ML20046A241
ML20046A241 | |
Person / Time | |
---|---|
Site: | Peach Bottom ![]() |
Issue date: | 12/31/1992 |
From: | Miller D PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
References | |
CCN#93-14100, NUDOCS 9307270134 | |
Download: ML20046A241 (58) | |
Text
CCN#93-14100-
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PIIILADELPHIA ELECTRIC COMPANY t[3 PEACil 110TIDM ATOMIC POWER STATION j
R. D.1, liox 208
%fedhlD Delta, Pennsylvania 17314
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erxn norsuu-Tur row in or rx< ti.s.:N(t (717)4547014 D. B. Miller, Jr.
Vice President July 20, 1993 Docket Nos. 50-277 50-278 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
SUBJECT:
Peach Bottom Atomic Power Station (PBAPS)
Annual 10 CFR 50.59 Report For The Period 1/1/92 through 12/31/92
Dear Sir:
Enclosed is the 1992 Annual 10 CFR 50.59 Report as required by 10 CFR 50.59.
Should you have any questions, or require further information, please contact us.
Sincerely, NJ Gd DBM/AJ)W/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 B.S. Norris, USNRC Senior Resident inspector H.C. Schwemm, Atlantic Electric C.D. Schaefer, Delmarva Power
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Docket Nos. 50-277 50-278 1992 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT This report is issued pursuant to reporting requirements for Peach Bottom Atomic Power Station Units 2 and 3 (Facility Ucense Numbers DPR-44 and DPR-56 respectively). This report addresses, but not limited to, tests and changes to the facility and procedures as they are described in the Peach l
Bottom Final Safety Analysis Report. This report consists of those tests and changes that were completed in 1992. A summary of the safety evaluation for each item has been concluded that no unreviewed safety question, as defined in 10 CFR 50.59 (a) (2), were involved.
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l PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 AND 3 DOCKET NOS. 50-277; 50-278 l
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l 1992 ANNUAL 10 CFR 50.59 REPORT TABLE OF CONTENTS 6
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Table of Contents Miscellaneous Safety Evaluations
................... Egge 1 0-V R R -3........................................... 1 Aux."A" Boiler contamination.
1 Aux."B" Boiler Contamination.............
1 CRD Blades 34% Depletion...........................
1 Core Design Report 2
Core Design Report..................................
2 Core Operating Limits Report 2
j Corrosion / Erosion 3
F/W Tracer Results implement...........................
3 I
HPCI Response Time Change......................
3 i
00001409 3
RHR Heat Exchanger Operability..
4 j
Reactor vessel 68 degree limit.
4 Temp. Breach Hazardous Barriers 4
l Unit 1 Transformer (750KVA).
4 Valve Closure Time 5
1 Modifications 0257 5
0625 5
0693 6
0800A 6-0955C..............
6 0955G..........
7 0955P 7
8 j
0964 1078 8
1111 8
1122 9
1309 9
1398 9
1488 10 1536A 10 1576 10 1665A 11
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i Table of Contents Modifications (continued)......
Paae i
1750B 11 1752 11 1795 12 1825 12 1829 12 1830 13 1909 13 1928 13 2T1 14 2071 14 2075 14 2115..................
15 2122 (A) 15 2185
.......................... 15 2225 16 2285 16 5032 16 5095 17 5123 17 5129 17 5164 18 5218 18 5221 18 5223 19 5228 19 5231 19 5233 20 5235 20
- 523o,
...................20 i
5238 21 I
5240 21 5244 21 5249 22 5255 22 525b
.............................. 22 5276A
. 23 5277 23 5339 23 5340.........
24 5343.
.........e i
24 5344 l
Table of Centents Modifications (Continued).......................
Paae 5353 25 5388 25 5393 25 5401 26 86-069......
26 Non Conformance Reoorts i
P880055 26 I
P880116 27 P890260
... -27 P890262 27 P900042 27 i
P900076 28 P900163 28 P900202 28 1
P900205 29 P900214 29 i
P900216 29 P900221 29 P900225 30 P900244 30 j
P900357
'30 P900446 30 P900490 31 P900579 31 P900622 31 P900789 32 P910023 32 P910158 32 P910175 32 P910218................
. 33 P910222 33 P910294 33 P910323.....................................
33 P910327 34 P910359 34 P910375 34 P910377 34 P910389 35
I Table of Contents Non Conformance Reoorts (Continued)
Pace P910491 35 P910540 35 P910543 35 P910572 36 P910573 36 P910665 36 P910694 36 P910695 37 P910713 37 P910714 37.
P910715......................................
37 P910716 38 P910721 38 P910722 38 P910781 38 i
P910830 39 P910869 39 P910971 39 l
P920009 40 P920209 40 P920222 40 P920297 40 P920489 41 i
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Procedures AO-10.9-2(3) 41 AO-13.2-2(3) 41 AO-14.1-2(3) 41 AO-23.3-2(3) 42 Emergency Plan Table 4.1 42 M-003-216 42 SP-1422 42 SP-1442...
43 SP-1443 43 ST-D-033-420-2(3) 43 ST-O-003-560-2 43 ST-X-07G-102-2(3) 44
1 Table of Contents Temocrarv Plant Alterations Paae 2-06-029.........................................
44 2-18-001 44 2-23-008 45 2-30-009 45 2-52-011 45 2-62-005 46 3-01 G-033 46 3-05-019 46 3-08-014 47 3-08-015 47 3-13-007 47 3-13-009 47 3-30-011 48-j 3-57-006 48 t
UFSAR Chanaes
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CR-9108447....
48 C R-9108590.......................
49 CR-9108594.
49 Section 12.2.10 49 l
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PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION UNIT 2 AND 3 DOCKET NOS. 50-277; 50-278 l
1992 q
ANNUAL 10 CFR 50.59 REPORT SAFETY EVALUATION SUMMARIES i
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MISC. 50.5910-VRR-3 Year implemented: U/2(1992) U/3(N/A)
This evaluation determined that RHR check valve CHK-2-10-29 be individually leak tested in series with motor operated gate valve MO-2-10-32, and the pair be considered a single pressure isolation boundary. These valves are in the Residual Heat Removal head spray line which is no longer used. Gate valves MO-2-10-032(033) are administratively blocked closed in their safety position. This change will not introduce any new operating modes or affect the safety of the plant. This change did not involve any plant hardware changes. Testing was enhanced.
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 Aux."A" Boiler contamination Year implemented: U/2(1992) U/3(1992)
This determination evaluated the radioactive contamination of the 'A' Auxiliary Boiler and the possible release to the environment. Releases were found to be within established limits. No new safety or environmental concerns were introduced.
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 Aux."B" Boiler Contamination Year implemented: U/2(1992) U/3(1992)
This determination evaluated the radioactive contamination of the 'B' Auxiliary Boiler and possible release to the environment. Releases were found to be well within established limits. No new environmental or safety concerns were introduced.
Based c.
- Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MISC. 50.59 CRD Blade.s_34% Deoletion Year implemented: U/2(1992) U/3(N/A)
The evaluation addressed a concern with several control rod blades which were replaced during the Unit 2 Ninth Refuel Outage because they had exceeded 35% B-10 depletion.
No actual safety concerns were created as the limits of safety were not exceeded. No affect on plant safety was realized.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MISC. 50.59 Core Design Reoort Year Implemented: U/2(1992) U/3(N/A)
This evaluation addressed the Unit 2 Core Design Report for CYCLE 10 operations. The core load was of 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 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.
MISC. 50.59 Core Design Reoort Year implemented: U/2(1992) U/3(N/A)
This evaluation addressed the Unit 2 Core Design Report for CYCLE 10 operations. The revision was needed due to the replacement of a leaker bundle (LY6215) with a slightly less reactive reload 6 bundle (LY6215). The original shutdown margin analysis is still valid since the new bundle is less reactive than the original bundle. No new 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 Core Ooerating Limits Reoort Year Implemented: U/2(1992) U/3(N/A)
This evaluation addressed the Unit 2 CORE OPERATING LIMIT Report for CYCLE 10 operations. It provided APLHGR, MCPR, Kf, LHGR and RBM flow bias setpoints. These j
values have been determined using NRC-approved methodology and are established such 1
that all applicable limits of the plant safety analysis are met. No safety concerns
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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.
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT 1
MISC. 50.59 Corrosion / Erosion Year Implemented: U/2(1992) U/3(1992)
This evaluation allowed for an upgrade replacement of carbon steel piping and fittings on the Off-gas Recombiner System which have experienced excessive wall thinning due to i
erosion / corrosion damage. The upgraded material allowed per this safety evaluation is 1-1/2% Chrome - 1/2% Moly alloy steel. This change will enhance piping life and reliability.
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 F/W Tracer Results Imolement.
Year implemented: U/2(1992) U/3(1992) l This evaluation supported recalibration of the feedwater flow points for the process computer and the digital feedwater control system. This change will enhance accuracy of feedwater flow inputs and reactor power indication. The activity is an enhancement to operations. No new 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.
MISC. 50.59 HPCI Resoonse Time Change Year Implemented: U/2(1992) U/3(1992)
The High Pressure Coolant Injection (HPCI) response time was changed from the current 25 seconds specified in the UFSAR to 30 seconds. The time was increased in order to provide a more realistic and less restrictive licensing basis for HPCI performance.
No safety concerns were created as a result of this change. System performance has been maintained.
Based on the Safety Evaluation and the above information, it was determined that these change-N not constitute an Unreviewed Safety Question.
MISC. 50159 00001409 Year implemented: U/2(1992) U/3(1992)
This evaluation addressed the station's responsibility to PORC approved safety evaluations associated with Technical Specification violations. This is administrative in nature and did not affect the safety 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.
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FEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MISC. 50.59 RHR Heat Exchanoer Ooerability Year implemented: U/2(1992) U/3(1992)
This evaluation supported the operability of the Residual Heat Removal Heat Exchangers. Pitting has occurred in the tubes, however, results from heat transfer tests support the continued use of these exchangers, however, the equipment is capable of fully performing its designed safety function. No safety concerns are created by this evaluation.
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 vesse! 68 deoree limit Year implemented: U/2(1992) U/3(N/A)
This evaluation discussed the impact of reactor coolant temperatures dropping below 68 degrees specified for Peach Bottom Unit 2 Reload 9 Cycle 10. This review concludes that this change does not affect the 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 Temo. Breach Hazardous Barriers Year implemented: U/2(1992) U/3(1992)
This evaluation incorporated lists of hazard doors / hatches in different reactor modes of operation. This evaluated the impact of blocking open, removing, or blocking closed doors and hatches. Recommended compensatory measures were made when doors / hatches are not in designated position during a postulated event. This change does not introduce any new modes or affect the safety 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.
MISC. 50.59 Unit 1 Transformer (750KVA)
Year Implemented: U/2(1992) U/3(1992)
This installation of a new 750kVA transformer was completed to support the Technical Support System. The existing transformer and switchgear were removed. This improvement minimizes effects associated with aged equipment. This installation poses no new safety 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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MISC. 50.59 Valve Closure Time Year Implemented: U/2(1992) U/3(1992)
This evaluation addresses and provides justification for increased valve closure times for the Torus to Reactor Building Vacuum Breaker Isolation Butterfly Valves.
The increased valve closure time has been analyzed and is adequate to ensure any leakage is within specified limits. There is no impact on plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MQD_Q252B Year Implemented: U/2(1s92) U/3(N/A)
This modification was completed to bring Peach Bottom into compliance with an Environmental Protection Agency request to modify the Spill Prevention Control Counter measures Plan so that additional oil spill protection be provided for in the form of oil spill collection systems. Contoured dikes were installed at the No. 2 and No. 6 fuel oil unloading areas to contain spills or hose leaks. Discharges from the diked areas will be routed through oil / water separators prior to release into the normalwaste system. The modification did not affect the function of equipment required to mitigate plant events.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
EQD 0625 Year implemented: U/2(1992) U/3(1992)
This modification was installed to improve the reliability and testability of the instrument Nitrogen System. The work included: Installation of test and vent connections at the ADS air supply check valves; replacement of MSIV air supply check valves with test and vent connections inside the drywell; replacement of instrument Nitrogen isolation check valves at penetrations N-22,52F, and 218A (with relocation of A03968 torus instrument nitrogen valve on Unit 3); replacement of MSIV air supply check valves with test and vent connections outside the drywell installation of seismic nitrogen supply piping inside the drywell; installation of seismic rJtrogen supply piping outside the drywell; replacement of containment isolation talves on the backup, safety grade nitrogen supply. This enhancement did not cause any safety concerns.
Reliability and testability of the system was enhanced.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 0693 Year implemented: U/2(1992) U/3(N/A)
This revision of the original safety evaluation for the construction involved changes in operation to the facility and to enhance clarity. The modification constructed a low-level, on site radwaste storage facility to contain 2.5 years of radioactive resins, radioactive trash, radioactive equipment, and some clean material. 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.
NOD 0800A Year Implemented: U/2(1992) U/3(1992)
This modification installed zone restriction circuitry on both Turbine Building Cranes so they cannot enter the hatchway area without using a key override. This will reduce the stopping distance of the Turbine Crane so that the restricted zone may be returned to its original dimensions. This modification was done in compliance with NUREG 0612. This modification ensures maximum operating flexibility while ensuring that loads are not inadvertently handled above the hatchway. Critical service water piping runs under i
the floor below the hatchway. 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 Ouestion.
MOD 0955C Year implemented: U/2(1992) U/3(1992)
This modification installed the P!cnt Monitoring System (PMS) computer equipment in the new computer room on the fourth floor of the Administration building. Work covered included: penetrations, ductbanks, raceways, conduit and cables that are needed to connect equipment to peripherals, distribution panels, multiplexers, and consoles located at various locations throughout the plant. This modification also included alarm cables and connections to allow for control room monitoring of air conditioning and uninterrupted power supply power status (via PMS), and battery room ventilation (via Administration building HVAC alarm panel) associated with the operation of the PMS equipment in the Administration building. The PMS equipment does not create a safety concern with installed plant systems. The modification overall provides better monitoring of plant equipment and therefore is a safety enhancement.
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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r PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT 1
MOD 0955G Year implemented: U/2(1992) U/3(1992)
This modification installed a new Rod Worth Minirnizer, new computer controlled trend recorders and digital displays and new Plant Monitoring System (PMS) digital inputs.
All Plant Process Computer (PPC) digital and analog outputs were disconnected from the PPC and reterminated on the new PMS. This modification also provided new consoles for the operators in the control room and upgraded communication systems in the control room. This change is a safety enhancement due to better monitoring afforded plant operators. The plant changes did not adversely affect any installed plant equipment.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 0955P Year Implemented: U/2(1992) U/3(1992)
This modification consisted of miscellaneous work necessary to complete installation of the Plant Monitoring System (PMS). Work included: control room monitoring of reactor water level during the flooding-up and refuel activities; sixteen drywell temperature elements to ellow the new Safety Parameter Display System (SPDS) to calculate drywell bulk average temperature; containment Atmosphere Control (oxygen concentration "high" range) to support SPDS calculations / indications; Drywell pressure support to SPDS; 24 VDC power supply fuses so that fuse status may be monitored and validation of any digital input be made if the associated fuse is blown. Other work included: providing each class 1E multiplexer cabinet with a " mini" uninterruptible power supply and back up ventilation to ensure continuous power during plant power transients, replacing all Change-of-State digital input cards and associated surge cards with new cards that accurately time-tag contact closures from "high contact bounce" relays; disconnecting all wiring between the old Process Computer and PMS; ensuring all 125 VDC distribution panel bus voltages are made available to PMS; rewiring to establish Change-of-State points and replacing 6 digital input cards to create 6 digital Change-of-State cards for Post-Trip-Review. This change is a safety enhancement due to better monitoring afforded plant operators. The changes did not adversely affect any installed plant equipment.
Based on the Safety Evaluation and the above inforrnation, it was determined that these changes did not constitute an Unreviewed Safety Question.
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT -
MOD 0964 Year implemented: U/2(N/A) U/3(1992)
This modification involved the removal of the Continuous Water Level Monitoring System for the scram discharge headers and the automatic system to initiate control rod insertion on low pressure in the control air header. These systems were necessary as an mterim until design deficiencies were corrected. Piping between the SCRAM discharge volumes and the instrument volume has been replaced with larger piping. Completion of the long-term fix eliminated the need for the system removed. Because the modification resulted in enhanced original system performance, there was no safety concerns created by this j
change. Inferior equipment was removed as it was no longer necessary.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1078 Year Implemented: U/2(1992) U/3(N/A)
This modification added flush and drain connections to the Residual Heat Removal (RHR) piping and several flow elements. The purpose of these connections is to facilitate flushing the accumulated radioactive crud from the space between the flow element and the RHR pipe, for ALARA considerations. These connections consist of a 1 inch diameter schedule 80 piping with sockets welded to fittings. A welded cap was installed after the initial flushing to assure integrity of the pressure boundary and to facilitate flushing at this location in the future. This modification maintains the capability of the system The addition of the taps does not cause any significant potential of a leak while allowing allowing the system to be flushed and will increase overall system performance and therefore the safety of the system is enhanced.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1111 Year implemented: U/2(1992) U/3(1992)
This modification installed bypass switches in the fire detector circuitry of the automatic code transmitter. This will eliminate the practice of installing and removing temporary jumpers which are frequently required for maintenance and construction to perform work activities which utilize ignition sources. The installation of a second bypass switch in the fire alarm circuitny will eliminate the nuisance alarms caused by periodic testing.
This modification enhanced safety by decreasing impact on operations personnel and improving testing flexibility.
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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PEACH BOTTOM ATOMIC POWER STATION -
UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 1122 Year implemented: U/2(1992) U/3(N/A)
This modification replaced 2A and 2C low pressure turbine rotors, two jib cranes, and a vertical fall guidance system. The turbine rotors were replaced because the old rotors developed disk cracking. The new rotors are of an improved quality material and the wheels were designed with radial keys between disks. Previous wheels utilized axial keys at the higher stress level. This modification is an overall improvement to turbine operation. Because the change is an overall enhancement to turbine operations, plant safety is enhanced. No new safety concerns were created.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewea Safety Question.
i RIOD 1309 Year Implemented: U/2(1992) U/3(1992)
This modification installed fire dampers in 53 ducts penetrating fire barriers in the Reactor, Turbine, and Radwaste buildings. These dampers were installed to comply with Appendix R to 10CFR50, This modification was an enhancement to safety and safe shutdown capabilities in the event of a fire.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1398 Year Implemented: U/2(N/A) U/3(1992)
This modification installed a 220 to 13kV transformer at the North Substation. This will provide a second source of startup and emergency power to 3SU regulating transformer switchgear increasing the reliability of the #3 start up source. The transformer was designed to supply voltage comparable to the existing offsite sources and is a compatible replacement for transformer 2SU if it should fail and need replacing. This change provides increased system reliability and enhances safety.
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 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3
-l DOCKET No. 50-277 & 50-278
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199210 CFR 50.59 REPORT MOD 1488 Year implemented: U/2(1992) U/3(1992)
This modification replaced the 2B,3A,3B, and 3C, low pressure turbine rotors. The I
design of the new LP rotor shafts have no shrunk-on discs (wheels), grooves, or keyways, thus eliminating the steam condensing mechanism which contributed to the initiation of stress corrosion cracking of the low pressure rotor discs. This modification is an overall enhancement to turbine operation. Because this change is an overall improvement to turbine operations, plant safety is enhanced. No new 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 1536A Year implemented: U/2(1992) U/3(N/A)
This modification provided power for the Nuclear Engineering trailer. Power was obtained from an existing transformer. All components affected by this modification have been designed to handle outage power demands greater than those created by the additional trailer load. No safety concerns with the electrical power 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 1576 Year implemented: U/2(1992) U/3(N/A)
This modification provided field verification, corrected print discrepancies, and created as-built electrical drawings of the Station Fire Alarm System. These document changes better reflect the as-is plant configuration. No hardware changes were implemented. The UFSAR was revised to reflect these changes.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 l
199210 CFR 50.59 REPORT i
MOD 1665A Year implemented: U/2(1992) U/3(N/A) l This review addressed the impact of "short stroking" motor operated valves. Short i
stroking complies with the PECo staff position requiring valve closure times at least one full second less than UFSAR or Technical Specifications requirements. This change does not affect the valve's ability to perform its safety function. By ensuring stroke time meets requirements, plant safety is maintained.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 17508 Year Implemented: U/2(1992) U/3(N/A)
This modification reduced a 12 inch underground Fire Protection line to 10 inches using sleeves. The reduction will not affect the System Operation Performance of the Fire Protection System. The line reduction was necessary because the Fire Protection line ran adjacent to the electrical cables. This enhancement provided additional j
insurance against possible leakage of water around the electrical cables. The sleeves I
were provided to protect the cables from such leakage. Station documentation were revised to reflect these changes. No impact on plant safety was caused 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 1752 Year implemented: U/2(N/A) U/3(1992)
This modification rerouted cabling for a drywell radiation monitor (RE-91038). This allowed the 'B' Source Range Monitor to be reterminated on an alternate cable within drywell penetration N-100C. This modification was necessary because of the failure of the Unit 3 'B' Source Range Monitor cable within the drywell. No affect on safety was created by this change. The change maintained the operational capability of the affected equipment.
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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PEACH BOTTOM ATOMIC POWER STATION.
UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 1795 Year implemented: U/2(1992) U/3(1992)
This modification replaced the instrument and Service Air System Compressors and upgraded the system with new dryer, purification equipment, instrumentation, and provided other system improvements. Improvements resulted in a system that has better reliability, requires lower maintenance and enhanced operations. This did not create a safety concern. Although the instrument Air System is not required for plant safety, increased reliability and overall enhancement to plant safety was created.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1825 Year Implemented: U/2(1992) U/3(1992)
This modification installed new pressure switches and connected these switches to existing fuel deluge detection circuitry to detect viater flow and provide an alarm in the control room for the Condensate Pump Transformer Deluge System, the Auxiliary Transformer Deluge System, the Standby Gas Treatment Charcoal Filter Fixed Water Spray System, the Fixed Water Spray System for the charcoal filters in the Recombiner building and the Hydrogen Seal Oil Deluge System. This change affected documentation addressed in the SAR. This modification enhances detection of water flow within the systems and aids operators in detecting actuation of the system. This does not introduce any new operating modes or affect the safety 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.
MOD 1829 Year Implemented: U/2(1992) U/3(N/A) i This modification provided automatic sprinkler protection for the Reactor Feed Pump areas. These sprinklers are designed to initiate when ambient iemperatures rise to the melting point of fusible material on the sealed sprinkjer heads The flow of water energizes a pressure switch which transmits an alarm condition to the fire protection panci in the control room. This modification was completed in compliance with NFPA 13, " Standard for the Installation of Sprinklers". The change affected documentation addressed 'n 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.
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 i
199210 CFR 50.59 REPORT MOD 183Q Year Implemented: U/2(1992) U/3(N/A)
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 melting point of the fusible material on the sprinkler heads.
The flow of 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 concerns were created as a result of this change. No adverse affects on safety related equipment was created by inis change.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1909 Year implemented: U/2(1992) U/3(1992)
This modification reconfigured existing valve stuffing box arrangements. This will
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provide a more efficient and reliable stem seal with less frictional drag. This reconfiguration uses graphite packing, carbon sleeves, live load glands and requires r
less packing depth. Independent laboratory testing and numerous nuclear valve installations have repeatedly demonstrated that this system is more effective than the previous methods. No safety concerns were created. This change resulted in an enhancement to safety due to less potential for valve leakage.
Based on the Safe y Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 1928 Year implemented: U/2(1992) U/3(1992)
This modification involved the replacement of the existing Main Condenser, tubesheets, and support plates. The new material will be titanium tubes and tubesheets with new carbon steel support plates fabricated into a module. This modification was designed to mitigate Main Condenser tube failures and to improve plant chemistry. This change affected documentation addressed in the SAR and provides an overall operations enhancement. 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 an Unreviewed Safety Question.
13
PEACH BOTTOM ATOMIC POWER STATION j
UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 2001 Year implemented: U/2(1992) U/3(N/A)
This modification installed trailers on the south end of the Administration building to house various station groups. There were no adverse effects to the station electrical system and no degradation to the sewer or water systems.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 2071 Year implemented: U/2(1992) U/3(N/A)
This modification replaced all the Emergency Diesel Generator Starting Air System receiver tank inlet check valves and one of the outlet check valves. In addition, this activity also added 3/4" test connections with shut-off valves on the compressor dischargo lines, and added a new 1/2" block valve on the instrument line that cross-ties the 1 1/2" headers of the starting air control station. These replacements improved system operation and reliability. The change affected documentation addressed in the SAR. 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 _2Q75 Year implemented: U/2(1991) U/3(1992)
This modification added block valves and test connections to ensure that local leak rate tests (LLRT) of containment isolation valves are performed in full compliance with 10CFR50 Appendix J, and also reverses manual block valves located outside primary containment between the primary containment and the inboard containment isolation valve to include the valve stem packing in the LLRT boundary. This activity affected documentation addressed in the SAR. No new operating modes were introduced the safety of the plant was not affected. The ability to properly test these valves is an enhancement to plant safety.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 2115 Year implemented: U/2(1992) U/3(1992)
This modification installed permanent shielding in the Traversing incore Probe (TIP) rooms. Lead bricks were installed over grating in these rooms to prevent high radiation exposure to personnel. This activity affected documentation addressed in the SAR. No new operating modes were introduced and there was no impact on plant safety. The installed lead can not adversely affect plant operations and therefore, overall plant safety is maintained.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 2122 (A)
Year implemented: U/2(1992) U/3(1992)
This modification identified suitable replacement solenoid valves for the Containment Atmosphere Dilution analyzers and other safety related valves and dampers. The original valves can no longer be maintained as O equipment because replacement parts are unavailable. The replacement parts are functionally equivalent and do not create any safety concerns therefore, overall plant safety is maintained.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 2185 Year implemented: U/2(1992) U/3(N/A)
This modification installed temporary Health Physics trailers at the north entrance to the Turbine building. These trailers will be utilized for radiological access control. Power was obtained from an existing power supply. No adverse conditions were introduced to the station electrical system.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 2225 Year implemented: U/2(1992) U/3(N/A)
This modification installed a training simulator and provided the necessary power and fire protection for new simulator area. There were no adverse effects to the station.
electrical system. This equipment is located outside the plant protected area. There are no affects on plant opeernon.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 2285 Year implemented: U/2(1992) U/3(N/A)
This modification provided minimum flow protection to RHR pump 2DP35 in the event of an Appendix R fire in Fire Area 6 South. An alternate power feed to RHR minimum flow bypass valve MO-2-10-16D was installed and the electrical interlock between valve MO-2-10-16D and MO 2-10-17 was removed. This restores the original design state of valve MO-2-10-16D from normally open to normally closed._ This modification does not make any changes to the design function or modes of operation of the RHR System.
Because this change enhances the plant capability in the event of an Appendix R fire, overall plant safety is enhanced.
Based on the Safety Evaluation and the above information, it was determined that these -
changes did not constitute an Unreviewed Safety Ouestion.
MOD 5032 Year implemented: U/2(1992) U/3(1992)
This modification enhanced the 4KV and 13KV vertical lift switchgear breakers position switches. The circuit breaker position switches were located in the rear of the breaker compartments and were mechanically operated by a device located on the breaker. This device could have become bent due to interference with the circuit breaker. The interference could cause a distortion in the breaker position switch operators and further operation of the breakers would have resulted in completely damaged operators and inoperative breaker position switches. The new switches are more rigid and are protected by a guide to prevent interference with the circuit breakers. This modification is an overall enhancement to operations and maintenanco.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT I
J MOD 5095 Year implemented: U/2(1992) U/3(N/A)
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 fail. 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.
MOD 5123 Year implemented: U/2(N/A) U/3(1992)
This modification provided access to the Reactor Water Clean-up Isolation 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 activity affected drawings addressed in 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 5129 Year Implemented: U/2(1992) U/3(N/A)
This modification installed an air condition unit on the Plant Security System remote multiplexers, and also relocated the power feeds. These air conditioning units are required to provide additional cooling of the electronic components associated with this equipment. These changes enable the security system to maintain the required on-line reliability and availability requirements under extreme high ambient temperature conditions. No i,afety or security concerns were created by this activity.
The overall reliability of plant security was increased.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 5164 Year Implemented: U/2(1992) U/3(N/A)
This modification provided a break trailer for operations personnel outside the Turbine building roll-up door. This provides operations with a break area outside the radiation area, but near the power block. There were no adverse conditions introduced into the station electrical system. No equipment was affected, therefore, there is no impact on plant operations.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5218 Year Implemented: U/2(1992) U/3(1992)
This modification replaced Bussmann Model JHC-100 fuses with Bussmann model LPJ-100SP fuses. These are equivalent, environmentally qualified fuses and will provide the function required by Appendix R of 10CFR50. This modification eliminates the manual restorative interirn procedures, and improves the overcurrent device coordination of AC electrical distribution system. This activity affected documentation addressed in the SAR and is an enhancen.ent to the plant.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5221 Year implemented: U/2(1992) U/3(1992)
This modification altered the refuel interlock relay circuit to permanently defeat the interlocks associated with the service platform jib crane. The jib crane with its load sensing switches has been removed and the refueling interlocks no longer exist.
The circuit will continue to perform the required function to prevent movement of the remaining refueling equipment and the control rods. 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 an Unreviewed Safety Question.
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT l
MOD 5223 Year Implemented: U/2(1992) U/3(N/A)
This modification replaced the 2A, 2B, and 2C low pressure condenser neck feedwater heaters and associated shell-side relief valves during the Ninth Cycle Refueling Outage. This improves erosion / corrosion resistance performance, and availability of the equipment. The design thermal performance and operation of the equipment was not changed. This was essentially an in-kind replacement. This activity was an overall enhancement to operations and maintenance.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute sn Unreviewed Safety Question.
MOD 5228 Year Implemented: U/2(1992) U/3(1992)
This modification added a high discharge pressure alarm to the Steam Jet Air Ejector.
The addition of the alarm will alert the operator of abnormal conditions within the system. The alarm will occur prior to a system isolation. This activity affected documentation addressed in the SAR. The activity is an enhancement to operations. 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.
i MOD 5231 Year implemented: U/2(1992) U/3(N/A) i This modification replaced obsolete instrumentation in the condensate flow loops with j
state-of-the-art instrumentation. These replacements consist of flow transmitters, square root extractors, summers, indicators, controllerc, 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.
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 5233 Year Implemented: U/2(1992) U/3(N/A)
This modification removed the Reactor Core Different!al 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 concerns were created. There is no impact on plant capability.
Based on the Safety Evaluation and 59 above information, it was determined that these changes did not constitute an Unre.
M Safety Question.
l MOD 5235 Year implemented: U/2(1992) U/30 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 modification did not affect platform structural integrity or load carrying capabilities. No new safety concerns 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 l
changes did not constitute an Unreviewed Safety Question.
MOD 5236 Year implemented: U/2(1992) U/30 This modification was installed 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 RHR system, 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 concerns were created. This change was made to comply to 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 these changes did not constitute an Unreviewed Safety Question.
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d PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MQD_5238 Year implemented: U/2(1992) U/3(1992)
This modification replaced an existing 3/4" sample connection on the manway cover of each diesel generator fuel storage tank with a 4" sample connection. This new 4" sample connection allows for more accurate tank level measurements and also improved tank sampling and water removal. All applicable codes and piping specifications were adhered to. There is no impact on the storage tank integrity, required fuel volume or the Emergency Diesel Generator. Fuel performance is enhanced 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 5240 Year Implemented: U/2(1991) U/3(1992)
This modification removed a steam line drain level switch and replaced it with a functionally equivalent level switch. A change to the drain line piping configuration was necessary to accommodate the new type switch. This modification was necessary r
because the old switch was obsolete. The new switch performs the same function as the original switch. No safety concerns were created. Overall plant operability was maintained.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5244 Year Implemented: U/2(1992) U/3(1992)
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 is no system interface changes which would affect plant safety. This change results in improved monitoring capability.
Based on the Safety Evaluatan and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question, i
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PEACH BOTTOM ATOMIC POWER STATION -
UNITS 2 & 3 DOCKET No. 50-277 & 50-278 t
199210 CFR 50.59 REPORT i
MOD 5249 Year implemented: U/2(1992) U/3(N/A)
This modification installed a General Electric Zinc injection Passivation (GEZIP) system. Disposition 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 soluble zinc 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.
MOD 5255 Year implemented: U/2(1992) U/3(1992)
This modification removed a level transmitter and level indicator which monitored i
level in the clarified water storage tank and provides a signal for main control room level indication. This equipment is no longer necessary because monitoring is now performed in the water treatment building. This modification also relocated local level indicator Ll-70095 from a pipe mount near the floor to a wall mount. This mounting is more secure and will redJCe the likelihood of damage to the level indicator. No safety concerns were created. Equipment important to safety was not 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 5265 Year implemented: U/2(1992) U/3(N/A)
This modification installed a decontamination unit in the Turbine Building on elevation 116' at the south end of the hot tool room, for the decontamination of tools. Service Air and Demineralized Water piping, support systems, and related hardware were also supplied to the unit. The addition of the unit enhances the plant's capacity to keep contamination and radiation levels ALARA. There is no affect on installed plant equipment.
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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT l
MOD 5276A Year Implemented: U/2(1992) U/3()
This modification replaced Leeds and Northrup (L&N) multipoint Model W recorders with functionally equivalent Chessel Model 4200, and removes reco.ders that are i
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.
P MOD 5277 Year imp'emented: U/2(1992) U/3(1992)
This modification redesigned, and replaced the process components required to add corrosion treatment chemicals to the Emergency Service Water piping. The previous stainless steel components had deteriorated and needed to be removed from service.
The new piping, tubing, fittings, injection points, branch connection, and the i
replacement of stainless steel with titanium will enhance the systems performance and I
reliability. This modification makes no changes to the intended functions of the Service Water or Emergency Service Water and was completed to conform with applicable codes and original design bases.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
AiOD 5339 Year implemented: U/2(N/A) U/3(1992)
This modification changed the 120VAC feed to the auxiliary control circuit for the D-RHR and D Core Spray cooler room fans from a non-safety power source to a safety power source. This modification meets the applicable environmental, seismic, design, material, and construction standards. This activity is an enhancement to component reliability. No safety concerns were created. This change meets the original design intent of the system.
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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PEACH BOTTOM ATOMIC POWER STATION -
UNITS 2 & 3 l
DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 5340 Year implemented: U/2(1992) U/3(1992)
This modification installed seismically qualified storage racks inside the Torus rooms for the storage of scaffold tubing, miscellaneous scaffolding accessories, scaffold planks, plywood and scaffold toeboards during non-outage periods. Storage eliminates the cost associated with decontamination and transportation of this material from the i
Torus Room. The scaffold storage racks have been designed to withstand seismic events and flood forces. There is no effect on plant equipment important to safety.
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(N/A)
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 r
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.
MOD 5344 Year implemented: U/2(1992) U/3(N/A)
This modification added vibration instrumentation to the HPCI 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 will be 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 these changes did not constitute an Unreviewed Safety Question.
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 5353 Year implemented: U/2(1992) U/3(N/A)
This modification replaced the carbon steel Emergency Service Water Supply and return piping for the Residual Heat Removal purnps seal water coolers with larger diameter piping. This modification also involves the deletion of the test taps across the shutoff throttle valve on the return line of each seal water cooler, the deletion of the thermowells on the supply and return lines, and the addition of isolation valves on the supply and return line for the seal water coolers. This activity affected drawings addressed in the SAR. The changes increased the flow margin, and replaced the piping with material which is more resistive to corrosive environments. This change was an overall improvement to plant system reliability.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 53B8 Year implemented: U/2(1992) U/3(1992)
The existing office facilities within the Turbine Building were insufficient to i
support maintenance activities during plant outages. New facilities were provided in the Turbine building. This enclosure provides computer terminals electric, telephone communications, air conditioning, lighting and a fire protection system. This office i
does not adversely affect the station electrical system or create any safety concerns with installed plant equipment.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
MOD 5393 Year implemented: U/2(1992) U/3(N/A)
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 removed 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 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 an Unreviewed Safety Question.
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PEACH BOTTOM ATOMIC POWER STATION i
UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT MOD 5401 Year implemented: U/2(1992) U/3(N/A)
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 3
documentation addressed in the SAR. No design limits or safety concerns 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 86-069 Year implemented: U/2(N/A) U/3(1992)
This modification installed pressure gauges and vent valves on the Recirculation Motor Generator set lube oil systems, increased the size of the sensing lines for the lube oil pressure switches, and replaced and reopened the orifices in the fluid drive bearing oil header with a smaller one. This change was made to make the four M/G sets uniform, provide indication for performance monitoring, and add test points to install pressure gauges. This change affected drawings addressed in the SAR. This activity is an enhancement to operation and testing. 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 Ouestion.
NCR P880055 Year implemented: U/2(1992) U/3(N/A)
This NCR approved the removal of electrical power supplies for Residual Heat Removal Head Spray isolation Valves. This was accomplished by determination of the power and 1
control cables at their respective MCC's. The equipment had been abandoned in place and was closed under a previous modification. This change affects documentation 1
addressed in the SAR. The disposition does not affect the safety of the plant or create any seiety 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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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P880116 Year Implemented: U/2(1992) U/3(1992) -
This NCR was dispositioned to use as-is for the as-found condition of both Sola Regulating Transformer and a circuit panel supplying a portion of the Security System load. This was different from the design information drawing E-29 which showed load being supplied from another source. These changes are considered administrative in nature 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.
l NCR P890260 Year implemented: U/2(1992) U/3(1992) l This NCR was dispositioned to make changes to Process and instrumentation Diagram M-331. These changes were editorial in nature with the exception of the addition of two pressure indicators. These changes will enhance operation, testing, and maintenance of the hydrogen analyzer panels.
Based on the Safety Evaluation and the above information, it was determined that these i
changes did not constitute an Unreviewed Safety Question.
NCR P8902fi2 Year Implemented: U/2(1992) U/3(1992) i This NCR was dispositioned to correct discrepancies between the Process Instrument Diagram M-307 and the as-built condition of the plant. None of these items required a i
physical change to the plant. These changes were considered administrative in nature 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 P900042 Year Implemented: U/2(1992) U/3(1992)
This NCR addressed correcting the O-Status, line size and type, and locked status of the lines and valves to the ILRT Portable Console. These changes are considered administrative in nature 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.
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i PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P900076 Year implemented: U/2(N/A) U/3(1992)
This NCH was dispositioned to make changes to Process and instrumentation Diagram M-351. These changes show an as-installed vent connection and valve in the Startup Recirc Flowpath. This was a document change only and does not impact the design basis or function of any other equipment or systems shown. These changes were considered administrative in nature 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 P900163 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to revise drawings to reflect the as-built conditions of two 13.8KV switchgear for cocling tower lift pumps. In addition, replacement of several 6 Amp fuses to 3 Amp fuses will ensure fuse coordination in the power factor relay circuits of the pumpo. No safety concerns were created. This equipment is not safety related.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P900202 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve several discrepancies identified between the as-built Turbine and Extraction Steam System and Process and Instrument Diagram M-304.
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These are editorial changes only and do not affect the capability of the system or any interfacing system to perform their required function. This change is administrative in nature. 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.
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P900205 Year Implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve discrepancies identified between the as-built conditions of the plant and the Process and Instrumentation Diagram M-308.
Documentation changes will be made to show the as-is condition. These changes are administrative in nature 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 P900214 Year Implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the as-built condition of the plant and the Process and Instrumentation Diagram M-352. Changes consisted of correcting valve numbers, drafting errors, and making other clarifications. These changes are administrative in nature 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 P900216 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the as-built condition of the Reactor Water Clean-up System and the Process and Instrumentation Diagrams M-354.
Changes made consisted of corrections of valve numbering, drafting errors, and clarifications. These changes were administrative in nature 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 P900221 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the as-built condition of the Core Spray System and the Process and Instrumentation Diagram M-362. These changes consisted of numbering changes, were administrative in nature, 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.
29
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT BCR P900225 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the Containment Atmospheric Dilution system diagram, and Process and Instrumentation Diagrams M-372
)
and M-388. None of the discrepancies require physical changes to the system or its components. These changes are administrative in nature 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 P900244 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to eliminate the Reactor Feedwater pumps ('B' and 'C')
differential pressure transmitters DPT-3123 and DPT-3124 from service. The feedwater lines leading to the instruments were cut and capped. An evaluation determined that these switches provided no useful input to operations. 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.
NCR P900357 Year implemented: U/2(1992) U/3(N/A)
This NCR was dispositioned to remove the area radiation monitor (RE-2-18-30BC) and associated equipment. This equipment was inside the vault in an area designated for new fuel, but never used. There are no plans to use this vault for fuel storage in the future since station procedures require placing new and used fuel directly into the fuel pool. 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.
NCR P900446 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve discrepancies betwee, plant configuration and to correct discrepancies between the OAD, the O-list and the Process and Instrumentation Diagram M-363. This documentation is administrative in nature and does not affect safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
30
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P900490 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the as-built configuration the plant and documentation. Corrections are a result of the changes made by a previous modification to the instrument nitrogen system, but not incorporated into documentation.
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 P900579 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to revise documentation to recognize the Rod Worth Minimizer control rod select blocks as nonfunctional and remove their reference from the UFSAR.
The insert and withdraw blocks meet design criteria. This is a document change only is administrative in nature, and reflects the as-is condition of the plant. This raise in temperature does not affect any accident or transient response or the ability of the equipment to perform their function. There is no affect on 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 NCR P900622 Year Implemented: U/2(1992) U/3(N/A)
This NCR was dispositioned to resolve discrepancies between single line drawings, as-built plant conditions, and the UFSAR. These changes are administrative in nature and do not change plant configuration or affect plant safety.
i l
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
l 31
I PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P900789 Year Implemented: U/2(1991) U/3(1992)
This NCR is dispositioned to resolve discrepancies between the as-built configuration of the plant and the Process and instrumentation Diagram M-318. Revisions show that the standpipe is connected to the 10" fire main at one connection only which is upstream of hose reel HR-116-19. These changes are administrative in nature 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 P910023 Year implemented: U/2(1992) U/3(1992)
This NCR is dispositioned to permanently increase the average analytical drywell air temperature to 145. Analysis of operating temperature data indicated that the previous 135 degree average was exceeded. Increasing the average temperature was evaluated to determine the effect on the drywell structure, plant equipment, and accident analyses. 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 P910158 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between UFSAR Chapter 14 text and Chapter 8 text, tables and various previously reported changes conducted over the years. These changes provide clarifications only. No physical changes were made.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i l
NCR P910175 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to enhance the Process and Instrumentation Diagram M-367 to -
read sample points instead of sample elements and adds outer root valves and inner root valves to the drawing. These changes were administrative in nature and do not afiect plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
32 i
, J
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 i
DOCKET No. 50-277 & 50-278 1
199210 CFR 50.59 REPORT l
i NCR P910218 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between Process and Instrumentation Diagrams M-317 and M-319 and plant configuration. This revision will show portable stop valve, conductivity equipment, and has assigned equipment numbers i
to and labeled equipment. This change is administrative in nature and does not affect plant safety.
i Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i NCR P910222 Year implemented: U/2(1992) U/3(N/A)
This NCR was dispositioned to install a sample connection which was installed on the Reactor Core isolation Cooling system. This connection is used during maintenance and provides a means of sampling the quality of oil within the system. It does not alter piping to which it is connected or affect the function or safety of the system.
Based on the Safety Evaluation and the above information, it was determined that these I
changes did not constitute an Unreviewed Safety Question.
-I NCR P910294 Year implemented: U/2(1992) U/3(1992)
This NCR was disposition to resolve discrepancies between the as-built configuration and the Process and Instrumentation Diagram M-317. This change involved drawings associated with the Clarified Water Storage Tank. This is a document change only and does not affect plant design or safety, Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910323 Year Implemented: U/2(N/A) U/3(1992)
This NCR concerned the 3C battery rack anchor bolts. One of the twelve bolts which anchor the rack to the floor was corroded. A seismic analysis review has been concluded with the recommendation to use-as-as. There is no affect on the performance of safety related equipment. No safety concerns were created as a result of this decision.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
33
.=.
l PEACH BOTTOM ATOMIC POWER STATION l
UNITS 2 & 3 l
DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P910327 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between documentation and as-built plant configuration. The existing turbine stop valve permissive in all modes will be
)
noted on appropriate documentation. This is an administrative change and will bring documentation into conformance with plant conditions, i
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910359 Year implemented: U/2(1992) U/3(1992)
This NCR identified various document changes to reflect as installed plant configurations on P&lD Drawings M-303 and M-331. Stop valve lines from the reactor feed pump turbines 1
were noted, details were added for a 6" piping nozzle, a branch connection was noted, details for the Y-strainer were added and branch connections are documented as upstream.
These changes were administrative in nature and do not affect plant safety or reliability.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
9 NCR P910375 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to use-as-is and approves the operation of the Residual Heat Removal motors and the Core Spray pump motors in the absence of certain EO required maintenance. This disposition does not affect the operation of equipment or plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safe'y Question.
i NCR P910377 Mr implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the as-built' configuration and documentation. It addressed penetrations that existed and were being tested, but were not shown on station documentation. Also addressed were editorial discrepancies for penetration N-24, Unit 2. Tnese changes were administrative in nature and hE.ve no affect on plant performance.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
4 34
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P910389 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve discrepancies between as-built configuration of the plant and the Condensate Off-gas Loop seal piping. These changes are administrative in nature only and do not affect system operation, performance, or safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
1 i
NCR P910491 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the as-built configuration of the plant. The change addressed the removal of the sight flow gauge on 10GB-24" header drain. This gauge was removed. Flow can be visually determined by observing flow through tha funnel to the Clean Radwaste drain. These changes are administrative in nature only and have no affect on 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 P910540 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between Process and Instrumentation Drawing M-310 and the as-built configuration of the plant. Documentation was revised to i
show the Marinelli Sampler, associated piping, valves, and off-gas stock sample panel 00C101. These changes are administrative in nature and have no impact on the sampling system or its ability to perform its design function.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910543 Year implemented: U/2(1992) U/3(N/A)
This NCR was dispositioned to resolve discrepancies between the as-built plant conditions 1
and station documentation. Revisions reflect the as-built configuration of the Safety Grade Instrument Gas system.- This change will document the as-is configuration of the vent control valves. The change is administrative in nature only and does not affect safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
35
I
-l PEACH BOTTOM ATOMIC POWER STATION I
UNITS 2 & 3 DOCKET No. 50-277 & 50-278
.i 199210 CFR 50.59 REPORT NCR P910572 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the O-list and station documents. No equipment hardware is affected. Documentation will be revised as necessary. These changes are administrative in nature and do not affect plant safety.
Based on the Safety Evaluation and the above information, it was determined that these l
changes did not constitute an Unreviewed Safety Question.
RCR P910573 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies discovered during a 0-list validation effort and to correct the associated Process and instrumentation Diagram.
RO 2 (3) A2706B,27078,2110B and 2111B were deleted from the O-list and TE designations were changed to TW. These changes are administrative in nature 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 P910665 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to eliminate Service Air and Service Water lines supported from Fire Protection Piping. The headers were originally installed for convenience to.
be used during Water-Box clean-up. These changes do not affect plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewes Safety Question.
NCR P910694 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between station documentation and the as-is condition of the plant. This change was administrative in nature and did 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.
36
PEACH BOTTOM ATOMIC POWER STATION -
UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P910695 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies and between station documentation and the as-is condition of tne plant. These changes are administrative in nature and j
do not affect the safety of the plant.
t Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910713 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies in documentation to reflect the as-built plant configuration of the Emergency Service Water and High Pressure Service Water systems. These changes are administrative in nature. No plant hardware was changed. No safety concerns were created.
BasM ~1 the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910714 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between station documents (M-872 and 0-list) and as-is plant condition. Document changes consisted of adding sub components, correcting component numbers and editorial changes. These changes are considered administrative in nature and have no affect on 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 P910715 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between station documents (M-872 and O List) and as-is plant conditions. Document changes consisted of adding sub components, correcting component identification numbers, and editorial changes. These-changes are considered administrative in nature and have no affect on plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
37
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P910716 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to resolve discrepancies between station documentation (M-367, OAD 867, and the O-list) and the as-built condition of the plant. These changes are administrative in nature 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 P910721 Year implemented: U/2(1992) U/3(1992)-
This NCR was dispositioned to resolve discrepancies between station documentation (M-372), the associated QAD, the O-list and the as-installed condition of the plant.
These are administrative changes only and do not make any physical change to the plant or 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 P910722 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between plant documentation (M-367 the associated QAD, O-list) and the as-is plant configuration. This activity is administrative only and does not affect the safety 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.
NCR P910781 Year implemented: U/2(N/A) U/3(1992)
This NCR identified discrepancies between station documentation (M-367, OAD M-815, the O-list) and the as-built plant configuration. These documents will be changed to represent the as-built configuration. These changes are administrative in nature and will 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.
38
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P910830 Year implemented: U/2(1992) U/3(1992)
This NCR identified that the as-found configuration of the 480V AC power feeds to the 2B Low Pressure Coolant injection swing bus does not agree with the single line drawing and schematic. The as-found wiring indicates that the preferred power is from MCC 20B37 (E224 R-B) circuit breaker 52-3791 and the alternate feed is from 20B39 (E424-W-A) circuit breaker 52-3922. Documentation was changed to reflect the correct configuration. This NCR was dispositioned to also change the labeling of " preferred" power source to " normal" power source for all four swing busses. These are -
administrative changes only and will 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 P910869 Year implemented: U/2(N/A) U/3(1992)
This NCR was dispositioned to replace the existing DC motor on MO-3-23-016. with a DC motor that will be capable of delivering the required torque under degraded voltage conditions. This new motor was required so that the Limitorque operator could develop sufficient thrust. The new motor does not impact test procedures, plant requirements, or safety concerns.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
NCR P910971 Year Implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to resolve discrepancies between the as-built TIP drive and station documentation. 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.
b i
39
.g PEACH BOTTOM ATOMIC POWER STATION UNrrS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT NCR P920009 Year implemented: U/2(1992) U/3(1992)
This NCR was dis itioned to change documentation to agree with the as-built c
configuration of the Orywell Torus Vacuum Breaker System. An existing hand valve, sub component number, and component numbers will be added to drawings. These changes are administrative in nature only and do not create a change to the facility or a safety concern.
Bised on the Safety Evaluation and the above information, it was determined that these tanges did not constitute an Unreviewed Safety Question.
NCR P920209 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to revise documentation to agree with the Core Spray System Instrument range specifications as they exist in the plant. This is an administrative change only 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 P920222 Year implemented: U/2(1992) U/3(N/A)
This NCR was dispositioned to disconnect and remove a dimaged lightning arrestor at a location on the Unit 1 main transformer. This arrestor was no longer needed as the transformer has been removed.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
LtC_R2920297 Year implemented: U/2(1992) U/3(N/A)
TI e A CR was determined that several cooling tower intake screen level switches and associated transmitters are not required and can be abandoned in place. Documentation will be revised as appropriate. No safety concerns were created as a iesult 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.
40 1
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT fJCR P920489 Year implemented: U/2(1992) U/3(1992)
This NCR was dispositioned to increase the allowable ambient room temperature in the Emergency Switchgear and Battery Rooms from 105 degrees to 118 degrees. Evaluation confirmed this has no affect on safety related equipment controlled in these rooms.
Based on the Safety Evaiuation and the above information, it was determined that these i
changes did not constitute an Unreviewed Safety Question.
FROCEDURE AO-10.9-2(3)
Year implemented: U/2(1992) U/3(1992) 1 These procedures established a method to be used to either transfer or restore the Residual Heat Removal System Room Coolers to their desired configurations. This activity affected documentation addressed in the SAR. The change will not..koduce any new operating modes or affect the safety of the plant.
Based on the Safety Evaluation and the chove information, it was determined that these changes did not constitute an Unreviewed Safety Question.
. PROCEDURE AO-13.2-2(3)
Year Implemented: U/2(1992) U/3(1992) l This Safety Evaluation reviewed and approved the procedures used to either transfer or restore the Reactor Core isolation Cooling Systi. Room Coolers to their desired configurations. This activity affected documen ' <. 'ddressed in the SAR. The change did not introduce any new operating mc.s of affect the safety 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.
fBQ7 DURE AO-14.1-2(3)
Year Implemented: U/2(1992) U/3(1992)
These procedure established a method to be used to either transfer or restore the Core Spray System Room Coolers to their desired configurations. This activity affected docurnentation addressed in the SAR. The change did not introduce any new operating modes or affect the safety of the plant.
Based on the Safety Evaluation and the above information, it ~Re etermined that these changes did not constitute an Unreviewed Safety Question.
~
41
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278
' 199210 CFR 50.59 REPORT PROCEDURE AO-23.3-2(3)
Year Implemented: U/2(1992) U/3(1992)
These procedures established a method used to either transfer or restore the High Pressure Coolant injection System Room Coolers to their desired configurations. This activity affected documentation in the SAR. The change will not introduce any new operating modes or affect the safety 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.
PROCEDURE Emergencv Plan Table 4.1 Year implemented: U/2(1992) U/3(1992)
This procedure was revised to add changes associated with Modification 5238 (Torus Hardened Vent). These changes do not adversely affect the safety of the plant, but provide additional actions required because of this modification. No safety concerns were created as a result of this activity. This change is an overall enhancement to plant safety for plant events beyond design bases.
Based on the Safety Evaluation and the above info. iation, it was determined that these changes did not constitute an Unreviewed Safety Question.
PFOCEDURE M-003-216 Year implemented: U/2(1992) U/3(1992)
This new procedure was developed to inspect the Control Rod Drive Hydraulic Control Unit gate valve wedge. This activity created a small opening in the SDV for a short period of time. The scram discharge volume is not affected. This does not affect the overall safety 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.
PROCEDURE SP-1422 Year implemented: U/2(1992) U/3(1992)
This evaluation addresses the conversion of a procedure for control rod withdraw for maintenance during core shuffle which is already in use on Unit 3 to a procedure for l
Unit 2. This allows the use of jumpers previously made and adds extra operational verification steps. No safety concerns are created by this procedure.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
42
PEACH DOTTOM ATOMIC POWER STATION UNITS 2 & 3 l
DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT PROCEDURE SP-1442/1443 Year implemented: U/2(1992) U/3(1992).
This procedure provides a method for performing feedwater flow measurement using radioactive sodium (Na24) tracer. This test was performed to verify line flow measurement instrumentation by using a radioactive tracer to verify whether there was any discrepancy in the feedwater flow measurement. The injection of the small amount of codium nitrate into the feedwater was negligible. The increase in main steam line radiation was kept within normal operating limits. No safety concerns were created by this procedure.
Based on the Safety Evaluation and the above information, it was determined that these changes dio not constitute an Unreviewed Safety Question.
PROCEDURE ST-D-033-420-2(3)
Year Implemented: U/2(1992) U/3(1992)
This procedure provides a method to ensure that the Emergency Core Cooling Systcm and Reactor Core isolation System Room Coolers are maintained in an operable status during Emergency Service Water activities. No safety concerns were created as a result of this procedure.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
PROCEDURE ST-O-003-560-a Year Implemented: U/% 992) U/3(N/A) i This procedure change involved the testing of control rod drive 40-31 because of rod withdraw problems. In addition wording will be changed to enhance the quality of the test. This change will eliminate the unneeded distraction of an alarm sounding during the performance of these procedure steps.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
i i
43
-l
I PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT PROCEDURE ST-X47G-102-2(3)
Year implemented: U/2(1992) U/3(1992)
This procedure was changed to improve format and establish consistency among tests.
This test involves various portions of the Primary Containment isolation System logic.
Abnormal plant conditions and isolations are simulated by installing jumpers, lifting leads and pulling fuses. This activity is addressed in the SAR. The original intent of logic system functional testing included testing to the relay level only. This test verifies proper operation of each contact in the logic. This level of testing is not possible without the use of temporary circuit alterations. No safety concerns were created.
Based on the Safety Evaluation and the above information, h was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 2-06-Q29 Year Implemented: U/2(1992) U/3(N/A)
This TPA was required to support passive monitoring during testing of the newly installed Replacement Feedwater Control System modification. This did not adversely impact operation of the system or create a safety concern. Installed instrumentation did not affect system performance.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
P TPA 2-18-001 Year implemented: U/2(1992) U/3(N/A)
This TPA defeated the refueling interlocks associated with bridge position over the reactor and control rod withdraw. This was necessary so that work could be performed over the reactor shield plug. During this activity the reactor was assembled and reactor shield plugs were in place. This change affected documentation addressed in the SAR This did not create a safety concern. Because no vessel work was in progress, this change does not 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.
I 44
PEACH BOTTOM AYOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT TPA 2-23-008 Year implemented: U/2(1992) U/3(N/A)
This TPA was necessary to complete High Pressure Coolant injection troubleshooting on the hydraulics during a maintenance outage. It replaced gages with pressure transducers to support system monitoring. The change affected figures and drawings addressed in the SAR. The configuration of the system and the modes of operation were not affected. Installed instrumentation did not affect system performance.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 2-30-009 Year implemented: U/2(1992) U/3(N/A)
This TPA provided a cooling water supply to the Service Water bearing oil coolers while the high pressure lube water pump was out of service. This activity affected documentation addressed in the SAR. This activity did not introduce any new operating modes or affect plant safety. Service Water pumps are not safety related. The change maintained operational capability for Service Water Systems.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Questinn.
TPA 2-52-011 Year Implemented: U/2(1992) U/3(1992)
This TPA replaced two existing needle valves in the pilot air lines to the air start valve with straight piping per interim disposition of an Non Conformance Report. This was necessary to support operation of the E2 Emergency Diesel Generator and because parts for these needle valves were not available. This change did not affect system pressure or impact safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
45
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT TPA 2-62-005 Year implemented: U/2(1992) U/3(N/A)
This TPA was installed to jumper a control rod position interlock in order to complete Reactor Manual Control System logic. This was done to satisfy the " full in" condition necessary for the fuel movement while in the refuel mode. The control rod blade was full in and no safety concerns were created.
Based on the Safety Evaluation and the above information, it was deterrained that these changes did not constitute an Unreviewed Safety Question.
TPA 3-01G-033 Year implemented: U/2(N/A) U/3(1992)
This TPA was necessary because a bad thermocouple was creating a failure upscale causing a nuisance alarm. This was a temporary condition until repairs cot';d be made.
during an outage. This equipment was used for monitoring valve leakage only. There were no safety concerns and other methods to measure leakage were also available. The installed equipment caused no affect on plant operation. Increased monitoring was an enhancement to plant safety.
Based on the Safety Evaluation and the above information, it was determined that these i
changes did not constitute an Unreviewed Safety Question.
TPA 3-05-019 Year implemented: U/2(N/A) U/3(1992)
This TPA installed a temporary feed skid to the Condensate Demin system to support the addition of precoat for longer demin operation. Tie-ins included demin water supply, electrical feed to a power pack and 3/8" stainless steel tubing to the condensate demin inlets. This activity affected documentation addressed in the SAR.
This TPA did not change or degrade any existing plant equipment or impact safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
46
PEACH BOTTOM ATOMIC POWER STATION UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT TPA 3-08-014 Year Implemented: U/2(N/A) U/3(1992)
This TPA was necessary to defeat the alarming function of an off gas pressure switch.
Ufting the lead on this pressure switch removed the false alarm indication. This did not create a safety concern as a redundant alarm was still active.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 3-08-015 Year implemented: U/2(N/A) U/3(1992)
This TPA installed thermocouples on the offgas piping. This was necessary to measure pipe temperature and ambient air temperature on the offgas pipe so that trouble shooting could be completed. These thermocouples were strapped on the outside of ths pipe only and did not interfere with the intended function of the Offgas system or its associated piping, therefore, there was no impact on plant safety.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 3-13-007 Year Implemented: U/2(N/A) U/3(1992)
This installation was initiated because the Reactor Core Isolation Cooling steam line temperature switches were inoperable. Jumpers were installed to place the 'D' channel in a safe condition as required by the Technical Specifications. This change will not introduce any new operating modes or 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.
j i
TPA 3-13-009 Year implemented: U/2(N/A) U/3(1992)
This installation was initiated because the Reactor Core isolation Cooling steam line temperature switches were inoperable. Jumpers were insta!!ed to place the 'D' channel in a safe condition as required by the Technical Specifications. This change will not introduce any new operating modes which 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.
47
PEACH BOTTOM ATOMIC POWER STATION
. UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT TPA 3-30-011 Year implemented: U/2(N/A) U/3(1992)
This TPA installed hoses from discharge of Service Water pumps to supply cooling water-to bearing coolers while the high pressure lube water pump is out of service.
Personnel performed monitoring during the time this TPA was in effect. This activity affected documentation addressed in the SAR. There were no new safety concerns created as a result of this activity. The ability to maintain the Service Water pumps is important to plant safety. This activity assured overall plant safety was maintained.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
TPA 3-57-006 Year Implemented: U/2(N/A) U/3(1992)
This TPA installed a monitoring system on the alternate shutdown Topaz inverters.
Topaz inverters have experienced failure due to overvoltage conditions. A system was installed to monitor and detect such conditions. No changes were made to the system or equipment. No new modes of operation were introduced by this activity. This change was an enhancement to plant monitoring capabilities.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
t UFSAR CR-9108447 Year implemented: U/2(1992) U/3(1992)
This evaluation addressed a change to the Updated Final Safety Analysis Report to more accurately describe post accident Main Steam Isolation Valve operational capabilities.
This change affected documentation addressed in the SAR. No hardware changes were made. This change was administrative in nature only and does not introduce any new safety concerns.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
48
PEACH BOTTOM ATOMIC POWER STATION 1
UNITS 2 & 3 DOCKET No. 50-277 & 50-278 199210 CFR 50.59 REPORT.
1 UFSAR CR-9108590 Year implemented: U/2(1992) U/3(1992)
This change was made to allow better methods of analysis for boron and chloride in Post Accident Sampling System camples. These methods employ current industry standards and will allow sampling at any time without exceeding the radiological guidelines. No new safety concerns were created. This does not change the operation of equipment, but improves methods of ar.alysis.
Based on the Safety Evaluation and the above information, it was determined that these changes did not constitute an Unreviewed Safety Question.
UFSAR CR-9108594 Year Irnplemented: U/2(1992) U/3(1992)
This evaluation addressed 10CFR21 a deficiency in the coaxial cable interconnecting integral components of the Primary High Range Radiation Monitoring System. The determination that was made that the cable can provide its design indication function as it exists. Original equipment capabilities were maintained. No affect on plant operations was realized. 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.
UFSAR Section 12.2.10 Year implemented: U/2(1992) U/3(1992)
This evaluation addresses a revision to the UFSAR to indicate that the circulating water pump structure water tight exterior flood doors are normally closed. The plant has been maintained in this condition but not documented as such. 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.
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