ML20059E712
| ML20059E712 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/30/1993 |
| From: | Boyce R PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9401120180 | |
| Download: ML20059E712 (37) | |
Text
, q 10CFR50.59(b)
PHILADELPIIIA ELECTRIC COMPANY LIMERICK GENERATING STATION P. O. BOX 2300 f SANATOGA, PA 19464-23(X)
L (215) 327-1200 EXT. 2000 ROBERT W. BOYCE December 30, 1993 PLAN 7 MANAGE 9 uMEntCK GENERATING STRION 50-353 License Nos. NPF-39 NPF-85 l
l U.S. Nuclear Regulatory Commission Attn: Dor iment Control Desk l
Washington, DC 20555 subject: Limerick Generating Station, Units 1 and 2 Annual 10CFR50.59 Report For the Period July 1, 1992 Through June 30, 1993 Attached is the Annual 10CFR50.59 Report as required by 10CFR50.59(b).
If you have ar.y questions or require additional information, please contact us.
Sincerelyy -
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BN:cah Attachment f cc: T. T. Martin, Administrator Region I, USNRC I N. S. Perry, USNRC Senior Resident Inspector, LGS I
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PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 -
ANNUAL 10CFR50.59 REPORT JULY 1, 1992 THROUGH JUNE 30, 1993 This report provides a brief description of changes to the facility and procedures as described in the Safety.
Analysis . Report, tests, and experiments that were implemented between July 1, 1992 and June 30, 1993. ~
A summary of the safety evaluation for each item concluded that an unreviewed safety question, as defined in 10CFR50. 59 (a) (2 ) , was not involved.
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1 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 ,
ANNUAL 10CFR50.59 REPORT JULY 1, 1992 THROUGH JUNE 30, 1993 TABLE OF CONTENTS l l
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PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 ANNUAL 10CFR50.59 REPORT TABLE OF CONTENTS Modifications . . . . . . . . . . . . . . . . Page 5001 . . . . . . . . . . . . 1 ;
5241 . . . . . . . . . . . . . . . . . . . . . 1
. 5248 . . . . . . . . . . . . . 2
, 5983 . . . . . . . . . . . . . . 2 6065 . . . . . . . . . . . . . . 3 6130 . . . . . . . . . . . . 3 6133 . . . . . . . . . . . . . 4 6135 . . . . . . . . . . . . . 5 6136 . . . . . . . . . . . . 5 6137 . . . . . . . . . 6 6138 . . . . . . . . . . . . . . . 6 6139 . . . . . . . . . . . . . 6 6144 . . . . . . . . . . . . . . . . . . . . 7 6145 . . . . . . . . . . . . . . . . 7 6147 . . . . . . . . . . . . . . . . . 7 6154 . . . . . . . . . 8 6155 . . . . . . . . . 8 l 1
6161 . . . . . . . . . . . 8 6163 . . . . . . . . . . 9 6165 . . . . . . . . . . 9 6167 . . . . . . . . . . . . 10 6168 . . . . . . . . . 10
. 6171 . . . . . . . . . . . . . 10 6183 . . . . . . . . . . 11 6187 . . . . . . . . . . . . . 11 6188 . . . . . . . . . . . . . . . 12 6189 . . . . . . . . . . . . . . . 12 6191 . . . . . . . . . . 12 6207 . . . . . . . . . . . . . . . 13 6218 . . . . . . . . . . . . . . . . 13 6220 . . . . . . . . . . . . . . 14 6221 . . . . . . . . . . . . . . . . 14 6226 . . . . . . . . . . 15 Moncon f ormance Ror> ort s l
91-164 . . . . . . . . . . . . . . . . . . . 15 '91-267 . . . . . . . . . . . . 15 92-086 . . . . . . . . . . . . . 16 92-109 . . . . . . . . . . . . . 16 92-196 . . . . . . . . . 17 l 92-207 . . . . 17 92-230 . . . . . . 18
TABLE OF CONTENTS Nonconformance Reports (Continued)92-233 . . . . . . . . . . . . . . . . . . . 18 92-234 . . . . . . . . . . . . . . . . . . . 19 po_737
. u . . . . . . . . . . . . . . . . . . . . 19 '92-248 . . . . . . . . . . . . . . 20 co_u,c;4
-u . . . . . . . . . . . . . . . . . 20 co_oro su ua- . . . . . , . . . . . . . . . . . . . u1 0-92-270 . . . . . . . . . . . . . . . 21 92-306 . . . . .. . . . . . . . 22 92-353 . . . . . . . . . . . . . . . . . . 23 92-355 . . . . . . . . . . . . . . . . 23 93-036 . . . . . . . . . . . . . . . 24 93_no,UOu e . . . . . . . . . . . . . .
9 u4 93-092 . . . . . . . . . . . . . . . . 25 93-119
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~6 93-137 . . . . . . . . . . . . . , 27 93-203 . . . . . . . . . . 27 93-204 . . . . . . . . . . 28 93-214 . . . . . . . 28 q,
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93-252 . 79 q3_or, w au . . . . . . . . . . . . . o Procedures, Tests, or Experiments '
SP-127 . . . . . . . . . . . . . . . . 30 SP-S-080 . . . . . . . . . .. . . . 30 i SP-HF-010 . . . . . . . . . 31 j i
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PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 ANNUAL 10CFR50.59 REPORT JULY 1, 1992 THROUGH JUNE 30, 1993 SAFETY EVALUATION SUMMARIES
i LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 5001 Unit 1 x Unit 2 _ Common _
This modification installed vent lines and valves in the Reactor Water Cleanup (RWCU) system between the pump suction block valves and the pump discharge block valves. These vents were added to allow venting of the pumps following maintenance since the RWCU system high point vents are not located in the section of piping !
isolated during pump maintenance. System venting is needed to avoid water hammer and introduction of air into the reactor following RWCU system maintenance. This modification affected a figure in the UFSAR. The modification resulted in an overall enhancement to the operation of the RWCU system.
MOD 5241 Unit 1 x Unit 2 _ Common _
This modification replaced several temporary plant alterations l (TPAs) to the computer input circuits of the Emergency Response Facility Display System (ERFDS). During the startup and power ascension of Unit 1, temporary ERFDS computer inputs (startup inputs) were installed and subsequently modified by TPAs to provide convenient tap locations and/or proper signals for input to the ERFDS computer, or to determinate ERFDS temporary (startup) cables in order to maintain separation criteria, or to determinate ERFDS temporary (startup) cables to maintain safe shutdown criteria. This modification closed out these TPAs, provided a permanent installation for a certain number of these i temporary inputs and removed the remainder of the temporary inputs. Systems affected by this modification include Electrical Distribution (220 kV and 500 kV Substation), Residual Heat Removal, Reactor Core Isolation Cooling, High Pressure Coolant Injection, Reactor Protection, Reactor Recirculation, and Nuclear Instrumentation. The monitoring of these systems by ERFDS ,
provides the plant operators and plant staff with additional information to monitor the performance of the affected systems.
This modification affected various figures in the UFSAR. The modification resulted in an overall enhancement to operations.
This modification does not compromise safety or reliability and will not adversely affect the original design intent of the Unit 1 ERFDS or Class IE systems monitored by the ERFDS computer.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 l
l MOD 5248 Unit 1 _ Unit 2 3 Common _
This modification upgraded various analytical instrumentation with more reliable instrumentation. Specifically, this modification replaced conductivity instrumentation in the Reactor i Water Cleanup inlet / outlet, Feedwater Condensate Pump discharge, Makeup Demineralizer, Condensate Filter Demineralizer inlet / outlet, and Radwaste Demineralizer effluent sampling systems. This modification also added dissolved oxygen monitoring to the Reactor Water Cleanup inlet and Feedwater ,
sampling systems. Sample temperature conditioning was included !
for the Reactor Water Cleanup inlet and Feedwater samples.
Turbidity instrumentation was removed from the Feedwater, Reactor Water Recirculation, and Condensate Filter Demineralizer sampling systems. These changes provided more reliable and accurate monitoring and managing of the plant water chemistry. This modification affected several figures of the UFSAR. The changes were selected to envelope or duplicate previous instrument ranges with required accuracy to monitor the quality of the water.in plant systems and provide operations with information for taking any necessary corrective actions. This modification did not provide any interlocks or any other safety-related function.
Ch?'ges in seismically qualified panels do.not create any di.ierent type of accident or malfunction because the modification only replaced recorders, deleted signal resistors, .
and determinated wiring. The modification is an overall !
enhancement to operations.
MOD' 5983 Unit 1 3 Unit 2 3 Common _
This modification installed eight Deep Bed Condensate Demineralizers on each unit. The primary purpose for installing the Deep Bed Condensate Demineralizer System (DBCDS) is for the more complete removal of copper ions from the condensate. Copper is known to cause fuel element degradation, via a mechanism known as crud-induced localized corrosion. The copper comes from admiralty condenser tubes. The Deep Bed Demineralizers were installed in the feedwater system between the Condensate Filter /Demineralizers and the Feedwater Heaters. This modification also installed the necessary support equipment and j instrumentation to operate the new system. This modification I affected several sections, figures, and tables of the UFSAR. No safety related systems are affected by this modification. The probability of a failure in the DBCDS which would inhibit the restoration of feedwater following a loss of feedwater event is insignificant 1y low. An accident initiator different then those l in the SAR does not exist because of this modification. Active and passive equipment that respond to an accident will not be changed or degraded. This modification is an overall enhancement to operations and safety.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 6065 Unit 1 _ Unit 2 _ Common x This modification installed vendor designed safety indicator panels for the Control Enclosure chiller units. The mechanical / electrical equipment protection logic system of the previously existing equipment precluded Operations personnel from determining the cause of a chiller trip in a timely manner. A trip condition can be initiated from one of eleven input signals from various monitoring devices and troubleshooting the actual cause can be unnecessarily difficult and time consuming. The new '
indicator panel is capable of providing first trip indication for the trip functions. This modification affected two figures in the UFSAR. The modification is an overall enhancement to operations.
MOD 6130 Unit 1 _ Unit 2 _ Common 3 This modification installed an in-plant office facility for use by the floor operators. The facility is used for break periods and for performing planning work associated with daily operations. Two permanent offices were installed within the Turbine Enclosure. The immediate area surrounding the new facilities is a non-radiologically controlled area (non-RCA).
This involved the relocation of an existing Personnel Contamination Monitor (PCM) and the addition of railing and signage to isolate the area. This modification affected several sections and figures in the UFSAR. This modification is an overall enhancement to operations and reduces some unnecessary .
Main Control Room traffic.
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l LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 l
MOD 6133 Unit 1 _ Unit 2 3 Common _ i This modification replaced the Riley Temperature Monitoring Instrumentation associated with the Steam Leak Detection System ;
(SLDS) with General Electric Co. (GE) Nuclear Measurement '
Analysis and Control (NUMAC) Leak Detection Monitors (LDM's).
These monitors are part of the GE microcomputer based instrument family designed for various applications in nuclear power plants. l This modification also made changes to the power feeds of the SLDS and replaced a Residual Heat Removal system temperature switch. The Riley Temperature Monitoring System was replaced l because it had been a source of several Licensee Event Reports (LER's) due to spurious trip signals which have caused system isolations on both units. Additionally, the Riley ambient and '
differential temperature transmitter switches required the lifting of the thermocouple leads to perform the monthly functional testing. The lifting of the leads was labor intensive and has also contributed to several LER's. This modification affected several sections, figures, and tables in the UFSAR. The design function of the SLDS system was not changed by the replacement of the Riley Temperature Monitoring System. The NUMAC LDM's perform the same design function as the previously existing instrumentation. The installation of the new equipment eliminated the need to physically determinate and reterminate thermocouple wires in order to accomplish a monthly surveillance ,
test procedure. Additionally, the updated electronic components contained in the NUMAC design are not susceptible to electrical ;
noise which contributed to spurious false trip signals in the previously installed Riley instruments.
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l LIMERICK GENERATING STATION l UNITS 1 AND 2 '
DOCKET NOS. 50-352 AND 50-353 i LICENSE NOS. NPF-39 AND NPF-85 j I
MOD 6135 Unit 1 _ Unit 2 x Common _
This modification modified and enhanced the existing Turbine Generator Fire Detection and Suppression System to provide the ,
following: l
- 1. Extended existing suppression system (sprinkler and rate i compensated heat detectors) around the perimeter of the l generator beneath the appearance lagging.
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- 2. Extended existing suppression system (fusible element l nozzles and a rate compensated heat detector) for generator / exciter bearing (Bearing No. 10).
- 3. Relocated and added the rate compensated heat detectors for Bearings 2, 3, 4, 5, 6, 7, 8, and 9 so that each bearing has ]
an individual heat detector.
l This modification was installed to enhance the generator and generator / exciter bearing suppression system to minimize property damage and unit downtime resulting from a postulated Turbine l Generator fire. This modification affected several sections, l figures, and tables of the UFSAR. The Turbine Generator Fire j Detection and Suppression System is not important to safety and :
the inadvertent operation of the system is minimized by the use l of a preaction sprinkler system.
MOD 6136 Unit 1 _ Unit 2 x Common _ l This modification added CO2 fire protection for the Turbine Generator Exciter Bearings 11 and 12 and heat and smoke detectors
- in the Turbine Generator Exciter housing for fire detection.
This modification was installed to provide the Turbine Generator ,
Exciter with a fire protection system to minimize property damage I and unit downtime resulting from a postulated Turbine Generator Exciter fire. This modification affected several sections, figures, and tables of the UFSAR. The added components and systems are nonsafety-related and have no impact on safety-related systems. This modification did not affect any active or !
passive safety-related systems as described in the SAR. l l
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 6137 Unit 1 x Unit 2 3 Common 3 This modification installed an addition to the Radiation Area Access Control (RAAC) system. The RAAC system will provide permanent stations in close proximity to eight (8) radiological area access points. These will allow Health Physics (HP) personnel to set up access control stations. Each station provides corporate computer and telephone access and the capacity for future installation of an electronic dosimetry system. Each field station provides HP personnel the ability to view and print essential radiological exposure data and radiation work permit (RWP) data. This modification affected several sections and a figure of the UFSAR. Installing this system in the turbine, control, radwaste, and reactor buildings did not impact any active or passive safety-related system, as described in the SAR.
This modification enhanced the radiation monitoring capability of the plant personnel.
MOD 6138 Unit 1 E Unit 2 _ Common _
This modification installed test taps on the three Reactor Feed Pump suction lines. These taps were installed to provide injection points for hydrogen injection in the event a hydrogen water chemistry control program is implemented in the future.
This modification affected a figure in the UFSAR. This modification does not affect the operation or safety of the plant.
MOD 6139 Unit 1 g Unit 2 _ Common _
This modification removed snubbers from Unit 1 calculations and the plant which correspond to Unit 2 calculations which had snubber reduction reviews performed on them previously. ,
Modifications to the pipe supports and civil structure ensure that the design commitments for the items are maintained due to '
increased snubber reduction program loads. The specific systems involved are Main Steam (Outside Containment), Diesel Generator, '
Standby Liquid Control, RPV Head Vent, Main Steam Drain, and Fuel Pool Drain piping systems. This modification eliminated unnecessary mechanical snubbers from the piping systems and had i
no impact on the snubber testing program as addressed in Technical Specifications. This modification affected a figure in the UFSAR. This modification does not affect the safety of the plant and was done in accordance with a specific USNRC approved ASME code case.
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UNITS 1 AND 2 l DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 6144 Unit 1 _ Unit 2 3 Common _
This modification replaced underrated power fuses in various non-1E 125/250V DC equipment. The replacement fuses are properly rated and capable of meeting the design requirements for voltage rating and interrupting capability. This modification affected several figures of the UFSAR. The modification did not alter any assumption previously made in evaluating or mitigating the radiological consequences of an accident as described in the SAR. i The original function of the DC Power Distribution System is j maintained by the replacement with the properly rated fuses !
capable of meeting the design requirements for voltage rating and interrupting capability. This modification maintains the DC Power Distribution System performance necessary for reliable operation of equipment. This change does not reduce system redundancy or independence, or impose more severe testing requirements than previously evaluated in the SAR.
MOD 6145 Unit 1 _ Unit 2 E Common _
This modification altered the removable reactor cavity ladder to make it possible to install the ladder in a location where it will not interfere with the removal of the drywell head. This will allow the ladder to remain in place while the drywell head is being removed or installed. The reason for this proposed change is to reduce the refueling floor critical path and to eliminate the safety concern posed by not having a ladder available to access or exit the reactor cavity when the drywell ,
head is being lifted. This modification affected a figure of the UFSAR. This modification is an overall enhancement to maintenance. I MOD 6147 Unit 1 _ Unit 2 x Common _ ;
This modification installed a cross connection between the Residual Heat Removal (:RHR) and the Fire Protection systems to provide an alternate source of water to the drywell spray mode of ,
operation of the RHR system. This cross connection would be used in the event of a " severe beyond design basis" accident in which all low pressure Emergency Core Cooling System (ECCS) systems fail. At such a time, the cross connection would be made via a 6 ,
inch fire hose and ability to spray the drywell would be maintained. This modification affects several figures of the UFSAR. The RHR connection has the required isolation of a locked '
close gate valve along with a safety grade check valve and a cap. '
During all modes of operation as described in the SAR, the RHR/ Fire Protection cross connection will remain inactive. It would only be used for an accident beyond the design basis.
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l LIMERICK GENERATING STATION I UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 l
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MOD 6154 Unit 1 _ Unit 2 _ Common x This modification involved a complete renovation of a warehouse located outside of the Protected Area Boundary. The renovations included taking 24 feet off of the West end, upgrading the ;
exterior building shell and landscape, providing a new office l area within the building, changing the layout of the warehouse and receiving areas, and providing new warehouse racking and storage systems. This will allow enhanced cost and inventory I control. This modification affected several figures of the ;
UFSAR. The facilities and utilities being added are nonsafety-related structures located in nonsafety-related areas. !
l MOD 6155 Unit 1 x Unit 2 x Common x j l
This modification re-routed the drains from the Unit 1 and Unit 2 Condensate Storage Tank retaining dike area drains from the radwaste floor drain system to the radwaste laundry facility.
Both retaining dikes accumulate rain water during periods of rainfall. Due to the presence of high content of silica, total organic carbon (TOC), and other impurities, this water cannot be sufficiently processed through the radwaste systems to meet the condensate water quality requirements. This modification affected several sections and a figure of the UFSAR. All systems and components associated with this modification are not involved in mitigating the radiological consequences of an accident described in the SAR. This modification does not increase the potential for radioactive releases, nor does it introduce a new method for radioactive release, as the radwaste laundry facility effluent is monitored for radioactivity.
MOD 6161 Unit 1 x Unit 2 x Common _
This modification eliminated the nuisance low service water flow alarms from the glycol and vault cooling refrigeration machines which annunciate locally and in the Main Control Room. This modification will make an interim modification a permanent modification by removing the low service water flow Switches FSL-70-110A/B and Alarms FAL-70-110A/B for the glycol refrigeration machines and the vault cooling refrigeration machines and installs flange spool pieces in the service water lines. These machines are utilized in the Offgas Treatment System and the changes will not affect operation of this system. These alarms annunciate frequently during cooler months when service water is throttled below the 2 gpm setpoint due to system parameters.
This modification affected several figures of the UFSAR. This modification is an overall enhancement to operations.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 1
MOD 6163 Unit 1 _ Unit 2 x Common _
This modification deleted the automatic transfer capability of the Unit 2 offgas after condenser drainage from the main I condenser to clean radwaste during high conductivity and -
eliminated the Unit 2/1 difference. This is accomplished by ;
removing the interlock (electrical) between conductivity switch CITS-69-255 and control valve CV-69-256. This will prevent the backup of condensate in the aftercondenser due to a slow drainage from the aftercondenser to clean radwaste (CRW). This modification affected a figure of the UFSAR. An aftercondenser tube leak would not affect reactor safety. It would only affect ,
plant economics. The condensate demineralizers would remove the '
dissolved solids introduced into the condensate by an aftercondenser tube leak so that they would not enter the reactor and cause increased activity levels or corrosion problems. This modification does not affect the performance or operation of the condensate system, liquid radwaste system or any other safety or non-safety related systems or equipment. Draining of the aftercondenser condensate to the main condenser or to the CRW does not involve any accident previously evaluated in the SAR.
l MOD 6165 Unit 1 x Unit 2 x Common _ l This modification replaced the undervessel cabling and connector ;
hardware for the Local Power Range Monitors (LPRM's). The !
subpile room portion of cabling and associated connectors for l each of the 172 LPRM detectors was replaced with mineral l insulated (MI) stainless steel sheathed cable and quick disconnect connectors. This new type of cabling resists degradation from the harsh thermal and radiological environment in the subpile room that, as was evident in the previous installation, destroys organic based cabling. The new connectors are Loss of Coolant Accident (LOCA) environment and Class 1E qualified and meet all applicable criteria for LPRM's. This modification affected one section of the UFSAR. Replacement of the subpile room portion of the LPRM cabling with improved MI type cabling and quick disconnect connectors maintained the ability of the LPRM's to detect neutron flux excursions and enable a reactor trip, and improved the overall reliability of the system. This modification reduces the time necessary to connect / reconnect the cables during maintenance and reduces the amount of repairs necessary and therefore enhances efficiency and reduces occupational exposures.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 6167 Unit 1 _ Unit 2 _ Common x ;
This modification removed the trip capaoility of the Standby Gas Treatment (SGTS) heaters' manually reset- overtemperature switches. This will eliminate the operability problem caused by inaccessibility of the SGTS rooms following the Design Basis !
Accident (DBA). The overtemperature switches require manual resetting after tripping. This modification will make an interim modification permanent by installing permanent jumpers in the heater panels, removing wires to Annunciator Auxiliary Relays, abandoning the manual reset switches and relays in place inside the local panels. This modification affected a table of the UFSAR. These manual overtemperature cutout switches are redundant to existing automatic reset overtemperature cutout switches. This modification is an overall enhancement to operations.
MOD 6168 Unit 1 3 Unit ? E Common _
This modification installed a high point vent on the suction piping of each Unit 1 and Unit 2 Reactor Feed Pump (RFP) A, B, and C, downstream of the suction valve of each pump. There was no method available to vent entrapped air from RFP suction piping from the suction valves to the pump inlet. This modification affects a figure in the UFSAR. This modification is an overall enhancement to operations and maintenance.
MOD 6171 Unit 1 _ Unit 2 3 Common _
This modification relocated the process tap locations for the Instrument Air system low pressure alarm and indication to a e common process tap downstream of the Instrument Air Dryer Package on the air headers. This tap is also used for the local pressure l indicator added by this modification. The modification was )
implemented to address the plants' concern of the absence of a !
low pressure alarm and indication in the Main Control Room for Instrument Air Header pressure, downstream of the Instrument Air l Dryer Package. Lack of the information would preclude timely I intervention for corrective action during instrument air system perturbations. This modification affected a section and several figures of the UFSAR. This modification is an overall enhancement to operations.
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LIMERICK GENERATING STATION l UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 i
MOD 6183 Unit 1 _ Unit 2 x Common _
This modification increased the Secondary Containment blowout panel actuation setpoints to eliminate inadvertent actuations of the blowout panels. A Reactor Enclosure ventilation transient resulted in an overpressurization transient that caused a blowout panel to actuate and was reported in a License Event Report (LER). This event resulted in the plant being in an unanalyzed condition. Specifically, breaching the Secondary Containment would impact the performance of the Standby Gas Treatment System (SGTS) by not allowing proper drawdown of the Secondary ,
Containment and subsequently impacting its ability to filter all i of the radioactive releases. The actuation setpoints were increased for four blowout panels in the Reactor Enclosure. i These blowout panels are part of the Secondary Containment ,
boundary. This modification affected several figures of the UFSAR. The new blowout panel actuation setpoint (0.5 psid) is above the maximum pressure that can be developed by the Reactor Enclosure supply fans. Pressurizing Secondary Containment to 0.5 psid will not adversely affect the Reactor Enclosure or systems and components within the Reactor Enclosure and the Reactor Enclosure can structurally withstand this pressure differential.
Evaluations of the pressure-temperature transient analysis for the affected blowout panels was performed. Compartment peak :
pressures and temperatures determined in these evaluations are not greater than the design values given in the UFSAR. This modification is an overall enhancement to operations and safety.
MOD 6187 Unit 1 x Unit 2 x Common 3 J
This modification provided a new permanently installed 480-volt power feed to the Turbine Operating Floor from Site Perimeter i Substation No. 10 to support turbine generator outage work. The power feed will supply three new permanently installed disconnect switches. The disconnect switches will be 600 amp, non-fused, and will be located on the north wall of the Turbine Operating !
Floor. This modification affected a table in the UFSAR. This modification is an overall enhancement to maintenance.
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LIMERICK GENERATING STATION UNITS 1 AND-2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 6188 Unit 1 _ Unit 2 _ Common x This modification installed a second intake screen backwash air compressor with auxiliaries at the Perkiomen Pump Station. The second compressor was added to act as a backup to the existing compressor should it or its associated motor control center fail.
The second compressor also operates in tandem with the existing compressor to recharge the backwash receiver tank faster. The modification also increased receiver maximum pressure from 125 to 140 psig, for better screen cleaning. This modification affected a figure of the UFSAR. This modification does not involve any safety related equipment and is an overall enhancement to operations.
MOD 6189 Unit 1 x Unit 2 3 Common _
This modification changed low temperature trip instrumentation of the Turbine Enclosure Ventilation System (TEVS) to prevent spurious supply fan trips caused by cold air stratification.
Other existing instruments are used which insure a more representative measurement of air temperature. This modification affected a figure of the UF9AR. The function and operation of the TEVS is not different m_ om that described in the UFSAR. No other equipment functions are changed as a result of implementing this modification. This modification is an overall enhancement to operations.
MOD 6191 Unit 1 x Unit 2 x Common _
This modification installed a manually operated selector switch to disable the low pressure alarm for an out-of-service Steam Jet Air Ejector (SJAE) train. This modification was implemented to eliminate a Main Control Room alarm that was unnecessarily alarming during normal plant conditions. This modification affected a figure of the UFSAR. No malfunctions to equipment !
important to safety pertaining to the SJAE system are discussed in the UFSAR. This modification is an overall enhancement to operations.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 6207 Unit 1 x Unit 2 _ Common _
This modification replaced approximately 180 feet of flawed 3 inch Emergency Service Water (ESW) piping from the High Pressure Coolant Injection (HPCI) room unit coolers to the common return line of the 'B' loop of ESW. The replacement piping is austenitic stainless steel (25% Ni, 6% Mo) of the same diameter and thickness as the existing pipe and follows the same pipe routing utilizing all existing restraints. In addition, the modification installed a manual 3 inch valve which provides isolation capability for this section of ESW piping during future maintenance activities. A freeze' seal was necessary to implement this modification. This modification affected several figures of the UFSAR. The piping and valve associated with this modification improved the isolation capabilities and corrosion resistance of a portion of the ESW system. The only malfunction that would be applicable to the design basis of the ESW is a medium energy line break (MELB) which was considered in the original design. Although there is increased probability of j flooding in the service water pipe tunnel associated with the use of a freeze seal, the service water pipe tunnel has been ;
evaluated for MELB and its effects. All systems will continue to function during any accident as previously assumed and there will be no change to the onsite or offsite radiological effects above ,
those previously approved in the SAR. This modification is an overall enhancement to operations and maintenance.
MQD 6218 Unit 1 _ Unit 2 x Common _
This modification removed the reactor recirculation pump stator high temperature trip. This trip function has caused a spurious trip of the 1B recirculation pump in the past. This modification affected a figure of the UFSAR. The stator high' temperature-trip is provided for pump motor protection. A high temperature alarm ,
and temperature indication is available to the operator to enable ;
corrective actions to be taken without risking plant transients ;
due to a spurious recirculation pump trip caused by erroneous or defective temperature instrumentation. Additionally, ground fault current and phase overcurrent trips are provided to protect
, the pump, wiring and penetration assemblies. The only type of equipment malfunction that can occur is the loss af a reactor recirculation pump. The temperature switches affected by this modification are classified non-safety related. This modification is an overall enhancement to operations.
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I LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 l
MOD 6220 Unit 1 _ Unit 2 3 Common _
This modification relocated the condensate reject (Control Rod Drive (CRD) water supply) supply location, from the discharge j side of the new condensate demineralizers to the discharge side of the newly installed deep bed demineralizers. With the implementation of the deep bed demineralizers, a higher quality deaerated water will be supplied to the CRD mechanisms and the !
reactor vessel. This modification will not affect the operability of the condensate, deep bed demineralizer, or CRD systems. This modification affected a figure of the UFSAR. The new supply location will provide higher quality water to the CRD l l
system which may result in reduced CRD mechanism degradation.
This modification is an overall enhancement to operations and maintenance.
MOD 6221 Unit 1 x Unit 2 _ Common _
This modification installed a corrosion monitoring system for the Residual Heat Removal (RHR) Heat Exchangers (HXs). Analysis of ,
the results of eddy current testing performed after disassembly l and hydrolyzing of the Unit 1 'A' and 'B' RHR HXs during l refueling outage 1R04 showed indications of pitting on the tube inside diameters. This corrosion monitoring system will expose specimens of 304 L tubing material to similar water chemistry and temperatures that will be experienced by the RHR HXs. This j capability will provide information which is essential to support the continued operability of the Unit 1 'A' and 'B' RHR HXs.
This modification affected several sections, figures and tables of the UFSAR. Although this modification increased the RHRSW inlet temperature to the '1B' RHR HX by 1.5 degrees F, this ;
increase insignificantly impacts overall heat removal or equipment performance or reliability for all normal modes of plant operation. Furthermore, this modification did not result in any change in the inlet Residual Heat Removal Service Water (RHRSW) temperature to the '1B' RHR HX during Loss of Coolant Accident (LOCA) and/or Loss of Offsite Power (LOOP) events, and it will have zero impact on RHRSW inlet temperature to the 1A RHR 2
HX. Although the corrosion monitoring system will bypass a small fraction of normal RHRSW flow for the RHR HX "B", this bypass ;
will not reduce RHRSW flow to the Spray Pond; only system volume '
will change by a minute amount. This change will not impact any operating characteristics of the system. This modification is an overall enhancement to safety.
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I LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 MOD 6226 Unit 1 _ Unit 2 x Common _
This modification added fire detection to three plant locations where Thermo-Lag 330 fire barriers are installed. This j modification was necessary to reduce continuous fire watch )
requirements as mandated by Technical Specifications for areas where deficient Thermo-Lag 330 exists and fire detection does not. This modification affected several sections, figures, and tables of the UFSAR. The existing fire detection logic will not be affected. This modification does not change either the design function or the modes of operation of the fire detection system.
This modification was installed to reduce operating costs.
NCR 91-164. Rev. 4 Unit 1 x Unit 2 x Common _
This NCR identified a discrepancy in the Main Steam Isolation Valve Leakage Control System (MSIV LCS) inboard blower capacity i
., test readings conducted during performance of a surveillance test procedure. The blower capacity readings did not meet Technical Specification Section 4.6.1.4 requirements of 15" water gage at 100 SCFM for the inboard blower. The flow readings were determined by the calibration curve of a Flow Element. This Flow -
Element is installed during a test and subsequently removed after i the performance of Surveillance Test ST-1-040-400-2, "MSIV LCS ;
Functional Test". Alternate flow measurement readings were taken via a rotating vane anemometer. These readings correspond closely to the typical blower performance curve. Additionally, ;
during inspection and examination of the blower, station '
maintenance personnel did not detect any obvious internal or external problems. The disposition of this NCR is use-as-is and allows the use of alternate flow measuring devices to verify operability of the blower. A change to the procedure is required to address use of alternate flow measuring devices.
NCR 91-267. Rev. 1 Unit 1 3 Unit 2 _ Common _
This NCR identified two motor operated valves that have not had the required Environmental Qualification (EQ) preventive maintenance (PM) work performed since 1987. These two valves are
- associated with the steam condensing mode of the Residual Heat i
Removal (RHR) system. This mode of RHR was previously eliminated by a modification but the valves were not removed and the EQ requirements are still applicable. This NCR has been dispositioned use-as-is. The naintenance requirements described i in the EQ report are to ensure post accident operability. l Because these valves do not have any post accident operability l requirements, this maintenance activity is not required. It is l concluded that these MOVs do not require any EQ maintenance l actions. i i
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 NCR 92-086. Rev. 1 Unit 1 3 Unit 2 _ Common _ )
The Residual Heat Removal (RHR) system outboard shutdown cooling return valve, HV-051-1F015A, was disassembled for failure of a local leak rate test. The retaining ring was found to be missing from the stem. The retaining ring could not be found in the valve body or the accessible regions of the adjacent piping. The i retaining ring was considered to be in either the 'A' loop of the '
RHR system, or the reactor vessel. The NCR was dispositioned to use-as-is since there was no operability concern with the 'A' l loop of the RHR system, and the existence of this lost part and !
the cumulative effects of previous lost parts was determined not j to compromise safe reactor operation. Since the SAR assumes that lost parts do not exist, the existence of this lost part is different than what is described in the SAR. The lost retaining ring would not adversely affect the chemical or metallurgical l environment, would not block control rod operation, would not -l cause fuel bundle flow blockage, would not adversely affect RHR .
system operability, and would not cause damage to other reactor '
internal components.
NCR 92-109. Rev. 3 Unit 1 x Unit 2 _ Common _
Examination of the Main Steam Line Control Valves in the area i above the seat drain lines, revealed three fittings with wall thickness below minimum allowances. This reduction is believed to be the result of two phase (water and steam) flow. This NCR i was dispositioned to replace the carbon steel piping and fittings l with those upgraded to a 1 1/4% Chromium - 1/2% Molybdemum (1 1/4 ;
Cr - 1/2 Mo) alloy material to significantly reduce the rate of I wall loss due to erosion / corrosion damage. This material substitution made changes to the piping material as shown in the drawings of the SAR. The upgraded replacement resulted in a decreased erosion rate, and therefore, did not affect the integrity of the pipe. The configuration of the replacement pipe i did not change, and the piping material was upgraded without changing the mechanical properties of the material.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 '
NCR 92-196. Rev. O Unit 1 _ Unit 2 x Common __ ,
The General Electric (GE) Company instruction manual for the reactor pressure vessel N7 nozzle specifies a modified small tongue and groove (T&G) closure flange for the nozzle. The >
special height requirements were specified to prevent crushing the flexitallic gasket contained in the flange beyond its design compression. Plant design drawings showed a standard small T&G flange, and the as built condition is a standard small T&G flange. Also, a discrepancy was identified between the UFSAR and the installed plant condition. The UFSAR stated that all reactor pressure vessel top head nozzles were provided with large groove :
facings. This was in conflict with the GE design drawings, the plant design drawings, and the installed plant condition. The interim disposition of the NCR was use-as-is until final rework of the flange could be completed. A slight over compression of the gasket would not adversely affect the sealing characteristics '
of the gasket, since the gasket was contained inside the groove of the flange. The original flanged joints were installed in accordance with the applicable codes for flange fittings.
Successful hydrotests performed after installation verified that the flange was not leaking, and would perform its design function. The final disposition of the NCR was to rework the piping flanges to conform to the GE requirements. Installation was per the GE design and conformed to ANSI Standard B16.5-as required by ASME B&PV Code Section III for pipe flanges and flange fittings. A UFSAR change was initiated to properly reflect plant configuration.
NCR 92-207. Rev. 0,1 2 2 Unit 1 x Unit 2 _ Common _
Eddy current testing of the RHR Heat Exchanger detected pitting of the inside of some of the tubes of between 20% and 99%. The cause of the pitting was determined to be under deposit corrosion caused by an oxygen concentration cell created by a manganese i rich deposit aggravated by the presence of biological activity and the use of an oxidizing biocide. The short term solution was to plug all tubes with 80% or greater through wall defects. '
Additionally an on line corrosion monitoring system has been installed (MOD 6221 - Page 14 of this report) that will provide data to support continued operability of the heat exchanger. The heat exchanger's tubes will be replaced during the next refueling outage, scheduled for January 1994.
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j LIMER.ICK GENERATING STATION ,
UNITS-1 AND 2 )
DOCKET NOS. 50-352 AND 50-353 '
LICENSE NOS. NPF-39 AND NPF-85 1 i
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NCR 92-230. Rev. 0 Unit 1 x Unit 2 3 Common _ !
The pressure control valves, PCV-056-1F035 and PCV-056-2F035, which are installed in the High Pressure Coolant Injection (HPCI) systems for Units 1 and 2, respectively, contain certain parts which were not manufactured to ASME B&PV Code,Section III, Class 2 requirements. The valve bonnet flange and bolts were not originally classified as part of the ASME Code,Section III pressure boundary, but were subsequently reclassified by the manufacturer. Therefore, the valves were not in agreement with the information in the SAR. The disposition of this NCR is use- l as-is. Although the valve bonnet flange and bolt materials are :
not ASME Code,Section III, a Certified Material Test Report was !
provided for the flange and a Certificate of Compliance was '
provided for the bolts. The ultimate tensile strength of the bonnet flange material is comparable to that for the typical ASME Code,Section III flange material. The design for the load >
carrying capacity of the bolts includes a safety factor of approximately 10. The disposition of this NCR is for a short term basis, i.e., approximately eight months, at which time the !
permanent repairs will take place. The Unit 2 valve has already e 4
been repaired.
NCR 92-233. Rev. 0 Unit 1 _ Unit 2 3 Common _
Flow verification testing of the Emergency Service Water (ESW) system flowrate to the Unit 2 Residual Heat Removal (RHR) pump ,
room unit cooler 2GV210 revealed a reduction in flow to below the l original design value of 115 gpm specified in the UFSAR. The flow was measured as 90 gpm. The disposition of the NCR was_use-as-is based on an analysis of heat transfer test results which showed that the unit cooler was capable of removing the required design heat loads given a minimum indicated ESW flowrate of 84 gpm or greater. This would maintain the room temperature below 125 degrees F, which is above the UFSAR speciiied limit of 115 degrees F, but is below the qualified temperature of the equipment in the room. Operability of the '2C' RHR pump is supported by this analysis. The RHR pump room temperature is not an accident initiator, and the increased room temperature would not prevent the safety-related equipment located in the room from j performing its safety-related design function. This temporary l design change to the RHR pump room temperature does not involve i any physical changes to equipment and does not introduce any new failure modes to existing equipment. This is a temporary change to the facility until flow balancing of the 'A' loop of the ESW system can be performed.
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LIMERICK GENERATING STATION-UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 NCR 92-234. Rev. 0 Unit 1 _ Unit 2 3 Common _
Flow verification testing of the Emergency Service Water (ESW) system flowrate to the Unit 2 Residual Heat Removal (RHR) pump room unit cooler 2CV210 revealed a reduction in flow to below the original design value of 115 gpm specified in the UFSAR. The flow was measured as 95 gpm. The disposition of the NCR was use-as-is based on an analysis of heat transfer test results which showed that the unit cooler was capable of removing the required design heat loads given a minimum indicated ESW flowrate of 84 gpm or greater. This would maintain the room temperature below 125 degrees F, which is above the UFSAR specified limit of 115 degrees F, but is below the qualified temperature of the equipment in the room. Operability of the '2C' RHR pump is supported by this analysis. The RHR pump room temperature is not an accident initiator, and the increased room temperature would not prevent the safety-related equipment located in the room from performing its safety-related design function. This temporary design change to the RHR pump room temperature does not involve any physical changes to equipment and does not introduce any new failure modes to existing equipment. This is a temporary change to the facility until flow balancing of the 'A' loop of the ESW system can be performed.
NCR 92-237, Rev. 0 Unit 1 _ Unit 2 x Common _
An inspection revealed the presence of a hole in a pipe fitting on the downstream of the #4 feedwater heater on the "C" Main Steam Line. This condition is similar to leaks recently detected on the #3A feedwater heater start up vent line. In each case, this condition was attributed to an erosion / corrosion mechanism of material degradation. The interim disposition is use-as-is with a leak repair clamp attached to the leaking fitting. The effects of attachment of the leak repair clamp on the subject fitting has been evaluated and the weight of the leak repair clamp would not affect the structural integrity of the piping cn:
associated supports. This fitting was replaced at the next refuel outage. The replacement utilized 1 1/4 CR, 1/2 MO alloy steel. This material upgrade would result in this fitting being much more resistant to erosion / corrosion damage than the original carbon steel.
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LIMERICK GENERATING STATION UNITS 1-AND 2 DOCKET NOS. 50-352 AND 50-353 l LICENSE NOS. NPF-39 AND NPF-85 NCR 92-248. Rev. 0 Unit 1 x Unit 2 x Common _
A Review of the In-Service-Testing (IST) program revealed the Emergency Service Water (ESW) supply check valves to the Emergency Diesel Generators (EDG) were incorrectly classified as ASME components that required both forward and reverse flow quantitative testing on a quarterly basis. The motor operators on these stop-check valves are normally in the open position and have a safety function to remain open. Because the motor operator does not change position to perform its safety function, it is considered a passive function and can be classified as a ASME component that does not require quarterly quantitative reverse flow testing. The inttrim and-final disposition of this NCR is administrative only in tnat changes to documentation is all that is required.
NCR 92-254. Rev. 1 Unit 1 _ Unit 2 x Common _
The RCIC pump room unit coolers are required by the UFSAR to maintain the RCIC pump room at a maximum temperature of 120 degrees F when the RCIC pump is in operation. Heat transfer and ,
flow verification test results and associated engineering analysis for the Unit 2 2BV208 Reactor Core Isolation Cooling (RCIC) pump room unit cooler indicated that the maximum post-accident RCIC pump room temperature could be as high as 134 degrees F, provided the 2BV208 unit cooler was the only cooler ,
operable. The disposition of this NCR was use-as-is until-cleaning and retesting of the unit cooler could be accomplished.
This disposition was based on an engineering evaluation of the qualification of the equipment in the RCIC pump room which verified that the equipment is qualified and would remain operable at temperatures greater than 140 degrees F. The RCIC .'
pump room temperature is not an accident initiator, and the increased room temperature would not prevent the safety-related equipment located in the room from performing its safety-related design function. This temporary design change to the RCIC pump room temperature does not involve any physical changes to equipment and does not introduce any new failure modes to existing equipment.
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LIMERICK GENERATING STATION .
UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 NCR 92-259. Rev. 0 Unit 1 _ Unit 2 3 Common _
Flow verification testing of the Emergency Service Water (ESW) system flowrate to the Unit 2 Residual Heat Removal (RHR) pump room unit cooler 2AV210 revealed a reduction in flow to below the original design value of 115 gpm specified in the UFSAR. The flow measured was 90 gpm. The disposition of the NCR was use-as-is based on an analysis of heat transfer test results which showed that the unit cooler was capable of removing the required design heat loads given a minimum indicated ESW flowrate of 84 gpm or greater. This would maintain the room temperature below 125 degrees F, which is above the UFSAR specified limit of 115 degrees F, but is below the qualified temperature of the equipment in the room. Operability of '2A' RHR pump is supported by the 2EV210 unit cooler. The RHR pump room temperature is not an accident initiator and the increased room temperature would not prevent the safety-related equipment located in the room from ,
performing its safety-related design function. This temporary 1 design change to the RHR pump room temperature does not involve )
any physical changes to equipment and does not introduce any new failure modes to existing equipment. This is a temporary change to the facility until flow balancing of the 'A' loop of the ESW system can be performed.
Unit 1 _ Unit 2 3 Common _
NCR 92-270. Rev. O. 1 A faulty air start solenoid was discovered on the Unit 2 D22 Emergency Diesel Generator (EDG) air start system. Replacement parts were not available. This NCR evaluated the operability of the EDG with one air start subsystem in service and the other removed from service until repairs can be made. This NCR was dispositioned interim use-as-is with a final disposition of rework. The single air start subsystem is capable of independently starting the EDG within the required time.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 NCR 92-306. Rev. O Unit 1 x Unit 2 3 Common _
During preparation of the Limerick Generating Station 13KV Design Baseline Document, a discrepancy was discovered between the UFSAR and the startup checkoff lists. The checkoff lists 1S91.9.A and 2S91.9.A for Units 1 and 2, respectively, are related to the lineup of the 13KV unit aux buses to the offsite sources. The UFSAR indicated that both unit aux buses for Unit 1 are supplied by power from one offsite source and both unit aux buses for Unit 2 are supplied by power from the other offsite source. The checkoff lists indicate that one 13KV unit aux bus from each unit is supplied power from each offsite source. Actual plant configuration is in accordance with the checkoff lists. The disposition of the NCR was use-as-is and to change the UFSAR to-correctly identify the alignment of the 13kV unit aux buses when they are being supplied power from the offsite sources.
Supplying power to the two 13kV unit aux buses for a particular unit from different offsite sources provides more stability than supplying power to both buses from the same source. If both buses are aligned to transfer to different offsite sources, a trip of one of the Main Generators would have less impact on-voltage regulation if the buses transferred to opposite sources rather than having both buses transfer to the same offsite source. Aligning one 13kV unit aux bus per unit to each offsite j source does not render the electrical distribution system ;
inoperable because the less than 0.5% change in loading is '
bounded by the existing analysis. l l
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. s LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICFNSE NOS. NPF-39 AND NPF-85 NCR 92-353. Rev. 0 Unit 1 x Unit 2 3 Common _
During an engineering review of the heat-up calculations for the High Pressure Coolant Injection (HPCI) system pump room, it was noted that the HPCI turbine steam leakage assumed in the design calculation was different than the steam leakage noted on the HPCI process flow diagram. The leakage assumed in the calculation was 345 lbm/hr while the process flow diagram indicated 500 lbm/hr. Operability of the HPCI system was not impacted since a sensitivity study was performed which indicated that even with a HPCI total steam leakage flow of 500 lbm/hr, the HPCI room temperature would not exceed the equipment qualification limit of the safety-related equipment located in the room. A review of the discrepancy concluded that the actual HPCI total turbine steam leakage flow should be 395 lbm/hr. The disposition of the NCR was " document change only." The change involves only a design parameter change which requires a change to one design drawing and a design calculation. No hardware changes were required. The concern with the steam leakage flow rate is related to a failure of the HPCI barometric condenser which would result in heat input to the HPCI pump room. This change involves reducing the steam leakage flow rate from 500 lbm/hr to 395 lbm/hr which reduces the potential heat input to the HPCI pump room in the event of a failure of the HPCI barometric condenser. The slight increase in the flow rate from the existing calculation's value of 345 lbm/hr to 395 lbm/hr is bounded by the sensitivity study described previously.
NCR 92-355. Rev. O Unit 1 x Unit 2 3 Common _
i During a system walkdown, it was discovered that certain normal waste drains were connected into the storm drain system on elevation 313' of the reactor enclosure for both Unit 1 and Unit 2 in the reactor enclosure HVAC air supply fan areas. .These storm drain lines are then routed to either the holding pond, during normal operation or the Possum Hollow Creek during periods of heavy rainfall. Due to the contamination possibility from two known sources this situation could result in an unmonitored release if the drains are directed to the Possum Hollow Creek.
This NCR has been dispositioned interim use-as-is with a modification required to resolve the concern. Administrative controls have been established to avoid an unmonitored release of contaminated liquid. Maintenance activities have been established to minimize the potential for contamination reaching the normal waste drains in the fan areas.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 NCR 93-036. Rev. 1 Unit 1 _ Unit 2 3 Common _
Refueling floor inflatable seal number 4 was found to be leaking during surveillance testing. This NCR was initiated to evaluate the acceptability of continuing refueling activities with inflatable seal number 4 leaking. The final disposition of the NCR was to rework the seal prior to the next refueling outage.
The use of operable inflatable seal number 3.as the single water sealing mechanism was determined to be acceptable until seal number 4 is reworked, provided the seal drain is operable. The i redundant ring seals contained in the annular space between the reactor cavity and the refueling ring prevent water f rom entering the reactor building when the reactor cavity is flooded. Seal number 3 is covered by and rests on removable plates above seal number 4. This configuration prevents vertical seal displacement even when the seals are fully deflated. Any leakage from the seals is captured by a trough which drains to the equipment drain collection tank in the radwaste building. Fuel would not be uncovered due to seal leakage because the transfer canal is above the top of the fuel stored in the spent fuel pool and the reactor pressure vessel would remain flooded. Adequate fuel pool cooling is available via the Emergency Service Water system. Adequate core cooling is available because existing Emergency Core Cooling Systems are unaffected by the leakage.
NCR 93-082. Rev. 3 Unit 1 3 Unit 2 _ Common _
A small pinhole leak on a weld upstream of the inlet valve to the 1EV210 RHR pump room unit cooler was identified during a walkdown of the 'A' loop of the Emergency Service Water (ESW) system.
Also, sediment was found just upstream of another weld which was indicative of a second pinhole leak. Revisions 0 and 1 of the NCR were dispositioned use-as-is. An engineering evaluation of the subject piping with small through wall holes (less than 1/8 inch) was performed which confirmed the structural integrity of 1
, the piping was maintained. The piping met the seismic category I ,
requirements and the piping stresses remained within ASME Code l allowable limits. The hydraulic effect on the design flow rates '
to the Unit 1 RHR compartment unit coolers, the RHR pump seal, and RHR pump motor oil coolers was insignificant when compared to the actual flow rates since the holes were small and the leakage rates were negligible. The leakage was monitored once a day to confirm that the leakage was not increasing. Revision 2 of the NCR was issued to address the degraded condition of the. leaking ESW system piping upstream of the unit cooler. Revision 2 of the NCR was dispositioned as rework as soon as possible. Revision 3 of the NCR was a documentation change only to correct erroneous indications in the ASME Section XI fields of the NCR. The t
subject piping was replaced in kind. !
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l LIMERICK GENERATING STATION i UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 ;
LICENSE NOS. NPF-39 AND NPF-85 NCR 93-092. Rev. O Unit 1 3 Unit 2 _ Common _
During the performance of troubleshooting, the Emergency Service Water (ESW) system flowrate to the Unit 1 Residual Heat Removal (RHR) pump room unit cooler 1HV210 was found to be 109 gpm which is below the design value of 115 gpm as specified in the UFSAR.
This condition occurred during a special ESW system alignment which was required to support the performance of a special procedure during maintenance activities associated with the RHR Service Water system. This special alignment is more restrictive than the normal ESW system lineup. The disposition of the NCR was use-as-is based on an assessment of the cooler's ability to remove design heat load at the reduced flow rate while in the more restrictive system alignment. This would maintain the room temperature below the qualified temperature of the equipment in the room. Operability of '1D' RHR pump is supported by the 1DV210 unit cooler. The RHR pump room temperature is not an accident initiator and the increased room temperature would not prevent the safety-related equipment located in the room from performing its safety-related design function. This was a temporary change to the facility during the performance of the special procedure.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 NCR 2_3-119. Rev. 1 Unit 1 _ Unit 2 _ Common 3 A discrepancy between the measured delay time and the delay time specified in the UFSAR for the toxic gas analyzers was >
identified. The response time specified in the UFSAR was 40 seconds. The response time provided by the vendor and ,
demonstrated during troubleshooting was nominally 90 seconds.
This discrepancy brought into question whether adequate time i existed for the Main Control Room (MCR) operators to respond to a high toxic gas alarm with the MCR ventilation system in the normal mode of operation. A calculation was performed to quantify the largest analyzer delay time that would be permitted and still allow 120 seconds as recommended by Regulatory Guide 1.78 for MCR operators to don self contained breathing ,
apparatuses (SCBAs). The NCR was dispositioned as repair to reduce the purge time of the analyzers. During the purge period of the sample cycle, the air that has just been sampled is exited from the analyzer. The purge cycle time contributes to increasing the response time of the system as there is no analysis of air while the purging takes place. The purge time ,
was reduced in accordance with advice from the manufacturer and is of adequate duration to insure a true representative sample of the next air entering the measurement chamber. The UFSAR was revised to reflect the amended analyzer delay time. The toxic gas detection system is a stand alone system with no physical boundary with safety-related equipment, and therefore, does not affect the ability of safety-related equipment to perform its safety function. The consequences to the operators of a toxic gas release is not increased since the new design basis calculation allows the operators 120 seconds to don the SCBAs.
NCR 93-123. Rev. 0 Unit 1 _ Unit 2 3 Common _
The results of heat transfer testing, performed to satisfy the i requirements of Generic Letter 89-13, indicated that the heat transfer capacity of the 2BV209 HPCI pump room unit cooler would result in a HPCI pump room temperature of 121 degrees F which is i higher than the UFSAR specified maximum temperature of 120 degrees F. The interim disposition of the NCR was use-as-is until the HPCI pump room unit coolers could be cleaned and retested. An engineering evaluation determined that the safety-related equipment in the HPCI pump room has been qualified to and will remain functional at temperatures higher than 140 degrees F.
The temporary change to the HPCI pump room maximum temperatures would not prevent the safety-related systems located in the room i from performing their safety-related design function. This temporary change did not involve any physical changes to the plant, did not involve any new accident initiators, and did not create any new functional interfaces with other systems.
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LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 NCR 23 '_37. Rev. 1 Unit 1 _ Unit 2 3 Common _
- While removing main steam line plugs during the Unit 2 second ,
refueling outage, two parts of the main steam line plug '
installation tool, a steel pin, and an aluminum spacer plate with two stainless steel dowel pins, were lost. They were presumed to be inside of the Unit 2 reactor vessel. An evaluation of this condition determined that these parts would not compromise safe reactor operations provided the following stipulation was met.
During reactor startup, the reactor must be maintained at a temperature greater than 450 degrees F and a power level less than 6% for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. This would ensure that the aluminum material disintegrates and would eliminate the possibility of causing fuel damage. An evaluation concluded that the lost objects would not interfere with control rod motion or fuel bundle coolant flow, would not create-a chemical or corrosion conce rn, and would not create the potential for damage to the reactor internal components.
NCR 93-203. Rev. O Unit 1 _ Unit 2 3 Common _
By letter dated March 11, 1991, regarding Qualification Fuel Bundles (QFBs) in operating cycle 2 of Unit 2, Philadelphia Electric Company (PECo) committed to having margin between the QFBs and the limiting fuel bundles in the core during steady.
state operation. During the initial startup of Unit 2 for operating cycle 3, the QFBs became the limiting Maximum Fraction of Limiting Power Density (MFLPD) fuel bundles. The disposition of the NCR was to continue operation with the QFBs leading the MFLPD until a control rod pattern adjustment could be performed.
The QFBs are designed to be physically, thermal-hydraulically, and neutronically compatible with the non-0QFB fuel in the reactor core. The thermal limits for the QFBs are determined by NRC approved methods the results of which demonstrate that the QFBs are conservatively bounded by the General Electric (GE) GE9 fuel design. The commitment to keep the QFBs from leading the core in thermal limits is an extra level of conservatism and is not required to insure safety.
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LIMERICK GENERATING STATION UNITS 1 AND 2' DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 l NCR 93-204. Rev. 0.1 Unit 1 _ Unit 2 _ Common x !
The minimum Emergency Service Water (ESW) and Residual Heat Removal Service Water (RHRSW) flow rates, using flow balance data, were identified to be less than the flows assumed in the calculations for the spray pond heat rejection rate. At the lower flow rates the spray nozzles may reduce the spray efficiency and lower the heat rejection from the spray pond.
This NCR was dispositioned to use-as-is based on a revision to the calculation using realistic meteorology and spray droplet i test data. The conclusion was that the heat rejection rate of the spray networks is acceptable at the lower flow rates.
i NCR 93-214. Rev. 0.1.2.3 Unit 1 x Unit 2 _ Common _
A replacement pressure relief valve was installed on the air starting system for two Unit 1 Emergency Diesel Generators (EDGs). The replacement valve is identified in NRC Information Notice No. 90-18 as being not suitable for use in a seismic I >
application. This NCR was dispositioned use-as-is since each EDG has a redundant air starting subsystem that is unaffected by the potential failure of the suspect relief valve. The piping was also reanalyzed and the section of pipe containing the compressor ;
discharge relief valves was downgraded to a non-seismic class.
The air starting subsystems are considered operable with the ;
suspect relief valves. l l
NCR 93-224. Rev. 0.1 Unit 1 x Unit 2 _ Common _
A small pinhole leak on the Emergency Service Water (ESW) outlet of the piping 1HV210 RHR pump room unit cooler was identified during a PT/UT exam. Revision 0 of the NCR was dispositioned '
interim use-as-is. An engineering evaluation of the subject piping with small through wall holes (less than 1/8 inch) was j ,
performed which confirmed the structural integrity of the piping 1 was maintained. The piping met the seismic category I l requirements and the piping stresses remained within ASME Code {
allowables. The hydraulic effect on the design flow rates to the Unit 1 RHR compartment unit coolers, the RHR pump seal, and RHR pump motor oil coolers was insignificant when compared to the actual flow rates since the holes were small and the leakage rates were negligible. Revision 1 of the NCR was issued to provide rework instructions which called for brazing the unit cooler tubing.
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LIMERICK GENERATING STATION ;
UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 l NCR 93-252, Rev. O Unit 1 3 Unit 2 3 Common _
Upon review of safeguard battery Surveillance Tests (STS), it was ,
discovered that the requirement to torque each battery inter-cell i and terminal connection was removed from the STS in 1988. The requirement was removed due to concerns about battery post deformation. Upon further review of the basis for Technical Specifications (TS) Surveillance Requirement (SR) 4.8.2.1.C.2 it became unclear as to whether the connections had to be torqued to the manufacturer's recommended torque value every 18 months or if the connections had to be verified tight as specified in TS SR 4.8.1.2.C.2 and as recommended in IEEE 450-1975. Based on a review of the license commitments, the TS SR requires a tightness check, not a torque check. This NCR was dispositioned use-as-is.
The batteries were declared operable following tightness verifications.
NCR 93-262, Rev. 0 Unit 1 _ Unit 2 3 Common _
By letter dated March 11, 1991, regarding Qualification Fuel Bundles (QFBs) in operating cycle 2 of Unit 2, Philadelphia Electric Company (PECo) committed to having margin between the QFBs and the limiting fuel bundles in the core during steady state operation. During the initial startup of the Unit 2 for operating cycle 3, the QFBs became the limiting Maximum Fraction of Limiting Power Density (MFLPD) fuel bundles. The disposition of the NCR was to continue operation with the QFBs leading the MFLPD until a revision to the limits in the process computer can be made. The QFB limits are bounded by the GE9 reload bundle limits. The process computer limits are based on the GE9 bundle limits. Once the actual limits for the QFB are used in the '
calculation of'the actual limiting bundles, the QFB's would no longer be the limiting bundles. The QFBs are designed to be physically, thermal-hydraulically, and neutronically compatible with the non-00FB fuel in the reactor core. The thermal limits for the QFBs are determined by NRC approved methods the results .
of which demonstrate that the QFBs are conservatively bounded by the General Electric (GE) GG9 fuel design. The commitment to keep the QFBs from leading the core in thermal limits is an extra level of conservatism and is not required to insure safety.
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~ 4 o LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 PROCEDURE SP-127. Rev. O Unit 1 _ Unit 2 x Common _
This procedure titled, " Tie-in of Temporary Compressors to Compressed Air System," is to connect temporary compressors and provide the proper configuration of the Compressed Air System to support removal of Instrument & Service Air Compressors for service and maintain normal A/B Instrument Air and Service Air header pressure. Use of this procedure does change the configuration of the Instrument Air System as described in the UFSAR. The procedure was implemented during a Unit 2 refueling outage in support of service water, Instrument Air Compressor cooling water, and air compressor maintenance. The instrument Air System is non-safety related. Inadvertent loss of the air system will not adversely affect the ability of safety related equipment to perform their design function. The temporary compressors met the UFSAR specifications of the permanently installed compressors.
PROCEDURE SP-S-080. Rev. O Unit 1 _ Unit 2 x Common _
This procedure titled, " Unit 2 Drywell Chilled Water to Reactor Water Cleanup (RWCU) for Decay Heat Removal," is to provide chilled water to the RWCU non-regen heat exchanger to support RWCU as an alternate decay heat removal method. The backfeeding of Drywell Chilled Water to cool the Reactor Enclosure Cooling pumps is not addressed in the SAR. Additionally, normally interlocked Primary Containment Isolation Valves (PCIV's) will be de-energized and manually positioned to provide the required flowpath. This procedure will only be utilized during plant shutdown periods. The Loss of Shutdown Cooling Accident analysis as described in the UFSAR is not impacted since this procedure will support a reliable alternate heat removal mechanism and the Emergency Core Cooling System (ECCS) equipment is not impacted.
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a d e LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 LICENSE NOS. NPF-39 AND NPF-85 l
PROCEDURE SP-HF-010. Rev. 0 Unit 1 x Unit 2 _ Common _ j This special test procedure titled, "EHC Pressure Control System Stability Test for Power Rerate Capability Determination Unit 1,"
is to simulate EHC Pressure Control System conditions equivalent to expected steady state operation after Power Rerate. This test will facilitate data collection to substantiate design data and calculations that the EHC system can adequately control Reactor Pressure with the #4 Control Valve (CV) modulating at a nominal 45% open position. The data will be evaluated and incorporated into the Power Rerate Project at a later date. Reactor pressure will be systematically lowered such that the #4 CV will modulate at a nominal 45% open verses the current nominal value of 30%.
During the test, the reactor pressure will be reduced to 990 psia, which will allow control valve volumetric flow to increase 3% above rated with the reactor power not exceeding rated. This test is not described in the UFSAR. Equipment is to be operated within design parameters. No new component / system interaction ,
that could lead to an accident is created. No new challenge to ,
equipment is involved with this collection of data. No equipment that is assumed to fail in an accident is affected. All equipment will remain within design parameters. The consequences of all the transients will not cause the Minimum Critical Power Ratio (MCPR) safety limit to be exceeded. However, if at any :
time during the test, divergent pressure oscillations accur, the reactor power will be reduced or reactor pressure increaaed to ,
respective levels previously achieved and at which the unit has !
been proven to be stable. The margin of safety to the MCPR t safety limit will be maintained. The test requires the control !
valve position to go beyond the present experience range of the control system to simulate power rerate conditions. However, Philadelphia Electric Company and General Electric Company have completed a study of the unit's design and operational data and ,
concluded that the turbine control system will be adequate for '
operation at rerated conditions. Plant safety will not be affected by the test and the test will not involve an unreviewed safety question.
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