126 86-137 Process Computer 68 j 145- Rod Worth Minimizer 35 Proccdure Title Page l

HPO/CO - 18 Processing Liquid Radioactive Waste 129 HPO/CO - 18 Processing Liquid Radioactive Waste 130 S.2.3.1.A Startup of a Recirculation Pump i ST-9.17-2 Reactor Coolant Leakage Test - Unit 2 134 i ST-9.17-3 Reactor Coolant Leakage Test - Unit 3 136 1

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UNIT 2 l

Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report 1 Safety-Related Valve and Safety Valve Postion Monitoring Upgrade Modification No.:- 575 n

)

A. ' System: Safety-Relief Valves & Safety Valves B.

Description:

Y The. safety-relief valve and safety valve position indicating monitoring system has been upgraded to include a diode which prevents indicator light dimming.

C. Reason for Change:

This change was implemented to enhance the readaollity of the safety-relief valve and safety valve position indicating lights.

D. Safety Evaluation Summarv: '

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 was originally designed in accordance with NUREG-0578, to i prevent false indication of an open valve.

The additional features provide for the enhanced ability to monitor valve position.

Failure of the diode does not affect the operation of safety valves or the safety relief 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. Each valve had an existing channel of valve position indication. A diode was installed in the circuitry to prevent control room indicator lamp dimming. Failure of the diode does not affect the operation of the safety valves or the safety relief valves.

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Peach Bottom Atomic Power Station f Unit 2 ]

Docket No. 50-277 Annual Plant Modification Report Safety Relief Valve and Safety Valve Position Monitoring Upgrade (Continued)

,j lii) Does this modification reduce the margin of safety as 1 defined in the basis for the Technical Specifications? -

Answer: No, because this modification does not adversely affect the operation of the safety-relief valves and safety valves. These l changes enhance the operators' ability to accurately monitor valve position; hence, the ---

4 margin of safety as defined in the basis of the Technical Specifications is not reduced.

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report l Bnckup' Loss-of-Field Generator Protection Modification-No.: 880 A.- System: Main Generator.

B.

Description:

A backup loss-of-field relay and associated timing relay was

' installed in the existing current transformer and potential 4 transformer protection circuits of the main generators.

C. Reason for Change:

The backup relays were installed to provide redundant generator protection.

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 redundancy provided by the backup relays reduces the possibility of a plant transient without adversely affecting-any existing generator protective features or safety-related equipment. The backup relays initiate the existing auxiliary relays.

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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 main generator does not serve any safety-related purpose; thus, changes to the generator protection circuits do not affect the possibility of an accident.

Failure of the backup relays would not be a different type than previously evaluated because they are similar to the existing '

relays.

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Peach Bottom Atcmic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification. Report

. Backup Loss-of-Field Generator Protection (Continued) lii)- Does.this.modifi' cation reduce the margin of safety as defined in the basis for the Technical Specifications?-

Answer: No, because the main generator protection circuits are not discussed in the Technical Specifications bases and'this change of the protection circuits does not affect a safety-related equipment.

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Peach Bottom Atomic Power Station-Unit 2 Docket No. 50-277 Annual Plant Modification Report Trip Setting Changes for Unit 2 Load Center 480V Breakers and 4 kV Breaker Protective Relays Modification No.: 1029G

-A. System: 4 kV and 480V AC Power -

B. -Description:

a The trip settings for ten 480V breakers and three.4 kV breaker protective relays were changed to provide breaker coordination such that breakers for individual equipment wil1~

trip before the source breaker for the bus trips. The concrete dikes in front of the Load Centers 20B11 and 20B13 ---

were removed to facilitate the completion of this modification.

C. Reason for Change:

This modification was performed in accordance with the requirements of 10 CFR 50, Appendix R, to facilitate safe shutdown in the event of a design basis fire. The breaker and relay setting changes will permit continued operation of equipment necessary for safe shutdown during faults of equipment which exist on the same line feeds.

-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 new breakers allow the continued operation of equipment necessary for safe shutdown even though this equipment may be fed from a bus that also feeds potentially faulted equipment. The new breaker and relay settings do not adversely affect breaker operation under previously evaluated circumstances. Therefore, this modification reduces the probability of occurrence and consequences of an accident or malfunction of safety-related equipment.

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 l

Annual Plant Modification Report l

l Trip Setting Changes for Unit 1 Load Center 480V Breakers and 4 kV Breaker Protective Relays (Continued)

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 these changes were designed in accordance with the applicable safety-related equipments, including Appendix R, without changing the plant electrical load. In addition, the reliability of the equipment essential for safety shutdown has been enhanced without creating any new failure modes. Removal of the dikes do not violate separation requirements, have no effect on .;

plant safety, and allows the mechanical ?

removal of the Load Center breakers. e Therefore, the work can be completed without worker safety concerns pertaining to equipment de-energization. Therefore, this change does not create the possibility for an accident or malfunction of any type.

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

Answer: No. Changing the breakers and relay settings permits breaker trip coordination such that safety-related equipment may continue to 4

operate though fed by the same feed as faulted equipment, thus enhancing the reliability of equipment necessary for safe shutdown. Additionally, although the breakers themselves are not discussed in the Technical Specifications, the safety functions of their associated equipment are not adversely affected. Therefore, no margins of safety, as defined in the bases of the Technical Specifications, have been changed.

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Replacement of High Density Spent Fuel Storage Racks Modification No.: 1140 A. System: Spent Fuel Pool 1~

B.

Description:

The aluminum high-density spent fuel racks were replaced with stainless steel high density spent fuel racks.

C. Reason for Change:

The new racks are capable of withstanding higher fuel loads and will increase the previous spent fuel storage capacity.

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 storage racks are passive structures which are designed to maintain subcriticality and coolable geometry under all previously evaluated conditions. All of -

the appropriate analyses have been performed including:

o thermal-hydraulic o structural, seismic, NUREG-0612 a o spent fuel pool cooling capacity, alternate fuel pool cooling, and solid and gaseous waste generation.

o subcriticality Additionally, calculations were performed to evaluate the radiological consequences of postulated accidents, the spent fuel pool structural capacity, and shielding adequacy.

Based on the results of these analyses, it has been determined that the addition of the .)

new spent fuel racks does not increase the probability of occurrence or the consequences of an accident or malfunction of any safety-related equipment.

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Peach Bottom Atomic Power Stction Unit 2 Docket No. 50-277 Annual Plant Modification Report Replacement of High Density Spent Fuel Storage Racks (Continued)

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: No, because the new storage racks being passive structures, interact only with the pool floor and stored fuel in the same manner as the existing racks. The new racks do not - -

introduce any new or~different types of accidents or malfunctions. Existing fuel and l cask handling accidents addressed in the UFSAR remain unchanged.

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

Answer: No, because the margin of safety identified by the Technical Specifications for suberiticality is the difference between a K-effective of 1.00 and the maximum allowable K-effective value specified in the Technical Specifications which is 0.95. This margin of safety is not reduced by the new racks.

Adequate spent fuel pool water level is maintained as discussed in section 3.10 to provide sufficient shielding and cooling.

The new racks do not affect the pool water-l level, therefore, the margin of safety, as defined, is not affected by this modification.

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Removal of Pump Trip Relays Modification No.: 1187 A. System: Reactor Water Cleanup (RWCU)

B.

Description:

1 The 16A-K38A, B and C relays were removed. The purpose of (

these relays was to trip their associated RWCU recirculation- j pump when the reactor return valve, MO-12-68, was in the '

closed position.

C. Reason for Change:

'This change was made to allow the RWCU system to operate in the " condenser dump" mode with the MO-12-68 valve in the closed position.

.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 analycis report?

Answer: No, because the pump control circuit is not safety-related and the pressure integrity of the system was not affected. This change merely eliminates the need for temporary jumpers by removing the~ relays; a low flow pump trip exists to protect the pumps.

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 RWCU system is safety-related only for primary coolant system pressure boundary purposes, and the ability of the system to isolate was not affected. Thus, no r.ew types of malfunctions of equipment important to safety or accidents were created by this modification.

10

Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Removal of Ptimp Trip Relays (Continued)

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

Answer: No, because the operation of the RWCU system is not discussed in the bases for the

)

Technical Specifications and primary containment isolation capability was not affected. No Technical Specifications were affected.

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Replacement of Recirculation Pump "2A" Lower Lube Oil Level Switch Modification No.: 1192 A. System: Recirculation Motor-Generator Lube Oil B.

Description:

The motor oil level switch (LS-4501A) on the lower lube oil reservoir on the "2A" recirculation pump motor was replaced with a switch with a larger differential range.

C. Reason for Change:

The purpose of the modification is to prevent premature low level alarms which had been occurring on the "2A" motor only.

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 new level switch will perform the same non-safety function as the original switch and will not adversely affect any safety-related equipment. The large differential range of the switch will help prevent low level alarms and increase the reliability of the motor oil pump and related systems. The low level alarm point has been returned to its original design point.

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

Answer: No. Should the switch fail, the worst results would be failure of the recirculation ,

pump motor, causing a reduction of recirculation water. The reduction of )

recirculation water may cause a feed water trip due to high reactor water level and i subsequent reactor scram on low level. Since ;

this condition has been addressed in the '

FSAR, this modification does not present the possibility of an accident or failure condition different than previously evaluated.

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 Docket No. 50-277 Annual Plant Modification Report Answer: No, because this modification is not safety related; thus, the margin of safety, as defined in the bases of the Technical Specification has not been reduced. '

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report 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 High Pressure Service Water 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.

C. Reason for Change:

This change was made to supply an alternate discharge flow '

path for the High Pressure Service Water to remove decay heat from the reactor in the event that the block valves or common ,

RHR line to the HPSW discharge line are removed from 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 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.

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report

' Alternate High Pressure Service Water Alternate Flow Path (Continued) !

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

-evaluated.previously in the safety analysis report?

Answe": 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 requirements to protect against RHR room flooding. Furthermore, the alternate HPSW flow path is to be used only during cold shutdown since secondary containment will be brcached. Therefore, this modification does not create the possibility of a new accident or malfunction, iii) Does this modification reduce the margin of safety as defined in the_ basis for the Technical Specifications?

Answer: No, because neither the function of the HPSW system has been changed, and secondary containment will only be breached during operation of this modification in cold' _

shutdown. Therefore, no margins of safety as defined in the bases of the Technical Specifications were reduced.

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Peach Bottom Atomic Power Station :

Unit 2 Docket No. 50-977 Annual Plant Modification Report.

i "2A" huctreulation Pump Motor Lower Bearing Oil Reservoir Addition 1 Modification No.: 1246 A. System: Recirculation Motor-Generator Lube Oil B.

Description:

A An. auxiliary oil reservoir was installed for the existing small' capacity lower bearing oil reservoir for the "2A" Recirculation Pump Motor-Generator (M-G) set.

C. Reason for Change:

The purpose of this modification is to make up small leakage or vaporization losses and permit continued operation during the normal operating cycle.

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. auxiliary oil reservoir is not cafety related and does not adversely affect any safety-related equipment. Increasing the capacity of the reservoir will not alter the operation, function or flow path of the existing oil lubrication and cooling system, and therefore will not increase the probability of the occurrence or consequences of an accident.

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 installation of the reservoir was designed to withstand a seismic event of -

0.07g vertical and 0.659 horizontal. The design was analyzed to introduce no '

additional fire or missile hazard.

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Peach Bottom Atomic Power StGtion ~

Unit 2 Docket No. 50-277-Annual Plant Modification Report "2A" Recirculation Pump Motor Lower Boaring Oil Reservoir Addition (Continued) iii) 'Does this modification reduce the margin of safety as <

defined in the basis for the Technical Specifications?

Answer: No. Although the recirculation pumps are s addressed in the Technical Specifications, the lower bearing oil reservoir and the auxiliary oil reservoir.are not addressed.

Additionally, the auxiliary oil reservoir ensures sufficient oil supply for operation throughout the cycle. Therefore,.no margins of safety, as defined in the bases of the Technical-Specifications, have been reduced. ,,

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o Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Leak Repair of Residual Heat Removal System Valve Modification No.: 1945 A. System: Residual Heat Removal B.

Description:

An elastomeric sealant compound provided by Leak Repairs, Inc., was injected into a clamp chamber which was attached to the Residual Heat Removal (RHR) Recirculation Loop 'B' Return Valve (MO-2-10-25B) to temporarily repair a leak which developed between the 1/2-inch vent nipple and the valve ---

body.

C. Reason for Change:

This temporary repair was installed to remedy valve leakage and prevent the 1/2 inch capped nipple from separating from the valves, thereby becoming a projectile and causing a plant shutdown due to reactor coolant leakage.

l D. Safety Evaluation Summary:

i

1) 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?

l Answer: No, because the sealant compound is not considered to have any hazardous or adverse effects on the valve or other materials which it may contact. Thus, the compound and fixture will not prevent the valve from performing its safety function.

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 sealant will satisfactorily 1 withstand all anticipated plant operating conditions. The fixture prevents the nipple from becoming a potential projectile hazard.

Thus, this modification does not present the possibility of a new type of accident or failure condition.

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Pecch Bottom Atomic Power Station Unit 2 (

Docket No. 50-277 i Annual Plant Modification Report Leak Repair Heat Removal System Valve (Continued) 1 iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

,i

?.nswer:

No, because this modification does not adversely impact the function or functional testing of the valve. In addition, structural integrity / design verification stress limits of AGME Section VIII, 1983 Edition, which are comparable to the original valve design code limits, have been maintained. Thus, the margin of safety, as defined in the bases of the Technical Specifications, has not been reduced.

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P3ech Bottoa Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Linitorque Internal Wiring Replacement Modification No.: 2062 A. System: Various Systems B.

Description:

The internal wiring'for 83 Limitorque Motor Operated Valves (MOVs) was inspected and replaced with environmentally qualified wire.

C. . Reason for Change:

.This modification was implemented in response to the NRC IE Notice 86-03, which identified potential situations where environmentally qualified wiring may have been replaced by wiring which may not have the required documentation.

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 wiring in all of the Limitorque motor operators has been replaced with the required environmentally qualified wiring, which enhances the reliability of these valves even under severe accident conditions.

J 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 1

Answer: No, because replacing the MOV wiring with environmentally qualified wiring fulfills the original design requirements, assuring that these valves will be able to perform their function. This modification did not change the operation or design of the MOVs in any way. Therefore, this modification does not l create the possibility of an accident of any i kind. i 1

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i I Peach. Bottom Atomic Power Station Unit 2 L

Docket No. 50-277 Annual Plant Modification Report Limitorque-Internal Wiring Replacement (Continued) lii) .Does.this modification reduce the margin of safety as defined in-the basis for the Technical Specifications?

Answer: No,.because this' modification restores the

i. . motor operated valves to their. original i design condition, thereby assuring that=they will perform as prescribed in the Technical Specification. Therefore, no margins of safety as described in the Technical Specifications have been changed. -

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= _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _

Peach Bottom Atomic Power Station i Unit 2 Docket No. 50-277 Annual' Plant Modification Report Installation of Test Valves on Condensate Demineralized Bypass Valve Between Seat Drains Modification No.: 80-007 A. System: Condensate Demineralized B.

Description:

Two 1-inch test valves were installed on the condensate demineralized bypass valve between seat drain line.

C. Reason for Change:

This modification was installed to provide a means of checking for condensate demineralized bypass valve through leakage.

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 l related, and has no adverse affects on any ;

safety related equipment. This modification merely facilitates testing of the normally closed bypass valve for through leakage.

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 operation and function of the bypass valve has not been changed in any way due to the installation of the test valves.

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

Answer: No, because this bypass valve is not addressed in the Technical Specifications, nor does it affect any safety related equipment. Therefore, no margins of safety as defined in the Technical Specifications have been reduced.

l 22

Pasch Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Recirculation Pump Lubrication Oil Pump Logic Modification Modification No.: 83-041 A. System: Reactor Recirculation

B.

Description:

This modification involves the installation of a diode in series with contacts 5-6 of' Relay 2A-K31A in the Recirculation Pump lubrication oil pump logic.

C. Reason for Change:

The purpose of this modification is to eliminate the possibility of running the recirculation motor generator set with low-low oil pressure,'by correcting a circuit deficiency.

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 this modification does not involve any safety-related equipment. The deficiency in the logic circuit allowed an Emergency Lube Oil Pump running condition to incorrectly energize relays inhibiting any AC Lube Oil Auto-Starts. The addition of the diode prevents the incorrect actuation of these relays, allowing the pump system to function as designed. 1 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 this change allows the pump oil system to perform its function as designed; thus, no new failure conditions are introduced.

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

Answer: No. The modification permits proper operation of the system will eliminate any possible hazards present during its previously incorrect operation. Thus, the margin of safety is not reduced.

23 D

Pasch Bottom Atomic Power Station-

. Unit 2 Docket No. 50-277 Annual Plant Modification Report Fuel Pin' Puncturing Process Area Exhaust-Line Modification No.: '83-159 A. System: . Ventilation B. .

Description:

l A 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-l 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 to eliminate frequent airborne radioactive releases to the 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, because 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 1 l

line can be manually isolated from the SBGTS at the block valve.

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 exhaust line is designed specifically 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 radiation level, which is a condition bounded by the F3AR.

Therefore, no new failure conditions have been introduced.

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I Pasch Bottom Atomic Powar Station

' Unit 2 Docket No. 50-277 Annual Plant Modification Report Fuel Pin Puncturing Process Area Exhaust Line i

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

l l 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.

25

Porch Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Replacement of Transmitters Modification No.: 85-074 A. System:, Reactor Recirculation D.

Description:

The Foxboro model 611 DM recirculation drive flow transmitters (PT 2-2-Il0A,B, C, D) and recirculation pump differential pressure transmitters (DPT 2-2-Illa and B) were replaced with Rosemount models of a newer vintage.

C. Reason for Change:

One of the flow transmitters was damaged and had to be replaced. All of these transmitters were obsolete and difficult to maintain, so they were all replaced with better models.

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 new transmitters are "one-one" replacements; only minor piping and hardware changes were required to install them. The replacements perform the same function and meet or exceed all of the original specifications. In addition, the Rosemount models are more reliable and ease testing and maintenance; and experience with similar Rosemount models has indicated that they experience less drift than the Foxboro model.

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. These transmitters are not part of the primary system boundary and their output signals are constantly monitored by redundant signal comparators. Thus, the replacements which are comparable to the original ones do not create the possibility for an accident or malfunction of a different type.

26

Pasch Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Replacement of Transmitters (Continued) iii) Does this modification reduce the margin of safety as

'I defined in the basis for the Technical Specifications?

1 Answer: No, because these transmitters do not perform

.a a safety related function; they provide indication and alarm functions. However, the flow transmitters supply a signal for Average Power Range Monitor flow bias trips, but this - ---

function was improved because the Rosemounts have been shown to experience.less drift.

Thus, no margin of safety was reduced.

l 27

Peach Bottom Atomic Power Station Unit 2 Docket No. 50-2*7 j Annual Plant Modification Report Condensate Storage Tank and Floor Drain Sample Tank Tie-In Check Valve Modification No.: 85-118 A. System: Condensate Storage Tank and Radwaste B. ~

Description:

The tie-in check valve No. 229, between the Condensate Storage Tanks (CST) and the floor drain sample tank on the Radwaste System, was removed to permit chemically clean water to be pumped to the-CST when additional process time is required to reduce activity.

C. Reason for Change:

This modification was made to prevent the reprocessing of chemically clean water in the Radwaste System.

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 no safety-related equipment is -

associated with or adversely affected by this modification. This change merely prohibits the release of radioactive water to the river and redundant processing. Therefore, the probability of an accident of malfunction of safety-related equipment is not increased.

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. There is no difference in the purity of the water in the CST and floor drain tank, 1

and the increased activity level in the piping will be minimal (0.0001 uti). Thus, this modification does not create a different type of accident or malfunction than previously analyzed.

28

-Peach Bottom Atomic Power Station Unit l2 Docket No. 50-277 Annual Plant Modification Report Condensate Storage Tank and Floor Drain Sample Tank Ti_e-In Check Valve (Continued) lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because no major changes have been made to che radwaste system and no-new material or l

radioactivity has been introduced to the system. Therefore, the margin of safety, as defined in the Technical Specifications, has been been reduced.

i 1

29

Pmach Bottom Atomic Powar Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Residual Beat Removal Pump Impeller Rebuild Modification No.: 86-001 A. System:

Residual Heat Removal (RHR)

B.

Description:

Pump impellers removed from other Peach Bottom RER pumps were machined and installed in the Unit 2 'B' and Unit 2 'D' RHR pumps. The machining slightly reduced the outside diameter of the impeller eye. A specially designed, thicker impeller eye wear ring from the pump manufacturer was installed in -'

both pumps. .An pump casing volute wear ring from another Peach Bottom RHR pump was slightly machined and installed in the Unit 2 'D' RHR pump.

C. Reason for Change:

The wear ring land of the impeller eye was scored as a result of IGSCC wear ring failures. To restore the impellers.to original design specifications, the wear ring land of the impeller eye had to be machined smooth and a slightly thicker wear rings had to be installed. '

D. Safety Evaluation Summary:

Does this modification increase the probability of

1) '

occurrence or the consequences of an accident or malfunction of equipment important to safety as s previously evaluated in the safety analysis report?

Answer: No, because this modification conforms to the original design specifications of the RHR pumps such that the pump performance was not i

affected. Reducing the impeller eye outside diameter has an insignificant effect on its strength. Therefore, the pump will perform as assumed for the PSAR analyses.

30

Prach Bottom Atomic Pow 3r Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Residual Heat Removal Pump Impeller Rebuild (Continued) 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 conforms to the original pump design specifications and, thus does not create any new failure modes or accident precursors. The pumps will operate within the bounds of the FSAR analyses.

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

Answer: No, because the RHR pumps will still deliver rated flow to provide adequate core cooling in the event of a LOCA as discussed in the Bases of the Technical Specification.

31

1 1

Peach Bottom Atomic Power Station ]

Unit 2 Docket No. 50-277 i Annual Plant Modification Report i

i Installation of Lead Shielding Wall for High Pressure Service Water Radiation Monitors Modification No.: 86-008 A. System: Miscellaneous B.

Description:

A lead and steel shieluing wall was installed in the vi'ainity of the High Pressure Service Water (HPSW) radiation monitors.

C. Reason for Change:

This change was made to reduce radiation fluctuations in the area due to radwaste cask loading activities which caused transients on the monitors, thereby rendering them ineffective for HPSW radiation monitoring.

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 no safety-related equipment is associated with or adversely affected by this modification. Therefore, there is no increase in the probability of an accident or malfunction of any 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, because the design is within the safety limits of the FSAR. In addition, there are no safety-related equipment or components within the area of the wall. This modification merely reduces the spurious, high radiation signals previously detected on the HPSW radiation monitors. Thus, no new accident conditions or failure modes have been created.

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

Answer: No, because the shielding permits the radiation monitors to operate more effectively. Therefore, the margin of safety, as defined within the Technical Specifications, has been increased.

32

Pasch Botton Atomic Powar Stction Unit 2 Docket No. 50-277 Annual Plant Modification Report j Temperature Indicating Switch Replacement Modification No.: 86-009 A. System: Reactor Water Clear.up (RWCU)

B.

Description:

The temperature switch which trips the 'C' RWCU pump on high pump cooling water temperature, TIS-2-12-89C, was replaced with a new, superior model (Fenwal Model Eumber 55-101140-391) from the same manufacturer. No wiring or logic changes were made. The power requirements of the replacement are the ~ - - -

same.

C. Reason for Change:

The old switch, Fenwal Model Number 56100-2, was damaged and could not be repaired. This model is obsolete and is no longer manufactured; so a new model had to be chosen.

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 RWCU system is important to safety only for primary system pressure boundary purposes and to isolate on boron control solution injection and these functions are not adversely affected by the replacement of this switch. By replacing the failed switch with an equivalent switch, the high cooling water trip for protecting the pump is maintained as designed.

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 switch could, at l worst, allow the pump to overheat and fail causing a leak of primary system coolant which is covered by the FSAR evaluations.

Therefore, replacing this switch with a new model cannot create a new or different type of accident.

33

__________________ - _ a

Pasch Bottom Atomic Power Station -l Unit 2 l Docket No. 50-277 Annual Plant Modification Report Temperature Indicating Switch Replacement (Continued) lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. The RWCU system is important to safety only for pressure boundary purposes and'to isolate on boron control solution injection.

~

Because these safety functions are not affected by this modification, no Technical Specification margins of safety are reduced.

I 1

i 34 I

Prach Bottom Atomic Pow 2r Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Rod Worth Minimizer Shutdown Sequence Update Modification Nos.: 86-020, 86-048, 86-67,86-109, 86-145 A. System: Rod Worth Minimizer (RWM)

B.

Description:

The Process Computer Shutdown RWM sequence was updated to reflect changes to the beginning of Cycle 7 control rod patterns.

C. Reason for Change:

This update was made to assure that the RWM would enforce proper group match operation during shutdown.

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 update enables the Process Computer to enforce the proper RWM sequence, augmenting the Rod Sequence Control System reactivity worth control as described 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, because 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, because 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.

l 35

Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report.

Lead Test Assembly Program Termination Modification No.: 86-021 A. System: Process Computer B.

Description:

The Lead' Test Assembly (LTA) Program, which was utilized by General Electric to obtain data for developing process computer software packages, was discontinued. Thus, the LTA program initialization was removed from the process computer.

C. Reason for Change: --

This change was made to reduce the process computer load.

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 LTA program is not required by the FSAR and does not perform any safety function. Removing the LTA program does not adversely affect the process computer performance or functions. Therefore, the probability of an accident or malfunction of safety related equipment has not been increased.

11) Does this modification r.?v.(e the possibility for an accident c- ma; funct iem of a different type than any evaluated p:?vi;;tay in tJe safety analysis report?

Answer: he, br : gun this modification reduces the comyui "; ioad, thus improving the process computers performance and speed in computations required by the FSAR.

Therefore, this modification does not create the possibility of an accident or malfunction of any type.

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

Answer: No, because the LTA program performed no safety function, and is not specifically discussed in the Technical Specifications.

Therefore, no margins of safety, as defined in the bases of the Technical Specificati.ons, have been reduced.

36

Paach Botton' Atomic Pcwar Station-Unit 2 Docket No. 50-277-Annual Plant Modification Report Minimum Critical Power Ratio Limit Change Modification No.: 86-111 A. Systems . Process ComputerL 1B.

Description:

The minimum critical power ratio (MCPR) limits were changed in the process computer.

C. ReaEon for ChanQes Technical Specification 3.5.K.2.a and Table 3.5.K.2 require that if scram time testing has been satisfactorily completed, Option B MCPR limits must be used. Accordingly, 2000 MWD /T before the end of cycle, the MCPR limit must be adjusted.

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 this change allows the process computer to perform correct MCPR calculations in accordance with the Technical

. Specifications.

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 change was required by the Technical Specifications and has been previously evaluated. Therefore, no new types of accidents or malfunctions have been created.

)

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

Answer No, because this change is required by Technical Specifications for MCPR limits.

Therefore, the margin of safety bas not been reduced.

37

1 I

l UNIT 3 I

1 l

Peach Bottom Atomic Power. Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Replacement of Backup Nitrogen-Supply Valves for Safety-Related Valves Modification No.: 625G A. System: Instrument Nitrogen B.

Description:

The temporarily installed backup Nitrogen supply valves (SV-9130A and B) to the Automatic Depressurization System (ADS) Safety Relief Valves (SRVs) were replaced with new two-way isolation valves manufactured by Valcor Engineering Corp. (Model No. V526-5295-111).

C. Reason for Change:

The previously installed valves (manufactured by Target Rock) would not isolate in both directions, as specified by the design input document.

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 installation of the valves which satisfy the bi-directional isolation criteria increases the reliability of the Instrument i Nitrogen supply for the ADS SRVs. This modification was implemented pursuant to NUREG-0737 requirements for long-term ADS valve pneumatic supplies. Therefore, this modification decreases the probability or consequences of an accident or malfunction by increasing the reliability of the ADS valves.

The power requirement of the new valves is less; thus, there is no adverse effect on the electrical systems.

39

Peach Bottom' Atomic Power Station Unit 3 Docket No. 50-278

^

Annual Plant Modification Report I

Replacement of Backup Nitrogen Supply Valves for Safety Relief Valves (Continued)

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 affect 1 the operation of the ADS SRVs and does not affect any accident analyses. The new valves-are part of the recently installed backup Nitrogen supplies and are normally clostd.

.The upgrade of these isolation valves decreases the probability that the backup supply could have an adverse effect on the ADS I SRVs.

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

Answer: No, because the replacement of these valves increases the probability that the ADS will perform as designed without affecting any other safety equipment.

l 40

Psach Bottom Atomic Pow 2r Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Purge and Vent Valve Increased Angle of Opening Modification No.: 1007 A. System: Containment Atmosphere Control B.

Description:

The valve shaft and tapper pins on nine (each unit) purge and vent valves were replaced with shafts and pins of a higher strength steel to increase the allowable angle of opening.

Several of these valves were also reoriented to increase flow rates or facilitate local leak rate testing without valve bonnet modifications. Additionally, the valve opening angle stop nuts have been tack welded in place to prevent micadjustment.

C. Reason for Change:

Previously, inerting or deinerting of primary containment took approximately lE hours. This modification was implemented to reduce the time required to about 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

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 design of these changes and the choice of higher strength parts ensure that the valves will still isolate primary containment if an accident occurred during inerting or deinerting.

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 Fisher Controls performed an analysis which identified the valve components that required replacement and documented that the modified valves are able to close in worst-case accident conditions. Thus, the valves will still function within the bounds of the original safety evaluation.

41

Peach'Botton Atomic Pow 3r Station-Unit 3-Docket-No. 50-278 Annual Plant Modification Report Purge and Vent Valve Increased Angle of Opening (Continued) lii) Does this modification reduce the margin of safety as defined in.the-basis for the Technical Specifications?

Answer: No, the operation of these valves is not directly discussed in the bases for the Technical Specifications; however, their ability to isolate was maintained. Also, the ability to satisfy Technical Specification 3.7.A.5.b, which requires that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from entering the "RUN" mode containment oxygen concentration be 4% or less, was improved by reducing the time required to inert.

I 1

42

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification-Report

'Installati'on of Process and Diagnostic Instrumentation-Modification No.: 1029E-A. System Reactor Core Isolation Cooling (RCIC), Residuul Heat Removal (RHR), High Pressure Service Water (HPEW),

Emergency Service Water (ESW), Core Spray (CS),

Condensate Storage Tank (CST)

B. Descr'iption:

Diagnostic' instrumentation has been provided for the systems associated with'the three methods of safe shutdown, at the control room. In addition, process instrumentation for reactor I

water level reactor pressure, drywell pressure, Suppression Poy1 water level, Suppression Pool temperature, and Condensate Storuge Tank'(CST) water level for each fire area has been made available in the control room. In the event of a fire in the Turbine Building, control room indication for the CST water level and Emergency Service Water.(ESW) discharge pressure would not be available. Therefore, two complete inst'rument loops were installed for CST water level and ESW discharge pressure monitoring.

C.. Reason'for Change:

These changes were implemented in accordance with the requirements of 10 CFR 50, Appendix R, to ensure safe plant shutdown capability in the event of a design basis fire. i D. Safety Evaluation Summary:

l

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

Answer: No. The changes were designed to ensure proper' monitoring of safety-related systems for safe shutdown. All of the components and equipment affected by this modification are of the appropriate safety grade and meet the proper qualifications for the functions that they serve. Cables and wires associated with safety-related systems were rerouted using the appropriate electrical separation requirements. The remaining cables and wires affected by the rerouting serve no safety control function. Therefore, there are no adverse effects on any safety-related equipment.

43

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-279 Annual Plant Modification Report i

1

. Installation of Process and Diagnostic Instrumentation 4 (Continued) i l

1

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

evaluated previously in the safety analysis report?

Answer: No. The indicators, transmitters and other l

equipment installed for this modification were designed to operate in parallel with existing j i

equipment, without creating any new types of I failure modes. Therefore, these changes cannot create the possibility of an accident l or malfunction of a different type.

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

1 Answer: No. The safety functions of the associated safety-related equipment have not been-degraded; only indicating instrumentation was involved which cannot affect performance.

Therefore, no margins of safety were reduced.

i 44

h Peach' Bottom Atomic Power Station Unit 3

~ Docket No. 50-278

' Annual Plant Modification Report Trip' Setting Changes for Unit 3 Load Center 480V Breakers and 4 kV Breaker Protective Relays Modification No.: 1029H.

A.- System: 4 kV and 480V AC Power B. .

Description:

The trip ~ settings for the Load Center for six 480V breakers and for seven protective relays on the 4kV breakers were changed to provide breaker coordination such that the individual equipment breakers will trip before the source breaker for the bus trips._

C. Reason for Change:

This modification was performed in accordance with the requirements of 10 CFR 50, Appendix R, to facilitate safe shutdown in the event of a design basis. The breaker and relay setting changes will permit continued operation of equipment necessary for safe shutdown, during fault conditions of equipment that exist on the same line feeds.

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 new breakers allow the continued operation of equipment necessary for safety shutdown even though this equipment may be fed from a bus that also feeds potentially faulted equipment. The new breaker and relay settings do not adversely affect breaker operation under circumstances previously evaluated.

Therefore, this modification reduces the probability of occurrence and the consequences of an accident or malfunction of safety-related equipment.

45

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Trip Setting Changes for Unit 3 Load Center 480V Breakers and-4 kV Breaker Protective Relays (Continued)

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 these changes were designed in accordance with the applicable safety-related requirements, including Appendix R, without changing the plant electrical load. In addition, the reliability of the equipment essential for safe shutdown has been enhanced without creating any new failure modes. .j iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. Changing the breaker and relay settings permits breaker trip coordination such that safety-related equipment may continue to operate though fed by the same feed as faulted equipment, thus enhancing the reliability of the equipment necessary for safe shutdown.

Additionally, although the breakers and relays ,

modified are not discussed in the Technical i specifications, the safety functions of their associated equipment are not changed. <

Therefore, no margins of safety, as defined in the bases of the Technical Specifications, have been changed. '

i 46

Peach Bottom Atomic Power Station' Unit 3 Docket No. 50-278 Annual Plant-Modification Report-8-Hour Battery Pack Emergency Lights Modification No.: 1029R s

A. System: Emergency Lighting DC B.

Description:

Twenty 8-hour battery pack emergency ~1ighting units were installed along access routes to, and inside of areas essential for safe plant shutdown.

C. Reason for Change:

This change was made to meet the safety requirements of 10 CFR 50, Appendix R to assure accessibility to areas needed for safe j plant shutdown in the event of a design basis fire. '

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 this modification has been designed and installed in accordance with the requirements of Appendix R, and does not involve any safety-related equipment.

Therefore, the probability of an accident or malfunction safety-related equipment is not increased.

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 change merely assures emergency lighting in access routes and in areas necessary for operating equipment needed for safe plant shutdown. Thus, this change does not create the possibility of an accident or malfunction of any type.

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

Answer: No, because these lighting packs are not addressed in the Technical Specifications and do not adversely affect any safety-related equipment. Therefore, this modification does l not reduce the margin of safety as defined in ]

the bases of the Technical Specifications. i l

47

- A

, Peach Bottom Atomic Power Station Unit 3-Docket No. 50-278 Annual Plant Modification Report Installation of Instrumentation and Control at the Main Alternative Control Station Modification No.:. 1353B A. System: Residual Heat Removal (RHR) High Pressure Service Water (HPSW), High Pressure Coolant Injection (HPCI)

B.

Description:

Instrumentation was installed at the Main Unit 3 Alternative Control Station. Flow indication was provided for the 'D' RHR loop, 'D' HPSW loop and HPCI. Pump discharge pressure indication was provided for the 'A' ESW loop and HPCI. Differential - ---

pressure indication was provided for the 'D' RHR heat exchanger.

Alarms were provided for HPCI turbine exhaust high pressure and HPCI pump suction low pressure. Also, HPCI turbine speed control was provided.

C. Reason for Change:

This modification was implemented in accordance with the requirements of 10 CFR 50, Appendix R, to ensure safe plant shutdown capability in the event of 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, because the changes were designed to ensure control of the HPCI System and to facilitate monitoring of other safety-related systems from the ACS for safe shutdown. All of the components and equipment involved affected by this modification are of the i appropriate safety grade and meet the proper qualifications for the functions that they serve. Cables and wires which are associated with safety related systems, were rerouted using the appropriate separation requirements. ;

The remaining cables and wires affected by the )

l. rerouting serve no safety control function.

l Therefore, there were no adverse affects on j any safety related equipment.

l 48 I

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Installation of Instrumentation and Control at the Main Alternative

-Control Station (Continued) 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? q Answer: No, because the indicators, transmitters and other equipment installed for this modification were designed to operate in parallel with existing equipment without creating any unreviewed failure modes or unreviewed operations. Therefore, these changes cannot create the possibility of an accident or malfunction of a different type than evaluated in the FSAR.

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

Answer: No, because the safety functions of the associated safety related equipment have not been degraded. The additional manual HPCI turbine control and the added indicators and alarms enhance the safe shutdown capability from the ACS by permitting controlled reactor vessel depressurization. Furthermore, this modification does not adversely affect any bases of the Technical Specifications.

Therefore, no margins of safety as defined the bases of the Technical Specifications have been reduced.

1 49

Paach Bottom Atomic Powar Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Alternative Shutdown Control for Safety-Relief Valves Modification No.: 1353G A. System: Automatic Depressurization System (ADS)

B.

Description:

This modification involves rerouting safety-related circuits to transfer / isolation switches and to el,ternative control switches for SRVs and two backup solenoid valves which supply nitrogen to the SRVs.

C. Reason for Change:

This modification was implemented in accordance with the requirements of 10 CFR 50, Appendix R to ensure safe plant shutdown capability in the event of a design basis fire.

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 does not change the operation of the SRVs 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 required to anticipate a LOCA ,

or seismic event and concurrent loss of offsite power, defeating the control of all SRVs except A, B and K and transferring control of these and the nitrogen supply valves to the 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?

Answer: No. The safety functions of the ADS and Primary Containment Isolation System are not affected by rerouting cables to safety-related transfer / isolation switches to provide alternative shutdown control. Isolation of the SRV cables from their normal controls is done only for a fire in Fire Area 25 in which case control of the SRVs could not be guaranteed.

50 l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Peach Bottom Atomic Power Station Unit.3 Docket No.-50-278 Annual Plant Modification Report-Alternative Shutdown Control for Safety-Relief Valves (Continued)

~

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

Answer: No. Providing an alternative means of SRV and nitrogen supply valve control necessary for.

safe shutdown does not reduce ai7 Technical Specification margin of safety because the reliability of~the equipment was-increased.

...eMak-51 1

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual. Plant Modification Report

' Alternate Control Station Process Monitoring Instrumentation Modification No.: 1353H-A. System: Miscellaneous Instruments B.

Description:

Various monitoring instruments were installed for the Unit 3 High Pressure Coolant Injection (HPCI) Alternative Control Station (ACS) located in the recirculation M-G set room. The instrumentation installed for the ACS consisted of'the following:

a reactor water level transmitter in parallel with the existing level transmitter and an annunciator for'high water level; a reactor pressure indicator / transmitter to replace the existing pressure indicator PI 3-2-3-60B; a drywell pressure transmitter in parallel with the existing pressure transmitter; a thermocouple and a protective conduct seal to monitor.drywell .

temperature; a Condensate Storage Tank water level transmitter in' parallel with existing level transmitter and an annunciator for low water level; a Suppression Pool water' level transmitter in parallel with the existing level transmitter, and an r.nnunciator for high water level; and a thermocouple next to the existing temperature element inside the drywell to measure Safety Relief Valve (SRV) discharge temperature on the RV 3-2-71A discharge piping. . In addition, a thermocouple from temperature element TE 3442B, which monitors Suppression Pool temperature, was rerouted to the ACS. TE 3442B has two thermocouple; therefore, installing a new thermocouple was not required.

C. Reason for Change:

These changes were implemented in accordance with the requirements of 10 CFR 50, Appendix R, to ensure safety shutdown capability from the HPCI ACS located at the recirculation pump M-G set room in the event of 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?

52

Pecch Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report

-Alternate Control Station Process Monitoring~ Instrumentation (Continued)

Answer: No, because these changes were designed to meet the requirements of 10 CFR 50, Appendix R in the event that the Control Room becomes uninhabitable during a design basis fire. All of the instruments installed at the ACS for this modification operate in conjunction with l

existing instruments and equipment. The additional instruments provide remote monitoring capability without changing the operation of the equipment or adversely affecting the original instrumentation. Thus, the probability of occurrence or the consequences of an accident or malfunction of safety-related equipment are not increased,

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 instruments installed for this modification do not perform any control function, nor do they affect the control or operation of any associated equipment. The environmental and seismic requirements have been evaluated and determined to be acceptable since the instruments do not actively perform any safety function. Furthermore, the electrical considerations have been analyzed and it has been determined that the additional bus load is acceptable. Thus, this l modification does not create the possibility l for an accident or malfunction of a different l type than presented in the FSAR.

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

1 Answer: No. Additionally, the electrical load of the new instruments has been evaluated and determined to be acceptable. Although some of the instrumentation is associated with safety- i related equipment discussed in the bases of the Technical Specifications, there are no changes to or adverse affects on the operation or control functions of the associated l equipment. Therefore, this modification does I not reduce the margin of safety as defined in the bases for the Technical Specifications.

53

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POnch Bottom Atomic Powar Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Replacement of Refueling Platform Modification No.: 1533 A. System: Fuel Handling B.

Description:

The refueling platform was replaced with a platform procured from another plant and modified by the manufacturer to conform to Peach Bottom requirements. The replacement platform is equipped with a refrigerant air dryer system and uses a " boundary zone computer" system to restrict motion of the platform, trolley and --

hoist to within the permissible zone.

C. Reason for Change:

The previous platform was unreliable and, on several occasions, caused delays in refueling.

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 replacement conforms to the seismic design requirements for Peach Bottom and the floor loading and seismic accelerations have been verified to be acceptable. The replacement platform meets all design bases ;

and is expected to be more reliable.

Therefore, the probability or consequences of an accident or malfunction were not increased,

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 replacement platform will operate within the bounds of the FSAR analyses. The differences between the replacement and the original platform do not alter its operation such that no new type of failure modes or accident scenarios are created.

54

Peach Fottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Replacement of Refueling Platform (Cont:inued) ill) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because the replacement platform satisfies the design basis for the Peach Bottom refueling platform and none of the refueling interlocks required by the Technical Specifications were changed.

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Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Rupture Disk Support Modification No.: 1603 A.

System: High Pressure Coolant Injection (HPCI)

B.

Description:

Support pieces were installed on the HPCI turbine exhaust line rupture disks.

C. Reason for Change:

The Unit 3 HPCI system has experienced several rupture disk failures under vacuum conditions in the steam line; consequently, the support pieces were installed to protect the rupture disks from failure.

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?

l Answer: No, because the addition of the support piece was analyzed and it was determined that all performance specifications are still satisfied. Thus, the reliability of the HPCI system was increased 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. Modification of the HPCI rupture disks cannot create the possibility of an accident because the HPCI system's purpose is to mitigate the consequences of a LOCA, but does not affect the possibility of a LOCA.

Furthermore, the modification of the rupture disks does not change the performance of HPCI as evaluated in the PSAR or introduce any new

~

failure modes.

56

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Rupture Disk Support Pieces-(Continued) iii) Does this modification reduce the margin of safety.as defined in the basis for the Technical Specifications?

Answer: No, because the modification does not affect HPCI system performance and, therefore, does

-not reduce any Technical Specification margins of safety associated with HPCI's ability to provide coolant to the reactor without depressurizing. The rupture disks are not addressed in the bases of the Technical ---

Specifications.

57

1 Perch Bottom Atomic Powar Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Crack Arrest Verification System Equipment Modification No.: 1763 A. System: Reactor Water Sampling B.

Description:

A Crack Arrest Verification (CAV) system was installed downstream of recirculation system outboard sample isolation valve AO-351-40 to monitor crack growth in 304 and 316L stainless steel piping.

C. Reason for Change:

The CAV system was installed to determine the effects of reactor water chemistry on primary system pipe crack propagation.

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 operation of safety-related equipment was not affected and the entire CAV system process is outside the "O-boundary" of the primary system piping. Additionally, the electrical feeds are from non-safety related power panels.

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: No, because the CAV system sample and return taps have been installed outside of the Q-boundary; thus, no safety-related piping is involved which could compromise primary 4 containment integrity. Loss of primary coolant through the CAV system would not be a different type than previously evaluated.

Therefore, this change cannot create the

,I possibility of an accident or malfunction of any type.

58

Pasch Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Crack Growth Analysis Machine (Continued) lii) Does this modification reduce the margin of safety as defined.in the basis for the Technical Specifications?

Answer: No, because the CAV system does not adversely affect any piping or valves discussed in the

' bases of the Technical Specifications.

Therefore, no margin of safety has been reduced.

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Peach Bottom Atomic Power Station Unit 3-

, . Docket No. 50-278 Annual Plant Modification Report Residual Heat Removal Heat Exchanges Shell Cover Plange Laak Repair Modification No.:- 1850

,A. System: -Residual Heat Removal B.

Description:

~

A pressurized sealant compound provided by Leak Repairs, Inc.,

was injected into gaps in the Residual Heat Removal (RHR) heat exchanger shell cover flange gasket through cap nuts which have replaced the original'hexnuts on one side of the flange. A metal band was then placed around the outside of the flange.

C. Reason for Change: 1 This modification was implemented to remedy a shell cover flange leak without having to replace the flange.

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 sealant compound will not degrade due to radiation exposure. . In addition,'should any of the sealant become injected past the gasket into the reactor, it ,

would have a negligible affect on the reactor and reactor components.

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 consequences of the failure of the sealant compound are no different from the failure of the shell cover flange gasket, ,

which is accounted for in the plant's design. '

Therefore, this modification does not create the possibility for an accident or malfunction of a different type.

)

I 60

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Patch Bottom Atomic Powsr. Station-Unit 3 Docket No. 50-278 Annual Plant Modification Report Residual Heat Removal Heat Exchanges Shell Cover Flange Leak Repair (Continued) iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because this modification has no adverse affects on the safety function or operation of the RHR system. Thus, this modification does not reduce the margin of safety as defined in the Technical Specifications.

61

l Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 {

Annual Plant Modification-Report I Beginning of Cycle 7 Process Computer Update J Modification No.: 85-141 A. System: Process Computer l B.

Description:

The Process Computer Data bank was updated to reflect the new' core configuration at the beginning of Cycle 7.

C. Reason for Change:

This update was made to assure that the Process Computer makes accurate core thermal calculations.

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 this modification merely enables the Process Computer to do the calculations necessary to evaluate core thermal limits evaluated in the FSAR.

11) Does this modification create the possibility for an accident or malfunction of a different type than any evaluated previously in the c5fety analysis report?

Answer: No, because this update does not change the 'j way the calculations are done by the Process Computer; the data for the calculations was merely updated to reflect the new core configuration.

iii) Does this modification reduce the mergin of safety as i defined in the basis for the Technical Specifications? a Answer: No, because the Process Computer was updated to reflect the revised Unit 3 Technical '

l Specification fuel limits, such as MFLPD, MAPLHGR and MCPR.

62

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i Peach Bottom Atomic Power Station '

Unit 3 Docket No. 50-278 Annual Plant Modification Report R0 placement of Flow Transmitter j Modification No.: 86-003 A. System: Reactor Recirculation B.

Description:

The flow transmitter for the No. 17 jet pump, FT 3-02-3-064T, was replaced with a new model (Rosemount Model No. Il51DP).

C. Reason for Change:

The old transmitter (General Electric Model No. 555) was defective and is no longer manufactured, thuc a new, comparable model had to be 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, because the new transmitter performs the same function as the old transmitter and meets or exceeds all of the original specifications; therefore, the probability or consequences of and accident or malfunction were not increased.

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 changing the model of the transmitter without changing its operation does not introduce any new failure modes and thus cannot 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, because changing the model of the transmitter did not degrade its performance, or reliability and, therefore, no Technical Specification safety margins were reduced.

The accuracy of core flow measurement, which is used by the process computer for core thermal margin calculations, was not reduced.

63

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report

' Rod' Worth' Minimizer Shutdown Sequence Update Modification No.: 86-036, 86-049 A. System: Rod Worth Minimizer (RWM)

B.

Description:

The Process Computer Shutdown RWM sequence updated to reflect changes to the beginning of Cycle 7 control rod patterns.-

C. Reason for Change:

This update was made to assure that the RWM would enforce proper group notch operation during shutdowns.

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 update enables the orocess Computer to enforce the proper RWM sequence,

. augmenting the Rod Sequence Control System reactivity worth control as described in the FSAR.

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 update does not change the scope or function of the Process Computer RWM sequence. The update merely revises the RWM control rod sequences 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? 1 l

Answer: No, because this update enables the RWM to enforce the proper control rod sequences, reducing control rod worth to minimize the effect of a rod drop accident as defined in the bases for the Technical Specifications.

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Peach Bottom Atomic Power Station.

Unit 3 Docket No. 50-278 Annual Plant Modification Report Total Core Flow Drive Process Computer Update Modification No.: 86-074 A. System: Process Computer B.

Description:

The total core flow vs. recirculation drive flow curve data'was h updated in the process computer at the beginning of cycle 7. The data was obtained by performing a core flow calibration test (ST 13.30-2).

C.. Reason for Change:

This update is made once per operating cycle to assure that the computer will perform accurate heat balance and thermal limit calculations.

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 this update assures that the computer has accurate data to perform calculations so that the plant will operate within the bound of the FSAR 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, because this update does not change the calculations or limits discussed in the FSAR, it merely provides the computer with the data to perform the calculations accurately. Thus, no types of accidents or failure conditions have been created. ,

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

Answer: No, because This update increases the accuracy of the calculations of the limits defined in the bases of the Technical Specifications so ;

that margins of safety are maintained. !

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Peach Bottom Atomic Power Station

, Unit 3 Docket No. 50-278 Annual Plant Modification Report Replacement of the 3C Condensate Pump Motor Modification No.: 86-099 A. System: Condensate B.

Description:

The 4kV, 3C condensate pump motor was temporarily replaced with a 13.2 kV motor, requiring the removal of step-down transformer 3CX07 in order to supply 13.2 kV power directly to the new motor.

C. Reason for Change: .-

The original 4 kV motor experienced rotor, stator, and lower guide bearing damage which necessitated temporary motor replacement. The 13.2 kV motor was borrowed from Limerick Unit 2 and will be used in the interim until repairs can be completed on the original 4 kV motor.

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 condensate pump does not serve a safety-related purpose and no safety-related equipment was affected by this change. The motor power supply is not safety-related.

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 condensate pump provides coolant flow for normal power operation and, if failed, would result in a transient (reduced feedwater suction pressure / flow) <

which has been evaluated. The operation of the condensate system, pump performance characteristics, and motor control features have not been affected. Therefore, this modification does not create the possibility for an accident or malfunction of a different type.

66

Peoch Bottom Atomic Power Station Unit 3 Docket No. 50-278

-Annual Plant Modification Report Replacement of the 3C Condensate Pump Motor (Continued) lii)

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

Answer:

No, because no Technical Specification margins of safety involve the condensate pumps. Thus, no margins of safety, as defined in the bases of the Technical Specifications, have been reduced by this modification.

I 67

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Peach Bottom Atomic Power Station Unit 3' Docket No. 50-278

- Annual Plant Modification Report Removal of a Control Rod from the Rod Worth Minimizer Sequences Modification No.:- 86-137 )

J A.. System: . Process Computer

'B.

Description:

Control Rod 10-47 was removed from the Rod Worth Minimizer .('RWM) y startup and shutdown sequences in the process computer.

C. Reason for Change:

-Due.to an upcoupling problem, rod 10-47 was removed from the RWM startup and shutdown sequences to permit reactor operation with the rod in the full-in position.

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 reactor operation with rod 10-47 in the full-in position has been evaluated and determined to be bounded by the Rod Drop Accident Analysis in the FSAR.

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 leaving rod 10-47 in the full-in position does not create any new type of failure condition or accident precursor than any evaluated in the FSAR.

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

Answer: No, because it has been determined that the RWM will still satisfy the Rod Drop Accident energy limit with significant margin. ..

68

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F UNITS 2 & 3 i

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Pccch Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Radiation Monitoring Sample System Modification No.: 21C A. System: High Pressure Service Water-(HPSW)

B.

Description:

r Two radiation monitor sampling systems (A&B) and a control room high radiation alarm and recorder were installed for each unit c..

the High Pressure Service Water (HPSW) system discharge, downstream of the Residual Heat Removal (RHR) heat exchangers.

C. Reason for Change:

This modification provides the control room operator with an alternate method of detecting leaks in the RHR heat exchangers during periods when the HPSW is in operation and leakage occurs from RHR to HPSW. d 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 radiation monitoring system itself is not safety related. Thus, the potential for pipe break or other failure of the sample system has been evaluated and is considered to !

have an insignificant impact upon the capability of the HPSW and Emergency Cooling Systems to function because of the system's physical placement, operating pressure, and engineered features.

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 evaluation concludes that malfunctions or failures of the sampling i system will not adversely affect the HPSW or l Emergency Cooling Systems.

I 70

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f Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant. Modification Report Installation of Radiation Monitoring Sample System

-(Continued) 111) Does this modification reduce.the~ margin of' safety as defined in the basis for the Technical Specifications?

Answer: No. This modification merely provides the control room operator with the ability to monitor HPSW radiation levels without adversely affecting any safety-related equipment. Thus, the margin of safety, as defined in the Technical Specifications, has not been reduced.

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j Pacch Bottom Atomic Power Station I Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Oil Spill Containment Systems at the No. 6 and No. 2 Fuel Oil Unloading Areas

{

i Modification No.: 257B d

}

A. System: . Oily Waste System j; B. Description

i k

Contoured dikes were installed at the No. 2 and No. 6 fuel oil unloading areas to contain spills or hose leaks. Discharges from the diked areas are '-2ted through oil / water separators prior to release into the nors : waste system.

C. Reason for Change:

This modification provides the PBAPS Spill Prevention Control and Countermeasures (SPCC) Plan with additional oil spill protection in accordance with the request of the Environmental Protection Agency.

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, nor does it involve any safety-related equipment or components.

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 revealed that the modification will not adversely affect the Diesel Generator building. The modification will not affect other 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 is not safety related, l

nor does it adversely affect any safety-related equipment required by the Technical Specifications. Therefore, this modification does not affect the margin of safety as defined in the Technical Specifications.

72

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Psach Bottom Atomic Power Station i Units 2 & 3 Docket Nos. 50-277; 50-278 l4 Annual Plant Modification Report -

i Reducing the Low Water Level Group I Isolation Setpoint

]

Modification No.: 632 A. System: Primary Containment !

B.

Description:

The low water level isolation setpoint for the Group I Primary Containment Isolation Valves (Main Steam Isolation Valves, Main Steam Drain Valves, Main Steam Sample Valves and Reactor Water Sample Valves) was reduced from -48 inches to -160 inches. A correction factor was applied to the calibrated setpoint to account for the effect of worst-case accident drywell temperature on the density of water in the instrument lines.

C. Reason for Change: i This setpoint was decreased in response to the NUREG-0737 1 requirement to reduce challenges to safety relief valves. This change was made to reduce the number of unnecessary isolation during reactor transients and, thereby, reduce safety relief valve challenges, mitigate anticipated transient without scram events, improve control operators' control of reactor transients, and reduce the heat and thrust loads on the torus during a small break 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 equipment important to safety as previously evaluated in the safety analysis report?

Answer: No, because the effects of this setpoint change were analyzed and it was determined that, although the peak fuel cladding temperature of a postulated small break loss-of-coolant (LOCA) accident will be slightly higher, the consequences are still less severe than those of the limiting design basis accident break. It was determined that this change has no effect on the consequences of a large break LOCA due to the rapid depressurization of the vessel and has no ,

effect on emergency cooling system performance. Lowering this setpoint does not increase the probability of occurrence of an accident because the isolation will still occur at a level which will ensure that the fuel integrity limits are not exceeded.

73 i

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 l

Annual Plant Modification Report Reducing the Low Water Level Group I Isolation Setpoint (Continued)

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 setpoint change does not introduce a new accident scenario; it merely delays the isolation of the Primary Containment Isolation System Group I' valves without creating any new failure modes. Thus, this setpoint change does not create the possibility of a different type of malfunction or accident.

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

Answer: No, because this setpoint change only affects a small break LOCA; therefore, the margin between the limiting design basis accident (large break LOCA) peak cladding temperature and radiation release, and the 10 CFR 50, Appendix K and 10 CFR 100 limits is not changed. Therefore, the Technical Specification margins of safety are not reduced.

74

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Pacch Bottom Atomic Power ~'.stion Uni ;2&3 l Docket Nos. 50-277;30-278 ;

Annual Plant Modification Report '

l Permanent Access Platform Around Waste Demineralized Vessel Modification No.: 821 i

A. . System: Structural i

B.

Description:

This modification consists of the installation of structural steel and grating around.the waste demineralized vessel. The platform will be attached to the walls of the hatch.

C. Reason for Change:

This modification will provide safe and easy access to the ;

demineralized vessel. )

D. Safety Evaluation Summary:

i)' Does this modification increase the probability of !

occurrence or the consequences of an accident or j malfunction of equipment important to safety as previously ;

evaluated in the safety analysis report?

Answer: No. The physical situation of the platform is j such that neither its existence nor failure will increase the risks associated with the vessel.

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 physical structure performs no safety function. Additionally, its use will not introduce hazards to nucleat safety.

iii) Does this modification reduce the margin of safety as {

defined in the basis for the Technical Specifications?

Answer: No. The physical configuration of the i platform is such that in the event of its !

failure, the vessel and piping would not be affected, thus not decreasing the margin of safety.

75

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Replacement of Motor Brakes / Motor on Limitorque Valve Actuators Modification No.: 825 A. Systems: Core Spray, Residual Heal Removal and Reactor Recirculation B.

Description:

Limitorque valve operator motor brakes on various Core Spray, Residual Heat Removal, an'd Reactor Recirculation System valves were replaced with environmentally qualified brakes. In addition, the Unit 2 Core Spray MO-2-14-12B valve operator motor was replaced with an environmentally qualified motor.

C. Reason for Change:

The motor brakes, as well as the motor for the MO-2-14-12B valve lacked environmental qualification documentation and were replaced in order to meet the environmental qualification requirements of NRC I.E. Bulletin 79-01.

D. Safety Evaluation Summary:

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

Answer: No, because these valves were made more reliable to perform their intended safety functions by satisfying the environmental qualification requirements of I.E. Bulletin 79-01.

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 replacement components perform the same functions as the original ones. Thus, this modification does not create the possibility of an accident or malfunction of a difference type than previously evaluated.

76

Peach Bottom Atomic Power Station Units 2.& 3 Docket Nos. 50-277; 50-278 Annual. Plant Modifice. tion Report

- Replacement of Motor Brakes on Limitorque Valve Actuators (Continued)

.iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? l i

Answer: No,'because these valves were made more reliable to perform their. safety functions without changing their operation. Therefore, the margins of> safety, as defined in the bases for the Technical Specifications involving these valves, were not' reduced.

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Pasch Bottom Atomic Power Station Units'2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Drywell/Wetwell Differential Presure Instrumentation Modification No.: 826

-A. . System: Primary Containment B. -Description:

'A drywell to wetwell differential pressure instrument loop was-installed on each unit. An indicator was installed on the CO3 control room panel and an alarm window was installed in'the C209R control room panel. The alarm, " Low Drywell/Wetwell Pressure",

annunciated if drywell pressure becomes 0.11 psi lower than-wetwell pressure.

C.. Reason for Change:

The torus suppression pool swell load analysis associated with NUREG-0661 was based, in part, on drywell pressure not becoming less than wetwell pressure. Thus, an instrument loop was installed to provide assurance that this situation does not occur.

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 provides instrumentation which decreases the possibility of the plant entering a condition outside the bounds of previous accident analyses without adversely affecting any equipment important to-safety. The alarm function provides assurance that the hydrodynamic forces on the torus ring header and its supports, in the event of a LOCA, would not exceed the design limits. The increased bud load due to the new instrumentation is insignificant.

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 installation of this instrumentation does not introduce any new type of failure modes. This instrumentation merely provides indication and an alarm. It has no control function and is properly isolated from safety-related systems.

78

Panch Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Drywell/Wetwell Differential Pressure Instrumentation (Continued) iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because Sections 3.7.A and 4.7.A of the Technical Specifications address drywell and ,

wetwell atmosphere controls and are not '

affected by this modification. Also, this modification provides additional assurance that the integrity of the torus ring header is maintained so that it performs as described in the Technical Specifications Bases, in the event of a LOCA.

79

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Peach Bottom Atomic Powar' Station Units 2 & 3

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Docket Nos. 50-277; 50-278

~ Annual Plant Modification Report j

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Sodium Hypochlorite System Modification No.:. 829 l

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.A.. . System:- Water, Treatment JB. -Description: l The1 original: gaseous chlorination system was replaced with a ,

liquid sodium-hypochlorite system. The new system consists of a storage ~ tank,' injection metering pumps, controls, piping and-valves which. supply sodium hypochlorite to the circulating and service w&ter-systems. Additionally, the Wallace and Tiernan free chlorine monitors'were replaced with Orion chlorine

. analyzers.

C. Reason for Change:

The " Control' Room Habitability Study", which was performed _in response to NUREG-0737, identified that on-site storage of

' gaseous chlorine presents a threat to control room habitability.

The liquid system eliminates the threat.of a gaseous chlorine release. The free chlorine monitors had experienced numerous failures and had a1 poor calibration records. The total chlorine analyzers have been successfully used on other systems.

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 j.

evaluated in the safety analysis report?

Answer: Ik). The FSAR. analyses do not address the chemistry of the service _ water and circulating water systems, so the change in the physical form of the chlorine (from gas to liquid) will not impact the safety analyses. Eliminating the possibility of a chlorine gas release, which could affect control room habitability, may reduce the probability or consequences of an accident.- The increased accuracy and reliability of the new chlorine analyzers will improve water chemistry control, which will decrease the probability of bio-fouling and corrosion in these systems.

80

Pacch Cotto3 Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Sodium Bypochlorite System (Continued)

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 hypochlorite storage tank is located inside a retaining basin. This basin will contain the tank contents in the event of a failure, preventing uncontrolled chemical releases to the environment. The pump suction line is protected by a guard pipe, which drains into the basin in the event of a pipe failure. The electrical loading of the system will not have an adverse effect on safeguard power because it is powered by a non-safety-related source. The new chlorine analyzers will improve the reliability of the chlorination data collected. The systems involved are not safety related, and this modification will not adversely affect any safety-related systems.

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

Answer: No. The replacement of a gaseous chlorination system with a liquid system reduces hazards to station personnel. The modification also eliminates a potential for control room non-habitability, without adversely affecting safety-related equipment. The new chlorine analyzers will provide increased data reliability. The margins of safety, therefore, are not decreased.

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P cch Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Valve Operator Shaft Key Replacements Modification No.: 841 A. System: High Pressure Coolant Injection Residual Heat Removal and Condensate B.

Description:

The keys between the motor pinion gear and motor shaft on Limitorque operator models SMB-3 and SMB-4 with torque ratings in excess of 150 ft-lbs and on all SMB-5 models were replaced with the stronger keys as recommended by Limitorque.

C. Reason for Change:

I.E.

Information Notice 81-08 recommended this action as a result of key failures at several stations.

I 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 change does not affect any of the PSAR analyses because operation of the valves j was not changed. This change decreases the probability of a valve failure.

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

Answer: No, because replacing these keys does not }

create a new potential failure mode. Failure '

of the new keys would result in the same consequence as failure of the old keys and the new keys also will fail at a lower load than the shaft and gear will fail.

iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? i Answer: No, because this change merely reduces the probability of a valve operator failure which I may increase margins of safety associated with these valves.

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Perch Bottom Ato31c Pow 2r Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report l

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Removal of Scram Discharge Volume and Control Rod Automatic Insertion Temporary Modifications Modification No.: 964 A. System: Control Rod Drive i

B.

Description:

)

Scram discharge water level monitor, and the automatic control rod insertion O- low pressure in the control air header systems j were removed form service for Units 2 and 3. The Unit 2 panels, !

conduits, and cables remain as spares. The Unit 3 panels, conduits, and cables were removed.

C. Reason for Change:

These, systems were installed as temporary modifications. Both ;

nystems .are being removed because a permanent modificaiton han l been completed to replace both of these systems, thus making them no longer required.

1 D. Safety Evaluation Summary: !

(

i) 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, because these systems were installed temporarily to enhance the reliability of the scram discharge system in response to NRC '

Bulletin 80-17, Supplement 1. A permanent :

modification has since be implemented to j replace both of these systems. Thus, the ,

probability of any accident or malfunction has '

not been increased. ;

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 replacement modification for these systems has been designed to assure adequate draining of the Scram Discharge Volume. Therefore, removing these systems does not create the possibility of an accident or malfunction different than previously evaluated.

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Peach Bottom Atomic Power Station l Units 2 & 3 f Docket Nos. 50-277; 50-278 Annual Plant Modification Report Removal of Scram Discharge Volume and Control Rod Automatic Insertion Temprorary Modifications (Continued) lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because these systems were installed on a temporary basis only. The permanent replacement modification has been completed, thus allowing the removal of these temporary water level monitoring and automatic control rod insertion systems without any adverse effects to the function or operation of the scram discharge system. Therefore, the margin of safety, as defined in the Technical Specifications, has not been reduced.

84

P rch Bottom Atomic Pow 3r Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Cooling Fans in Control Room Cabiners Modification No.: 983 A. System: Ventilation B.

Description:

Cooling fans and louvers were installed in the control room cabinets and Rod Position Indication system cabinets in the cable spreading room.

C. Reason for Change:

The fans and louvers were installed to improve cooling of components in the panels.

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 fans and louvers will not hinder the execution of the saiety functions of these cabinet components. This modification will enhance safety by improving the reliability of the components. The components in the panels were designed to operate without fans; therefore, the addition i of fans and improved cooling will not increase f the probability of occurrence or the consequences of component failure.

)

I 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?

Answer: No. While the fans and louvers are not themselves safety-related, some of the cabinets in which they are installed do contain safety-related equipment. It has been determined that the normal mounting hardware j is sufficient to prevent the fans and louvers from falling on this equipment during a seismic event. The electrical loads presented l by these fans will not degrade plant l operability or reliability. !

85

Pasch Bottom Atomic Powsr Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Cooling Fans in Control Room Cabinets (Continued) lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. The additional cooling and circulation provided improves the reliability of the components in these cabinets by reducing the risk of overheating, thereby increasing the probability that the components will perform their safety-related function. Thus, the margin of safety, as defined in the bases of the Technical Specifications, have not been reduced.

t 86

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Safety-Relief Valve and Ventilation system Cable Rerouting Modification No.: 1029B A. System: Main Steam, Ventilation B.

Description:

Various Safety-Relief Valves (SRV) and Ventilation System control cables were rerouted, and in some cases, placed in their own ;

conduits. Other miscellaneous cables were also rerouted to facilitate this modification. The cables affected include the !'

following: 120 VAC feeds to Emergency Core Cooling System compartment; HVAC control panels; SRV control cables; Diesel Generator Motor Control Center 00B54 and OOB56 feed cables; and Core Spray outboard injection valves MO-2-14-12A and MO-3-14-12B feed cables. Additionally, new alternative power feeds were supplied to Battery Chargers 2BD03, 2DD03, 3AD03, and 3CD03.

C. Reason for Change:

These changes were implemented in accordance with the requirements of 10 CFR 50, Appendix R, to ensure safe shutdown capability in the event of 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, because these changes were designed to meet the requirements of 10 CFR 50, Appendix R, which increase the reliability of safety-related equipment necessary for safe shutdown capability. The design ensured compliance with FSAR electrical and physical separation criteria; therefore, the affected equipment will perform their safety function as addressed in the PSAR.

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 operation of the equipment is not affected. In addition, these changes were designed in accordance with the applicable safety requirements which improved electrical safeguard designations, supplied electrical conduits were needed to enhance the reliability of safety-related equipment, and did not significantly increase the electrical load on the system.

87

Peach Bottom Atomic Power Station Units 2 & 3-Docket Nos. 50-277; 50-278 Annual Plant Modification Report Safety-Relief Valve and Ventilation System Cable Rerouting (Continued) lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. Some of these cables are associated with safety-related equipment discussed in the Technical Specifications and the' cable rerouting increases the reliability of their associated equipment essential for safe shutdown. Also, surveillance requirements have been established to assure the operability of the new alternative power feed switches for the battery chargers. Thus, the '

margins of safety, as defined in the bases of the Technical Specifications, have been increased.

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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Diesel Generator Cooling Water Systems Cable Rerouting i Modification No.: 1029D 1 A. Systems: Diesel Generators, Emergency Service Water (ESW)

B.

Description:

Various control cables for the diesel generator cooling water systems were redesignated, rerouted and encapsuled. A redundant control cable was supplied to protect against cable damage that could result from a heavy load drop accident.

C. Reason for Change:

( This change vac made in compliance with the requirements of 10 CFR 50, Appendix R, to facilitate safe shutdown in the event of a design basis fire. The change prevents a single event from disabling the automatic start of the Emergency Service Water pumps.

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 modification was designed to meet the requirements of Appendix R, NUREG-0612, concerning control of heavy loads, and the applicable electrical separation requirements.

The control cables for the ESW pumps were rerouted and redesignated to enhance the reliability of the diesel generator cooling systems. Therefore, this modification does not create the possibility of an accident or malfunction of safety-related equipment.

89

0 Peach Bottom' Atomic Power Station-Units 2 & 3- 4 Docket Nos. 50-277; 50-278 Annual' Plant Modification Report

~

y Diesel Generator Cooling Water Systems Cable Rerouting (Continued) 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 equipment functions have not been changed. The modification was designed in accordance with the applicable safety requirements which improved the safeguard designations, provided cable encapsulation where needed, and protects against load drops, thus enhancing the reliability of the diesel generator cooling: water systems. Therefore, this modification does not create the possibility of an accident or malfunction different than previously evaluated.

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

Answer No, because this modification ensures diesel generator cooling water availability during a design basis fire. The rerouting, redesignating and encapsulation of the ESW pump control cables have no adverse effects on any safety-related equipment. Furthermore, neither the concrete dikes nor the cables are addrassed in the Technical Specifications.

Therefore, this modification does not reduce the margin of safety as defined in the basis of the Technical Specification.

90

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Parch Bottom Atomic Pow 2r Staticn Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Isolation of Corridor Behind the Emergency Switchgear Rooms Modification No.: 1029S A. System: Miscellaneous B.

Description:

Three-hour rated fire doors were installed at the end of the turbine building corridor between Unit 2 and Unit 3 to prevent fire from spreading from one Fire Area to another Fire Area.

C. Reason for Change:

This change is one step which enables both units to meet the requirements of 10 CFR 50, Appendix R, and the Safe Shutdown Report.

D. Safety Evaluation Summary:

1) Does this modification increase the probability of occarrence 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 installation of fire proof doors has no effect on any safety-related equipment. This change merely prevents the spread of fire between safe shutdown areas. [

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

Answer: No, because the new fire proof doors are I designed to mitigate the potential effect of a 1 design basis fire. Therefore, a fire in one safe shutdown area will not inhibit safe shutdown activities in an adjacent shutdown area. Thus, this modification does not create the possibility of an accident or malfunction 4 of any type.

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

Answer: No, because the installation of the fire-proof doors ensures that fires within the Method A shutdown area will be confined or adequately retarded from spreading to the two adjacent Method B shutdown areas. Thus, the margin of safety as defined in the bases of the Technical Specifications has been increased.

91

Peach Bottom Atomic Power Station Units 2 & 3 _

Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Jib Cranes in the Reactor Building Sump Room Modification No.: 1171 A. System: Structural B.

Description:

This modification will install a 1000 pound capacity lifting beam in the reactor building sump room.

C. Reason for Change:

The crane provides a safe and expedient way to move the HPCI and RCIC heavy components up the sump room stairs.

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 rigging beam satisfies the requirement of NUREG-0612, Section 5.1.1. The rigging, its loads and its load pack will not impact plant performance as evaluated with 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 repcit?

Answer: No. By meeting the requirements of NUREG-0612, the rigging will increase the margins of personnel as well as plant safety a safe method of transporting HPCI and RCIC equipment in the sump room stairs.

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

Answer: No. The modification does not enter the bounds of the Technical Specifications and does not affect the margins of safety therein.

92

P2rch Bottom Atomic Pownr Stction Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Monorail and Boist in Residual Heat Removal (RHR) Pump Rooms Modification Nos.: ll81A, 1784, 1861, 1904 A. System: Structural B.

Description:

A monorail and hoist was installed in the Unit 2 A, B and D RER Pump Rooms and in the Unit 3 A RHR Pump Room to facilitate lifting and movement of the RER pump and/or motor.

C. Reason for Change:

Previously, the concrete access hatches to the pump rooms, which are part of the secondary containment boundary, had to be removed to facilitate lifting and movement of the RHR pump and/or motor with a reactor building crane. Therefore, the RHR pump and/or motor could be moved only after breaching secondary containment, which required shutting down the plant to comply with the Technical Specifications. The monorails and hoists were installed in the pump rooms so that the pumps and/or motors can be moved during reactor operation.

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: No, because each monorail and hoist was designed and installed in accordance with the

" Guidelines for Control of Heavy Loads,"

NUREG-0612, Section 5. In addition, the new lifting devices will be operated in accordance with the NUREG-0612 guidelines. Therefore, the potential for a load drop is extremely small and the consequences of postulated load drops would be mitigated. l I

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 potential adverse consequences due to a failure of the new monorail and hoist are not of a different type than those due to a failure of the crane previously used. Plant i fire protection and safe shutdown l

considerations were evaluated and determined '

not to be impacted.

93

Pacch Bottom Atomic Powar Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Monorail and Hoist in Residual Beat Removal (RHR) Pump Rooms j (Continued) iii) Does this modif'ication reduce the margin of safety as defined in the basis for the Technical Specifications? (

Answer: No, because no Technical Specifications are affected, and the integrity of safety-related {

equipment is not jeopardized because of thct defense in-depth approach of the monorail ano ,

hoist design (in accordance with NUREG-0612). !

Therefore, no Technical Specification safety margins were reduced.

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A Peach Bottom Atomic Power Station Units.2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Linear Heat Detector Cable for Remote Shutdown

' Modification No.: 1351B A. System: Miscellaneous B.

Description:

Linear heat detection cables were installed in the cable trays in the Emergency Shutdown Panel area and in its associated control panels outside of the area.

C. Reason for Change:

This modification was implemented in accordance with the requirements of 10 CFR 50, Appendix R, to provide warningand protection for safety-related cables and to ensure safe shutdown capability in the event'of 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, because'the protection of the. cables in i the trays was enhanced by the installation of the linear heat. detectors and the applicable electrical separation requirements were .

I satisfied.

11) Does this modification create the possibility for an 4

accident or malfunction of a different type than any ;

evaluated previously in the safety analysis report?

Answer: No, because the control of safety-related equipment was not affected. Failure of these heat detection cables would not impact the operation of any safety-related equipment because these cables do not perform any control functions. Therefore, this change cannot create the possibility of an accident or malfunction.

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Peach Bottom Atomic Power Station

. Units'.2 & 3' Docket.Nos.-50-277; 50-278 Annual Plant Modification Report Linear Heat Detector-Cab 1'e for Remote Shutdown (Continued) 111 ) = Does this modification reduce the margin of safety as defined in the basis'for the Technical Specifications?

Answer: No, because the linear heat detection' cables increase the reliability of equipment

-essential for safe shutdown capability, thereby increasing the probability that equipment discussed in the Technical-Specification bases will perform as designed.

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96

Panch Bottom Atomic Powsr Stction Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of RCIC Turbine Monorails Modification No.: 1427 A. System: Structural B.

Description:

This modification will install a monorail in both RCIC rooms.

The monorail will be used with 4800 pound hoist to lift RCIC turbine inspection.

C. Reason for Change:

This modification will facilitate rigging operations during the 10-year RCIC turbine inspection.

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 monorail will not affect the operation of the RCIC or other safety 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. The nature of the modification is such i

that additional accident scenarios are not probable, iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. The modification does not enter the bounds of Technical Specifications and does not affect the margins of safety defined i therein.

97

Peach Bottom Atomic Power Station l Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Change of' Post-LOCA Containment Venting Method Modification No.: 1774 A. . System: Containment Atmosphere Dilution (CAD)

B.

Description:

The CAD System vent control valves on Units 2 and 3 were blocked i open. The CV-4957 and CV-5957 were blocked in the full open position to permit a 135 SCFM flowrate during post-LOCA venting operations at 30 psig containment pressure. The CV-4954 and CV-5954 were blocked in a partially open position to permit a 20 SCFM flowrate. The operating procedure for post-LOCA venting through these two-inch lines was revised to reflect this valve

' lineup and provide guidance for controlled venting in this configuration. Vent paths remain unchanged.

C. Reason'for Change:

This change was made because the control circuits for the vent control valves CV-4957 and CV-5957, and the vent line flow indication'are not environmentally qualified. Therefore, venting must be accomplished by relying on different equipment which is 3 environmentally qualified.

D. Safety Evaluation Summary:

1

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

Answer: No, because the new method of venting is I effective and satisfies.the CAD system design !

criteria and saf:Ly design basis described in the PSAR. This method of venting will result in a negligible change in offsite dose based on the model used in the FSAR. Backflow-

' preventer check valves in the sample outlet lines protect the CAD oxygen analyzers from the 1 psig or " slight" pressure increase in the lines during rapid venting. Additionally, the Standby Gas Treatment System is capable of handling the insignificant increase in flow I from the CAD system during venting.

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Pacch Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Change of Post-LOCA Containment Venting Method (Continued) ,

)

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 successful venting of containment is still ensured such that containment integrity will be maintained following a LOCA.

Blocking the control valves open does not create any unreviewed condition because there are redundant isolation valves upstream and flow is still restricted to within the bounds previously analyzed. Thus, there were not new types of potential accidents or malfunctions created. O iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because this method of venting still ensures that post-LOCA offsite dose would be less than the limit specified in the bases of the Technical Specifications, and that post-LOCA containment oxygen concentration and containment pressure are maintained as specified in the bases of the Technical Specifications. Thus, no margins of safety were reduced.

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Psach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Srparate Fuse for Condensate Panel Instruments Modification No.: 1926 A. System: Condensate D.

Description:

Individual power feeds were installed inside Condensate System Control panels 20C70A and 30C70A to provide individual fuses on various recorders and controllers.

C. Reason'for Change:

This change was made to prevent loss of power to all of the instruments in these cabinets due to a single fuse failure.

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 none of the circuits involved are safety-related and installing separate fused power feeds in these circuits does not adversely impact any safety-related equipment.

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 installing separate power feeds on these non-safety-related instruments does not create any new failure modes or affect any accident analyses. Therefore, this modification cannet create the possibility of an accident or equipment malfunction of a '

different type.

iii) Does this modification reduce the margin of safety as '

defined in the basis for the Technical Specifications?

Answer: No, because the affected equipment is not discussed in the bases for the Technical Specifications; therefore, no margins of safety were affected.

100

d Peach Bottom Atomic Power Station Unite 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Rnlocation of Smoke Detector Modification No.: 85-104 A. System: Pire Systems B.

Description:

. Smoke detector SD-34 was moved from the side of 2 beam in'the 'C' Residual Heat Removal room approximately one foot to the ceiling.

C. Reason for Change:

The smoke detector was difficult to test and calibrate in its previous location.

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 effectiveness of the smoke detector was not degraded. The detector was moved a small distance in the direction it faced which did not change the quality of fire l protection.

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? j Answer: No, because moving the detector did not affect fire protection or any other safety-related equipment and, thus, could not create the possibility of an accident or malfunction of a different type, iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because the quality of the fire protection was not reduced, thus, no margins of safety were reduced.

101

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4 Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Piping Supports for Pump Seal Cavity Feed Piping Modification No.: 85-145

.A. System: Reactor Recirculation B.

Description:

Piping supports were installed on the 3/8 inch tubing between the Control Rod Drive pumps and the Reactor Recirculation pump shaft seals.

C. Reason for Change:

It was recognized following installation of this piping (Mod. 79-28, 1985 Report) that more supports were needed on some spans of piping.

D. Safety Evaluation Summary:

1) Dees 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 more securing supporting this piping enhances the reliability of the recirculation pump seals without adversely affecting any 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, because this change does not affect operations or any safety related equipment; therefore, it does not create any new type of accident or malfunction.

iii) Does this modification reduce the margin of safety as defined in the basis-for the Technical Specifications? 6 Answer: No, because this change does not affect any safety-related equipment or Technical Specifications, i I

102

4 Peach Bottom Atomic-Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Rod Worth Minimizer Startup Sequence Update Modification No.: 86-019 A. System: Rod Worth Minimizer (RWM)

B.

Description:

The Process Computer Startup RWM sequence was updated to reflect changes to the beginning of Cycle 7 control rod patterns.

l .C. Reason for Chance:

This update was made to assurt that the RWM would enforce proper i group-match operation during startups.

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 update enables the Process Computer to enforce the proper RWM sequence, augmenting the Rod Sequence Control System reactivity worth control as described in the FSAR.

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 update does not change the scope or function of the Process Computer RWM sequence. The update merely revises the RWM control rod sequences 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 update enables the RWM to enforce the proper control rod sequenced, reducing control rod worth to minimize the effect of a rod drop accident as defined in the bases for the Technical Specifications.

103

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Mechanical Stops in Valve Actuator Modification No.: 86-032 A. System: Primary Containment B.

Description:

Mechanical stops were installed in the actuator of the 'A' torus vacuum breaker isolation valves AO-2502A and AO-3502A.

C. Reason for Change:

l These actuators were received without the stops, but normally are equipment with them. They were added to prevent the actuator i linkage from being overstressed.

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: No, because the change restores the actuators j to their design specifications and makes them i more reliable to perform their intended safety function. Therefore, the probability of a malfunction was decreased.

ii) Does this modification create the possibility for an ;

accident or malfunction of a different type than any 1 evaluated previously in the safety analysis report?

Answer: No. The stops do not create any new failure modes because the actuator were designed to include them. Therefore, this change cannot create the possibility of a new type of malfunction or accident because the actuators were merely restored to their design specifications.

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

Answer: No, because the change increases the probability that the valves will perform their isolation function as described in Techr.ical Specifications Bases.

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Peach Bottom Atomic Power Station-Units 2 & 3 Docket.Nos. 50-277;L50-278-

-Annual Plant Modification Report Main Generator Transformer Cooler Spray System Modification No'.: 86-080 A. System:- Main Generator Transformer

.B.

Description:

Spray systems consisting of.a 1 1/2" pipe and header, flow

' indicator, relief valve, block valve and Y-strainer were installed on the cooling fan units on the low voltage side of the Main Generator Transformers for Units 2 and 3. Cooling water for the spray system is taken from the Administration Building

~ domestic. water supply.

l C. Reason for Change:

l This change was installed as a result of Main Generator

. Transformer overheating during warm weather. This modification increases the effectiveness of the transformer cooling fans, without necessitating a plant shutdown for fan unit' cooler maintenance.

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 evaluatedEin the safety analysis report?

Answer: No,-because this modification is.not safety related and does not affect any safety-related equipment. This change merely protects the Main Generator Transformers from overheating by improving the fan cooler system capability.

Therefore, this modification does not increase the probability of occurrence or the consequence of an accident or malfunction of 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, because this change merely enhances the cooling capability ~of the existing fan cooler ;

units without requiring an outage for ,

maintenance. The spray water is supplied by domestic water from the Administration Building supply, which does not interfere with ,

any water supplies which are utilized for 1 plant operations or emergency purposes.

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Peach Bottom Atomic Power StGtion Units 2 & 3 g Docket Nos. 50-277; 50-278 l

Annual Plant Modification Report Main' Generator Transformer Cooler Spray System (Continued) lii) Does this modification reduce the margin of safety as defined in the~ basis for the Technical Specifications?

Answer: No . - The enhanced cooling capability of the fan cooler units. increases the reliability of the main transformers. Additionally, neither the water supply nor the main' transformers are discussed in the bases of the Technical Specifications. Thus, no margins of safety, as discussed in the bases of the Technical Specifications, are reduced.

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Peach Bottom Atomic Power Station

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Units-2 s 3 Docket Nos. 50-277;;50-278-Annual Plant Modification Report

MirrorfSymmetry Control Rod Pattern Software Modification Modification No.: 86-105 A. . : System: Process' Computer

.B.

Description:

j

.The Pl/ program has the capability'to perform thermal limitf calculations for both a mirror symmetric and rotational symmetric control rod pattern. The software change will permit P1 operation for a.mirrorfsymmetric control rod pattern only.

-C. Reason for: Change:

This modification was made to prevent the P1 from performing a rotational symmetric calculation for a mirror symmetric control rod pattern, which would result in incorrect thermal limit l calculations.

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 this change assures that the thermal limit calculations performed by the Process Computer will be based on the correct control rod pattern, thus preventing the thermal limits from being violated.

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 P1 calculations have not been altered, nor has the function of the Process Computer. Thus this change does not create the possibility of an accident or malfunction different from any type previously evaluated.

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

Answer: No, because this change auguments the Process computer's capability to accurately perform P1 mirror symmetric calculations in accordance with the Technical Specifications; therefore, the margin of safety as defined in the bases is maintained.

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COMMOM 1

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I Peach Bottom Atomic Power Station Common i Docket Nos. 50-277; 50-278 l Annual Plant Modification Report On-Site Low Level Radwaste Storage Facility Modification No.: 693 -

A. System Radwaste D.

Description:

A radwaste facility of modular type construction was built to accommodate approximately 2-1/2 years of low level radioactive l- waste for both Units 2 and 3. The modular construction will allow expansion of the facility to accept low level waste throughout the life of the plant.

C. Reason for Change: i This modification has been made to assure sufficient low level radwaste storage capability due to severely limited offsite disposal resources and periodic shipping interruptions, which could have a major impact on plant operations.

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 no safety-related equipment or '

components are involved in, or adversely'affected by, this modification. Therefore, the probability of occurrence or the consequences of an accident or malfunction of safety-related equipment has not been increased.

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 facility has been designed to be constructed and operated within the safety requirements set forth in the PSAR, as well as those established by the NRC. Waste material stored at the facility is not of a different type than the currently handled within the power plant and described in the PSAR. Therefore, this modification does not create the possibility of an accident or malfunction of a different type than previously evaluated.

109

Peach Bottom Atomic Power Station l Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report On-Site Low Level Radwaste Storage Facility (Continued)

~

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

Answer: No. The On-Site Radwaste Facility is not addressed

-in the Technical Specifications. In addition, it does not adversely affect any safety-related equipment or radioactivity limits specified in the Technical Specifications. Therefore, no margins of safety, as defined in the bases of the Technical Specification, have been decreased.

a 11 0

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j l

Peach Bottom Atomic Power Station Common I mn' Docket Nos. 50-277; 50-278 Annual Plant Modification Report Fan _ Drive Motor Replacements Modification No.: 832 A. System: Stbndby Gas Treatment (SBGT) 1 B.

Description:

The drive motors on the Standby Gas Treatment fans were replaced with environmentally qualified motors. The motor circuitry and i motor power; requirements were not changed. l C. Reason for Change:

1 The old motors lacked environmental qualification documentation l and had to be replaced to satisfy the requirements of I.E.

Bulletin 79-01B.

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 function of the fan motors is not affected by the replacement and the new motors may be more reliable because they were procured with environmental qualification documentation.

, ii) Does this riodification create the possibility for an

! accident or malfunction of a different type than any t- ,

evaluated previously in the safety analysis report?

Answer: No, because the operation of the fan actors is not changed; the replacement merely maken them more l reliable without creating any new failure modes.

L Therefore, this modification cannot create the possibility of any accident.

' i 1 lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because this modification increases the probability that the Standby Gas Treatment system ,

uill perform its safety function as described in 1 the Technical Specification Bases.

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Peach Bottom Atomic Power Station Common Dorket Nos. 50-277; 50-278 Annual Plant Modification Report Replacement of the Diesel Generator Ground Overcurrent Relays Mod?fication No.: 1029N A. System: 4 kV AC Power B.

Description:

The Ground Overcurrent Relays on the E-2 and E-3 Diesel Generators 4 kV bus supply breakers (152-1606 and 152-1704) were replaced with Westinghouse CO-7 Overcurrent Relays, Model No. 1456COSA09.

. The new relays have higher operating setpoints which allow the diesel generators to trip after t'e individual bus equipment breakers trip during fault condit ions.

C. Reason for Change:

This modification was implemented as part of the modifications pursuant to the requirements of 10 CFR 50, Appendix R, to ensure safe shutdown capability in the event of design basis fire. These relays were installed to permit continued operation of equipment necessary for safe shutdown even if faulted equipment exists on the same electrical feed.

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 new breakers allow the continued operation of equipment necessary for safe shutdown even though this equipment may be fed from a bus that also feeds potentially faulted equipment. The new relays do not adversely affect breaker operation under circumstances previously evaluated.

Therefore, this modification reduces the probability of occurrence or the consequences of an accident or malfunction.

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 reliability of the equipment necessary for safe shutdown has been enhanced without creating any new failure modes. In addition, these changes were designed in accordance with the applicable safety-related requirements, including Appendix R, without creating any additional plant electrical load.

112

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Replacement of the Diesel Generator Ground Overcurrent Relays (Continued) lii) Does this modification reduce the margin of saf'.ty as defined in the basis for the Technical Specifications?

Answer: No. Replacing the old relays with the new CO-7 overcurrent relays allows continued operation of safety-related equipment powered by the same feed as faulted equipment, thus enhancing the reliability of the equipment necessary for safe shutdown. In addition, the portions of the Diesel Generator circuits being modified are not specifically addressed in the Technical Specifications. Therefore, margins of safety, as defined in the Technical Specifications, have not been reduced. '

113

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report 230-13.8 KV Startup Source Power Transformer Modification No.: 1398 A. System: 13 kV Transformer B.

Description:

A 230-13 kV transformer (Number 343SU) was installed, connected through a circuit switcher to 230 kV Line 230-34, to provide a -

second source of startup and emergency power to the 3SU Regulating Transformer Switchgear.

C. Reason for Change:

The new transformer will provide an additional source of power for 3SU Regulating Transformer Switchgear and can be used as a replacement for Startup Transformer 2SU if it should fail.

l D. Safety Evaluation Summary:

i) Does this modification increase the probability of c-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 is not safety related #

and does not involve work on safety-related equipment. Therefore, this change does not increase the occurrence or the consequences of an accident of malfunction of any safety-related equipment.

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. Loss of offsite power was previously analyzed and the consequences have been accounted for in the FSAR. This change merely supplies a backup source of offsite power, capable of replacing either of the primary startup transformers. Thus, no new failure conditions have been created and the possibility of an accident of failure of any kind has been reduced.

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114

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report 230-13.8 kV Startup Source Power Transformer (Continued) lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because the new transformer has been designated as an alternate to the 3SU startup transformer and is designed as a direct replacement for the existing transformers when in operation. Thus, the --

margins of safety as defined in the Bases of the Technical Specifications have not been reduced.

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115

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of a Crosstie Between the Discharge Canal and Intake Canals Modification No.: 1413A A. System: Circulating Water B.

Description:

A corrugated metal arch pipe was installed in the dike between the discharge canal and the intake pond. A slide gate was installed to block flow through the pipe when recirculation is not desired.

C. Reason for Change:

The crosstie was installed to provide an alternate source of water to the intake pond for power operation, thereby giving operators additional time to gain control of a "frazil ice" incident.

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 crosstie does not adversely affect any safety-related systems or operations. The crosstie increases the probability that water will be available for the Emergency Service Water and High Pressure Service Water pumps. The liquid radwaste discharge procedure (HPO/CO-18) was revised to ensure proper dilution with the crosstie gate open. Therefore, this modification will not affect the FSAR analyses previously performed.

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. Neither the presence nor failure of the crosstle will affect the functioning of safety systems or degrade the control of liquid radwaste discharge. Therefore, the plant remains within the bounds of FSAR analyses.

116

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of a Crosstie Between the Discharge Canal and Intake Pond (Continued) lii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. The crosstie does not adversely affect any Technical Specifications for the Emergency Service Water System or Core and Containment Cooling Systems. Engineering calculations were performed and the necessary procedural changes were implemented to ensure compliance with the p Radioactive Materials Technical Specifications with the crosstie gate open. Thus, no Technical Specification margins of safety were reduced.

117

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Manual Trip of the Diesel Generators Modification No.: 1927A A. System: Diesel Generator .

B.

Description:

A time delay relay has been added to each diesel control circuit to de-energize the operating coil of the Maximum Credible Accident (MCA) relay on the diesel generators, after ten minutes of operation. Contacts from the dead bus start relays to the governor shutdowr, solenoid circuit and the generation excitation control circuit were also installed.

C. Reason for Change:

These changes were implemented to allow manual trip of the diesel generators from the control room once a LOCA signal has been received, and the control room operator has determined that the diesel generators no longer need to run. The reduction of run time of the unloaded diesel generators enhances their reliability.

This modification also allows the automatic restart of the diesels upon a subsequent loss of offsite power.

l D. Safety Evaluation Summary:

1 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. General Electric Company has reviewed the safety evaluation and concurs with its conclusions in that the modification will have no adverse effect on the Emergency Core Cooling System (ECCS) analysis. These changes will facilitate a manual diesel generator trip, yet they do not cause a trip. A manual diesel generator trip can only be i initiated after ten minutes of operation, which l allows resetting of the MCA relay. Once tripped, a i restart can occur automatically with only a ten-l second delay for bus power. Thus, this I modification does not increase the probability of an accident or malfunction.

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118

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Manual Trip of the Diesel Generators (Continued)

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 all of the proper design requirements have been met for the diesel generators including environmental qualification, seismic qualification, separation criteria, quality assurance, and ^--

testability. In addition, these changes have no' adverse effects on the performance characteristics of the diesel generator. Thus, this modification does not create the possibility for an accident or malfunction different than previously evaluated.

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

Answer: No, because this modification enhances the reliability of the diesel generators by allowing a reduction in the amount of time that the generators run unloaded which could increase the probability of damage, without reducing the effectiveness of the ECCS. Therefore, the margin of safety, as defined in the bases of the Technical Specifications, has not been reduced.

l 119

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Containment Atmospheric Control System Test Tap Modification No.: 84-037 A. System: Containment Atmosphere Control B.

Description:

A 3-inch test tap, block valve, and blind flange were installed upstream of relief valve RV-6553 on the containment atmospheric control (CAC) nitrogen injection line to facilitate Surveillance Testing of temperature switch TS-6536. --- ---

C. Reason for Change:

This modification was initiated in response to General Electric '

SIL No. 402 which requires that the CAC inerting system be tested j for low temperature shutoff capability.

D. Safety Evaluation Summary:

i) Does this modification increase the probability of .' q'% '

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 involve any safety related equipment. It merely enables the required Surveillance Testing of the inerting system's discharge temperature switch to determine its ability to detect low temperature.

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 this modification was designed to verify the nitrogen injection system's capaollity to measure low temperature, which is indicative of liquid nitrogen carryover that can result in cracking in the torus vent header. Therefore, this modification does not create the possibilty for a new type of accident or malfunction than previously evaluated.

120

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Installation of Containment Atmospheric Control System Test Tap (Continued) iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? -

Answer: No, because this change has no adverse affects on any safety related equipment or components.

Installation of the test tap enables the verification of the inerting system's capability to detect low temperature and isolate. Low temperature detection and isolation capability reduce the likelihood of a suppression system failure during a LOCA, which could result in overpressurization of the containment. Therefore,

the margin of sa'Jety as defined in the bases of the Technical Specifications has not been reduced.

121 k

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Agastat Relay Replacement Modification No-: 85-008 A. System: Standby Gas Treatment (SBGT)

B.

Description:

Relay No. 62-6023, which starts the 'B' SBGT system filter heaters 10 seconds after the inlet dampers open (Agastat Model No.

TR1413BC757) was replaced with a new model from the same manufacturer (No. ETR1413BC2002002). The power requirements and the contact arrangement of these models are the same. No wiring changes were made.

C. Reason for Change:

The original relay's coil failed and the manufacturer recommended replacement with the new model since the old model was not available.

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?

l Answer: No, because the new model is an acceptable replacement which will perform the same function as the original model. The performance of the SBGT system was not changed and the consequences of failure of this relay is independent of the model.

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 new relay model does not change the operation of the SBGT system and does not create any new failure modes of the SBGT system.

I Therefore, this modification cannot create the l possibility of an accident or malfunction of a 1 different type.

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122

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Agastat Relay Replacement (Continued) 111) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications? ,

Answer: No, because the replacement model performs the same function and is as reliable as the old model.

Therefore, no safety margins are affected.

123

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Continuous pH Monitoring System at the Discharge Canal Modification No.: 85-137 A. System: Circulating Water Sampling B.

Description:

A continuous pH monitor was installed downstream of the cl.rculating water recirculation sample pump to facilitate plant discharge to the river.

C. Reason for Change:

This modification was installed to meet the National Pollution Discharge Elimination requirements of once per day sampling when manpower is limited.

D. Safety Evaluation Summary:

i) Does this modification increase the probability of occurrence or the coneeouences of an accident or malfunction of equipmenE important to safety as previously evaluated in the safety analysis report?

l l Answer: No, because this modification is not associated with any safety-related equipment. Thus, the probability of an accident or malfunction of safety-related equipment is not increased.

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 has been installed downstream of any equipment required for discharge monitoring in the FSAR. Thus, sampling or equipment malfunction of the monitor will have no affect on any required equipment. Hence, this modification will not create the possibility of an accident or malfunction different than previously evaluated, iii) Does this modification reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. This modification has no adverse effects on the Liquid Radwaste Effluents or the Radiological Environmental Monitoring requirements. Thus, this modification does not reduce the margin of safety as defined in the Technical Specifications.

124

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Correction of OD-6 MCPR Program Error Modification No.: 86-093 A. System: Process Computer B.

Description:

The OD-6 program (GEXL) was changed to correct a programming error that occurred during a previous modification, which called data from an incorrect memory location for MCPR calculations.

C. Reason for Change:

This change was performed to assure that the computer will accurately calculate core thermal limits.

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 change enables the OD-6 program to accurately calculate the MCPR. This data, when combined with the information of the P1 Program, assures that the plant will operate within the bounds of the FSAR.

ii) Does thic 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 this modification does not change the i scope of function of the OD-6 program. It merely allows the OD-6 to accurately perform MCPR calculations by taking data from the correct memory locations. Thus, providing separate MCPR calculations (P1 and OD-6) for accurate evaluation.

Therefore, this modification does not create the possibility.of an accident of any type.

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

Answer: No, because this modification assures proper surveillance of the MCPR as defined in the Technical Specification by improving the accuracy of the OD-6 MCPR calculation. Thus, the margin of safety, as defined in the bases of the Technical Specifications has not been decreased.

125

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278 Annual Plant Modification Report Diesel Generator Control Fitting and Tubing Replacement Modification No.: 86-113 A. System: Diesel Generator Controls -

B.

Description:

The Parker " flared" compression type fittings and the associated copper tubing on the Diesel Generator governor booster servomotor were replaced with Swagelok fittings and stainless steel tubing.

C. Reason for Change:

The compression type fittings leaked. The Swagelok fittings have been found to be reliable, and are readily available in the storeroom. The stainless steel tubing was installed to prevent galvanic action between the stainless steel Swagelok fittings and old copper tubing.

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 new fittings and tubing enhance the reliability of the system by eliminating leaks without changing the performance of the Diesel Generators.

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 function and operation of the system is not affected by this modification and the reliability of the system has been improved. Thus, this change does not create the possibility of an accident or new type of malfunction.

126

Peach Bottom Atomic Power Station Common Docket Nos. 50-277; 50-278.

Annual Plant Modification Report Diese1 Generat'or-Control Fitting and Tubing Replacement (Continued) lii) Does this modification reduce the margin of safety as l

defined in-the basis for the Technical, Specifications?

Answer: Nc, because the system reliability'has been improved and.the new fittings and. tubing meet or exceed the pressure rating of the old fittings.and l- copper' tubing. Therefore, no margins of safety are i

reduced.

127

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PROCEDURES l

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Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Procedure No.: HPO/CO-18, Revision 17 A. Procedure

Title:

Processing Liquid Radioactive Waste B.

Description:

Steps were added to this procedure to ensure that the monthly >

projected dose of laundry drain tank releases are calculated and recorded.

C. Reason for Change:

These changes were made to provide additional assurance that the requirements of Technical Specifications 3.8.B.4 are satisfied for laundry drain discharges.

D. Safety Evaluation Summary:

i) Does this procedure 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 the revisions to the procedure did not affect the method of discharge as described in the FSAR. The additional steps provide additional assurance that allowable discharge limits are not exceeded.

11) Does this procedure change 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 physical method of discharge was not changed; therefore, no new situation was introduced which could create the possibility of a different type of accident or malfunction than previously evaluated in FSAR Section 9.2.4.

iii) Does this procedure change reduce the margin of safety as defined in the basis for the Technical Specifications? -

Answer: No, because the equation incorporated into the procedure for projected dose conforms to the Offsite Dose Calculation Model, and the steps added to the procedure increase the probability of compliance with 10 CFR 50, which is one of the purposes of Technical Specifications 3.8.B.

129 s.

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Procedure No.: HPO/CO-18, Revision 18 A. Procedure

Title:

Processing Liquid Radioactive Waste B.

Description:

This procedure was revised to include precautions and additional requirements to ensure proper dilution during liquid radwaste discharge with the discharge canal-to-intake pond crosstie gate open. Also, information was added to clarify the dose "

contribution beta emitters.

C. Reason for Change:

A crosstie gate was installed between the discharge canal and intake pond (Modification 1413A) and the procedure had to be revised to cover discharge when water is able to recirculate between the intake pond and discharge canal (gate open).

D. Safety Evaluation Summary:

1) Do this procedure change increase the probability of occurrence or the consequences of an accident or malfunction of equipment in.portant to safety as prevf ously evaluated in the safety analysis report?

Answer: No, because this procedure change decreases the probability cf discharging radioactive

! material in concentrations exceeding the allowable limits by ensuring proper dilution I as described in the FSAR.

ii) Does this procedure change 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 change to this procedure does not create any operation or condition which is outside the bounds of the FSAR analyses (Section 9.2.4). This change, in fact, provides the necessary precautions to ensure that discharge is within the bounds of the FSAR analyses.

130 ,

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Procedure No. HPO/CO-18, Revision 18 (Continued) 6 iii) Does this procedure change reduce the margin of safety as defined in the basis for the Technical Specifications? . ,

Answer: No. Engineering calculations were performed which indicate that with the crosstie gate f open, at least three circulating pumps should g be operating to provide dilution equivalent to -

that of one circulating pump with the gate closed. Three or more circulating pumps operating is required prior to discharge by this procedure change which ensures that the discharge dose calculations remain valid.

Therefore, the Technical Specification margins of safety (3.8.B) were not reduced.

b 131 .

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Procedure No.: S.2.3.1.A, Revisions 7, 8, 9, 10 and 11 j A. Procedure

Title:

Startup of a Recirculation Pump B.

Description:

Several minor technical and administrative changes were made to this procedure. The change:s involve: the speed of an operating pump when starting the other pump, pump cooling water indication,,

pump speed controller setting, " jogging" pump discharge valve open, equalizing temperature of operating and idle loops, and a core stability monitoring requirement.

C. Reason for Change:

These changes were made to improve the procedure by providing more operator guidance or clearer instructions, or to correct errors or inaccuracies.

D. Safety Evaluation Summary:

1) Do these procedure changes 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 significantly alter the method of starting a recirculation pump, and none of the changes contradict the FSAR description of recirculation pump startup. These changes increase the usefulness of the procedure and reduce the probability of operator error.

ii) Do these procedure changes create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer: No, because these changes involve items which are beyond the level of detail discussed in the FSAR and do not contradict any of the pump startup steps specified in the FSAR. These changes do not create any new potential failure modes or introduce any new type of potential core transient.

132 .

Peach Bottom Atomic Power Station Units 2 & 3 Docket Nos. 50-277; 50-278 Annual Plant Modification Report Procedure No. S.2.3.1.A, Revisions 7, 8, 9, 10 and 11 (Continued) lii) Do these procedure changes reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: llo, because these changes do not affect any Technical Specifications. The startup of a recirculation pump is not discussed in the Technical Specification Bases and these changes do not impact any safety-related systems that are discussed in the Technical Specification Bases.

133 .

Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Procedure No.: ST-9.17-2, Revisions 1, 2, 3 and 4 A. Procedure

Title:

Reactor Coolant Leakage Test - Unit 2 B.

Description:

The limits associated with drywell floor drain leakage (unidentified leakage) were revised to reflect leakage from a Reactor Core Isolation Cooling (RCIC) system valve packing.

Miscellaneous administrative changes were also made.

C. Reason for Change:

Leakage past the RCIC inboard isolation valve packing, MO 13-15, was quantified by backseating the valve and observing the reduction in the drywell floor drain sump pump-out rate.

To account for this leakage as " identified leakage" instead of " unidentified leakage", the procedure limits were revised.

D. Safety Evaluation Summary:

1) Do these procedure changes 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 change in the procedure leakage limits merely reflect increased leakage from a valve packing. Leakage which could potentially be from a primary coolant system pipe crack was still monitored based on leakage limits established in the FSAR (with the exception of " floor drain leakage not doubling since previous reading").

Therefore, these procedure revisions did not increase the probability of a primary coolant system pressure boundary failure.

l 11) Do these procedure changes create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report?

Answer: No, because primary coolant system pipe failures are analyzed in the FSAR (double-ended sheer of a recirculation system suction line).

134

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Peach Bottom Atomic Power Station Unit 2 Docket No. 50-277 Annual Plant Modification Report Procedure No. ST-9.17-2, Revisions 1, 2, 3 and 4 (Continued) iii ) Do these procedure changes reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No. Technical Specifications 3.6.C and 4.6.C are intended to ensure that primary coolant system pipe cracks are detected before they grow to a dangerous or critical size.

Because leakage from the RCIC valve packing was quantified and was, thus, no longer unidentified, increasing the unidentified leakage rate limit in the procedure by that quantity did not affect the Technical Specification margin of safety. Unidentified leakage was still monitored in accordance with the Technical Specification limit because the procedure's limit was merely increased to include the identified quantity from the RCIC valve packing.

135

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Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Procedure No.: ST-9.17-3, Revisions 1, 2, 3 and 4 A. Procedure

Title:

Reactor Coolant Leakage Test - Unit 3 B.

Description:

The limits associated with drywell drain leakage (unidentified leakage) were revised to reflect leakage from a Reactor Core Isolation Cooling (RCIC) System valve packing. Miscellaneous administrative changes were also made.

C. Reason for Change:

Leakage past the RCIC inboard isolation valve packing, M0-3 15, was quantified by backseating the valve and observing the reduction in the drywell floor drain sump pump-out rate. To account for this leakage as " identified leakage" instead of

" unidentified leakage" the procedure limits were revised.

D. Safety Evaluation Summary:

i) Do these procedure changes 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?

Answgr: No, because the change in the procedure leakage limits merely reflect increased leakage from a valve packing. Leakage which could potentially be from a primary coolant -

system pipe crack were still monitored based on the leakage limits established in the FSAR (with the exception of " floor drain leakage not doubling sincere previous reading").

Therefore, these procedure revisions did not increase the probability of a primary coolant system pressure boundary failure.

ii) Do these procedure changes 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 primary coolant system pipe l failures are analyzed in the FSAR (double-i ended sheer of a recirculation system suction line).

136 .

Peach Bottom Atomic Power Station Unit 3 Docket No. 50-278 Annual Plant Modification Report Procedure No. ST-9.17-3, Revisions 1, 2, 3 and 4 (Continued) lii) '

Do these procedure changes reduce the margin of safety as defined in the basis for the Technical Specifications?

Answer: No, because Technical Specification 3.6.C and -

4.6.C are intended to ensure that primary coolant system pipe cracks are detected before they grow to a dangerous or critical size.

Because leakage from the RCIC valve packing was quantified and was, thus no longer unidentified, increasing the unidentified

, leakage rate limit in the procedure by that quantity did not affect the Technical Specification margin of safety. Unidentified leakage was still monitored in accordance with the Technical Specification limit because the procedures limit was merely increased to include the identified quantity from the RCIC valve packing.

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