ML20063J921
ML20063J921 | |
Person / Time | |
---|---|
Site: | Peach Bottom ![]() |
Issue date: | 12/31/1993 |
From: | Rainey G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
References | |
NUDOCS 9402240384 | |
Download: ML20063J921 (31) | |
Text
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CersId R. Rtiney.
Vice President Peach Bortom Atomic Power Station t
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PECO ENERGY
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Delta, PA 17314-9739 717 456 7014 i
February 10, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Docket Nos. 50-277 50-278 Ucense Nos. DPR-44 DPR-56
SUBJECT:
Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Annual 10 CFR 50.59 Report For The Period January 01,1993 through December 31,1993
Dear Sir:
Enclosed is the 1993 Annual 10 CFR 50.59 Report as required by 10 CFR 50.59 (b).
if you have any questions or require additional information, please contact us.
Sincerely, GRR/AJW/GAJ Attachment cc:
R.A. Burricelli, Public Service Electric & Gas W.P. Dornsife, Commonwealth of Pennsylvania R.I. McLean, State of Maryland T.T. Martin, Administrator, Region I, USNRC W. Schmidt, USNRC Senior Resident inspector H.C. Schwemm, Atlantic Electric C.D. Schaefer, Delmarva Power CCN 94-14012 18'15a ggy7 94o224oss4 9st2s PDR ADoCK 05 coo 277 R.
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M. C. Kray J. A. Bernstein Commitment Coordinator Correspondence Control Desk E. J. Cullen A. J. Wasong T.J.Robb D. M. Smith CCN 94-14012
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Docket Nos.50-277 50-278 1993 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT i
This report is issued pursuant to reporting requirements for Peach Bottom Atomic Power Station Units 2 and 3 (Facility Ucense Numbers DPR-44 and l
DPR-56 respectively). This report addresses tests and changes to the facility and procedures as they are described in the Peach Bottom Final Safety Analysis Report. This report consists of those tests and changes that were implemented between January 1,1993 and December 31,1993.
A safety evaluation for each item has concluded that no unreviewed safety questions, as defined in 10 CFR 50.59 (a) (2), were involved.
CCN 94-14012
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PECO ENERGY COMPANY-PEACH BOTTOM ATOMIC POWER STATION UNIT 2 AND 3 DOCKET NOS. 50-277 AND 50-278 3
ANNUAL 10 CFR 50.59 REPORT JANUARY 1,1993 THROUGH DECEMBER 31,1993 TABLE OF CONTENTS F
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l PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION l
UNIT 2 AND 3 DOCKET NOS 50-277 AND 50-278 l
ANNUAL 10 CFR 50.59 REPORT j
TABLE OF CONTENTS J
l Miscellaneous Safety Evaluations...
.... Page MISC. 50.59 A0354012 1
MISC. 50.59 A0391485
..........1 MISC. 50.59 A0761522
.1 MISC. 50.59 Blocking open MO-3-10-25A
.1 1
MISC. 50.59 Core Design Report.....
..2 MISC. 50.59 Core Design Report...
2 MISC. 50.59 Core Design Report........
MISC. 50.59 Core Operating Limits Report.
.2 MISC. 50.59 Core Operating Limits Report 2
MISC. 50.59 LPRMs
.2 MISC. 50.59 Leaking Fuel Replacement
.3 l
MISC. 50.59 NSSS10.SDD.......
3 MISC. 50.59 Oxygen injection
...3 MISC. 50.59 PM-846 RHR Seal Flow Rate Calc
.....3 i
MISC. 50.59 RPV Pressure at 1000 psig.
...3 MISC. 50.59 RPV Shroud Cracking
.4 MISC. 50.59 Reactor vessel 68 degree limit 4
MISC 50.59 TIP # 1 Out of Service
.4 MISC. 50.59 TIP # 1 Out of Service
...4 MISC. 50.59 TIP # 3 Out of Service
....5 Modifications MOD 0887
...........................................5 MOD 1140
.5 MOD 1829
........5 MOD 1830
.6 MOD 1832
...6 MOD 5095.
6 MOD 5107 6
MOD 5123
.........7 MOD 5128
...7 1
MOD 5169
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MOD 5177 7
MOD 5195
..7 MOD 5205
...................................8 MOD' 5219
........8 MOD 5220 8
MOD 5229
..........8 MOD 5231
.8 MOD 5233...
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Modifications (continued)..
Paae MOD 5235
.9 MOD 5244
.........9 MOD 5249
........9 MOD 5260
................... 10 MOD 5269
. 10 MOD 5274
. 10 MOD 5276A......
10 MOD 52768.
. 10 MOD 5286
. 11 MOD 5290
. 11 MOD 5304
. 11 MOD 5336
.......... 11 MOD 5343
. 11 MOD 5347
. 12 MOD 5349
. 12 1
MOD 5355
..... 12 MOD 5393
.... 13 MOD 5394
. 13 MOD 5401
. 13 MOD P000128
. 13 MOD RWM Software changes (RWM003)
. 14 Non Conformance Reoorts NCR P900206.....
. 14 NCR P910165........
. 14 NCR P910165..
........................... 14 NCR P910591..
. 14 NCR P910592....
. 15 NCR P920134....
.............. 15 NCR P920254.....
. 15 NCR P920257...
. 15 NCR P920614.................
. 15 NCR P920777....
........... 16 NCR P920971...
. 16 NCR P930021......
. 16 NCR P930076.............
16 NCR P930130
. 16 NCR P930562..
. 17 NCR P930754.......
. 17 t
CCN 94-14012
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1 Proceduros PROCEDURE AO-20A.1
......... 17 PROCEDURE AO-52E.1
........ 17 PROCEDURE COL-14A.1.A-2(3)
..... 17 PROCEDURE Emergency Plan
..... 18 PROCEDURE OM43.2
. 18 PROCEDURE RT-D-33-600-2.............
..... 18 PROCEDURE SE-11 ATTACHMENTS A-H.J-N,P-Z
...... 18 PROCEDURE S12F-63M-761-XXC2(TC 93-270)
.. 18 PROCEDURE SO-14A.1.A-2(3)
...... 19 PROCEDURE SP-1413
................19 PROCEDURE SP 1444
.................... 19 PROCEDURE SP-1475
............. 19 PROCEDURE SP-1476 19 PROCEDURE SP-2002
. 20 PROCEDURE SP-2004
... 20 PROCEDURE SP-2006 20 PROCEDURE SP-2011
. 20 Temoorary Plant Nterations TPA 2-01G-032.....
.. 20 TPA 2-23410
. 21 TPA 2-30-012 21 TPA 2-50-022
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. 21 TPA 2-52-014 TPA 3-02-020
. 21 j
TPA 342-022
... 22 TPA 3-13411
............... 22
. 22 TPA 3-18-005 TPA 3-23409....
.......... 22 TPA 3-57-007
........... 22 23 TPA 3-57-007
...... 23 TPA 3-62444.
CCN 94-14012
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PECO ENERGY COMPANY l
PEACH BOTTOM ATOMIC POWER STATION l
UNIT 2 AND 3 DOCKET NOS. 50-277; 50-278 l
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.i ANNUAL 10 CFR 50.59 REPORT JANUARY 1,1993 THROUGH DECEMBER 31,1993 SAFETY EVALUATION SUMMARIES
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~I CCN 94-14012
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4 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MISC. 50 59 A0354012 Year Implemented: U/2(N/A) U/3(1993)
This activity allowed temporary operation of the Off Gas Recombiner system without steam flow indication and the low steam flow interlocks. In order to perform repairs of the instrumentation, the trips had to be defeated to ensui9 that the system remains on line. This activity affected documentation addressed in the SAR. No new arN9rse safety concerns were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC,50.59 A0391485 Year implemented: U/2(1993) U/3(1993)
This evaluation corrected the UFSAR Section 5.2.5.1 " Primary Containment integrity and Leak Tightness" which resolved discrepancies between the UFSAR and the Technical Specifications. This activity was administrative in nature and did not create any new operating modes or impact plant safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 50.59 A0761522 Year implemented: U/2(N/A) U/3(1993)
This evaluation reviewed and approved a proposal to maintain a constant reactor pressure by periodically increasing turbine throttle pressure setpoint during Unit 3 cycle 9 power coast down. This activity affected General Electric coast down analysis referenced in UFSAR section 3.2.4. The change will not adversely affect any existing operating modes or create any new operating modes or transient conditions. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 50.59 Ooeratina with MO-3-10 25A open Year Implemented: U/2(N/A) U/3(1993)
This safety evaluation addressed Low Pressure Coolant injection system operability while MO-3-10-25A is maintained in the open position verses the closed as specified in the UFSAR. This was a temporary condition until valve repairs were completed during the next shutdown outage. This change did not create any new adverse safety concems. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 50.59 Core Deslan Reoort Year implemented: U/2(N/A) U/3(1993)
This evaluation addressed the Unit 3 Core Design Report for Cycle 10 operations. The core load was of standard reload fuel and designed to be compatible with the existing fuel in the reactor. There was no impact on safety or increase in the probability of failure. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 1 of 23 CCN 94-14012
4 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT I
MISC. 50.59 Core Deslan Reoort Year implemented: U/2(N/A) U/3(1993)'
This evaluation addressed the Unit 3 Core Design Report for cycle to operations. The core load was designed to be compatible with existing fuel in the reactor. There was no impact on safety or increased probability of a failure. Based on the Safety Evaluation and the above information, it was determined that i
these changes did not constitute an Unreviewed Safety Question.
MISC. 50.59 Core Desian Reoort Year implemented: U/2(N/A) U/3(1993)
This evaluation addressed the Unit 3 Core Design Report for CYCLE 10 operations The core load was standard GE reload fuel and was designed to be compatible with the existing fuel in the reactor. There was no impact on safety or increase the probability of failure. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i MISC. 50.59 Core Ooeratina Limits Renort Year implemented: U/2(N/A) U/3(1993)
I This evaluation addressed the Unit 3 CORE OPERATING LIMIT Report for CYCLE 10 operations. It provided APLHGR, MCPR, Kf, LHGR, and RBM flow bias setpoints. These values have been determined using NRC-approved methodology and are established such that all applicable limits of the plant safety analysis are met. No safety concerns were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety.
Question.
I MISC. 50.59 Core Ooeratina Limits Reoort Year implemented: U/2(N/A) U/3(1993)
This evaluation addressed the Unit 3 CORE OPERATING UMIT Report for Cycle 10 operations. It provided the necessary reactor flux parameters. These values have been determined using approved methodology and are established such that all applicable limits of the plant safety analysis are met. No safety concerns were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
A MISC. 50 59 LPRMs Year implemented: U/2(1993) U/3(N/A)
The Safety Evaluation justified the operation of Unit 2 during cycle 9 with LPRM strings 56-41 and 56-33
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i having their cables swapped. It was assumed that during the refueling outage preceding cycle 9 operation, j
the cables for these LPRM strings were swapped in the subpile room where they connect to the LPRM. The impact of this cable swap to Tech Spec thermal limits, APRM and RBM operability and the impact to.
exposure accounting for fuel and control blade: has been reviewed. Data from cycle 9 show very little change from the LPRM swap and no impact to the Technical Specifications or core component exposure accounting. No safety concems were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 2 of 23 CCN 94-14012 I
I PEACH BOTTOM ATOMIC POWER STATION l
UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT l
MISC. 50 59 Leakina Fuel Reolacement Year implemented: U/2(N/A) U/3(1993)
This evaluation justified the continued use of the thermal operating limits as specified in the Unit 3 Cycle 9 Core Operating Limits Report for a revised core loading pattern and the replacement of several fuel bundles.
All original limits were maintained. No safety concems were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute
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an Unreviewed Safety Question.
f MISC. 50.59 NSSS10.SDD Year implemented: U/2(N/A) U/3(1993)
This activity modified the software associated OD-1 to support operation with an inoperable 3A TIP indexer.
This activity affected documentation addressed in the UFSAR. This change did not adversely impact plant operations or safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i MISC 50.59 Oxvaen iniection Year implemented: U/2(N/A) U/3(1993)
This Safety Evaluation allowed the permanent installation of equipment to support Oxygen injection during plant operations. This change affected figures and documentation addressed in the SAR. No new adverse safety concerns or new operating modes were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 5659 PM-846 RHR Seal Flow Rate Calc. Year implemented: U/2(1993) U/3(1993)
This calculation eliminated the need for ESW flow requirements to the RHR Pump Seal Coolers and testing of seal cooler flow rate is not longer required. This activity affected values specified in a NRC Safety Evaluation dated 12/23/91. This change did not adversely affect plant safety or create any adverse operating modes. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 50.59 RPV Pressure at 1000 osia Year implemented: U/2(N/A) U/3(1993)
This evaluation allowed the raising of Reactor pressure to 1000 psig which is within current analysis. This activity allowed reactor power to remain closer to 100% during the end of cycle coastdown. This activity affected documentation addressed in the SAR. No new adverse safety concerns were generated as a result of the increased reactor pressure at end of cycle coastdown. Based on the Safety Evaluation and the above information, it was determined that the changes did not constitute an Unreviewed Safety Question.
Page 3 of 23 CCN 94-14012
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MISC. 50.59 RPV Shroud Crackino Year implemented: U/2(N/A) U/3(1993)
This evaluation addressed the cracks identified in the Unit 3 Reactor Pressure Vessel Core Shroud and I
concluded that continued operation to the next Refueling Outage was acceptable. This activity affected documentation addressed in the UFSAR. This change does not significantly affect plant safety or the effects 4
of a plant accident or transient. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 50 59 Reactor vessel 68 decree limit Year implemented: U/2(N/A) U/3(1993)
This evaluation discussed the impact of reactor coolant temperatures dropping below 68 degrees as specified for Peach Bottom Unit 3 Reload 9 Cycle 10. This review concluded that this change does not affect safety of the plant or change the design functions. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 50.59 TIP # 1 Out of Service Year implemented: U/2(N/A) U/3(1993)
I The evaluation justified the operation of Unit 3 during cycle 9 with the "A" Traversing in-core Prove (TIP) out of service. The detector was not capable of performing its function to allow access to portions of the core to determine local power distribution. Symmetric TIP location traces were used as substitute values as long as the core was operated in an octant symmetric control rod pattern. No safety concerns were created.
e Based on the Safety Evaluation and the above information, it was determined that these changes did not
-i constitute an Unreviewed Safety Question.
MISC 50.59 TIP # 1 Out of Service Year implemented: U/2(N/A) U/3(1993)
This evaluation justifies the operation of Unit 3 during cycle 9 with the "A" Traversing in-core Probe (TIP) machine out of service and the TIP channel C-3 inaccessible. The indexer for the "A" TIP was not capable r
I of performing its function to allow detector access to portions of the core to determine local power distribution. Symmetric TIP location traces were used as substitute values as long as the reflected TIP' channels were unperturbed by local power differences caused by control rod insertion. The use of asymmetric control rods were reviewed on a case by case basis. No new adverse safety concerns were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i Page 4 of 23 CCN 9414012
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MISC. 50 59 TIP # 3 Out of Service Year implemented: U/2(N/A) U/3(1993)
This review justified the safe operation during Cycle 9 with the 'C' Traversing in-core Probe Channel 3 out-of-service. Operating with octant symmetric control rod pattems assures validity of substitute TIP traces for the out-of-service TIP channel. The TIP system cannot act as an initiator of any other type of transient and cannot cause failure of other equipment important to safety, therefore, no safety concerns were created.
i Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 0887 Year Implemented: U/2(N/A) U/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 j
improved operation and reliability. No safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety i
Question.
MOD 11g Year Implemented: U/2(1993) U/3(1993)
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This modification increased the storage capacity of the fuel pool. This enhancement affected documentation addressed in the UFSAR. No new adverse safety concerns or new operating modes were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1829 Year Implemented: U/2(1992) U/3(1993)
This modification provided automatic sprinkler protection for the Reactor Feed Pump areas. These sprinklers are designed to initiate when ambient temperatures rise to the melting point of fusible material on the sealed sprinkler heads. The flow of water energizes a pressure switch which transmits an alarm condition to the fire protection panel in the control room. This modification was completed in compliance with NFPA 13,
" Standard for the installation of Sprinklers" The change affected documentation addressed in the SAR. No safety concerns were created as a result of this change. No adverse affects on safety related equipment were created by this change. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 5 of 23 CCN 94-14012
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MOD 1830 Year implemented: U/2(1992) U/3(1993)
This modification provided an automatic fire protection system for the areas located under the turbine pedestals. These sprinklers are designed to initiate when ambient temperatures rise to the me{ ting point of the fusible material on the sprinkler heads. The flow energizes a pressure switch which transmits an alarm condition to fire protection panel in the control room. This modification was done in accordance with NFPA
- 13. This modification also removed a 4* pipe downstream of the isolation valve and replaced it with a 6" pipe to support additional water requirements. This change affects documentation addressed in the SAR.
No safety concems were created as a result of this change. No adverse affects on safety related equipment was created by this change. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1832 Year Implemented: U/2(1991) U/3(1993)
This activity installed an automatic fire protection sprinkler system over the front of the Main Turbine Generator. This change is an enhancement but did affect documentation addressed in the UFSAR. The new system will not adversely affect plant safety or any system operations. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5095 Year Implemented: U/2(1992) U/3(1993)
This modification upgraded the Emergency Cooling Water System. Vents were installed in the ESW/HPSW pump structures and level instrumentation standpipes. The electrical power supplies for LT-2804 'A', 'B' were upgraded by moving the 24 Volt D.C. supply from non-seismically qualified sources to qualified sources. Upgraded level controllers were also evaluated for use if the present controllers fall. Trip setpoint pressure for PS-0821 'A',
'B' were decreased from 2 psig to 12.1" Hg (vacuum) to mitigate tripping Emergency Service Water booster pumps. ESW flow was also returned to one ECT cell. This was done to ensure cavitation does not occur in other modes of operation and to enhance operations and reliability.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MQ,D 5107 Year implemented: U/2(1993) U/3(1993)
This modification resolved problems with the liquid waste drains in the lower level of the Reactor Buildings.
This activity was done to enhance use of the drains and to provide additional insurance that flood protection is maintained in these elevations. This change affected documentation addressed in the UFSAR. No adverse safety concerns or new operating modes were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 6 of 23 CCN 94-14012
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UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MOD 5123 Year implemented: U/2(1993) U/3(1992)
This modification provided access to the Reactor Water Clean-up lsolation Valve Room through the 'A' Reactor Clean-up Pump Room. This was done by removing a block wall and installing a door between the pump room and valve room. This increases personnel safety and productivity because ladders for access to the valve room will no longer be necessary. This actMty affected drawings addressed 'n the SAR. The structural integrity of the area, security and ALARA were maintained. There was no effect on plant performance or equipment capability. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5128 Year implemented: U/2(1992) U/3(1993)
This modification replaced instrument valves and reworked instrument line tubing associated with Reactor Pressure Vessel Pressure instnimentation. This actMty was done on the high point vents for ease of calibration. This change affected documentation addressed in the UFSAR. No adverse safety concems or new operating modes were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Ouestion.
i MOD 5169 Year Implemented: U/2(1993) U/3(not implemented yet) j This modification replaced the eight existing Exide battery chargers with new seismically qualified Class 1E charger assemblies. The existing chargers were approaching the end of their life. This activity enhanced the systems reliability. The change affected figures addressed in the UFSAR and no adverse safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5177 Year Implemented: U/2(not implemented yet) U/3(1993).
This modification relocated the Automatic Depressurization System (ADS) annunciator windows in the Main Control Room to improve human factors. This activity improved human factors for the control room operators. This change affected documentation addressed in the UFSAR. The modification did not adversely affect plant operations or create any safety concerns. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5195 Year implemented: U/2(not implemented yet) U/3(1993)
This modification provided a permanent instrument loop that is capable of an enhanced display of Reactor Water level during Refueling operations. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 7 of 23 CCN 94-14012 1
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MOD 5205 Year Implemented: U/2(1993) U/3(1993)
I This modification changed the configuration of the Reactor Core isolation Cooling system overspeed trip circuit contactor. This activity enhanced operations and maintenance. It affected documentation addressed in the UFSAR. This activity did not adversely affect system reliability or create any new adverse safoty -
concerns. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
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MOD 5219 Year implemented: U/2(N/A) U/3(1993)
This modification replaced several instantaneous circuit breakers with thermal type to increase reliability.
This change affected documentation addressed in the SAR. No new operating modes or adverse safety concerns were created as a result of this breaker change out actMty. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
l MOD 5220 Year implemented: U/2(not implemented yet) U/3(1993) i This modification installed new fuses for ine 2(3)0YO36 electrical distribution panel. This activity was done to support fuse configuration concerns. This change affected documentation addressed in the UFSAR. No adverse safety concerns or new operating modes were created as a result of this activity. Based on the i
Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5229 Year implemented: U/2(1993) U/3(1993)
This modification installed vibration instrumentation on the Emergency Diesel Generator Ventilation Supply i
Fans to support monitoring of fan vibrations. This change affected documentation addressed in the SAR.
No new adverse safety concerns or new operating conditions were created as a result of this change.
i Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5231 Year Implemented: U/2(1992) U/3(1993) '
This modification replaced obsolete instrumentation in the condensate flow loops with state-of-the-art instrumentation. These replacernents consist of flow transmitters, square root extractors, summers, indicators, controllers, and recorders. The overall function of the flow loops will remain the same. This activity is an enhancement to operation. No safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 8 of 23 CCN 9414012 i
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MOD 5233 Year implemented: U/2(1992) U/3(1993) i This modification removed the Reactor Core Differential Pressure Instrument loop. The indication loop was designed to provide a signal to the Indicator in the control room. This signal was originally designed as an alternate method to monitor core differential pressure during initial startup and testing and was no longer required or utilized. This activity affected documentation addressed in the SAR. No safety concems were created. There is no impact on plant capability. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i MOD 5235 Year Implemented: U/2(1992) U/3(1993)
This modification modified the refueling platform to improve reliability, reduce fuel handling time and ease future maintenance activities. This modification only affects the refuel platform and refueling interlocks. The l
modification did not affect platform structural integrity or load carrying capabilities. No new safety concems were introduced. Because reliability and efficiency were improved, this change is an enhancement to overall plant operations. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
I MOD 5244 Year implemented: U/2(1992) U/3(1993) l This modification added a valved demineralized water supply line and a liquid level gage for the loop seal on the offgas radiation monitor sample line drains. This modification was installed to provide a readily visible means to determine whether there is sufficient water in the loop seal. In addition, a connection to supply makeup water was provided if the level is insufficient. There are no system interface changes which would affect plant safety. This change results in improved monitoring capability. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5249 Year Implemented: U/2(1992) U/3(1993)
This modification installed a General Electric Zinc injection Passivation (GEZIP) system. Deposition of cobalt-60 in the primary piping system causes contact dose rates to increase and results in higher occupational radiation exposure during drywell maintenance activities. The GEZIP system which adds I
soluble zine to the BWR reactor water has been shown to considerably reduce Cobalt-60 buildup in the primary piping system. This modification meets all design, material, and construction standards applicable to the systems and structures affected. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 9 of 23 CCN 94-14012
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MQD#2@
Year implemented: U/2/1993) U/3(not implemented yet)
This modification installed a chemical injection system at the service water bay. This new system will allow the ability to inject a molluscicide into the service water systems. The system is interried to prevent the growth and accumulation of raw water macro fouling organisms such as zebra mussels and asiatic clams.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5269 Year Implemented: U/2(1993) U/3(N/A)
This modification added a new non safety related DC electrical system to Unit 3 to replace the "E" BOP battery and replaced non safety related loads from the safety related systems. No safety concerns were introduced as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5274 Year implemented: U/2(N/A) U/3(1993)
This modification replaced the existing CAC/ CAD analyzers with improved instrumentation to improve operations and reliability. No safety concerns were introduced as a result of this activity. Based on the Safety Evaluation and the above information it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5276A Year Implemented: U/2(1992) U/3(1993) i This modification replaced Leeds and Northrup (L&N) multbrant Model W recorders with functionally equivalent Chesset Model 4200, and removes recorders that ere functionally obsolete because of the installation of the Plant Monitoring System. The new recorders have no control function and are not required to mitigate the consequences of an accident. No safety concerns were created. This activity is an enhancement to operations. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i MOD 5276B Year implemented: U/2(1993) U/3(not implemented yet) l This modification replaced Leeds and Northrup (L & N) multipoint Model W recorders with functionally equivalent Chessell Model 4200. The L & N recorders were obsolete and replacement units or parts were no longer available. This activity affected documentation addressed in the SAR and no adverse safety concerns 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 Page 10 of 23 CCN 94-14012
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT j
MOD 5286 Year Implemented: U/2(1993) U/3(not implemented yet)
This modification retired the inoperable equipment from the original plant process computer and removed the computer cabinets along the West wall in the computer room. The balance of equipment remains in place. This does not impact safety as a new Plant Monitoring System has already been installed. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unroviewed Safety Question.
MOD 5290 Year implemented: U/2(not implemented yet) U/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 concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
ff.QD53Q4 Year implemented: U/2(1993) U/3(1993)
This modification involved the renovation of the interior of the Plant Services Building and the painting of the buildings exterior siding. The Plant Service electrical loading had no impact on the PBAPS Voltage Regulation Study or plant safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5336 Year Implemented: U/2(1993) U/3(1993)
This modification removed the " Lead Pump Start" alarm circuit and added an hour meter for the Main Condenser Water Box Scavaging Pumps. This activity affected documentation addressed in the UFSAR.
This change did not adversely affect plant safety or operations. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5343 Year Implemented: U/2(1992) U/3(1993)
This modification installed relief valves on the Reactor Feed pump Turbine Lube Oil Coolers. The valves are required to protect the tube side against thermal overpressurization of the Service Water. The service water supply and return lines to the coolers contain isolation valves. In the event that these valves were closed and hot lube oil was introduced into the shell side of the coolers, the tubes could be damaged due to the expansion of the entrapped Service Water. This change is an enhancement to system and personnel protection and does not affect plant safety equipment. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 11 of 23 CCN 94-14012
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MOD 5347 Year implemented: U/2(not implemented yet) U/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 locations to support testing and maintenance activities. No safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5349 Year Implemented: U/2(1993) U/3(1993)
This modification replaced several Radwaste system pumps with upgraded units to increase operations and improve maintenance. This change affected documentation addressed in the UFSAR. These replacements wie not change any operating modes or adversely affected plant safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Our,ction.
MOD 535ji Year implemented: U/2(1993) U/3(1993)
This modification provided rigging points on service platforms to improve rigging capabilities. This change affected documentation addressed in the UFSAR. No new adverse safety concerns or operating modes were created as a result of the new rigging points. Based on the Safety Evaluation and the above inforrnation, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5371 Year Implemented: U/2(1992) U/3(1993)
This modification provided the Main Generator with water in leakage detection equipment. The new instrumentation installed monitors the dew point of the hydrogen gas in the generator. This activity affected documentation addressed in the UFSAR. The change is an enhancement during generator operations and -
did not adversely affect plant safety or operations. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5383 Year Implemented: U/2(not implemented yet) U/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 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 an Unreviewed Safety Question.
l Page 12 of 23 CCN 9414012
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT MOD 5393 Year implemented: U/2(1992) U/3(1993)
This modification added 7 thermocouples to various locations inside the drywell to provide data on the performance of the Reactor water-level condensing chambers. In addition, the change remosed obsolete reactor level indicators which were a potential source of leakage, and installed instrument valves which can provide a possible means of restoring water level to condensing chambers during power operation in the event of a drop in the reference leg water level. No safety concoms 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.
MOD 5394 Year implemented: U/2(1993) U/3(1993)
This modification improved access Emergency Diesel Generator Fuel Oil Transfer Pumps to support 1
monitoring vibration levels. Large and heavy gratings were previously needed to be removed to allow access to these pumps. This change allowed easier access to _the area. This activity affected documentation addressed in the SAR. No safety concerns 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 1
an Unreviewed Safety Question.
MOD 5401 Year Implemented: U/2(1992) U/3(1993)
This modification permanently removed the missile shield which surrounds the Reactor Core isolation Cooling (RCIC) turbine to simplify the RCIC turbine maintenance and reduce overall system outage time.
The shield consisted of thick steel plates lag bolted to the RCIC turbine pedestal. It surrounded the RCIC turbine with very tight clearances on all four sides and overhead. An evaluation of the RCIC room has determined that the physical arrangement of the plant equipment protects safety-related equipment operability from missile hazards. This activity affected documentation addressed in the SAR. No design limits or safety concems are affected by this change. Based on the Safety Evaluation and the above Information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD P000128 Year implemented: U/2(1993) U/3(1993)
This modification was installed to provide a continuous backfill system to the reference legs associated with Reactor Water Level Instrumentation. This activity will enhance level Indication reliability. This change affected documentation addressed in the SAR. No adverse safety concerns or new operating modes 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.
Page 13 of 23 CCN 94-14012 i
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 1993 to CFR 50.59 REPORT MOD RWM Software chanaes (RWM003)
Year Implemented: U/2(1993) U/3(1993)
This modification changed the Rod Worth Minimizer software. The new software does not change the function of the RWM as described in the SAR, however, the insta!!ation and testing will remove the RWM from service. These activities did not impact plant safety or adversely affect operations while in this condition. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P900206 Year Implemented: U/2(1993) U/3(1993)
This NCR was dispositioned to resolve discrepancies between plant configurations and the Piping and Instrumentation Drawings (M-314). This drawing involved the Service Water systems for Unit 2 and 3. This activity affected drawings addressed in the UFSAR. This change does not create any new adverse safety concerns or create any new operating modes. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910165 Year Implemented: U/2(1993) U/3(1993)
This NCR was dispositioned to resolve discrepancies between plant configuration and documentation that occurred from the implementation of Modification 1950. Documentation has been changed to reflect the as-is condition of the Instrument Nitrogen and Containment Atmosphere Dilution Systems. These changes are administrative in nature only and do not affect plant safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910165 Year implemented: U/2(1993) U/3(1993)
This non conformance report resolved several discrepancies between the OADs for the Instrument Nitrogen and Containment Atmospheric Dilution Systems. These changes are editorial in nature and do not result in a physical change to any system or component in the plant. This activity does not create any safety concerns or result in any new operating modes. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910591 Year Implemented: U/2(1993) U/3(1993)
This Non Conformance Report evaluated a concern associated with two electrical panels not being part of the Environmental Qualification Program. This condition was dispositioned to use as is. This activity affected the EO plan which is referenced in the UFSAR. This activity did not adversely affect plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i Page 14 of 23 CCN 94-14012 P
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT NCR P910592 Year Implemented: U/2(1993) U/3(1993)
This Non Conformance Report evaluated a concern associated with two electrical panels not being part of the Environmental Qualification Program. This condition was dispositionod to use as is. This activity affected the EO plan which is referenced in the UFSAR. This activity did not adversely affect plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P920134 Year implemented: U/2(1993) U/3(1993)
This change removed the reference to position modulators (POM-2(3)804A&B) for MO-2(3)804A&B from drawings in the UFSAR. These position modulators are considered to be part of the motor operator and therefore do not need to be uniquely identified. This change is administrative in nature and does not affect plant safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P920254 Year Implemented: U/2(1993) U/3(N/A)
This non conformance resolved discrepancies between design documentation and the as-built configuration of the Unit 2 plant. These discrepancies involved the drywell and torus radiation monitoring systems. This did not result in physical changes to any structures, systems, or components. No safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P920257 Year Implemented: U/2(1993) U/3(N/A)
This NCR was dispositioned to resolve a discrepancy between design documentation and the as built configuration of the Unit 2 plant. This discrepancy involves the safety grade instrument gas system. It did not result in physical changes to any structures, systems, or components. No safety concerns were created.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P9?0614 Year Implemented: U/2(1993) U/3(1993) 6 This NCR was dispositioned to abandon instrumentation and associated isolation valves that provide indication of seal water level on the Low Pressure Turbine expansion joints. The subject instrumentation was determined not to be required. P&lD M-319 and associated documentation will reflect this change. This change does not impact the operability, function or design basis of the Main Condenser system. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 15 of 23 CCN 94-14012
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT NCR P920777 Year implemented: U/2(1993) U/3(1993)
This NCR was dispositioned to abandon the 1 A Meteorological Tower and its associated equipment because it is not operational and some of its equipment has been removed. Other meteorological equipment is still in operation. This change affected documentation addressed in the UFSAR. No adverse safety concerns or new operating modes have be created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P920971 Year Implemented: U/2(N/A) U/3(1993)
This NCR was dispositioned to eliminate 3 temperature switches (TS4643,6644, and 6645) and removed UFSAR figure 11.6.2 sheet 1 due to these changes. 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.
NCR P930021 Year implemented: U/2(1993) U/3(1993)
This Non Conformance Report increased the temperatures and pressures used as a bases for the High Energy Line Break (HELB) Analysis. This change affected documentation addressed in the SAR since the HELB Report is part of the UFSAR. This activity did not adversely affect plant safety or create any new adverse safety conditions. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P930076 Year Implemented: U/2(1993) U/3(1993)
This NCR was dispositioned to resolve discrepancies between the UFSAR Table 8.5.2B and Table 8.5.2C-L These tables address electric load criteria of the Residual Heat Removal Pumps and the Emergency Diesel Generators This change was administrative in nature. 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.
NCR P930130 Year implemented: U/2(1993) U/3(1993)
This NCR was dispositioned to update two single line piping drawings and figures in the Updated Final Safety Analysis Report. The change made valve positions of the Torus Water Cleanup Block valves conform with as built configurations. This was an administrative change only. No safety concerns were created.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 16 of 23 CCN 94-14012
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT NCR P930562 Year implemented: U/2(1993) U/3(1993)
This evaluation justified that Emergency Core Cooling System room coolers were operable with a Room Cooler water flow rate less than as specified in the station's Routine Test. This change affected documentation addressed in the SAR. This condition did not adversely affect plant safety or create any adverse operating modes. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
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i NCR P930754 Year implemented: U/2(N/A) U/3(1993)
This evaluation addressed the cracks identified in the Unit 3 Core Spray piping in the Reactor Pressure Vessel between the vessel wall and the shroud and concluded that continued operation to the next Refueling Outage was acceptable. This activity affected documentation addressed in the UFSAR. This change does not significantly affect plant safety or the effects of a plant accident or transient. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i PROCEDURE AO-20A.1 Year implemented: U/2(1993) U/3(1993) l l
This procedure involves the temporary removal and the installation of flood barriers in the Reactor Building drainage system. These flood barriers are required to be in place as specified in the Updated Final Safety Analysis Report. Administrative controls will be put in place whenever required flood barriers are temporarily removed. No safety concems are created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
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PROCEDURE AO-52E.1 Year implemented: U/2(1993) U/3(1993)
This procedure specifies what critical temperatures are needed to be monitored if the Emergency Diesel Generator low temperature alarm is received. This activity affects documentation addressed in the SAR regarding engine temperatures and operability. No new adverse safety concerns were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE COL-14A,1 A-2(3)
Year implemented: U/2(1993) U/3(1993) l This operating procedure was revised to allow the Torus Water Clean Up system and level control hand l
valve to be normally open instead of normally closed. This activity affected figures addressed in the UFSAR.
This change did not adversely affect safety or create any new adverse operating mode. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unteviewed Safety Question.
Page 17 of 23 CCN 94-14012
M' PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT PROCEDURE Emeroency Plan Year implemented: U/2(1993) U/3(1993)
This evaluation consolida*ed the Peach Bottom, Umerick, and Chesterbrook Nuclear Emergency Plans into a common Nuclear document. This actMty affected documentation addressed in the SAR. No new adverse safety concerns were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE OM-03.2 Year implemented: U/2(1993) U/3(1993)
This change to the Operations Manual allowed the Operations section the flexibility of manning the Shift Technical Advisor (STA) position with any STA qualified individual. This would include a Senior Reactor Operator (SRO) who is STA qualified. This change affected documentation addressed in the SAR. This activity is administrative in nature and will not reduce the margin of plant safety because the position will still be manned with a qualified individual. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE RT-D-33-600-2 Year implemented: U/2(1993) U/3(N/A)
This procedure change extended the testing frequency associated with the Emergency Service Water to the Emergency Core Cooling System Room Coolers and the Diesel Generator Room Coolers from once per month to once per six weeks. This was based on satisfactory trends from the monthly tests. This activity affected documentation addressed in the SAR and previous commitments with the NRC. This change did not adversely affect room cooler reliability. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE SE-11 ATTACHMENTS A-H.J-N.P-Z Year implemented: U/2(1993) U/3(1993)
This procedure is used to provide the operator with procedural guidance to mitigate events which include a loss of offsite power and various levels of emergency AC power degradation. This revision provided event specific guidance for plant control. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE S12F-63M-761-XXC2(TC93-270)
Year Implemented: U/2(1993) U/3(N/A)
This temporary procedure change allowed the use of test equipment with an accuracy less restrictive than previously specified in the test. The accuracy previously specified was too restrictive and not necessary for calibration checks associated with the radiation monitors. This activity affected documentation addressed in the SAR. No new safety concerns were created as a result of this change. Based on the Safety i
Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
5 Page 18 of 23 CCN 94-14012
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT PROCEDURE SO-14A 1.A-2(3)
Year implemented: U/2(1993) U/3(1993) -
This operating procedure was revised to allow the Torus Water Clean Up evstem and level control hand valve to be normally open instead of normally closed. This actMty affected figures addressed in the UFSAR.
This change did not adversely affect safety or create any new adverse ope ating modes. Based on the Safety Evaluation and the above information, it was determined that these chnges did not constitute an Unroviewed Safety Question.
PROCEDURE SP-1413 Year implemented: U/2(N/A) U/3(1993)
This Special Procedure allowed the operation of Unit 3 during cycle 9 with a final feedwater temperature reduction of up to 55 degrees F (down to 32 degrees F at full power) for cycle extension and during coastdown operations. This activity affected documentation addressed in the UFSAR. No adverse safety concerns or operation modes were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE SP-1444 Year implemented: U/2(1993) U/3(N/A)
This Special Procedure allowed the backfilling of the "2A* Reactor Water Level Condensing Chamber which 1
onsured that the reference leg water level is correct. This procedure will ensure that reactor water level trips and indication are consistent with actual water levels. This activity affected documentation addressed in the UFSAR. This procedure did not adversely affect plant operations or safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i PROCEDURE SP 1475 Year Implemented: U/2(1993) U/3(N/A)
This Special Procedure allowed the backfilling of the "2B" Reactor Water Level Condensing Chamber which ensured that the reference leg water level is correct. This procedure will ensure that reactor water level trips and indication are consistent with actual water levels. This activity affected documentation addressed in the UFSAR. This procedure did not adversely affect plant operations or safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
EROCEDURE SP-1476 Year implemented: U/2(1993) U/3(N/A)
This procedure installed temporary battery chargers for the Unit 2 24 volt DC loads while the old chargers were being replaced. This activity affected documentation addressed in the SAR. No new operating modes or adverse conditions were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 19 of 23 CCN 94-14012 f
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT PROCEDURE SP-2002 Year Implemented: U/2(1993) U/3(N/A)
This procedure allowed the performance of Flux Tilt Testing. The temporary sample chamber used during testing required the rerouting of some of the Off Gas sample station piping. This was performed by disconnecting the flexible hosing to the Off Gas Grab Sample Pump and connecting another flexible hose which was attached to the temporary sample chamber. This activity affected documentation addressed in the SAR. There were not adverse safety concerns or new operation modes created as a result of the testing.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE SP-2004 Year Implemented: U/2(1993) U/3(1993)
This procedure allowed the installation of a temporary battery charger on the Unit 3 electrical 125/250 VDC busses to support battery replacement during modification 5169. This activity affected documentation addressed in the UFSAR. This temporary change ensured that DC power supply was provided to the necessary equipment during the duration of the battery replacement modification. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
c PROCEDURE SP-2006 Year Implemented: U/2(N/A) U/3(1993)
This special procedure allowed the installation of a temporary electrical power cable from a DC electrical distribution panel to support outage activities associated with the 3A Recirculating Motor-Generator oil pumps. The act!vity affected documentation addressed in the UFSAR section 8.7. There were no new adverse safety concerns generated as a result of the activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE SP-2011 Year implemented: U/2(N/A) U/3(1993)
This activity installed a temporary Battery charger for 3BD028,3CD028, and 3DD028 during a design equivalent change out of the old battery chargers. This change affected documentation addressed in the UFSAR. No new adverse safety concerns or new operating modes were created as a result of this actMty.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 2-01G 032 Year implemented: U/2(1993) U/3(N/A)
This TPA was installed to defeat bellows leakage alarm input and indication for Safety Relief Valve (SRV-718).
The valve was considered inoperable at the time. The change allowed the other 10 Safety Relief Valves to be continuously monitored with the inoperable SRV-71B. No safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
Page 20 of 23 CCN 9414012 t
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT i
I TPA 2-23-010 Year implemented: U/2(1993) U/3(N/A)
This TPA installed a lumper in the High Pressure Coolant injection system Auxiliary Oil Pump control logic relay contact which will ensure that the bypass / starting resistors are lined up prior to the start of the pump.
I This activity did not introduce any new modes or attect the safety of the plant. Based on the Safety Evaluation and the above information, it was detemined that these changes did not constitKe an Unreviewed Safety Question.
TPA 2-30-012 Year implemerund: U/2(1993) U/3(N/A) i This TPA provided supplemental cooling water to the 2A Alterex Air Cooler by the installation of a temporary water line from the Service Water Return Header. This activity affected drawings addressed in the UFSAR.
i No now operating modes or adverse safety concerns were created as a result of this activity. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
l TPA 2-50-022 Year implemented: U/2(1993) U/3(N/A)
This TPA installed a dew point monitor in the main generator H2 monitoring line. The probe has an electrical connection to allow a monitor to be periodically booked up to obtain dew point readings. This does not adversely affect the electrical circuitry of the plant. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 2-52414 Year implemented: U/2(1993) U/3(1993)
This TPA installed a jumper to defeat the Emergency Diesel Generator Fuel Oil Transfer Pump trip on a high -
temperature condition. This activity was implemented since the temperature switches were not installed in accordance with station documentation. This trip function is discussed in the UFSAR and no adverse safety concerns or new modes of operation were created as a result of this logic change. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 3-02-020 Year implemented: U/2(N/A) U/3(1993)
This TPA was installed to remove the scoop tube brake annunciator and to remove the ability to monitor the status of the brake on the "3A" Recirculation Motor-Generator Set. This brake annunciator is addressed in the UFSAR. This change did not adversely plant safety or create any adverse operating modes. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
j Page 21 of 23 CCN 94-14012 j
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT TPA 3-02-022 Year implemented: U/2(N/A) U/3(1993)
This TPA removed the "3A* Recirculation M/G Set scoop tube positioner brake from service until parts become available for repairs. This activity affected a figure described in the UFSAR. This change does not adversely impact plant safety or create any new operating modes or transients. Based on the Safety l
Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 3-13411 Year Implemented: U/2(N/A) U/3(1993)
This TPA disabled the high temperature alarm function of the "B" Reactor Water Clean Up Demin Compartment due to equipment problems. Monitoring of area temperature is described in UFSAR section 4.10.3.4.b.
This change did not create any new operating modes or adversely affect plant safety or operations. Based on the Safety Evaluation and the above information, it was determined that these-changes did not constitute an Unreviewed Safety Question.
TPA 3-18-005 Year implemented: U/2(N/A) U/3(1993)
This TPA defeated the refuel platform Rod Block interiocks. This allowed the performance of refuel platform modifications (MOD 5235). These interlocks are only required while the plant is in the REFUEL mode with the reactor head removed. This modification was performed while in the RUN mode and the reactor head was installed. No safety concerns were created. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 3-23 009 Year implemented: U/2(N/A) U/3(1993)
This change installed a jumper in the High Pressure Coolant injection system Auxl!!ary Oil Pump control logic relay contact which will ensure that the bypass / starting resistors are lined up prior to the start of the pump. This activity did not introduce any new modes of operation or affect plant safety. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unre/lewed Safety Ouestion.
TPA 3-57-007 Year implemented: U/2(N/A) U/3(1993)
This TPA installed a temporary battery charger in parallel with 3AD027 to support the loads of 3AD028. This
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was accomplished during a modification to relocate 3AD028. This activity affected a figure addressed in the l
UFSAR. This TPA did not adversely affect plant safety or create any new adverse conditions. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an -
Unreviewed Safety Question.
Page 22 of 23 CCN 94-14012
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199310 CFR 50.59 REPORT JPA 3-57-007 Year implemented: U/2(N/A) U/3(1993)
This TPA installed a temporary a battery charger on the Unit 3 electrical 24 VDC busses to support battery replacement during modification 5169. This activity affected documentation addressed in the UFSAR. This temporary change will ensure that DC power supply was provided to the necessary equipment during the duration of the battery replacement modification. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 3-62-044 Year Implemented: U/2(N/A) U/3(1993)
This TPA installed a temporary electrical jumper across a defective Control Rod Drive magnetic reed switch to support movement of the refueling bridge over the reactor core with a defective reed switch. The associated control rod was verified to be properly inserted. This activity affected UFSAR section 7.6.3 which states that this control logic exists. This condition did not create any safety concerns or new operating modes.. Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
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Page 23 of 23 CCN 9414012
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