IR 05000352/1988017

Insp Rept 50-352/88-17 on 880706-0814.No Violations Noted. Major Areas Inspected:Plant Tours,Maint & Surveillance Observations & Review of Periodic Repts,Lers & Operational Events

ML20153C958
Person / Time
Site:	Limerick
Issue date:	08/24/1988
From:	Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20153C955	List: IR 05000352/1988017 ML20153C953
References
50-352-88-17, IEB-88-005, IEB-88-059, IEB-88-5, IEB-88-59, NUDOCS 8809010379
Download: ML20153C958 (48)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N Docket No. 50-352 License No. NPF-39 ' Licensee: -Philadelphia Electric Company 2301 Market Street Philadelphia, Pa 19101 Facility Name: Limerick Generating Station, Unit 1 Inspection Period: July 6 - August 14, 1988 Inspectors: T. J. Kenny, Senior Resident Inspector L. L. Scholl, Resident Inspector J. H. Williams, Project Engineer Reviewed by: [ P 28

  . Williams, Proje n neer Da'te /

Approved by: u[ JtC s Linville7 Chief, Pr a cts Section 2A P/J [[ Ddte ' Summary: .R ine daytime (122 hours) nd backshift/ holiday (16 hours) inspections of Unit 1 by the resident inspectors consisting of (a) plant tours, (b) observations of maintenance and surveillance, (c) review of LERs and periodic reports, (d) review of operational events and (e) system walkdown During this inspection period the licensee:

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Operated the plant at 85 to 90% power while monitoring the previously identified fuel leakag Submitted several LERs (section 6.0), the monthly operating report (section 5.0), safeguards event report 88-S04, and a special report on the seismic monitoring system operabilit Performed minimum flow tests on 'B' loop core spray pump Investigated an inadvertent single rod scram (section 2.3).

- Briefed NRC Region I staff on the fuel leak (section 8.0).

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i - 3  : i l DETAILS 1.0 Persons Contacted Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspection activit .0 Operational Safety Verification (71707, 70709, 71710 and 71881) 2.1 Documents Reviewed

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Selected Operators Logs

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Shift Superintendent's Log

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Radioactive Waste Release Permits-(liquid and_ gaseous) Selected Radiation Work Permits (RWP)

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Selected Chemistry Logs

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Selected Tagouts

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Health Physics Log 2.2 The inspector conducted routine entries into the protected areas of the plant, including the control room, reactor enclosure, fuel floor, and drywell (when access is possible). During the inspection, discussions were held with operators, technicians (HP & I&C), mechanics, security personnel, supervisors and plant management. The inspections were conducted in accordance with NRC Inspection Procedures 71707, 71709, 71710 and 71881 and affirmed the licensee's commitments and compliance with 10 CFR, Technical Specifications, license Conditions and Administrative Procedure No violations were identifie . Engineered Safety Feature (ESF)-System Walkdown: (71710) The inspectors verified the operability of tha selected ESF system by performing a walkdown of portions of the system to confirm that system lineup procedures match plant drawings and the as-built configuratio This ESF system walkdown was also conducted to identify equipment conditions that might degrade performance, to determine that instrumentation is calibrated and functioning, and to verify that valves are properly positioned and locked as appropriate. The inspectors also utilized methods prascribed in a study prepared for the NRC by-Brookhaven National Lat ocatory using the Limerick Probabilistic Risk Assessment (PRA). The study, entitled PRA-Based System

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Inspection Plan, dated May 1986,'provides inspection guidance by prioritizing plant safety systems with respect

,  to their importance to risk. Abbreviated system checklists which identify components,that are considered to have a high contribution to riskLas determined in the PRA are also provided. Accessible portions of the Loop
 'A' core spray system were inspecte The following procedures, drawings and. tests were also reviewed:

S52. Routine Inspection of the' Core Spray System S52.1.A (COL) Equipment Alignment for Core' Spray Loop 'A' Operation IV 52 Instrumentation Valve List Drawing M-52 P&ID Core Spray System Procedure A-8 Locked Valve List The inspector had no further questions concerning this syste .3 Inspector Comments / Findings (93702) On July 11, power was increased from 85*4 to 88* This increase was possible as a result of expected neutron flux redistribution in the core primarily as a result of gadolinia burnup. With the flux redistribution, this overall higher total core power is possible without generating any localized power peaks which would exceed General Electric's operational recommendations for preventing additional fuel failure At 3:53 a.m., on July 16, a security guard.was observed to be inattentive on post. This was identified by the security shift assistant who immediately compensated the area until the guard was relieved. Security sweeps were made with no discrepancies note The NRC was notified via the EN At 5:30 p.m., on July 17, the 'C' and 'D' main control room chlorine detectors spiked to 3.0 ppm resul. ting in a mai_n control room isolation and a control room emergency fresh air system initiatio Chemistry samples did not indicate the presence of any chlorine and che spike was apparently due to wetting of the detector probes during a thunderstorm. The NRC was notifed via +he EN On July 18, PEco announced it has awarded the Se::urity Contract for Peach Bottom and Limerick plants to Protection Technology, In (PTI). PTI has provided the security at Limerick since March 1987.

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At 13:27, on July 20, the plant. security computer failed and the licensee ton compensatory measures as described in security procedure When the inspector approached an area for which access is controlled by the security computer, he observed workers passing through'the inoperative equipment. However after the workers were aware of the inspector,.many of them would not_ go through the equipment and comments from workers clearly indicated that they were'not familiar with the correct procedure for entering and exiting the area'with the security computer inoperabl The security computer was returned to service that day at 15:39. The inspectur discussed the problem with licensee security personnel who acknowledged a problem existe The licensee indicated they would evaluate solutions to the problem including instructions in GET training and/or posting signs near the equipmen The inspector will follow licensee activities to correct this proble On July 21, at 5:38 p.m., a control room isolation occurred when the

'D' channel chlorine detector spiked upscale. The' spike was apparently caused by the detector pc >e becoming wetted during a thunderstorm. At the time of the isolation the Control Room Emergency Fresh Air System started as designed. The inspector had discussions with licensee management regarding the chlorine detector problem and the following actions wera identified. The system is being redesigned to activate on a signal .nat is more conservative thar b,e current one of four signal. The new design will less m n saurious signal In the interim, the detectors are being placed in the ventilation ducting such that rain will not cause them to get we At 11:57 a.m., on July 24, a telephone threat was received by .he Pennsylvania State Police, Lancaster Barracks, over the emergency 911 telephone numbe The caller statad, "the Philadelphia Electric plant has one hour to close or we will close it dow This is not
' idle threat, you have until 1:00 p.m." The caller further stated, sust sit there and listen, we will execute this. Check the first chapter of Proverbs". The police contacted the Peach Bottem Shift Manager regarding this threat. The Shift Manager contacted Peach Bottom security. The Peach Bottom Shift Security Assistant contacted Limerick security, corporate security, and the PECo load dispatche Security threat procedures were implemented at both plants. The licensee determined the threat to be nonspecific and not credibl '

The Pennsylvania State Police have a tape recording of the threat and have made it available to PEC On Aug;st 10, a radwaste operator was lifting a full radwaste liner out of the 'B' fill station transfer cart using the radwaste crane in tne remote mode. Unbeknown to the operator, the transfer cart stuck to the liner. As the cart and liner were lifted into the air, the cart came loose and fell approximately two feet. Damage was caused to the drive cart motor and to the fill station room electrical cable l

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The.'unciamp' solenoid on the cart vacuum pump'is suspected to have-become clogged with oil, thus vacuum between the liner and cart was never broken. Damage is suspected to be in excess of $2,000 thus the event was reported via the ENS in accordance with 10 CFR 20.403(b)(4).

' 2. Mispositioned Control Rod On July 17, at 3:00 a.m.', control rod 46-55 was found to be mispositioned. During the' performance of a periodic core thermal limit' check utilizing the P-1 computer program, the rod was discovered to be fully. inserted rather than being at its previous full out position. During the check of core thermal limits the maximum fr '* ion of limiting critical pc :: atio (MFLCPR) increased 0 : 3 0.889 to 1.02 The reason for the change was that with an tajmetric control rod pattern the computer everestimated local power. values in certain areas'of the ce:e. The actual effect on core power was minimal.since 46-55 is a peripheral ro Upon discovery of the mispositioned rod, power was reduced by 100 megawatts electric and three additional control rods inserted to achieve a symmetric rod pattern in accordance with procedure ON-104, Control Rod Problems. A core thermal limit check was then performed with no abnormal readings note Followup investigation revealed that roo 46-55 inserted at 8:26 a.m. on July 16 during the performance of a surveillance test which checks the reactor protection system 'A' channel scram function. Prior to this test the

 'B' scram pilot valve power supply fuse for rod 46-55 had blown resulting in a single rod scram during the 'A'
,  channel test. The cause for the t.lden fuse has not yet been determined however the circuit-functioned normally upon installation of a new fuse.

' At the time of the scram, no indication of rod insertion or alarm was observed due to the following: The Process Computer does not update control rod movement if the rod was not selected on the rod select matri l The rod drift alarm and annunciator are not activated during single rod scrams due to the high speed of rod insertion during a scra . i

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. ~ 7 The scram accumulator trouble alarm ant.' annunciator were not activated with the charging vater isolation valve in its normally open positio Lightsuon the Full Core Display ind'.cated that the-rod was scrammed and at the full-in position, but were not noticed by operators _who rely heavily on the process computer video-display for control rod positio The video display does not update'unless the process computer is update At 0:24 a.m. , on July 17, the control room operator performed an 00-7 option 2 to verify all rod position indicators are operable. This program also updated the-- red positions in the process computer :ausing the next P-l' program (at 3:00 a.m.) to indicate erroneous core _ thermal limit problem Since the problem described here could occur at other plants with similar computer software, the licensee disseminated the above information to other'BWR owners via the INPO notepa After discovery of the problem, the rod positions were updated at least hourly by the operator demanding an 00-7 Option 2 printout. The' licensee subsequently revised the computer software so that rod positions are automatically updated every 10 minutes. Surveillance tests are also being revised to clarify how to properly verify that no rods are inadvertently inserted during half scram testing. Additional modifications are under review which will improve the ability to rapidly detect inadvertent rod position change .4 (Closed 88-07-04) Seismic Calculations for HV-11-101G The licensee has reviewed the results of the original analysis supporting acceptability of the Core Spray Cooler Valves HV-11-101A through HV-11-101H and have determined that the as-built configuration is acceptabl The minimum value of acceleration to which these valves have been qualified is 6.7 .9 The highest calculated value from the original ant.lysi s i s 1.63 Even though the support configuration appears to br inadequate, the as-built analysis verified that the installation in acceptabl The inspector had no further questions. This item is close .

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. 8 3.0 Surveillance /Special Test Observations (61726) During this. inspection period, the inspector reviewed in progress surveillance testing as well as' completed surveillance packages. The inspec:or verified that surveillances were performed in accordance with licensee approved procedures and NRC regulation The' inspector also verified that instruments used were within calibration tolerances an that qualified technicians performed the surveillance The following surveillances were reviewed: ST-6-095-450-1 Division I through IV, 125/250 VDC Safeguard Power Distribution Alignment and Voltage Test ST-6-107-590-1 Daily Surveillance Log ST-6-092-311-1 Monthly D-11 Diesel Run SP-ST-016 B Loop Core Spray Pump Minimum Flow Verification Inspection SP-T-001 CRD/HCU 46-55 Troubleshooting RT-6-047-490-1 CRD Withdrawal Stall Flow Test ST-6-107-760-1 Control Rod' Exercise Test SP-ST-016 perforaed flovt rate checks on the 'B' and 'D' Core Spray pump minimum flow lines utilizing a clamp-on ultrasonic flow indicato The

'B' and 'O' pumps share a common minimum recirculation flow path. The test was written to gather data to investigate the potential for pump damage due to inadequate minimum recirculation flow as identified in NP Information Notice No. 87-59: Potential RHR Pump loss. The test resul show that the miniram flow with the pumps operating at_ shutoff head meet the vendor recommendations for the 'O' pump, however, the 'B' pump flow:

was slightly below the recommended value (300 gpm flow vs 320 gpm recommended).

Upon review of the test data by PECo and General Electric engineers, it was determined that pump operability is not adversely affected, however, as a conservative measure, pump operation at reduced flows will be minimized' until adjustments to the flow control valves can be mad The subject of adequate minimum pump flows has subsequently become the subject of NRC Bulletic No. 88-04: Potential Safety-Related Pump los The inspectors will follow the licensee actions related to this bulleti No violations were identifie .0 Maintenance Observations (62703) The inspector reviewed the following safety related maintenance activities to verify that repairs were made in accordance with approved procedures, and in compliance wit 6 NKC regulations and recognized codes

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and standards. The inspector also verified that the replacement part and quality. control utilized on the repairs were in compliance with the licensee's QA progra Work Order Number Description-8804145 Standby Liquid Control System level Transmitter LT-48-1N01C Calibration 8804389 '0' Main Control Room Chlorine Detector Repai .1 The inspector observed work performed on standby. liquid control system (SLC) level transmitter LT-48-1N01C. The function of thi instrument is to sense SLC tank level and to automatically stop the

 'C' injection pump on low tank leve The inspector noted that.a detailed procedure was not available for the calibration of the transmitte The data sheet referenced a generic instrumentation calibration procedure which was present at the work site. Although the technicians appeared to be knowledgable of the calibration method, the lack of a detailed procedure precluded the technicians from having specific written direction on asoects such as special precautions, test equipment setup and restoration to servic These concerns were identified to plant management and the instrumentation and controls department supervision who agreed that a detailed procedure is appropriate and will be writte The inspectors will continue to monitor plant activities for i procedural adequacy and personnel adherence to procedure .0 Review of Periodic and Special Reports (90713)

Upon receipt, the inspector reviewed periodic and.special r'eports. The review included the following: inclusion of information required by the NRC; test results and/or supporting'information consistent with design predictions and performance specifications; planned corrective action for resolution of problems, and reportability and validity of report information. The following periodic report was reviewed:

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Unit 1 Monthly Operating Report - June 1988 The inspector had no questions concerning this repor The inspector reviewed a special report written in accordance with Technical Specifications Sections 3.3.7 2 and 6.9.2 which require l reporting whenever certain instrumentation is inoperable. On June 2, ( 1988 six triaxial time history accelerographs of the Seismic Monitoring

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System were. declared administrative 1y inoperable due to inadequate 18 month calibration surveillance tests. The= licensee was informed by the supplier of the equipment that the-previous calibration method, which'the supplier recommended, was'no_ longer valid and that a new method had been developed. Until the licensee can develop the new procedure and perform the necessary calibrations, the. system, which is backed up by other seismic monitoring systems, is technically inoperabl _ The licensee will test four of the six accelerographs, which are accessible during operation, when the surveillance tests are approved and the remaining two, which are inacces,!ble during operation, during the next refueling outage. In the interim the control room receives other signals of possible seismic activity and has procedures in place to appropriately deal with the situation should it be necessar The resident inspector had no further questions regarding this repor .0 Licensee Event Report Followup (90712, 92700) The inspector reviewed the following LERs to determine'that reportability requirements were fulfilled, immediate corrective action was taken, and corrective action to prevent recurrence was accomplished in accordance with technical specification This LER reports the inadvertent start of D-13 Emergency Diesel during the installation of test equipment to perform a special tes This event was previously discussed in NRC inspection report 50-352/88-15. The inspector reviewed the LER, which accurately described the event, and the inspector also confirmed the licensee's actions to prevent recurrence. The inspector has no further questions or concerns with regard to this LE This LER reports the actuation of 'C' Emergency Service Water Pump (an engineered safety feature) during relay calibration. This event was previously discussed in NRC Inspection Report 50-352/88-15. The inspector reviewed the LER, which accurately describes the even The inspector also confirmed the licensee's actions to prevent recurrence. The inspector has no further questions or concerns with regard to this LE This LER reports the failure to stroke test Primary Containment Isolation Valves HV-061-132 and HV-061-112 (one and a half inch

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valves) following maintenance as require'd by Technical-Specification 3.6.3. Preventive maintenance had_been performed on the motor control center breaker for the two valves and the operations personnel incorrectly determined that the work performed did not constitute maintenance repair or replacement work as defined by-Technical Specifications, however, the nature of the work performed on the circuit breakers' constituted some disassembly of the breaker which would have required stroking the valve to prove operability of the system including the breaker. The event was discovered by the Shift Technical Advisor while conducting a review of the work performed (41.5 hours and 15 hours respectively after the event).

The licensee has demonstrated the valves operable and_the stroke time testing was less than the maximum allowable. The individuals were counseled and'the operations department personnel were instructed, through a memo, to be aware. of the significance of the even The inspector determined that this is a-licensee identified violation which meets the criteria for 10 CFR 2 appendix C for not issuing a Notice of Violation (50-352/88-17-01).

The inspector has discussed the event with station management and has no further questions or concerns regarding this even This LER reports the initiation of Reactor Enclosure HVAC 'B' during the performance aof a surveillance test. This event was previously discussed in NRC inspection Report 50-352/88-15. The inspector reviewed the LER; which accurately described the event. The inspector also confirmed the licensee's actions to prevent recurrence including the replacement of instrument isolation valve (IBVL-76-171.1). The inspector will review the procedure changes when issue The inspector has no further questiens at this time regarding this even ! , No violations were identifie (Safeguards Event Report) This report addresses the June 21, 1988 incident of inattentiveness by a security force member as discussed in section 2.3 of Inspector Report 50-352/88-15. The inspector had no further questions regarding this event, i

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7.0 Systematic Assessment of Licensee Performance The NRC Systematic Assessment of Licensee Performance (SALP) Report was issued on July 7, 1988. This . report assessed performance for the period of February.1, 1987 to April 30, 1988. The functional area ratings were: Plant Operations 1 Radiological Controls 1 Maintenance 1 Surveillance 1 Engineering / Technical Support 2 > Emergency. Preparedness 2 Security and Safeguards 1 Safety Assessment / Quality 1

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Verification The definitions of these performance categories are: Category Reduced NRC attention may be appropriate. Licensee management attention and involvement are aggressive and oriented toward nuclear safety; licensee resources are ample and effectively used so that-a high level of performance with respect to operational safety is being achieve Category 2. NRC attention should be maintained at normal level Licensee management attention and involvement are evident and concerned with nuclear safety; licensee resources are adequate and reasonably effective such that satisfactory performance with respect to operational safety is being achieve A SALP meeting was conducted on July 27 at which time the licensee was-provided the opportunity to discuss their plans to improve performanc .0 Licensee and NRC Meeting in Region I Pertaining to Limerick Fuel Leak On July 27, 1988, the licensee met with NRC management and staff members-to discuss the Limerick i fuel leak. (Attachment A). The licensee presented data and assessments (Attachment B) regarding the cause of the leak, actions taken since the discovery of the leak; abort term corrective actions, intermediate term corrective actions and the.long term corrective actions planned. The following is a brief summary of those action Short Term The licensee identified the leak then took steps to pin point the leak by control rod manipulations. The licensee then reduced power to below GE recommended guidelines for fuel leaks that may be CILC (CRUD Induced Localized Corrosion) related. The licensee also began a program to enhance the precoating of the condensate polishing units in order to remove more copper from the syste . .-  ; i

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Intermediate Term The licensee has identified a plan for the reloading of the reactor

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for cycle 3 and is continuing to improve the' copper removal qualities of the condensate polishing unit Long Term The licensee has outlined a plan to reinove all the copper components in the feed and condensate system by the replacement of the main condensers in both Limerick units with titanium tubed condenser .0. Bulletin 88-05 "Nonconforming Materials Supplied by Piping Supplies Inc."

(92703) This bulletin identified concerns regarding piping materials sold to power plants by the above supply compan The bulletin specifically directed licensees to identify any materials bought through the company and if installed in the plant, report to the Commission within 48 hours and perform a justification for continued operation analysis. The resident inspectors are monitoring these reports and will collectively summarize the status of them in the monthly repor The following components were identified this report perio Two flanges in the 16 inch header that connects the spent fuel pool cooling system with the RHR system were identified to have insufficient hardness. This line is normally not in use and the flanges are isolated on both sides with administratively locked valves. The line is normally depressurized and is only used as a backup in the event that the Spent Fuel Pool Cooling System is nonoperationa Three 150 pound raised face flanges exhibited high hardness when tested. The flanges are installed in the one inch piping to the RHR Service Water radiation monitoring syste This is a low pressure syste Three flanges were identified to have high hardness values. Two, one and one quarter inch 150 pound flanges are located in the 011 diesel generator jacket water cooling system on a pipe between the jacket water expansion tank and the air cooler coolant pump. The third two inch 150 pound flange is located on the discharge of the 1A diesel oil transfer pump for 011 diesel generato One 1 inch 600 pound flange was identified in an inaccessible area during operations. The flange is installed in the primary containment leak test pipin .

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Five three inch 600 pound flanges were found in the suppression pool level instrumentation which is.used to divert pump suction in the event of suppression pool high leve By the end of this report period, 46 flanges had been tested, and.13 of these exhibited hardness values which were outside of.the expected rang Three additional flanges were determined to be inaccessible. .The justification for continued operation (JCO) related to two inaccessible RHR service water flanges was also reviewed. The inspector had no further questions concerning the JCO On August 3, Supplement 2 to NRC Bulletin 88-05 was issued temporarily suspending the testing, field measurements, records review and JC0 preparation previously requeste The inspectors will continue to follow NRC and licensee actions concerning this issu .0 Assurance of Quality During this inspection period, the resident inspectors observed the assurance of quality and thorough followup of the licensee's investigation and management oversite in the following area Mispositiont j control rod 46-55. Section 2.3.1 of this repor Presentation of licensee's actions to the region regarding the fuel leak. Section 8 of this repor However, an area where effective assurance of quality was not demonstrated is presented in section 4.1 of this report, when the inspector observed that detailed procedures were not available for the calibration of a transmitter in the Standby Liquid Control Sycte .0 Exit Interview (30703) The NRC resident inspectors discussed the issues in-this report throughout the inspection period, and summarized the findings at an exit meeting held with the Plant Manager, Limerick Generating Station,' on August 12, 1988. No written inspection material was provided to licensee representatives during the inspection perio .- . -

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  ' LIMERICK FUEL MEETING July'27, 1988L
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![RC Name   Title 14. Butle Director, Project Directorate I-2 R. Clark   Limerick Project Manager W. Johnston  Acting Director, Division of Reactor Safety G. Sjoblem  Acting Director, Division of Radiation-Safety and Safeguards E. Wenzinger  Chief,' Projects' Branch No. 2 J. Linville  Chief, Reactor Projects Section 2A H. Williams  Project Engineer T. Dragoun   Senior Radiation Specialist T. Kenny   Senior Resident-Inspector L. Scholl   Resident Inspector PEco K. Carrabine  Project Management Division G. Barley   Plant Supervisory Chemist R. Scholz   _ Supervisory Engineer G. Hunger, J Manager, Mechanical Engineering L. Corsi   Engineering L. Rubino   Fuel Management Section R. Dubiel   Support Plant Services G. Leitch   Vice President, Limeric M. McCormick, J Plant Manager Pennsylvania D. Ney   Department of Radiation Protection

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AGENDA LIMERICK FUEL PRESENTATION NRC Regional Office, King of Prussia-Wednesday, July 27, 1988

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          .M. P. Gallagher UNIT 1, CYCLE 2, OPERATING STRATEGY     Limerick Reactor Engineer-L. F. Rubino UNIT 1, RELOAD 2,. STRATEGY
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Superintendent - Fuel Management COPPER REDUCTION STRATEGY: R. J. Scholz Introduction Supervisory Engineer Nuclear Engineering G. K. Barley Background /Short Term Actions Limerick Plant Supervisory Chemist Condensate Filter /Demin R. J. Scholz Performance Optimization CONDENSER RETUBING STRATEGY: L. Corsi Project Description Engineer - Nuclear Engineering Schedule K. Carribine Engineer - Nuclear

Engineering

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     '

o 4-  !

       *

O . Z i j 3- i i

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     ,   ;

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        :

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0- , i , , i i i e i i i i i 10 22 24

  ~

10 12 14 1'6 20 NODAL EXPOSURE (GWD/ST) O HF2 R2 + VY R3 o H2 R1

   .
. -

        - -     "

_6 --

 .
.. .
-

l l

  .
*                i Figure 2         t I
                '

LIMERICK 1 CYCLE 2

      ,

POWER vs EXPOSURE l l

                !
              .
                '

100 . i

       ,
         .
           .  .
            .

t  ; i . 4 _

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      (  I     i gg .
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           .
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           ,  ;
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       -     ;    :
     - >-  -  - - - - - - -  - - - - - - - -  - - -

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l 94 . O 9 - w .

            .
            :

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            . . . . . . .
      '

o . - . - . .k, /

         '

i

     - %

86 0- ~ k. /-- - - - - - - - - - - - - - - --

   ,I          :

I

         -- - -~  - - - -  *
   - ' - - -  - - - - - + - - - -
             >  .

8 4 ~l' . .

             !
     -      .

_l --*------- - - - - - - - - - - -- - - - - - - - - - - - - S21 + - --

         -
             :

l l  :

!   80 ,    ,  ,  ,  ,  ,

Apr 25 May 19 Jun 30 Aug 22 Oct 11 Nov 30 Dec 26 Jan 14 1 1988-89 ( l e

    *             l

_ , . _ _ . _ . - _ . _ , . . _ . . _ _ _ _ _ _ _ , . _ _ _ _ _ _ _ _ _ _ , . _ , . . . _ . , _ , . , _ _ _ _ _ . , _ _ _ _ , . . . . . , _ , _ , . , , , _ , , _ _ .,,_,_ ,_ _, , , . . _ _ _ _ _ _ .

. . _ . . . ~ .

   - - - - -
. .
.
.

e LIMERICK 1 CYCLE 3

RELOAD STRATEGY l l

_ _ , . _ . . _ _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ .- . _ _ _ _ , . _ . , _ -

_- . _ _ - l . .

.
.

LIMERICK 1 CYCLE 2 LOADING

   % HEAT TREAT BUNDLE
  # BUNDLES TOTAL  Go TYPE 38  33 IC-163  116 308 86  41 IC-248 33  0 IC- 27 8,  72a 268 100  100 R1-320 764 CYCLE 3 LOADING (BASE CASE)

BUNDLE TYPE # BUNDLES % HEAT TREAT IC-248 200

   .

MAXIMlZE IC-278 72e 268 100 R1-320 224 100 R2-318/322 i

764 l

_ _ _ _ _ _ _ - . _ . _ - - ._ __ __. ... - - _ _ . . _ _ , _ _ . _ _ _ -.._. - _,

. PRIORITIZED INSPECTION PROGRAM  : l

* SIPPING TO BE PERFORMED DURING CORE OFFLOAD
# 648 BUNDLES WILL BE SIPPED
# FUEL INSPECTIONS TO CLEAR RETURNING IC BUNDLES
 * INSPECTION BASED ON 10% SAMPLING OF MAJOR ROLL'S IN 2.48 AND 2.78 FUEL ,
 * MAJOR ROLilNVOLVES A MINIMUM OF TEN RODS
 * FUEL MUST BE VISUAL STANDARD 4 OR BETTER FOR RETURN
* PRIORITIZED INSPECTION PLAN
 * CLEAR THE MOST BUNDLES POSSIBLE IN A GIVEN TIME l
 * MODIFICATION OF PLAN MAY BE NECESSARY AS INSPECTION DATA IS OBT AINED
 * PLAN MUST HAVE BUILT IN FLEXIBILITY TO ACCOMODATE CHANGE

 * CAN INSPECT UP TO 38 BUNDLES DURING CORE OFFLOAD AND SIPPING

. _ _ . -- .. - - - . -
   ---. - _ _ . . . .

  .-  - ._
'
:
.

PRIORITIZED INSPECTION PROGRAM (cont.)

. FORIC-248

 # BUNDLES   # BUNDLES TO INSPECT  FOR CLEARED 10    13

' 15 34 20 99 25 138 30 205 35 261 40 294 43 306 FORIC-278

 # BUNDLES   # BUNDLES TO INSPECT  FOR  CLEARED 10    11 15    43 20    70 22    72
*
* IC-248 SHOULD BE INSPECTED PRIOR TO IC-278 (MORE BUNDLES CLEARED FOR A GIVEN NUMBER INSPECTED WHILE MINIMlZING CRITICAL PATH IMPACT)

,

--- , - . . , - - - - - ,-,,e.-,-n-..,e.-
   -
    , - - , , , , , , .,,n...._..~, , - - , ,,- - - - - n,- - - , - , , - , , - . , , _ -

     - - - - - -  - - - _ - - -
-  . _   ..
.
.
.

e i l

l LOADING SCENARIOS

   -

6 I l .,

           .

.I __ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . _ _ _ _ _ . - _ _ _ . _ _ _ _ _ _ _ ._ _ _ , , _ _ _ _ . - , , _ , . . _ . ~ _ _ , . _ , _ , , . . . , , _ _ _ , . - .

'
.
.
.

OPTIONS SUMMARY COASTDOWN INCREMENTAL ADDl'iioNAL OUTAGE REPLACEMENT FUEL CYCLE LICENSING EXTENSION TOTAL ENERGY COST COSTS COSTS # COST 3 COSTS SCENARIO (10'S) (10'S) (10 S)

INSPECTED (10' S) (10' S)

1. 1 BASE CASE O O O 65 .40 2001C-248 (9 DAY)

7210-278

1. 2 REINSERT CYCLE 1 0 .6 3.50 DISCHARGED FUEL (44 D AY)

TO EDGE 1081C-248 7210-278 9 210-094e (REIN)

1. 3 DISCHARGE ALL 2.78 0.25 .05 27 0 8.50 REINSERT CYCLE 1 FUEL & USE LGS 2 FUEL 1081C-248 9 2 (C-094e (REIN)

72 lC-094e (L2 FRESH) e4 DISCHARGE ALL 2.78 0 .05 36 0 6.85 USE LGS 2 FUEL 20010-248 72 lC-094e (L2 FRCSH)

1. 5 DISCHARGE ALL 2.78 0.25 =0 0 43 .25 272IC-248 (14 D AY)

85a DISCHARGE ALL 2.78 0.38 .28 REINSERT CYCLE 1 FUEL 180 IC-248 9 210-054e (REIN) e6 FUEL SHARING 0.33 .2 36 0 1.23 j 20010-248 72 IC-094e L2 FRESHIN L1 72 IC-094e (REIN) FROM  ! L1 IN L2 INITIAL CORE '

l

__

.

.

.

. DECISIONS l

* DO NOT USE LGS 2 INIT;AL CORE FUEL (HIGH COST) l
* DO NOT USE FUEL SHARING   l
* LGS 2 LICENSING RECORD IS CLOSEO - SHARING WILL
    !

ALLOW REOPENING

* NO COST ADVANTAGE OVER SCENARIO #5
* STRIVE TO ACHIEVE SCENARIO #5 i
* INSPECT IC-248 BUNDLES FIRST
* CLEAR AS MANY IC-248 AS POSSIBLE OFF CRITICAL PATH
* DISCHARGE 11 SUSPECT AND SYMMETRIC COUNTERPARTS
* USE LOW EXPOSURE REINSERT FUEL AS MAKEUP IF INSUFFICIENT IC-248 ARE CLEARED
* LICENSE A LOADING WHICH WILL BE BOUNDING UNDER GESTAR FOR THE FINAL LOADING
* LOADING WILL MAXIMlZE THE NUMBER OF HEAT TREATED GD RODS IN THE CORE INTERIOR = 96% ALL RODS HT,67% GD HT l
    :
* GE WILL PROVIDE DETAILED (ROD BY ROD) INSPECTION LIST I l

THIS FALL

* GE WILL PROVIDE A LIST OF BUNDLES CLEAPED EACH DAY DURING INSPECTION

          .

Reactor Coolant Copper , O

              .

Limerick Generating Station g 40 , v

O t Z 20- - - - - - - O N m t_ A e C D 0

         -

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g a s s aga s a s p i a a gi i a a sg s s a a ge a a s g a a a sg i a a u

,

DEC JAN FEB MAR APR MAY JUN JUL l 1985 1986

..         Cyr:]e 1
              ,
.

___-___-_________________--___ _ ________ - --- -_- ___. --

             **

Beactor Coolant Copper ,"

             .

Limerick Generating Station g 40 a O La Z " 2G-

            --
    -

O I

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e D D-- t 1004l- y

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7~- 7

         -
          , ----- -.

o 75 - l-

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=  2s-         - --

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. .

    .
    ....... ,,..  ,.........,....,....,....,,...,.

AUG SEP OCT NOV DEC JAN FEB MAR APR , 1986 1987 o Cycle 1 e,

, _ . _ _ _ - _ - - _ _ _ _ _ _ _ _ - _ _ _ - - - _ _ - _  - _ - - _ _ _ _ - _ _ _      -

__ _ _

                 . ..

Reactor Coolant Oopper

                 .

Limerick Generating Station

  ^40- O a
   &

v

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s j k

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p \ _ _ . C D O-i b 10 0 - -- ~ --- - - II y- i r \

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+-
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C l D ................................................................................... Q .......................

5 5 5 5 g5 5 5 5 g5 5 5 5 Ig 5 5 5 5 gI 5 5 5 g5 5 5 5 3 5 5 5 5g5 5 5 5 y 5 OCT NOV DEC JAN FEB MAR APR MAY 1987 1988 Cycle 2

     .- .. - -.  .  -- - . -. - - - . - . - . . _
                --. .

_ _ _ _ _ _ . - _ - . _ _ . _ _ _ . _ _ - _ _ _ - - __ -_ _ __ _ _ _ - - _ _ .- v

         . ..

Feedwater Copper

         .

Limerick Generating Station 2- - O_ ^ v _ sn Na O 1 -- N < L_ C

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5

        -- - -

r - - n e w rv

       -

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      - - -  - -
-
      - - - *

,

.2 25-C 011-      -

i=i' 'r 'i=' i' i =i 'i' r si 1986 1987 ( cycle 1 ' -

 -___ __-_ _ _ _ _ _ _ _

                . ..

Feedwater Coppar

                .

Limerick Generating Station m 2 _o Q_ v _ s

   <a      '      a O y_ .... .... . . . . . . .. .. . ........
           .....l...
             ................j eA p
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     -

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M 50- - i-------------------------- -

            .
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,

c 1 i a o_ . .. ............ ...... .. ....... .... ............. . ... ............ _ m.............................

               ,
     ............,,....,...,...,.............,......a SEP OCT NOV DEC JAN FEB MAR APR MAY JUN 1987     1988 Cycle 2

__ _ _ . _ _ _ _.._ _ I

        . _ _ _ . _ . . _ _

i - __

. _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ - _   _ _

_ _ ._ _

             * ' *

Condensate Filter Dmtinerclizcr Conductivity

             .
;    Umerick Generating Station 2....             l
,

w E

:

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a g

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i is 13 i gi i .iig .i ii3 i i - . OCT NOV DEC JAN FEB MAR APR " 1986 1987 > Cycle 1 g )

. .. .-   .- - -. - - . - . . _ . .  --.  .
 .- _         _ _ _ _ _ _ _ . _ _ . - . _ _ _ _ _ _ _ _ _

Cbndensate Filter DemirEralizer Conductivity ,

             -
             ,o
             .
              .
;      Umerick Generating Station
            -
      . . . _ . . . . . . .
...    . _

I- l .

.

 . ... ........... ... ..........................................................    .......N.................

2 ..... E \ s VW

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  .
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.
.
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             .

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*

i : ,, n......................r..Il..............1.r........7................................-n

 ....
   .

y.,........y ................... ..................... ................... ....... ....................

          ...... ......................
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              -
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          ..... .......................

i....................... - g .. ................................................................ .................. O Ff NF * F g s y a I I I I[ T I 'Y I gs iIQF y a e F"{ a s

           : l5 s e i

~ OCT NOV DEC JAN FEB MAR APR MAY

1987 1988 m a

i Cycle 2 7 h-

 ._ _ .-  ._ .

_ _

      -

SHORT TERM COPPER REDUCTION

      .
* INTRODUCTION   -

R J SCHOLZ

* HISTORIC BACKGROUND    -

G K BARLEY

* SHORT TERM ACTIONS    -

G K BARLEY

     .
* CONDENSATE F/D   OPTIMIZATION -

R SCHOLZ l * RESULTS OF OPTIMIZATION - R J SCHOLZ i i hoacts er

    " ' 'I"
-- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _

.
.
.
.

_ _

  -
  -
  -
  -
  -
  -

_

  .
  -
  -
  -
  -
 *

N - O

  .
  -

I _ T t s C e n r e n U 4 _1 6 L

  .

D mm O h ew e R T - N C I C S A I

 ~
 /

C _ C C _ _ S _

. L I G  -

C I -

 * e
  -
  -
  -
  -

_ _

.l , l ! I

_ _ -

;    ..

i

;    -

U l

, HISTORICAL BACKRROUND

) * PRIMARY COOLANT COPPER HISTORY '

i ! * FEEDWATER COPPER HISTORY

i

* CONDENSATE F/D PERFORMANCE l

i l ! l i i

) i i ! ! ! ! .

.
  -

P900MCD ], BT l l WileC L*. (Devent

. _ .. . -
        . .
        . ..
      '
         .
       ,

SHORT TERM ACTIONS

l !

       <
* CONDENSATE F/D SYSTEM CHECKOUT

! ! , l * REDUCE HIGH AIR INLEAKAGE . . ! I . !

l

}

l i .

      -

numan 1, w i uusc D, inmel i _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _. -

       .- _ - _ _

_ - - . _ . _ . - - - _ - .

       . ..
        -

i CONDENSATE F/D OPTIMIZATION

:

   * EXPERIENCE FROM SWEDEN

'

i * EXPERIENCE FROM GERMANY

-l ! * US EXPERIENCE i . i l ! ! ! i i

!

i .

_

     -

r ,

     .1 l      w- v-

_ _ - - - _ - _ _ _ - - _ - _ _ - - _ - _ - - - -_ --

- . . - - _ .
    * P 4
    .
    .
   -

US EXPERIENCE l OPTIMIZATION HORK BEING DONE AT :

 .

, e FITZ PATRICK

  ~

ji e DRESDEN

  -

l . i e HATCH i l ! * MONTICELLO i ! ! '

* QUAD CITIES
'

I IMPROVEMENT SEEN IN WATER QUALITY AND~ FILTER RUN LIFE

   '

! 1 .

 =L:. - . _ _ __ _ _- _ _ -

_ _ _ - -- .. _- _ _ . ..

  . ..
   -

   '

i I i

i SUMMARY OF EXPERIENCE l l . l CRITICAL FACTORS

* PRECOAT MATERIAL l

i * PRECOAT PROCEDURE / METHOD l

! * ELEMENT CONDITION

* BODY-FEED HIGHLY EFFECTIVE i

l i

  -

e k ]

. .- _ .
  . ..
   .
   .

LGS OPTIMIZATION PROGRAM e PRECOAT MATERIAL

  '

I e IMPROVED PRECOAT PROCEDURE i I '

* PRECOAT LOADING
  '

! e REPLACEMENT ELEMENT STUDIES , ' i l * CARBOXYLIC RESINS l

i i

l I _ )

 ":" K " .

l

_ _ _ _ _ __ . _ . _ _ _ _- . ._ . . ..

           .
           .
'

RESULTS OF OPTIMIZATION

)

L AMOUNT OF DATA ! J ! * LIMITED , )

HATER QUALITY SEEN i ! * IMPROVED . ' l

IN RUN LENGTH SEEN

! * INCREASE I i i b i I L l ! !

! } l ! I __ WE b (DWMil ! _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

-- ; e -.

O

.c o

i l

:.,.    -
    ,
    ';

f

  ,  l
    !

'

    ,

t

    '
   '9
   ,
    ;

PHILADELPHIA ELECTRIC COMPANY-CONDENSER RETROFIT PROJECT .

   '

LIMERICK GENERATING-STATION

    ,

t

    ;
   .  -
   .

TYPE

    :
 - MODULAR TITAN!UM RETUBING-
    ,

I i

    !
    :
    '!

i

    !
    !
    !
    '!

l I

    !
    .;

l

    !
    . . _ . .

    .. _ _ . -
, . .

o

:

O ADVANTAGES  :

 -

MINIM 1ZE POTENTIAL FOR CILC DAMAGE TO NEW FUEL l

 -

IMPROVE REACTOR VATER CHEMISTRY ,

 -

OPTIMlZE THERMAL PERFORMANCE

 -

INCREASE UNIT AVAILABILITY AND RELIABILITY

      -
       ,
 -

REDUCE INLEAKAGE TO CONDENSER

       ,
 -

REDUCE CW PUMP MAINTENANCE

 -

MINIMlZE LOAD REDUCTION DURING HIGH CW INLET TEMPERATURE

 -

REDUCE SUSCEPTIBILITY OF IGSCC/lASCC

 -

ELIMINATE FUTURE RETUBINGS

       :

l r

       !
       ,

_ s- + w --- --- + %

-

m --- -- ym-*- -

  -r-u - 7- - - ywg w- -- -r -n, y
    -
     ,v7y-- ,--- -w

  .
. - .
'o

.

f DESIGN FEATURES

- TITANIUM TUBES
- TITANIUM TUBE SHEETS
- ROLLE0 AND VELDED TUBE TO TUBE SHEET JOINT
- ELIMINATED POTENTIAL FOR TUBE V!BRATION  ,
    -
- OPTIMIZE CONDENSER PERFORMANCE
 - TUBE SIZE / LAYOUT
 - NEW C0ATED WATERBOXES

- REDUCE CIRCULATING WATER (CW) PRESSURE DROP THRU

, BUNDLE

    :
- REDUCE HP SHELL BACKPRESSURE AT RATED LOAD a
    ,

i l l i l _-________-_-_-___

_ ,

 .

s t

              !
              !
              .
              )-
              ;

ASSOCIATED ENGINEERING ISSUES I

             -
          - ON-LINE CLEANING SYSTEM
          - C0ATING OF CW AND CROSS-AROUND PIPING  l i
          - CATHODIC PROTECTION SYSTEM '
              !
           - WATERBOXES
           - CROSS-AROUND PIPlNG   l t

i

          - STEAM AND CW SIDE WATER CHEMISTRY-l
              ,
              !
              .

l _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ . _ - - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _

. ..

b

:

o

:

PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION CONDENSER RETROFIT PROJECT UNIT 2 -- PLANNING FOR PRE-FUEL LOAD IMPLEMENTATION i l BIDS REC'D FOR TITANIUM TUBES , TITANIUM TUBE SHEETS l l MODULE FABRICATION 08/01/88 -- PURCHASE TUBES / TUBE SHEETS 09/01/88 -- RELEASE MODULE FABRICATION l 02/89 -- DECISION TO IMPLEMENT ' 05/89 -- RECEIVE MODULES 08/01/88 -- LOAD FUEL I I l UNIT 1 -- EARLIEST OPPORTUNITY FOR RETROFIT IS i THIRD REFUELING OUTAGE

4 }}