ML20087H825

Annual 10CFR50.59 Rept,Jan-Dec 1994

ML20087H825
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1994
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20087H824	List: ML20087H820 ML20087H825
References
CCN-95-14012, NUDOCS 9505050029
Download: ML20087H825 (39)

Text

{{#Wiki_filter:. s i i I l i PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 AND 3 DOCKET NOS. 50-277 AND 50-278 l t i t ANNUAL 10 CFR 50.59 REPORT l JANUARY 1,1994 THROUGH DECEMBER 31,1994 TABLE OF CONTENTS i l CCN 95-14012 9505050029 950426 PDR ADOCK 05000277 1 R PDR l

1 t .,1 .1 _ v~ ?l i ' PECO ENERGY COMPANY j PEACH BOTTOM ATOMIC POWER STATION 'i UNIT 2 AND 3 DOCKET NOS. 50-277 AND 50-278 ANNUAL 10 CFR 50.59 REPORT r i TABLE OF CONTENTS l a 1 ? Miscabansous Safety Evaluations.................................... Eage a Engineering Change Request P948013................................ 1 i Engineering Change Request P948319................... ..............1 [ erg::1rs Change Request P948320....................'............. 1 l Engineering Change Request P949058................................ 1 i Engineering Work Request A0793697.................................. 1 j Er$::1rs Work Request A0887997 ..................................2 16-VRR.1 Relief Request............................................ 2 l 2D Residual Heat Removal Hatch Breach................................ 2 I AB8 Qualifications................................................ 2 ARTS MEl.LLA................................................... 3 j Containment Atmosphwie DRution N2 Use During Opwations................. 3 - Control Rod To Exceed 58% Depletion................................. 3 l - Core Design Report............................................... 3 ) Core Operating Umits Report........................................ 4 l Core U/3 CYCLE 10 Core Operating Limits Report......,.................. 4 i Core U/3 CYCLE 10 Core Operating Limits Report......................... 4 i Elimination of Licensee Event Report Commitment......................... 4 j Foodwater Temperature 55 degree F Reduction........................... 5 Feedwater Temperature 20 degree F Reduction........................... 5 i Fuel Bundle Movement........'..................................... 5 l Logic System Functional Testing Performance at Power..................... 5 i Low Level Radweste Factity......................................... 6 i Nuclear Er@:: irs Department Organization ...................,.......6 Oligas Troubleshooting leak.......................................... 6 Operation of P 1.................................................. 8 Operation of PI in Symmetric Mode.................................... 7 POWER RERATE................................................. 7 -l Performance of P t s.............................................. 7 Transverse-Incore-Probe # 2 Out of Service............................. 7 Unit 1 Diesel Generator Removal from UFSAR............................. 8 Unit 1 Security Fence Changes....................................... 8 Coast Down Opwating Map......................................... 8 Dome Pressure during Coastdown.................................... 8 Control Rod Boron Depletion........................................ 9 .l i Operation of Low level R/W Factity.................................. 9 i l l Modifications.................................................. faga j l MOD 0087...................................................... 9 1 MOD 1 106B.................................................... 9 i

En i t L.,* 4 1 t t Modifications (continued)......................................... fage i NKM) 1843..................................................... 10 h4DD 51 30..................................................... 10 h4DD 5151..................................................... 10 NK)D 5109..................................................... 10 [ h4DD 51 73..................................................... 10 hH)D 51 77..................................................... 1 1 i bH)D 51 94..................................................... 1 1 WK)D 5195.................................................... 11 r bH)D 5207...................................................... 12 M @ S M...................................................... Is n MOO 5236.................................................... 12 MOO S248............ ........................................12 bH)D 5252..................................................... 13 bK)D 5269 ................................................... 13 l bH)D 5274 ...................................................13 + h4D D 5 280..................................................... 13 MOD 5281..................................................... 14 hMO D 5290.................................................... 14 MODN.....................................................14 kH)D 5347..................................................... 14 h4D D 53 54.................................................... 15 } h4D D 5357..................................................... 15 I bH)D 5 300..................................................... 15 hH)D 53 70..................................................... 15 i bH)D 5371.................. ..................................16 } MOD M 74..................................................... 16 NK)D 53 75..................................................... 16 hM)D 5383.................................................... 16 hMOD 5387.................................................... 17 NH)D 53 96..................................................... 17 ~ NK)D 5397..................................................... 17 i NK)D P000128.................................................. 17 e WK)D P000190.................................................. 18 M OD P000207................................................. 18 i hK)D P0002 43.................................................. 18 hM)D P000270 ( 5362 ) ...........................................18 bK)D P000287.................................................. 19 Non Conformance Reports........................................ fage i NCR P000559............ ......................................19 NCR P000561 .................................................19 NCR P920317 ......... 19 NCR P900326................................................... 19 NCR NM0741................................................... 20 NCR P930SM) ........... 20 NCP M42 ... M NCR P93DSN!................................................... 20 I CCN 95-14012 i e t 4 j' P e m--

cr. -.. i ..a. .\\ Non Conkumance Reports (Continued)............................... faGR NCR PM0055.................................................. 21 NCR PM0079.................................................. 21 NCR PM0150.................................................. 2, NCR PM0218................................................... 21. NCR PM0252.................................................. 22 1 NCR PM0278.................................................. 22 c NCR PM0344.................................................. 22 NCR PM0374.................................................. 23 NCR PM0888.................................................. 23 3 h e Station Events................................................. Page i Performance Enhancement Program 10001332........................... 23 } Event Investigation Report 2 279.................................. 23 i e Procedures.................................................... fagt PROCEDURE A-12.1............................................ 24 ) PROCEDURE AO-058C.1-2........................................ 24 l PROCEDURE AO463L1 ........ 24 PROCEDURE ERP-101 - ...........................................24 PROCEDURE Emergency Plan...................................... 25 j PROCEDURE Emergency Plan...................................... 25 PROCEDURE Emergency Plan..............................,........ 25 PROCEDURE M-57-013............................................ 25 PROCEDURE MAT-22548.......................................... 26 i PROCEDURE ON.125 & BASES...........................,......... 26 ~ PROCEDURE RT-W-20A-980-2(3).................................... 26 PROCEDURE SP-1368............................................. 26 f PROCEDURE SP 1458............................................ 27 PROCEDURE SP-2014............................................. 27 PROCEDURE ST.C495-835-2....................................... 27 PROCEDURE ST-O-052-110-2 TC 94 884............................... 27 PROCEDURE ST.0452110-2 TC 94-885............................... 27 j f 1 Temoorary Plant Alterations...................................... fage l TPA 2-12-008.................................................... 28 TPA 2-12-008.................................................... 28 i TPA 2-18402................................................... 28 TPA 2-2341 1................................................... 28 { TPA 2-23 -012................................................... 29 i TPA 2-23413................................................... 29 TPA 2 33 005................................................... 29 TPA 2 33408.................................................. 29 tea 2 37437................................................... 29 TPA 2 37 039................................................... 30 I i I CCN 9514012 i i i ~,, 3

m 6- .'l i e . i, Temocrary Plant Alterations (Continued) ..................... fa98 TPA 2-50433......................... .....................30 l TPA 2-52-016.......................... ....................... 30 TPA 2-80419.................................................... 30 TPA 24341 1.................................................... 31 l TPA 341 -041................................................... 31 TPA 342 423................................................... 31 TPA 342425................................................... 31 { TPA 34b 017................................................... 32 t TPA 346419................................................... 32 TPA 3-16 001................................................... 32 TPA 3-23410.................................................. 32 i TPA 3-23 41 1................................................... 33 TPA 342445................................................... 33 k i h t i 4 I t .t k ,I I; t t Y l I I I -L I t t t t I k 6 1 1 i 1 l f 1 l i j CCN 9514012

m_. __m-1 i i PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 AND 3 DOCKET NOS. 50-277; 50-278 ] ANNUAL 10 CFR 50.59 REPORT JANUARY 1,1994 THROUGH DECEMBER 31,1994 SAFETY EVALUATION SUMMARIES i CCN 95-14012

4 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT ECR P946913 Year implemented: Unit 2 (1994) Unit 3 (N/A) This change permanently removed the spool piece on the Unit 2 head spray piping. The piping was replaced with two blind flanges. This change affected a drawing in the Updated Final Safety Analysis Report. This activity did not create any new adverse accidents or safety concems. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. ECR P948319 & P948320 Year implemented: Unit 2 (1994) Unit 3 (1994) These items changed the setpoint of the low differential pressure switch setpoints for the backup instrument nitrogen to Automatic Depressurization System from 0.0 psid to 20.0 psid. This change affected documentation specified in the Updated Final Safety Analysis Report. No new accidents or operating modes or adverse safety conditions were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question. ECR P949058 Year implemented: Unit 2 (1994) Unit 3 (1994) This activity reviewed and evaluated the use of a new type Reactor Pressure Vessel head strong back / carousel and accessories for optional use during a recent refueling outage. The use of this unit was not consistent with the system description spect'ied in the Updated Final Safety Analysis Report. This change has been fully analyzed and did not create any now or different type accident or malfunction. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. EWR A0793597 Year Implemented: Unit 2 (1994) Unit 3 (1994) This evaluation supports the temporary breaching of one of the Main Control Room doors. This was done to support maintenance activities associated with the Main Control Room chillers. The change affected door closure requirements as described in the Updated Final Safety Analysis Report. This condition was evaluated to ensure that the Main Control Room would still be habitable under design events. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Page 1 of 33 CCN 95-14012

g? s b-PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT EWR A0867997 Year implemented: Unit 2 (1994) Unit 3 (1994) This ovaluation justified Emergency Core Cooling Systems, Reactor Core Isolation Cooling, and Diesel Generator operability for river temperatures up to 93 degrees F. The Updated Final Safety Analysis Report specifies that the design inlet temperatures for the Residual Heat Removal heat exchanger be no more than 90 degrees F. The activity will not create any new accidents or adversely affect the consequences of any existing or new type accident. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. 16-VRR-1 Year implemented: Unit 2 (1994) Unit 3 (1994) This safety evaluation revised the in Service Testing (IST) Program Specification (M-710) to add a forward exercise test for the Automatic Depressurization System Instrument Nitrogen check valves. This new testing provided additional Insurance that these check valves would function properly In the forward direction. This change affected the IST Program which is addressed in the Safety Analysis Report. The testing is performed while the system is not required to be operable so no new or additional adverse conditions were created. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. ? 2D Residual Heat Removal Hatch Breach Year Implemented: Unit 2 (1994) Unit 3 (N/A) This Safety Evaluation reviewed the temporary breaching of the Unit 2 Secondary Containment to allow the replacement of the "2D' Residual Heat Removal Pump Motor. This change affected the Secondary Containment system descriptions as specified in the Updated Final Safety Analysis Report. The condition was evaluated and considered to be of minimal consequence based on the additional compensatory actions taken. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. 1 ABB Oualifications Year implemented: Unit 2 (1994) Unit 3 (1994) This evaluation justifies operation of Unit 2 dunng cycle 10 for a limited time with ABB Oualification Fue8 Bundles leading the core in the MAPLHGR thermal limit during steady state operation. This activity affected correspondence between PECO and the NRC. The change did not adversely affect plant operations or safety. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Page 2 of 33 CCN 95-14012

n~; i ..j g 24, j PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 -] Docket No. 50-277 & 50-278 i 199410 CFR 50.59 REPORT ARTS MELLIA Year Implemented: j ) Unit 2 (N/A) Unit 3 (1994) j This evaluation justified the change to the maximum extended load line limit analysis and'the ' Average Power Range Monitor, Rod Block Monitor, Technical Specification" (ARTS) Improvement Program Analysis for Unit 3 cycle 10. The change affected documentation addressed in the Safety Anafysis Report. No new adverse safety concems were created as a result of this activity. Based on the Safety Evaluation and the i above information, it was determined that this change did not constitute an Unreviewed Safety Question. CAD N2 Use During Operations Year implemented: Unit 2 (1994) Unit 3 (1994) l This evaluation justified the use of the Cmtainment Atmospheric Dilution system for Nitrogen makeup during i normal operations. This activity affectac' the system description as specified in the Updated Final Safety Analysis Report. The change did not adverdy affect normal operations or the system's functions during an accident. Based on the Safety Evaluation axi Se above information, it was determined that this change . did not constitute an Unreviewed Safety Question. l Control Red To Exceed 56% Depletion Year implemented: Unit 2 (1994) Unit 3 (N/A) l This evaluation justified the operation of Unit 2 in cycle 10 with control rod 30-23 (A163) at an exposure up to 56% B-10 depletion on the second segment. This change affected depletion limits in NEDE-30931-2-P. Operation in this manner did not result in exceeding the control rod's end of life B-10 equivalent depletion limit. Therefore, no new accidents or consequences were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an l Unreviewed Safety Question. .f Core Desion Report Year implemented: Unit 2 (1994) Unit 3 (N/A) This evaluation addressed the Unit 2 Core Design Report for Cycle 11 operations. The core load was of standard reload fuel and designed to be compylble with tne existing fuel in the reactor. There was no Impact on safety or increase in the probability c faRure. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. l Page 3 of 33 CCN 95-14012

u PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT Core Operatina Umts Reoort Year Implemented: Unit 2 (1994) Unit 3 (N/A) This evaluation was revised to address the Unit 2 CORE OPERATING UMIT Report for CYCLE 10 operations. The revision was made to meet the commitment to the NRC that the Quality Fuel Bundles will not lead the cord wkh respect to thermal limits. This activky affected documentation addressed in the Safety Analysis Report. The change did rot adversely affect any safety liml*.s or operating modes. Based on the Safety Evaluation and the above information, R was determined that this change did not constitute an Unreviewed Safety Question. Core U/3 CYCLE 103QLS Year Implemented: Unit 2 (N/A) Unit 3 (1994) This evaluation justified the changes to the Unk 3 cycle 10 Core Operating Umit Report regarding MAPLHGR limits for all GE-88 fuel. The activity affected documentation addressed in the Safety Analysis Report. No new adverse reactivity er plant safety concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Core U/3 CYCLE 10 COLR Year Implemented: Unit 2 (N/A) Unit 3 (1994) Th ;, evaluation made a revision to the Unit 3 cycle 10 Core Operating umit Report to relax Rod Block Monitor setpoints. The activky affected analysis documented in the Safety Analysis Report. No new reactivky or safety concerns were initiated as a result of the change. Based on the Safety Evaluation and the above infomut:an, it was determined that this change did not constitute an Unreviewed Safety Question. Elimination of LEli Commitment Year implemented: Unit 2 (1994) Unit 3 (1994) This activity eliminatec a commitment which was created as a result of on LER in 1991. The commitment was to perform High Pressure Coolant injection and Reactor Core isolation Cooling Exhaust line draining after turbine operations. Based on operating data and component performance, the draining of the exhaust line is not required and can be eliminated from the survellance tests. Based on the Safety Evaluation and the above information, it was determined that this chango did not constitute an Unreviewed Safety Question. Page 4 of 33 CCN 95-14012

l 1 i PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT i F/W Temoerature 55 dearee F Reduction Year implemented: Unit 2 (1994) Unit 3 (N/A) This safety evaluation allowed operation of Unit 2 with reduced feedwater temperatures of up to 55 degrees F prior to and during cycle extension. This change affected documentation specified in the Safety Analysis Report. The activity did not create any adverse operating modes or conditions or advusely affect plant safed/. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. F/W Temoerature 20 dearee F Reduction Year implemented: Unit 2 (1994) Unit 3 (1994) This evaluation justified the operation of Unit 2 during cycle 10 with a reduced feedwater temperature of up to 20 degrees F prior to cycle extension operation. This change affected Updated Final Safety Analysis Report section 14 transients. No new adverse concems or operating modes were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Fuel Bundle Movement Year implemented: Unit 2 (1994) Unit 3 (N/A) This evaluation addressed the movement of a fuel bundle with only 7 of the 8 tie rods intact. This activity affected system descriptions as described in the Updated Final Safety Analysis Report. No new accidents, transients, or adverse safety concems were created during this evolution. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. LSFT Performance at Power Year Implemented: Unit 2 (1994) Unit 3 (1994) This safety evaluation justified that the performance of Logic System Functional Testing at power is acceptable. This activity affected a commitment that specified that testing should be done while in cold shutdown. No new adverse safety concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. b Page 5 of 33 CCN 95-14012

t ( PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50 278 199410 CFR 50.59 REPGAT i Low Level R/W Facky Year implemented: [ Unit 2 (1994) Unit 3 (1994) l This review evaluated the handling and movement of waste from the area where it is received in processed and packaged from, to placement into storage at the facIlty. These activities affect the factity as described ( in the Updated Final Safety Analysis Report. FacRy operations will not affect the operating units in any way. Based on the Safety Evaluation and the above information, it was determined that those activities did not j constitute an Unreviewed Safety Question. f NED Organizational Changes Year implemented: Unit 2 (1994) Unit 3 (1994) i This evaluation reviewed the Nuclear Engineering Departmen; organizational changes at Chesterbrook. The changes involved the shifting of responsibuities to other organizations. This activity impacted wording in the Updated Final Safety Analysis Report chapter 13. These changes do not make any physical changes to the station and did not impact plant safety. Based on the Safety Evaluation and the above information, it was d6termined that these changes did not constitute an Unreviewed Safety Question. i Offaas Troubleshootina leak Year implemented: Unit 2 (N/A) Unit 3 (1994) l i The safety evaluation justified Off Gas system manipulations in order to identify the source of a leak. This activity operated the Off Gas system in a manner not specified in the Updated Final Safety Analvsis Report. No new accidents, transients, or adverse operating modes were created during the perfomance of this troubleshooting. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. i Ooeration of P1 Year implemented: Unit 2 (1994) j Unit 3 (N/A) This evaluation justified the operation of Us A 2 during cycle 10 with the P1 program in symmetric mode with l Group 18,20, and 21 asymmetric by no more than eight positions (four notches). This actMty affected system operations as described in the Updated Final Safety Analysis Report section 3.6. Plant respont* to a transient remained bounded by the safety analyses. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewad Safety Question. I i Page 6 of 33 i CCN 95-14012 ~

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT Ooeration of PI in Svmmetric Mode Year implemented: Unit 2 (1994) Unit 3 (N/A) This evaluation justified the operation of Unit 2 during cycle 10 with the P1 program in symmetric mode with groups 18,20, and 21 asymmetric by no more than six positions. This allowed asymmetric control rod pattem operation with the leaker suppression control rods inserted. This change affected section 3.6.5 of the Updated Final Safety Analysis Report. No new adverse concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Power Rerate Year implemented: Unit 2 (1994) Unit 3 (not implemented yet) This evaluation justified the revision of a General Electric Nuclear Energy document Power Rerate Licensing Report with respect to " Average Power Range Monitor, Rod Block Monitor, Technical Specification" (ARTS) . The change affected documentation specified in the Safety Analysis Report. The activity did not adversely affect plant safety or any operating modes or accidents. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Performance of Pts Year Implemented: Unit 2 (1994) Unit 3 (N/A) This evaluation justified the operation of Unit 2 during cycle 10 with the P1 Program in symmetric mode with groups 18,20,21, and 26 asymmetric by no more than 8 positions (4 notches). The activity affected the systems operational descriptions as specified in the Updated Final Safety Analysis Report. No adverse safety concems were created as a result of this change. Based on the Safety Evaluation and the atxrn ltrmation, it was determined that this change did not constitute an Unreviewed Safety Question. TIP # 2 Out of Service Year implemented: Unit 2 (N/A) Unit 3 (1994) This evaluation justifies the operation of Unit 3 during Cycle 10 with the 'B' Traversing in core Probe (TIP) out of service. The indexer for the 'B' TIP was not capable of performing its function to allow detector access to portions of the core to determine local power distributions. Symmetric TIP location traces are used as substitute values. The use of asymmetric control rods are to be reviewed on a case by case basis. No new adverse safety concems were created as a result of the activity. Based c1 the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Sa'ety Question. Page 7 of 33 CCN 95-14012 i

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT U/1 D/G Rergwal from UFSAR Year implemented: Unit 2 (1994) Unit 3 (1994) This evaluation reviewed the removal of all references to the Unit 1 Diesel Generator in the Updated Final Safety Analysis Report. The diesel was used as a backup electrical power supply to the Unit 1 Technical Support Center (TSC). This was done due to the recent completion of a modification (5396) which installed a new backup electrical power source to the Unit 1 TSC. The new power supply is for Station Blackout Requirements and is fed from a nearby hydroelectric plant. No new adverse safety or electrical concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question. U/1 Security Fence Chance Year Implemented: Unit 2 (1994) Unit 3 (1994) This 50.59 review addresses the proposed change to move the Unit 1 containment security fence to allow for training activities in the southem lay down area adjacent to the containment structure in the Unit 1 facility. This change allows for the security fence to be removed with no degradation in security at Unit 1. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. U/2 Coast Down Ooeratino Mao Year Implemented: Unit 2 (1994) Unit 3 (N/A) This evaluation justified expanding the allowable operating domain of the power / flow map on Unit 2 during cycle 10 coast down. This activity affected the allowable ranges for operation as described in the Updated Final Safety Analysis Report. This change did not reduce the margin of safety or adversely affect any accident initiators and no adverse safety conditions were created. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. U/2 Dome Pressure durino CD Year implemented: Unit 2 (1994) Unit 3 (N/A) l This safety evaluation justified constant reactor steam dome pressure of 1055 psig during Unit 2 cycle 10 end of cycle coastdown operat'ons. This was done to maximize total energy output at this point in the cycle. Operations in this mode is different than as specified in the Updated Final Safety Analysis Report. 1 No adverse safety concerns were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Page 8 of 33 CCN 95-14012

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT Control Rod Boron Deoletion Year implemented: Unit 2 (1994) Unit 3 (N/A) This evaluation allowed control rod blade A163 located at position 30-23 to exceed 56% boron-10 depletion on Unit 2 for cycle 10. This activity affected boron depletion limits as specified in the Safety Analysis Report. Fuels evaluated this condition and determined that it did not adversely affect plant safety or reduce the margin of safety. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Ooeration of Low Level R/W Facility Year implemented: Unit 2 (1994) Unit 3 (1994) This evaluation reviewed the current operating practices of the Low Level Radwaste Storage Facility at Peach Bottom. The review involved the handling and movement of waste in the facility. The process affected UFSAR Section 9.3.4. No new safety concems or possible unplanned exposure or releases were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that these operating practices did not constitute an Unreviewed Safety Question. MOD 0887 Year implemented: Unit 2 (1994) Unit 3 (1993) This modification upgrades the Reactor Recirculation A & B speed control loops, scoop tube positioner, master controller, dual speed controller, speed indication circuits, and the reset logics. The enhancements improved operation and reliability. No safety concems were created. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. MOD 11068 Year Implemented: Unit 2 (1994) Unit 3 (1994) This modification reworked the condensate phase separator interfaces and internals to minimizo plugging and improve resin processing operations. This actMty affected documentation addressed in the Safety Analysis Report. No safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Page 9 of 33 CCN 95-14012

1 I !y. l PEACH BOTTOM ATOMIC POWER STATION I UNIT 2 & 3 ) Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT MOD 1843. Year implemented. Unit 2 (1994) Unit 3 (1994) This modification replaced the existing Reactor Feed Water Control system with an digital control type. .i system and performed post modification testing. This activity affected the systems operation as described in the Updated Final Safety Analysis Report. No new adverse safety concems were created as a result of this change. Based on the Safety Evaluation and the above information, k was determined that this change did not constitute an Unreviewed Safety Question. j t l MOD 5130 Year implemented: [ Unit 2 (1994) [ Unit 3 (1994) This modification upgraded the existing Containment Atmosphere DHution instrumentation loops it also replaced instruments in the Containment Atmosphere Control instrumentation loops The upgraded components more accurately measure the loop processes. No safety concems were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question. l t MOD 5151 Year implemented: Unit 2 (1994) l Unk 3 (1994) j i This modification provided a time delay in the high flow isolation signal to the solenoid valves (SV) in the l nitrogen supply lines to the Automatic Depressurization System Safety Relief Valves. The piping downstream [ of the SVs depressurizes with time due to piping leakage. When the SVs are ope'wd, the initial rush of nkrogen to repressurize the piping automatically closes the SVs due to excessive high flow. The time delay will provide enough time to repressurize the piping without an umented isolation of the SVs and ensure the l a large piping leak would be adequately isolated. No safety concoms were created. Based on the Safety l Evaluation and the above information, it was determined that this change did not constitute an Unreviewed i Safety Question. l

i MOD 5169 Year implemented:

Unit 2 (19P3) j Unit 3 (1994) j This modification replaced the eight existing Exide battery chargers wkh new seismically qualified Class 1E charger assemblies. The existing chargers were approaching the end of their life. This ac.tivity enhanced 4 the system's reliability. The change affected figures addressed in the Updated Final Safety Analysis Report and no adverse safety concems were created. Based on the Safety Evaluation and the above information. - l R was determined that this change did not constituts an Unreviewed Safety Question. t Page 10 of 33 l CCN 95-14012

1 s: PEACH BOTTOM ATOMIC POWER STATION- \\ UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT l MOD 5173 Year implemented Unit 2 (1994) l i Unit 3 (1994) -1 5 This modification converted the Standby Gas Treatment system charcoal filter deluge system from automatic l operation to manual operation. The new system consists of a manuel block valve in an accessible location. l The object of the modification was to minimize the risk of inadvertent operation of the deluge system. No safety concems were created. Based on the Safety Evaluation and the above information, it was determined { that this change did not constitute an Unreviewed Safety Question. i I MOD 5177 Year implemented: Unit 2 (1994) Unit 3 (1993) t This modification relocated the Automatic Depressurization System annunciatorwindows in the Main Control i Room to improve human factors. This change affec'ed documentation addressed in the Updated Final I Safety Analysis Report. The modification did not adversely affect plant operations or create any safety concems. Based on the Safety Evaluation and the above information, it was determined that this change j . did not constitute an Unreviewed Safety Question. MOD 5194 Year implemented: Unit 2 (1994) - 1 Unit 3 (not implemented yet) l l This modification installed a pressure switch in the Shutdown Cooling Valve (MO-17) to prevent an unwanted spurious opening during power operation in the event of an Appendix R fire. The change did not create any new adverse safety concems or operating modes. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. j MOD 5196. Year implemented:' Unit 2 (1994) Unit 3 (1993) This modification provided a permanent instrument loop that is capahle of an enhanced display of Reactor Water level during Refueling operations. This change affected documentation specified in the SAR. No new refueling safety issues or concems were created as a result of this change. Based on the Safety Evaluation i and the above information, it was determined that this change did not constitute an Unreviewed Safety l Question. j 5 l Page 11 of 33 j CCN 95-14012 i

~ PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 I 199410 CFR 50.59 REPORT MOD 5207 Year implemented: Unit 2 (1994) Unit 3 (19M) i This modification upgraded the Raw Water and Domestic Water Systems. This change affected the system l descriptions as specified in the Updated Final Safety Analysis Report. No new adverse plant operating

onditions or safety concems were introduced by this activity. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question.

MOD 5232 Year implemented: Unit 2 (1994) Unit 3 (1994) This modificction installed new storage racks in the Low Level Radwaste Building. This activity will provide increased storage capabilities and the use of these racks does not create any adverse conditions that would effect plant operations or radiation exposures. Based en the Safety Evaluation and the above information, it was determined that this change did not constituto an Unreviewed Safety Question. I MOD 5236 Year implemented: Unit 2 (1992) Unit 3 (1994) This modification was inAalled to provide a hardened torus vent in accordance with Generic Letter 89-16. The main objective of the vent is to mitigate the consequences of a long term loss of decay heat removal. This was beyond the plant licensing basis and assumes, with the exception of the Residual Heat Removal system, that all other systems are operational and the core is not in a degraded condition. The design of the torus hardened vent has been analyzed in accordance with 10CFR50.59 and Generic Letter criteria. No safety concems were created. This change was made to comply with an NRC commitment and provides plant safety improvements in an outside of design bases event. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. MOD 5248 Year Implemented: Unit 2 (not implemented yet) Unit 3 (1994) This modification removed the area radiation monitor from service from the new fuel storage vault. The fuel storage vault has never been used since new fuel is placed directly into the fuel pool. This change affected documentation addressed in the Updated Final Safety Analysis Report. No new accident or adverse safety concerns were created during the modification. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Page 12 of 33 CCN 95-14012

M a e._4 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 l Docket No. 50-277 & 50-278 i 199410 CFR 50.59 REPORT l MOD 5252 Year implemented: Unit 2 (1994) L-Unit 3 (1994) l This modecation upgraded the radiation monkoring capabbity of the HPSW radiation montoring system. This activky provided a more reliable system. The change allected figures addressed in the Updated Final Safety Analysis Report and no adverse safety concems were created. Based on the Safety Evaluation and _ the above irdormation, k was determined that this change did not constitute an Unreviewed Safety Question. 9 MOD 5209 Year implemented: f Unit 2 (1993) Unit 3 (1994) l t This modification added a new non safety related DC electrical system to replace the existing BOP battery. it also moved non safety related loads from the oefety related DC electrical systems to the new non safety related DC systems. This change affected documentation specified in the Updated Final Safety Analysis Report No safety concems were introduced as a result of this activity. Based on the Safety Evaluation and j the above information, it was determined that this change did not constkute an Unreviewed Safety Question. i MOD 5274 Year implemented: I Unit 2 (1994) Unit 3 (1993) This modification replaced the existing CAC/ CAD analyzers with improved instrumentation to improve operations and reliab5ty. The activity made the system operation and layout different from that described In the Updated Final Safety Analysis Report. No safety concoms were introduced as a result of this actMty. Based on the Safety Evaluation and the above information !! was determined that this change did not constkute an Unreviewed Safety Question. MOD 5280 Year implemented: Unit 2 (1994) Unit 3 (1994) This modification replaced the existing seismic monkoring equipment wkh an improved system. The new system is functionally equivalent but provides improved operation and reliability. No safety concems were { created. Based on the Safety Evaluation and the above information, k was determined that this change did not constitute an Unreviewed Safety Question. l ? i P Page 13 of 33 { CCN 95-14012 p- ,r,., e w. v v w

l i PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT MOD 5281 Year implemented: l Unit 2 (1994) I Unit 3 (1994) This modification replaced the existing Main Control Room Ventilation radiation monitoring system with an f upgraded and more reliable system. The existing system was obsolete. This activity affected documentation addressed in the Safety Analysis Report and no adverse safety concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. MOD 5290 Year implemented: Unit 2 (1994) Unit 3 (1993) This modification installed safety related indicators located within several Reactor vessel pressure Instrumentation loops to verify loop functionality. Station technicians in the past needed to perform lengthy tests to determine loop functionality to satisfy the Technical Specification requirements. This change eliminates these tests since the operators can routinely survey the indicators during their normal shift checks. No safety concems were created. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. MOD 5344 Year implemented: Unit 2 (1992) Unit 3 (1994) This modification added vibration instrumentation to the High Pressure Coolant injection turbine-pump assembly. This instrumentation does not affect the pressure retaining boundaries of the High Pressure Coolant injection (HPCI) system or operational modes. The data available from this instrumentation was incorporated into the HPCI Surveillance Testing so that the origin of HPCI pump vibration can be determined. This activity will enhance system testing. There is no impact on system capability or operation. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Queshn. [ MOD 5347 Year implemented: Unit 2 (1994) Unit 3 (1993) The modification installed improved flow meters at several locations on the Emergency Service Water system. These flow meters are used to provide a means of determining the flow rates at various lccations i to support testing and maintenance activities. No safety concems were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question. 3 P Page 14 of 33 f l CCN 95-14012 I

=j q . PEACH BOTTOM ATOMIC POWER STATION I UNIT 2 & 3 1 Docket No. 50-277 & 50-278 l 199410 CFR 50.59 REPORT !l MOD 5354 Year implemented: Unit 2 (1994) Unit 3 (1994) I i This modification was designed to provide improved chlorine delivery in the Water Treatment System and added improved chlorine monitoring. This activity affected documentation addressed in the Updated Final [ Safety Analysis Report. No new adverse safety concems or new adverse operating conditions were created i as a result of this change. Based on the Safety Evaluation and the above information, it was determined ~ that this change did not constitute an Unreviewed Safety Question. j MOD 5357 Year implemented: I Unit 2 (1994) [ Unit 3 (1994) This modification provided remote operation to enable transfer of liquid radwaste from the water collector i tank by the addition of a Air Operated Valve instead of a Manually Operated Valve. In addklon, a manuel l valve was installed on the Floor Drain Collector Tank to facultate easier liquid transfers. This change affected l system operations as specified in the Updated Final Safety Analysis Report. The activity did not create any i new adverse operating modes or accidents. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. i MOD 5300 Year implemented: I Unit 2 (1994) Unit 3 (1994) This modification added a new Service Water Bay chemical injection system to treat the intake structure for macro fouling organisms before excessive growth occurs. The new equipment was permanently installed j to replace the portable system which was previously in use. This activity affected figures addressed in the i Updated Final Safety Analysis Report and no adverse ~*ety concems were introduced. Based on the Safety l 4 Evaluation and the above information, k was determr.ed that this change did not constitute an Unreviewed ) Safety Question. i MOD 5370 Year implemented: 7 Unit 2 (1994) i Unit 3 (1994) 'l [ This Reactor Water Cleanup system modification replaced the three existing 50% capacity pumps with two l 100% capacity purnps and uporaded the existing fRter domineralizer precost cycle control circuitry and l manual isolation valves. It also installed new piping to support the change. This modification improved the l Reactor Water Cleanup System by providing better overall system performance. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed l Safety Question. i j Page 15 of 33 CCN 95-14012 ~, ..,..-.-..,,-o n ~.,-, r

~. -_m q; R l ~- i PEACH BOTTOM ATOMIC POWER STATION i UNIT 2 & 3 Dccket No. 50-277 & 50-278 '199410 CFR 50.59 REPORT MOD 5371 Year implemented: Unit 2 (1994) I Unit 3 (1993) l ~ This modification provided the Main Generator wth water in leakage detection equipment. The new

instrumentation installed monkors the dew point of the hydrogen gas in the generator. This actMty affected.

l documentation addressed in the Updated Final Safety Analysis Report. The change is an enhancement during generator operations and did not adversely allect plant safety or operations. Based on the Safety i Evaluation and the above information, R was determined that this change did not consttute an Unreviewed j Safety Question. l MOD 5374 Year implemented: Unit 2 (1994) Unk 3 (1994) This modification allows the operation of the plant to include the extended operating region bounded by the i rod line which passes through the 100% power / 75% core flow point (approximately 121% rod line). The technical analysis is referred to as the Maximum Extended Load Line Limit (MELLLA) analysis. In addition, this modification also incorporated the Average Power Range Monitor, Rod Block Monitor, and Technical Specification improvement (ARTS) Program which increases plant operating efficiency by updating the - t thermal limits requirements and improving plant instrumentation responses and accuracy. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an j Unreviewed Safety Question. MOD fi2Z5_ Year implemented: Unit 2 (1994) i Unit 3 (1994) i This' modification enhanced several alarm annunciator logics to allow no alarms to be up during normal operations. Five alarm annunciators were normally up during power operations. This change allowed the alarms to be cleared during power operations. This affected a logic figure in the Updated Final Safety Analysis Report. This change does not create any new adverse operation conditions or adverse safety t oncoms. Based on the Safety Evaluation and the above information, R was determine 6 that these changes j did not constkute an Unreviewed Safety Question. .l 5 MOD 5383. Year implemented: l Unit 2 (1994) Unit 3 (1993) This modification added a manual block valve in the equalizer line of the Residual Heat Removal System testable check valve, which is a containment isolation valve. This will allow for the performance of a Local i Leak Rate Test to positively determine whether the testable check valve or equalizer valve is leaking. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute i an Unreviewed Safety Question. l t Page 16 of 33 CCN 95-14012 l . i

~. t l PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 L 199410 CFR 50.59 REPORT l MOD 5387 Year Implemented: Unit 2 (1994) - Unit 3 (1994) This modification replaced the existing DryweH Leak Detection System with an improved and more reliable i type monitor. The change affected documentation addressed in the Updated Final _ Safety Analysis Report._ l No new adverse safety concems or new operating modes / transients were introduced by this act'/ity. Based on the Safety Evaluation and the aixwe information, k was determined that this change did not constitute an Unreviewed Safety Question. l MOD 5396 Year implemented: Unit 2 (1994) Unit 3 (1994) This modification installed an addition bus and associated swkchgear to support station black out. This activky affected documentation specified in the Safety Analysis Report. No new accider ts, transients, or i adverse operating modes were created as a result of this change and electrical reliabilky was increased as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. MOD S397 Year implemented: Unit 2 (1994) Unit 3 (1994) l - This modification provided various design changes to the Emergency Diesel Generators. These enhancements involved the lube ou, auxNiary inter cooler, Jacket coolant pump, and the air start systems. No safety concems were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.~ j I MOD P000128 Year implenumted: Unit 2 (1994) j Unit 3 (1993) j i This modification was installed to provide a continuous backful system to the reference legs associated with l Reactor Water Level Instrumentation. This activity wRl enhance level indication reliabuky. This change affected dovumentation addressed in the Safety Analysis Report. No adverse safety concems or new 1 operating modes were created as a result of this change. Based on the Safety Evaluation and the above j information, t was determined that this change did not constitute an Unreviewed Safety Question. l Page 17 of 33 i CCN 95-14012

?, PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 ' Docket No. 50-277 & 50-278 l 199410 CFR 50.59 REPORT i MOD P000190 Year implemented; j Unit 2 (1994) Unit 3 (N/A) 'l This modification replaced seven valve motor actuators with larger type units. This was done to most increased thrust demand requirements. The change affected documentation addressed in the Updated Final Safety Analysis Report. No new adverse safety concems were created during this activity. Based on the Safety Evaluation and the above information, k was determined that this change did not constitute an Unroviewed Safety Question. i MOD P000207 Year implemented: j Unit 2 (1994). Unit 3 (not implemented yet) This modification changed the Feedwater / Recirculation Flow Control system instrumentation, alarms, and i runback logica. The change affected documentation specified in the Updated Final Safety Analysis Report section 7.9. The activity did not adversely affect plant operations or accident analysis. Based on the Safety i Evaluation and the above information, it was determined that this change did not constitute an Unreviewed .y Safety Question. MOD P000243 Year implemented-Unit 2 (1994) Unit 3 (1994) i This modification installed an equalization line and access platform for the Diesel Driven Fire Pump test i header valve. This activity affected documentation addressed in the Updated Final Safety Analysis Report. i No new adverse safety concems or operating modes were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an i Unreviewed Safety Questioit j t ? MOD P000270 (5362) Year Implemented: Unit 2 (1994) Unit 3 (not implemented yet) i This modification upgraded the Reactor Pressure Vessel Bottom Head Drain line to improve the accuracy. ] of the line temperature measurement to minimize unnecessary shutdowns, power reductions and i stratifications of the bottom head. This actMty affected documentation addressed in the Updated Final-Safety Analysis Report. This change is an enhancement and does not create any new adverse safety concems or new plant transients or accidents. Based on the Safety Evaluation and the above information,- It was determined that this change did not constitute an Unreviewed Safety Question. Page 18 of 33 CCN 95-14012

9.- PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docicet No. 50-277 & 50-278 199410 CFR 50.59 REPORT I MOD P000287 Year implemented: Unit 2 (1994) i Unit 3 (not implemented yet) This modification revised reactor water level measurement pressure compensation to account for increased operating pressures and increased ambient temperatures in the drywell and reactor buliding. This change affected documentation specified within the Safety Analysis Report. No new adverse system operations or now concems were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. i NCR P000559 & P000561 Year implemented: S Unk 2 (1994) Unit 3 (1994) p These Non Conformance Reports identified and resolved discrepancies between the plant and the Fire l Protection Program figures. This activity made the figures match the installed configuration. No new adverse safety concems were created and no new operating modes were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question. i NCR P920317 Year implemented: Unit 2 (1994) Unit 3 (1994) i This Non Conformance Report evaluation addressed what maintenance activities are required on the spare Emergency Core Cooling System Room Coolers to ensure Erwironmental Qualifications are maintained. This change affected documentation addressed in the Updated Final Safety Analysis Report. The activity did not affect plant operations or safety. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. NCR P930326 Year implemented: Unit 2 (1994) Unit 3 (1994) This_ Non Conformance Report evaluated the action levels associated with a river flooding scenario.' The i action levels were lowered by 2 feet due to the design of the Circulating Water Pump Structure. The structure was not desigrwd to support the current action levels. This change affected documentation in the Updated Final Safety Analysis Report. This change is conservative and did not affect plant safety. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute + an Unreviewed Safety Question. Page 19 of 33 CCN 95-14012

i, e PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT t NCR P930741 Year implemented: Unit 2 (1994) Unit 3 (1994) l This Non Conformance Report changed the Motor Driven Fire Pump Logic to prevent it from automatically loading on to the EDG on a loss of power condition. This change affected documentation addressed in the Fire Protection Plan. This activity did not create any new adverse safety concems or adverse plant safety conditions.' Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. NCR P930809 Year implemented: Unit 2 (1994) t Unit 3 (1994) This evaluation was performed to support the weld repair of the Emergency Service Water Booster Pump piping. This activity affected system descriptions as specified in the Updated Final Safety Analysis Report. j System line up and operations have been evaluated and no adverse safety concerns were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. i NCR P93QB4_2 Year implemented: Unit 2 (1994) Unit 3 (1994) { This Non Conformance Report addressed a change to the Emergency Service Water flow criteria for the Emergency Core Cooling System and Reactor Core Icolation Cooling Room Coolers. The change affected documentation previously specified in the Updated Final Safety Analysis Report, The change did not adversely affect plant operations, safety, or accident analysis. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. P NCR P930882 Year implemented: Unit 2 (1994) Unit 3 (1994) l This Non Conformance Report addressed discrepancies between the as found condition and the "E 1305 Panel Electrical Schedules". The Updated Final Safety Analysis Report was affected by this condition. An evaluation was performed which ensured that the as found conditions were acceptable and that electrical i margins were still acceptable. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Page 20 of 33 CCN 95-14012

4 ' 'J -l q i i PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 j 199410 CFR 50.59 REPORT l NCR P940055 ~ Year impiomonted: Unit 2 (1994) l Unit 3 (1994) ] This Non Conformance Report alkmed the use of Bussman JHC100 fuses in a Motor Control Center which . supplies power to a electrical distribution panel (20B3842). The fuses were installed as part of the initial Appendix R modification. This change affected fuse configurations as specified in the Fire Protection Plan. - No new adverse safety concems, accident modes, or equipment falures were created as a result of this 1 activity. Based on the Safety Evaluation and the above information, it was determined that this change did not constkute an Unreviewed Safety Question. i i NCR P940079 Year implemented: Unit 2 (1994) Unit 3 (N/A) This Non Conformance Report removed the Unit 2 Torus Vacuum Breaker three degree position swkches l until they are permanently replaced with a new type in Refuel Outage 2R11. These switches are speellically l 4 addressed in the Updated Final Safety Analysis Report. Removal of these switches wil not adversely affect plant safety or create any new accident or transient initiators. Based on the Safety Evaluation and the above l information, it was determined that this change did not constitute an Unroviewed Safety Question. j 1 l i NCR P940150 Year implemented: ] l Unit 2 (1994) .l Unit 3 (N/A) This Non Conformance Report replaced the 480 voit magnetic only type circuit breaker associated with a j Radwaste area radiation monitor (20B060-21) to a thermal magnetic type breaker due to multiple high j impedance fault considerations. This change affected protective device configurations as specified in the i Fire Protection Plan. No new adverse safety concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an I Unreviewed Safety Question. i NCR P940218 Year implemented: 1 Unit 2 (N/A) ' Unit 3 (1994) '] This Non Conformance Report evaluated a calculation associated with fuse configurations and required the installation of Bussmann FRS-R-150 fuses in series with several existing magnetic trip devices. This change affected fuse coordination figures located in the Fire Protection Program. No new accidents, transients, or adverse operating modes were created as a result of this change. Based on the Safety Evaluation and the at'ove information, it was determined that this change did not constitute an Unreviewed Safety Question I Page 21 of 33 i CCN 95-14012 i .j

. 4 _,j..

7 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 1 Docket No. 50-277 & 50-278 l l 199410 CFR 50.59 REPORT NCR P940252 Year implemented: Unit 2 (1994) Unit 3 (1994) l' This Non Conformance Report evaluated the Torus Narrow Range 1.svel Transmitters design range of 13.7 to 15.7. A recent survey indicated that level could only be measured on this instrument from 13.5 to 15.6, therefore, it could not be measured between 15.6-and 15.7. The Non Conformance Report was dispos 41oned to use as is. This activity affected Torus water level ranges as described in the Safety Analysis Report. Since levels are normally maintained between 14.5 and 14.F, no new accidents, transients, or malfunctions are created. Based on the Safety Evaluation and the above information, k was determined that this change did not constitute an Unreviewed Safety Question. l NCR P940278 Year implemented: Unit 2 (1994) Unit 3 (1994) This Non Conformance Report evaluated changing the Diesel Generator voltage regulator setting range between 4160 and 4400 volts. This change affected system descriptions as specified in the Updated Final Safety Analysis Report. This activky did not adversely affect system operations, plant safety, or the possitAlty of a transient, accident, or malfunction of equipment. Based on the Safety Evaluation and the above information, k was determined that this change did not constkute an Unreviewed Safety Question. NCR P940344 Year implemented: Unit 2 (1994) Unk 3 (1994) This review addressed the use-as-is disposition of a Non Conformance Report. The Non Conformance Report identified three improper assumptions for the Emergency Diesel Generator loading tables in the Updated Final Safety Analysis Report. The calculations associated with these tables have been revised. This - activky affected diesel loading calculations and ks associated tables in the Updated Final Safety Analysis Report. This change did not create any new adverse electrical loading concoms and did not affect plant safety or reliabhy. Based on the Safety Evaluation and the above information, k was determined that this change did not constitute an Unreviewed Safety Question. Page 22 of 33 CCN 95-14012 i

PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT NCR P940374 Year implemented: Unst 2 (1994) Unit 3 (1994) This Non Conformance Report evaluated and justified continued operation of the Unit 2 reactor with weld crack indications identified on the shroud. This activky made changes to the Updated Final Safety Analysis Report since initial review and approval were not based on these crack indications. Calculations and analysis performed on the as found conditions verified that no new adverse safety concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. NCR P940868 Year implemented: Unit 2 (N/A) Unit 3 (1994) This modification installed the VOTES Torque Cartridge on MO-J 10425A to perform monthly diagnostic testing. The activity removed the handwheel The handwheel is described in the Updated Final Safety Analysis Report for this valve. This change did not adversely affect the operation of the valve or associated system during normal or emergency system operations. Based on the Safety Evaluation and the above information, it was determined that this change did not consthute an Unreviewed Safety Question. Performance Enhancement Procram 10001M Year implemented: i Unit 2 (1994) Unit 3 (1994) This safety evaluation was written to evaluate the consequences of a missed fire watch. This condition resulted in a Technical Specification violation. The condition was temporary in nature due to a personnel error. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Event Investiaation Reoort 2-93-279 Year implemented: Unit 2 (1994) Unit 3 (1994) This evaluation addressed an event were a high radiation door was not locked and secured as specified in the Technical Specifications. No plant safety or operating concems were created as a result of this event. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question. Page 23 of 33 CCN 95-14012

l,* 4 I PEACH BOTTOM ATOMIC POWER STATION ' i UNIT 2 & 3 r Docket No. 50-277 & 50-278 f-199410 CFR 50.59 REPORT PROCEDURE A-12.1 Year implemorded: Unit 2 (1994) Unit 3 (1994) I This procedure revised the guidance invoMng actions for fire system impairments The change affected the associated actions specified in the Fire Protection Program. This actMty did not create any adverse safety l condition or create any new or different concems. Based on the Safety Evaluation and the above [ information, k was determined that this change did not constitute an Unreviewed Safety Question. j PROCEDURE AO-058C 1-2 Year implemented: l Unit 2 (1994) i Unit 3 (N/A) This procedure provided the instruction needed for supplying temporary electrical power to lighting panel 95L from a motor control center. This was done to avert the loss o! electrical power to Off Gas Stack Sampling systems and fire code devices during bus outage actMties which de-energize the power source I This change affected electrical lineups as specified in the Updated Final Safety Analysis Report. No unsafe i configuration was created as part of this actMty. Based on the Safety Evaluation and the above information, i it was determined that this change did not constitute an Unreviewed Safety Question. f 1 PROCEDURE AO4)S3L1 Year implemented: l Unit 2 (1994) l Unit 3 (1994) f This procedure provided the steps needed for supplying temporary electrical power to the DMsion i Main Control Room radiation instrumentation and Drywell Radiation instrumentation. The electrical system lineup was different than that specified in the Updated Final Safety Analysis Report. No new accident, transients, or operating modes were created by this lineup. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. I PROCEDURE ERP-101 Year implemented: Unit 2 (1994) i Unit 3 (1994) l t This procedure was revised to provided references and clarifications to an Emergency Response i implementing Procedure. The actMty affected documentation in the Updated Final Safety Analysis Report Appendix 0. This change did not adversely affect plant safety or the station's ability to respond during an f emergency. Based on the Safety Evaluation and the above information, it was determined that this change -{ did not constitute an Unreviewed Safety Question. l l Page 24 of 33 CCN 95-14012 i . ~ I

a PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT PROCEDURE Emeroency Plan Year implemented: Unit 2 (1994) Unit 3 (1994) This procedure was revised to address changes in the plant evacuation process and the removal of the Conference Center. Evacuating personnel wNI go to Unit 1 instead of the Conference Center. This change affected documentation in the Safety Analysis Report and NRC commitments The activity did not adversely affect plant operations or Emergency evacuation processes. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. PROCEDURE Emeroency Plan Year implemented: . Unit 2 (1994) Unit 3 (1994) This evaluation justified revisions to address the Power Rerate Project. The change revised the plan to remove any references to the station's maximum power capacity which is described in the Updated Final Safety Analysis Report. No new adverse safety concems were created as a result of this change. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. PROCEDURE Emeroency Plan Year Implemented: Unit 2 (1994) Unit 3 (1994) This evaluation justified revisions to the Emergency Plan to clarify Emergency Response Organization personnel responsibilities and authorities, corrected typographical errors, and incongruities. This change affected the previous revision to the EP Plan. This activity did not adversely affect plant safety or personnel response during an emergency. Based on the Safety Evaluation and the above information, it was i determined that this change did not constitute an Unreviewed Safety Question. i PROCEDURE M-57-013 Year implemented: Unit 2 (1994) Unit 3 (1994) This procedure provided the means to perform maintenance activities on safety related battery cells whle maintaining the battery and DC power supply operable. This activity affected documentation specified in Updated Final Safety Analysis Report section 8. This procedure did not adversely affect plant operations during procedure performance. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. 1 4 Page 25 of 33 CCN 95-14012

q. 1 e: 1 PEACH BOTTOM ATOMIC POWER STATION l UNIT 2 & 3 j Docket No. 50-277 & 50-278 i 199410 CFR 50.59 REPORT PROCEDURE MAT-22548 Yearimplemented: Unit 2 (1994) Unit 3 (1994) This evaluation supports the Modification Acceptance Testing (MAT) for installation of a third Off She i Electrical Power Source under Modification 2254. This activity is considered a Test not described in the j Salsty Analysis Report. This testing did not adversely allect plant operations and safety. Based on the l Salsty Evaluation and the above information, it was determined that this change did not consttute an j Unreviewed Safety Question. I i PROCEDURE ON-125 & BASES Year implemented: Unit 2 (1994) - Unit 3 (1994) l This procedure provided the necessary instructions to restore Shut Down Cooling following an inadvertent i Shut Down Cooling isolation whue the reactor pressure head is removed This activty allected the " Loss j of Shut Down Cooling

• scenario addressed in the Updated Final Safety Analysis Report. This change did not adversely affect any described accidents or transients Ramari on the Safety Evaluation and the above information, t was determined that this change did not constitute an Unreviewed Safety Question.

l l l l' PROCEDURE RT-W-20A-960-2(3) Year implemented: Unt 2 (1994) Unit 3 (1994) This procedure change incorporated modification P000421 which unplugged floor drains in the plenums for i the Turbine Bunding (TB), TB Equipment Cell, Reactor Bulding (RB), and the Refuel Floor (RF) supply vont i fans. These drains were unplugged because of past water and ice accumulation which resulted in a j significant slipping hazard for personnel. This change allected a 1981 commitment in response to IE Bulletin j 80-10 end an Inspection Report 92-07. No new adverse plant safety conditions were created as a result of l this change. Based on the Safety Evaluation and the above information, t was determined that this change did not constitute an Unreviewed Safety Question. PROCEDURE SP-1368 Year implemented: Unit 2 (1994) Unit 3 (1994) .j q - This procedure allowed the operation of the Core Spray pumps with the Emergency Service Water flow to the motor oR coolers isolated. This was done to obtain test data to support elimination of the ou coolers. This change affected system descriptions regarding Core Spray motor ou coolers. Operation in this mode did not create any new concems Based on the Safety Evaluation and the above information, k was determined that this change did not constitute an Unreviewed Safety Question. Page 26 of 33 CCN 95-14012 ,_ a

c. f. o *. ~ [' PEACH BOTTOM ATOMIC POWER STATION i UNIT 2 & 3 Docket No. 50-277 & 50-278 l 199410 CFR 50.59 REPORT PROCEDURE SP-1458 Year inpemented: Unit 2 (1994) Unit 3 (N/A) This procedure removed the 5th stage feedwater heaters from service on Unit 2 cycle 10 prior to cycle i extension allowing a final feedwater temperature reduction of up to 55 degrees F. This change affected documentation specified in the Safety Analysis Report. Operation in this condition has been analyzed and j . no new adverse condklons were created as a result of this activity. Ramad on the Safety Evaluation and the above information, it was determined that this change did not constkute an Unreviewed Salsty Question. f PROCEDURE SP-2014 Year implemented: Unit 2 (1994) Unit 3 (1994) This special test verified that the EHC control system can adequately control reactor pressure for power at a re-rated condition of 105%. This was a test not described in the Updated Final Safety Analysis Report. j PECO and GE have completed a study and concluded that the Turbine Control System is adequate for i operations at rerate condklons. Plant safety was not affected by the test. Based on the Safety Evaluation i and the above information, it was determined that this change did not constkute an Unreviewed Safety Question.

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PROCEDURE ST-C495-835-2 Year implemented: Unit 2 (1994) Unit 3 (N/A) i i This evaluation revised a Survegiance Test to allow the use of portable composite sampling systems on the j Circulating Water intake and Discharge. These portable units were required due to equipment problems associated with the permanently installed systems. This change affected the system descriptions in the Updated Final Safety Analysis Report. The new systems are more reliable and do not create any adverse safety concems. Based on the Safety Evaluation and the above information, it was determined that this j change did not constitute an Unreviewed Safety Question. j i i PROCEDURE ST-O-052-110-2 TC 94-884 & 885 Year implemented: Unit 2 (1994) Unk 3 (1994) l i This procedure, Diesel Generator Simulated Automatic Actuation and Load Acceptance Test, was changed 1 due to Unit 3 being shutdown with a Residual Heat Removal Shutdown cooling loop inservice The test 1 could not be adequately performed whBe in this configuration. A jumper was added to the circuit to facultate test performance. This activity affected system descriptions as specified in the Updated Final Safety Analysis Report. This change did not create any new type transient or malfunction which would affect plant safety. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. Page 27 of 33 CCN 95-14012

7[N 1 q PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 j Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT ] I TPA 2-12 008 Year Implemented: Unit 2 (1994) i Unit 3 (N/A) ' This Temporary Plant Alteration removed a flow glass lens on the Reactor Water Clean Up system and installed a steel plate to stop a sealing surface from leaking. The steel plate will be in place untN either a new flow glass or a straight pipe is installed This change affected figures in the Upda:ed Final Safety 3 Analysis Report. This activity did not adversely affect system operations or create any new safety concerns-Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. i TPA 2-12-008 Year Implemented: i Unit 2 (1994) Unit 3 (N/A) l This Temporary Plant Alteration deleted the low flow, high flow, and high vibration trips associated with the "2A" and "2C" Reactor Water Clean Up pump trip logics due to spurious tripping This configuration is different than that specified in the Updated Final Safety Analys's Report. The change wHl not create any new type accidents, transients, or malfunctions of equipment important to safety. Based on the Safety Evaluation and the above information, it was determined that this change did rect constitute an Unreviewed Safety Ouestion. I TPA 2-18-002 Year implemented: Unit 2 (1994) l Unit 3 (N/A) This Temporary Plant Alteration was installed due to the use of the GE In-Core sipping hoods which i necessitate the alteration of a refueling platform grapple interlock feature in order to un grapple the hood when it is pested on the core. The interlock involved is one which senses the elevation of the main hoist when the platform is over the core. This change affected system configurations as described in the Updated Final Safety Analysis Report. This condition was analyzed and no new adverse safety concems were created. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. TPA 2-23e t Year implemented: l Unit 2 (1994) i Unit 3 (1994) f ~ This Temporary Plant Alteration was installed to jumper several Motor Control Center start logic relays associated with the High Pressure Coolant injection system. This activity affected documentation addressed in the Safety Analysis Report. This change does not create any new adverse safety conditions but wRI i enhance High Pressure Coolant injection component reliabuity Based on the Safety Evaluation and the i above information, it was determined that this change did not constitute an Unreviewed Safety Question. 'l 4 Page 28 of 33 l CCN 95-14012 l ,m, [

--. f.b( i PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 . Docket No. 50-277 & 50-278 ( 199410 CFR 50.59 REPORT TPA 2-23-012 Year implemented: l Unit 2 (1994) ~ Unit 3 (N/A). This Temporary Plant Alteration was installed to jumper several Motor Control Center start logic relays associated with the High Pressure Coolant injection system. This activity affected documentation addressed in the Safety Analysis Report..This change does not create any new adverse safety conditions but wil enhance High Pressure Coolant injection component reliabutty. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. t i TPA 2-23-013 Year Implemented: Unit 2 (1994) .l Unit 3 (N/A) l This Temporary Plant Alteration installed a temporary drain line on the High Pressure Coolant injection f system turbine drain pot due to a leaking valve. This change affected the system description of turbine draining as described in the Updated Final Safety Analysis Report. The change did not adversely affect system performance or accident mitigation. Based on the Safety Evaluation and the above information, it t was determined that this change did not constitute an Unreviewed Safety Question. TPA 2-33405 & 008 Year implemented: Unit 2 (1994) Unit 3 (1994) These Temporary Plant Alterations failed open solenoid valves which supplied cooling water to Emergency ) Core Cooling Systems, the Reactor Core Isolation Cooling system, and the Emergency Diesel Generators j due to solenoid valve sticking problems. This was done untR a permanent solenoid replacement is i completed. This affected system operations as defined in the Safety Analysis Report. No new adverse i safety concems were created since all systems stui function per their design. Based on the Safety l Evaluation and the above information, it was determined that these changes did not constitute an j Unreviewed Safety Question. t i TPA 2-37437 Year implemented: Unit 2 (1994) Unit 3 (1994) This Temporary Plant Alteration was installed to isolate a ground on the Unit 3 Main Transformer heat detection circuit from the plant DC power system. This change affected system description as specified in l i the Fira Protection Program. The activity does not adversely affect plant operatione or the performance of any safety systems. Based on the Safety Evaluation and the above information, it was determined that this i change did not constitute an Unreviewed Safety Question. Page 29 of 33 CCN 95-14012

~ t. djf*.J e PEACH BOTTOM ATOMIC POWER STATION i UNIT 2 & 3' Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT i TPA 2-37 039 Year implemented: Unit 2 (1994) i Unit 3 (1994) This Temporary Plant Alteration was installed to isolate an electrical short on the Unit 3 Auxiliary Transformer j heat detection circuit and prevent an inadvertent discharge. This change affected system performance as described in the Fire Protection Program. The activity did not adversely affect normal plant operations or the function of any safety system. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Untm%wed Safety Question. TPA 2-50-033 Year implemented: Unit 2 (1994) ] Unit 3 (N/A) = This Temporary Plant Alteration was installed to allow the clearing of the ISOPHASE BUS TROUBLE alarm in the Unit 2 Control Room. This alarm was being brought in by a local Hi HUMIDITY alarm at the bus. This activity temporar#y defeated the HI HUMIDITY alarm to allow operators to be aletted to other potential isophase bus problems. This changa affected the system description as specified in the Updated Final Safety Analysis Report. The Temporary Plant Alteration did not adversely impact any safety system actuations or plant operations. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. TPA 2-5241,0 Year implemented: Unit 2 (1994) Unit 3 (1994)' This Temporary Plant Alteration removed the E-1 EDG Scavenging Air Cooler Standby Valve disc since it had separated from the valves stem. The system will be restored to its normal configuration in a future outage. This activity affected documentation addressed in the Safety Analysis Report. The valve is maintained in its safety position and did not create any new adverse conditions. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. TPA 240-019 Year implemented: Unit 2 (1994) ) Unit 3 (N/A) j This Temporary Plant Alteration allowed the monitoring of various system parameters with temporary equipment during power ascension testing. This change aflected system descrip*lons as specified in the Safety Analysis Report. Isolators have been installed as needed and no adverse safety concems were created duilng this activity. Based on the Safety Evaluation and the above information, it was determined that this changs did not constitute an Unreviewed Safety Question. Page 30 of 33 CCN 95-14012

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i PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 ~ Docket No. 50-277 & 50-278 J 199410 CFR 50.59 REPORT f 4 TPA 243 011 . Year implemented: Unk 2 (1994) Unit 3 (1994) This Temporary Plant Alteration installed a strip chart recorder to address concems associated with the Main Control Room Radiation Monitors experiencing spurious spiking The equipment recorded system fluctuations from a steady :date condition to be monitored. This activity affected documentation specified in the Safety Analysis Repc,rt. The change did not adversely affect system operations or plant safety in any. way. Based on the Safety Evaluation and the atxwe information, it was determined that this change did not constitute an Unreviewed Safety Question. l: TPA 3-01-041 Year implemented: Unit 2 (1994) Unit 3 (N/A) This Temporary Plant Alteration installed temporary push buttons in the Cable Spreading Room on the EHC pressure control circuits. This allowed the raising and lowering of the setpoint from this altamate loc' xm. This affected the system as described in the Updated Final Safety Analysis Report. No new adverse safety concems were created as a result of this activity. Based on the Safety Evaluation and the above Information, it was determined that this change did not constitute an Unreviewed Safety Question. l e TPA 3-02-023 Year Implemented: Unit 2 (N/A) Unit 3 (1994) This Temporary Plant Alteration was installed on Unit 3 Recirculating Motor Generator Set to monitor various operational parameters. This activity affected documentation addressed in the Safety Analysis Report. No adverse safety concems, new operating modes, or plant transients were created at. a result of this activity. Based on the Safety Evaluation and the above information, it was determined that this change did not j constitute an Unreviewed Safety Question. TPA 3-02 025 Year implemented: Unit 2 (N/A) Unit 3 (1994) This Temporary Plant Alteration was Installed to defeat a false alarm associated with an Excess Flow Check Valve which is masking several other alarm circuits. This change affected system descriptions as specified In the Updated Final Safety Analysis Report. No new accidents, equipment failures, or safety concems were l created as a result of this activity. Based on the Safety Evaluation and the above information, it was j determined that this change did not constitute an Unreviewed Safety Question. t Page 31 of 33 CCN 9514012

.,, y c% 2-e a PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199410 CFR 50.59 REPORT j -i TPA 34SA1Z Y6er Implemented-i Unk 2 (N/A) l Unit 3 (1994) j This Temporary Plant Alteration temporarey installed jumpers on ON Gas flow swkches to prevent inadvertera Off Gas system isolations during an effort to repair a steam leak on the Off Gas / Recombiner system. This i activity aNected Off Gas system descriptions as specified in the Updated Final Safety Analysis Report. No .l safety concems were introduced during this repair activity. Based on the Safety Evaluation and the above information, k was determined that this change did not constkute an Unreviewed Safety Caestion. 5 l ) TPA 3-08419 Year implemented: Unit 2 (N/A) Unit 3 (1994) This Temporary Plant Alteration provided an air increase into the Off Gas system to dRute oN gas concentrations and potentially reduce the impact of off gas leaks on airbome activity. This change affected l the system description as specified in the Updated Final Safety Analysis Report. No new accident modes or equipment failures were created as a result of this change. Based on the Safety Evaluation and the above j information, k was determined that this change did not constkute an Unreviewed Safety Question. l l TPA 3-18-031 Year impic:nonted: Unit 2 (N/A) i Unit 3 (1994) j This Temporary Plant Alteration installed a test tee connection and a block valve on the instrument nkrogen I compressor to support troubleshooting activities on the pressure rupture disc, This change affected figures j specified in the Updated Final Safety Analysis Report.' The Temporary Plant Alteration did not adversely

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affect plant safety or system operatFons. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unreviewed Safety Question. TPA 3-23-010 Year Implemented: Unit 2 (1994) Unk 3 (1994) This Temporary Plant Alteration was installed to jumper several Motor Control Center start logic relays associated with the High Pressure Coolant injection system. This activky affected documentation addressed { In the Safety Analysis Report. This change does not create any new adverse safety conditions but wRI enhance High Pressure Coolant injection component reliabRky. Based on the Safety Evaluation e.nd the above information, it was determined that this change did not constkute an Unreviewed S fety Question. l t l Page 32 of 33 i CCN 95-14012 l 1; J

- -.~ ~_ - - .t l o 1 l.! PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 l Docket No. 50-277 & 50-278 j 199410 CFR 50.59 REPORT TPA 3-23-011 -Yoar implemented. Unit 2 (N/A) Unit 3 (1994) This Temporary Plant Alteration was installed to jumper several Motor Control Center start logic relays j associated wth the High Pressure Coolant injection system. This activity affected documentation addressed j in the Safety Analysis Report. This change does not create any new adverse safety conditions but does. 1 enhance High Pressure Coolant injectica component reliabilty. Based on the Safety Evaluation and the above information, it was determined that this change did not constitute an Unnwiewed Safety Question. j t t } TPA 342-045 Year implemented: U d 2 W /A) i Unit 3 (1994) This Temporary Plant Alteration was installed to clear a faulty Control Rod Drift Alarm by the installation of } a jumper. This change affected the system description as specified in the Updated Final Safety Analysis Report. The activity did not create any new adverse operating modes, accidents, or transients. Based on l the Safety Evaluation and the above information, it was determined that this change did not constitute an j Unreviewed Safety Question. i i i i f i f ) i Page 33 of 33 CCN 95-14012 i l -.y.,_ m .i,,.y. r m..., __,}}